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235U enrichment or UF6 mass determination on UF6 cylinders with non-destructive analysis methods
ARTICLE IN PRESS Nuclear Instruments and Methods in Physics Research A 612 (2010) 309–319
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Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
235
U enrichment or UF6 mass determination on UF6 cylinders with non-destructive analysis methods
R. Berndt a, E. Franke b, P. Mortreau a, a b
The European commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, TP 800, Via Fermi, Ispra, Italy Consultant, Dresden, Germany
a r t i c l e in fo
abstract
Article history: Received 10 July 2009 Received in revised form 22 September 2009 Accepted 2 October 2009 Available online 1 November 2009
The measurement of the 235U enrichment and the mass of UF6 in 30B and 48Y transit cylinders are important safeguards verification tasks of IAEA. In the framework of a project aiming to establish an unattended measurement station at isotope enrichment facilities (Lebrun, 2007) [1], a study was carried out to describe the state-of-the-art of non-destructive assay methods applicable to UF6 cylinders. The objective of the present work is to provide a feasibility assessment study of all known NDA techniques applicable to the quantitative verification of all uranium categories involved in an enrichment processing plant. Based on this investigation, the most appropriate techniques were then investigated for suitability of use in an unattended measurement station. & 2009 Elsevier B.V. All rights reserved.
Keywords: Gamma ray spectrometry Uranium enrichment Neutron assay MCNP calculations UF6 Non-destructive assay methods
1. Introduction
2. Radiation characteristics of uranium hexafluoride
Nuclear safeguards inspections have to verify the amounts of declared nuclear materials such as uranium and plutonium. The 235U enrichment as well as UF6 weight determination in 30B and 48Y transit containers used in enrichment plants represents a significant portion of this program of verification. In the framework of a project aiming to establish an unattended measurement station at an isotope enrichment facility [1], a study was carried out to describe the state-of-the-art of non-destructive assay (NDA) methods applicable to UF6 cylinders. Well-established NDA methods as well as techniques based on the analysis of delayed neutrons and delayed photons were first investigated. The technical requirements (such as operation in unattended mode, counting time compatible with the need of plant operation, and implementation timeframe limitations), have then orientated the work towards a more specific investigation. Previous results of in field measurements as well as MCNP calculations, lead to a proposal for an unattended measurement station of UF6 cylinders.
This section describes the emission of gamma rays and neutrons from UF6 and from the daughter isotopes for U of natural origin or U from reprocessing. Since alpha and beta particles emitted by the uranium isotopes as well as uranium X-ray cannot cross the thick wall of UF6 cylinders, the only visible signature comes from gamma and neutron radiation. 2.1. Description of uranium isotopes 2.1.1. Natural uranium In its natural state, uranium consists of three isotopes 234U, 235 U and 238U. Extremely low quantities of 236U due to activation processes such as cosmic ray activation can also be present [2]. Mass spectrometry analysis of U samples shows that the isotopic composition of natural uranium varies slightly as a function of its geographical origin. By measuring a set of different U ores, it was shown in Ref. [2] that the range of natural variation of the three natural isotopes expressed in at% is 234
U
0:0054 0:0051 at%;
0:7201 at%;
Corresponding author.
E-mail address: [email protected] (P. Mortreau). 0168-9002/$ - see front matter & 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.nima.2009.10.060
238
U
235
U
0:7207
99:2748 99:2739 at%
2.1.2. Reprocessed uranium In addition to the three naturally occurring uranium isotopes, the reprocessed uranium may contain all the isotopes from 232U
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Table 1 Nuclide concentrations (g/t) of a PWR fuel with a burn up of 33000 MWd and an initial enrichment of 3.2%.
Counts
3
4
4.248E–04 4.796E–03 1.177E + 02 7.516E + 03 4.437E + 03 2.951E–05 9.434E + 05 0.000E + 00 2.203E–21
6.138E–04 5.118E–03 1.203E +02 7.516E + 03 4.438E + 03 2.969E–05 9.434E + 05 0.000E + 00 2.203E–21
7.233E–04 5.440E–03 1.230E+ 02 7.516E+ 03 4.437E+ 03 2.450E–05 9.434E+ 05 0.000E +00 2.203E–21
102.31 keV - Th231
98.44 keV - UKα1
1
95.87 keV - Pa Kα2
100000
2.810E–04 4.643E–03 1.163E+ 02 7.516E+ 03 4.437E+ 03 1.021E+ 01 9.434E+ 05 0.000E +00 2.203E–21
94.66 keV - UKα2
1000000
Discharge cooling time (year)
93.35 keV - ThKα1
0.000E + 00 0.000E + 00 2.574E +02 3.201E+ 04 0.000E + 00 0.000E + 00 9.677E +05 0.000E + 00 2.203E–21
92.30 keV - Pa Kα2 92.367 keV - Th234 92.792 keV - Th234
69.8a 1.592 105a 2.457 105a 7.037 108a 2.342 108a 6.75d 4.469 109 23.5 min 14.1 h
Discharge
89.95 keV - Th231 89.96 keV -Th K α2
234
84.21 keV - Th231 84.8 keV - PbKβ1
U U U 235 U 236 U 237 U 238 U 239 U 240 U 233
Charge
82.09 keV - Th231
232
Half-life
81.23 keV - Th231
U isotope
10000 high enriched Uranium
1000 low enriched Uranium
100 80
85
90
95
100
105
Energy (keV) Fig. 1. Uranium spectra taken with a germanium detector—80–105 keV region [4].
to 240U. Table 1 shows the example of the U isotopic composition of a LWR spent fuel reactor (burn up= 33000 MWd, initial enrichment= 3.2%) at the end of the irradiation. The isotopes 237U, 239U and 240U have a very short half-life and 233 U is present in very small quantities. 2.2. Gamma ray spectra Typical features of U spectra as well as the difference between enriched natural UF6 and enriched reprocessed uranium are illustrated by Figs. 1–3. The low energy part of the U spectrum (80–105 keV region—Fig. 1) is characterized by an overlapping of many X and gamma rays which contains the signature from 235U, and from 231Th and 234Th, daughter products of 235U, and 238U, respectively. The most intense lines (U Ka at 98.43 and 94.65 keV) are due to self fluorescence of the material itself. This region is used to determine the 235U enrichment with the intrinsic calibration method [3]. The 100–210 keV region (Fig. 2) shows the typical lines of 235U, the most intense line of which (185.7 keV) is used to determine the enrichment by application of the enrichment meter principle. Fig. 3 shows spectra of depleted natural UF6 and depleted reprocessed uranium stored in transit containers in the 50–2500 keV range. The 235U gamma lines are strongly attenuated in the container walls but the 185.7 keV is still intense enough for
the measurement of the 235U enrichment. 238U is identified with the gamma rays of its daughter nuclides 234mPa (at 766.4 and 1001.0 keV) and 234Pa (880, 883, 925 and 926.7 keV). That isotope is in secular equilibrium with 238U after several periods of half-life (24.1 days of 234Th). Its signature can only be used if the history of the individual drum is known which is not compatible with the use of automated systems. The depleted reprocessed uranium presents several specific lines emitted by the decay products of 232U (212Pb at 238 keV, 208 Tl at 583, 860 and 2614.6 keV) or by contribution of the scattered radiation of the 2614.6 keV line as well as the single and double escape peak associated with this line. 2.2.1. Other possible features in UF6 spectrum The spectra shown in Fig. 3 were measured with depleted UF6. However, when measuring high enriched UF6, the 1274 keV line from 22Na may become visible in UF6 spectra [5]. This line is produced from the 19F (a, n) reaction and could be used to reveal the presence of HEU material hidden behind a thin layer of LEU in a UF6 drum since the half thickness at 1274 keV is about 2.5 cm in solid UF6. The presence of lines from 137Cs and 233Pa (a daughter of 237U) were also observed during the measurement of UF6 product from reprocessing of spent fuel [6]. In spite of the fact that the backscattered peak of 137Cs is located at
ARTICLE IN PRESS
194.94 keV - U235
185.715 keV- U235
182.61 keV - U235
163.33 keV - U235
150.93 keV - U235
143.76 keV - U235
140.76 keV - U235
10000 high enriched Uranium
311
198.9 keV - U235 202.11 keV - U235 205.311keV - U235
Counts
100000
120.91 keV - U234
114.4 keV - UKβ22 114.6 keV - UKβ23
108.99 keV - ThKα2 109.178 keV - U235 110.41 keV - UKβ3 111.30 keV - UKβ1
1000000
105.362keV - ThKβ1 106.581keV - Th231
R. Berndt et al. / Nuclear Instruments and Methods in Physics Research A 612 (2010) 309–319
1000 low enriched Uranium
112.8keV - Th234 / U238
100 100
120
140
160 Energy (keV)
180
200
Counts
766.4 keV234mPa
100
Tl (2614 keV) peak of
208
1592 keV Double escape
Tl
peak of 208Tl (2614.6 eV)
X
2103 keV Single escape
X
860.37
185.7 keV
keV208
X
1000
727.33 keV212Bi
10000
583.4 keV208Tl
238.6 keV212Pb
Fig. 2. Uranium spectra taken with a germanium detector—100–210 keV region [4].
X
1001 keV-234mPa
X X
10
X
X UF6, U from natural origin enr = 0.3%, 48Y container UF6, U from reprocessing, enr = 0.252%,48Y container
1 0
500
1000
1500 Energy (keV)
2000
2500
Fig. 3. UF6 spectra, from natural or reprocessed Uranium-The peaks marked with an X appear only in reprocessed Uranium.
184.2 keV, the determination of the 235U enrichment based on the 185.7 keV line analysis in Ref. [6] for these cylinders were very good.
are mainly emitted by 238U. 235U cannot be measured directly with passive neutron measurements. Fig. 4 shows the example of a 100 group neutron spectrums for natural UF6.
2.3. Neutron output 3. Characteristics of the UF6 cylinders In UF6, the neutron output arises primarily from spontaneous fission of 238U, and by 234U alpha decay and subsequent 19F (a, n) 22 Na reaction. Table 2 gives the relative contribution of the different U isotopes to the primary neutron production and for four typical enrichments. For natural uranium, the 234U and 238U contributions are almost equal. The 234U-produced neutrons dominate the (a, n) intensity of low enriched 235U. The spontaneous fission neutrons
The most commonly used cylinders are the 30B (2.5 tons UF6) cylinders for low-enriched product and 48Y (12 tons UF6) cylinders for natural feed and 48G for depleted uranium storage only. The nominal length is 206 cm for the 30B and 380 cm for the 48Y. The cylinders are constructed of ASTM A-516 steel [8]. The measurements performed in Ref. [9] showed that the 30B
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Table 2 Calculated relative contribution of the U isotopes to the total output neutron signal [7]. Isotopes
Amount (%) Neutron source 1/s/ton UF6
Relative yield (%)
Spontaneous fission UF6(a,n) Depleted U 232 U 234 U 235 U 236 U 238 U
0 0.00234 0.3 0 99.6977
0 0 0 0 9.96E + 03
0 7.85E+ 03 1.86E+ 03 0 7.53E+ 03
0 28.9 6.9 0 64.3
Natural U 232 U 234 U 235 U 236 U 238 U
0 0.0056 0.718 0 99.2764
0 0 0 0 9.92E + 03
0 1.88E+ 04 4.45E+ 02 0 7.49E+ 03
0 51.3 1.2 0 47.5
LEU 5% 232 U 234 U 235 U 236 U 238 U
0 0.039 5 0 94.961
0 0 0 0 9.50E +03
0 1.31E+ 05 3.10E+ 03 0 7.17E+ 03
0 86.9 2.1 0 11.1
0 0 0 0 9.50E +03
1.66E+ 04 4.46E+ 05 3.10E+ 03 2.76E+ 04 7.04E+ 03
3.26 87.48 0.61 5.41 3.24
neutron 100 group source 1/s/ton UF6
LEU 5% repr. 232 U 7.60E–07 234 U 0.133 235 U 5 236 U 1.55 238 U 93.317
3.0E+03 2.5E+03 2.0E+03 1.5E+03
(Alpha,n) spontaneous fission total
1.0E+03 5.0E+02 0.0E+00 1E+04
1E+05
1E+06
1E+07
Energy eV Fig. 4. Hundred group neutron spectra for natural UF6 [7].
container thickness varies from 12.5 to 13.8 mm. These deviations influence strongly the enrichment determination by application of the enrichment meter principle which requires a precise knowledge of the steel thickness at the measurement position. A variation of 70.5 mm of the wall thickness leads to a minimum error of 75% of the intensity of the 185.7 keV line of 235U. If there is no possibility to measure the wall thickness of the container with an ultrasonic gauge (contact measurement), the uncertainty on the 235U enrichment determination is at least 5%. 3.1. UF6 filling profile As it will be discussed later, the response of a neutron detector in the vicinity of a container depends strongly on the filling profile. UF6 is characterized by a high coefficient of expansion in the liquid phase. When heated in the containment autoclaves, the transformation of solid UF6 at 20 1C to liquid UF6 at 110 1C goes with a volume increase of 53%. Consequently, to prevent
Fig. 5. (a) Profile after filling at 80 1C. (b) Profile after desublimation.
deformation and rupture of the cylinders, the maximum UF6 mass must not exceed 2/3 of the maximum possible load of the cylinders. The distribution of the material within the container depends on how it was filled, on the last operation made on it (for instance, sampling in liquid phase after homogenization) or how long and under which conditions (temperature, sunshine) it was stored. When it is filled in liquid phase or after sampling, UF6 solidifies using all possible heat exchange surfaces. Most of UF6 remains in the lower part. However, there is a deposit of UF6 of several centimeters on all the upper part (Fig. 5a). When the container is filled by desublimation in a cylinder cooled at 25 1C, UF6 is deposited in a uniform way on all the
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inner surface of the cylinder (Fig. 5b) and forms an annular solid ring. When the container is filled with UF6 gas at 80 1C in a cylinder at 15 1C, first a layer of solid UF6 is deposited on the entire inner surface. Later, due to the increasing temperature of the solid UF6, additional amount of UF6 will condense inside the shell of solid UF6, flows by gravity in the cylinder and then solidifies. Consequently, this material will fill the lower part inside the solid UF6 shell (it is assumed that 75% of the gas is liquefied). As a result, we have the same filling profile as in Fig. 5a. With the changes of temperatures during the day, the filling profile changes. The coating sublimes from the cylinder wall to cooler cylinder areas, thus decreasing the thickness. However, no important modification due to the daily heat cycle is expected. 3.2. Chemical and radiochemical impurities When the cylinder is emptied by off gassing the UF6, the nonvolatile daughter products of 238U (234Th and 234mPa) as well as uranium deposit in the form of UF6, UO2F2, UF4 and UF5 and also volatile UF6 remain in the ‘‘heel’’. Heels in excess of 22.7 kg for 48Y or 48G and 11.3 kg for 30B require removal by cylinder cleaning. On filling a freshly emptied cylinder with UF6, the daughter products clearly remain mainly plated on the cylinder wall. The heel is responsible for significant radiation levels observed for empty cylinders. Since the non-volatile products are not self-shielded any more for an empty cylinder, the radiation dose increases by a factor of 12 for a freshly emptied 48Y cylinder and a factor of about 6 for a 30B cylinder [8]. The spatial distribution of the heel influences the amplitude and the spectral shape of the gamma radiation background. The contribution to the spectrum of the Compton scattering of the high energy gamma rays from 234mPa as well as Bremstrahlung production from the 2.3 MeV beta particle emitted in the decay of 234m Pa are sensitive to the location of the material. In addition, since the 185 keV gamma ray used for determining the 235U enrichment has a half thickness of 0.13 cm, the deposit can mask the signal from the actual UF6 filling. In the measurements carried out in Ref. [9], the authors mentioned that the residual activity was often clearly defined as forming a lateral band along the length of the cylinder. When measuring the 235U enrichment, the detector position has to be chosen in a way that the deposit does not influence the result. As already mentioned in Section 2.2, [5,6] also mentioned the presence of 137Cs and 233Pa (daughter of 237U) during their measurements on UF6 cylinders.
4. Classical NDA methods Typical NDA measurements are currently carried out by inspectors and the measurement devices are placed in contact with the cylinders. 4.1. Enrichment determination by gamma spectrometry Two basic NDA methods are routinely used to determine the U enrichment. One is based on a self-intrinsic calibration with unfolding of complex energy region 80–110 keV whereas the other makes use of the enrichment meter principle applied to the 185.7 keV line. In the case of measurement of UF6 containers, the thick shielding of the cylinder precludes the use of method based on the unfolding of low energy region. Several attempts were made to overcome the previous limitation when measuring U through thick container wall
235
313
[10,11]. These papers refer to the analysis of the 120–1200 keV region by intrinsic calibration using, the 143.8–163.3, 185.7 and 205.3 keV of 235U and the 258.3, 766.6 and 1001 keV from 234mPa, with an additional gamma ray from 234U at 120.9 keV for [10] only. Using either a standard empirical relative efficiency curve or a physical model for the sample detection system, these two applications were tested with various uranium enrichment standards using in particular 13 to 16 mm steel absorber and different germanium detectors. This approach has delivered good results but was tested on a limited number of items. Precise measurements would require very long counting times, especially when the detector has to be at 60 cm distance instead of being in contact with the drum surface. In addition, the presence of nonvolatile daughter product of 238U (paragraph 3.2) remaining in the heel may bias the results. The classical method to determine the 235U enrichment is the enrichment meter principle [12] making use of the proportionality between the net peak area of 185.7 keV line of 235U and the enrichment for infinitely thick items. This method requires the use of reference samples and since UF6 is very absorbent, the measurement results indicate only the enrichment at the cylinder wall. On the whole, the limitations encountered with the application of this method are mostly defined as a function of either the performance of the detectors used (NaI, LaBr3, Ge or CZT) in terms of efficiency or resolution or of the algorithms used in the analysis codes. For instance, the necessity of making wall thickness corrections accounting for the thickness and type of material placed between the source and the detector or the presence of interfering lines precludes the use of NaI detector with an analysis based on two ROIs of interest [13]. This method is used routinely to determine the 235U enrichment on UF6 containers either with low resolution detectors [5,6,14] or high resolution detectors [9,15]. The described characteristics of the gamma radiation measurements do not allow enrichment measurements of the required quality nor the direct or indirect determination of the 235U mass in a UF6 container. 4.2. UF6 mass and 235U enrichment determination with passive and active neutron assay 4.2.1. Passive-neutron assay The passive neutron detection consists in measuring the fast neutron flux emerging from a UF6 cylinder. The primary flux is due to F(a, n) reactions induced by the 234U alpha activity. If a constant 235U/234U ratio can be assumed, the fast neutron flux is then a linear function of 235U. The total neutron source term from a mass of UF6 can be written as [7] Q ¼ Mu ð3230000 f32 þ496 f34 þ0:0917 f35 þ 2:64 f36 þ ð0:0112 þSF38 Þ f38 Þ where Q is the neutron source strength in n/s, Mu is the uranium mass in g, f is the fractional isotopic abundance of the subscripted isotope and SF38 is the spontaneous fission coefficient of 238U. If the terms with f32, f35 and f36 can be neglected, the previous expression becomes Q ¼ Mu ð496f34 þ ðð0:0112 þ SF38 Þf38 ÞÞ: If the ratio f34/f35 is known and constant, the neutron source strength can be written as Q ¼ Mu ða þ bf35 Þ:
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Several experiments [5,6] were carried out to find the relationship between the neutron detector response, the 235U enrichment and the total UF6 mass. The passive-neutron measurements were performed with shielded neutron assay probes (SNAP) on the container surface and in a low lateral position. The conclusions of Refs. [5,6] were:
There is a linear relationship between the detector response
and the enrichment for a given UF6 mass for an enrichment ranging from depleted to 5%. The variations in the neutron leakage (fraction of the source neutron that escape from the cylinder) per source neutrons caused by large change in enrichment and cylinders loadings are small.
A drawback of the passive-neutron assay is that a falsification of the 235U enrichment declaration, which cannot be detected by gamma spectrometry, could lead to the expected neutron count rate (i.e. according to the declaration) by compensating for the missing neutrons with HEU material hidden in the cylinder.
a linear relationship between the delayed neutron signal and the uranium enrichment in Ref. [18], 235 U enrichment with a calibration curve using the 235U weight and the ratio 974 keV(132Sb)/1031 keV (89Rb) [19].
the
These methods demonstrated a high sensitivity well adapted to the detection of hidden nuclear material and appear as a powerful and promising tool for determining either the U enrichment or U mass given a value for the other. However, the transposition of such experiments to the UF6 containers would require an experimental study which up to now has not been done.
6. Specific investigations with respect to the design of an unattended measurement station for UF6 containers The previous section showed that it is possible to determine: 235 U enrichment with a gamma spectrometer when the wall thickness can be measured at the measurement point, when the contribution of interfering gamma lines to the 185.7 keV line can be evaluated, and when 234mPa deposits do not mask the signal, The 235U content and the UF6 mass with passive neutron counting provided that the 234f/235f ratio is known. Contrary to the gamma spectrometric measurements, neutron-based techniques are not sensitive to deposits of protactinium, to the container wall thickness variation and to the presence of potential interference, The 235U enrichment with an active neutron system making use of special geometry conditions (source and detector in contact with the container).
The 4.2.2. Active-neutron assay A usual active measurement technique for the enrichment is the irradiation of the UF6 with thermal neutrons and the observation of the fast neutron from induced fissions of 235U. Contrary to the passive measurement case, there is no need to know the f34/f35 ratio. The mean-free-path for thermal neutrons in solid UF6 is a strong function of 235U enrichment varying from 24 cm for 0.3% to 4 cm for 4%. This leads to a correspondingly strong dependence of the detected fast neutron flux on 235U enrichment. Several experiments [5,6,16] were performed in the 70’s with techniques based on thermal neutron induced fission of 235U in UF6 cylinders (30B and 48Y). For the three experiments, the neutron detector was placed in contact with the cylinders. The results showed that, up to 3% the net count rate is a linear function of the enrichment for the three experiments whereas for the 0.3–4% enrichment range, results of Ref. [6] show a non-linear behavior which was then confirmed with MCNP calculations.
5. Analysis of delayed neutrons and delayed photons Both classical gamma spectrometric and neutron assay described in the two previous sections cannot provide any information about the inner part of the large UF6 cylinders. For this reason, alternative methods having the possibility of overcoming this limitation are of interest. Applications such as systems for border security monitoring or for radioactive waste management for which highly sensitivity methods are required lead to recent investigations making use of the analysis of delayed neutrons and b-delayed gamma rays emitted by uranium fission products to determine the 235U enrichment or 235U mass. In these experiments, the interrogation of small samples was made with either a 14 MeV pulsed generator [17], or with a pulsed neutron beam produced by a LINAC [18,19] or with thermal neutron flux [20]. The combination of an irradiation with very penetrating 14 MeV neutrons associated to the detection of delayed gamma lines above 3 MeV (also very penetrating) could be used. Proof of principle measurements were performed in the previously cited papers on small samples for short counting time and it was possible to determine
the
235
U enrichment by fitting the temporal die away tail in Ref. [17],
For the unattended measurement, a minimum distance between the cylinders and the measurement device has to be respected to allow safe movement of the cylinders. Consequently, the wall thickness cannot be measured point with ultrasonic gauge (contact measurement). Typical variation of the wall thickness from the nominal value easily leads [13] to an enrichment error of 6% which is above the maximum uncertainty specified in Ref. [1] for low enriched products. The use of active neutron system for determining the 235U is also prohibited in the sense that this technique makes use of a special geometry and cannot be applied easily if the cylinder is not in contact with the detector. Additional MCNP calculations [21] showed that activation with a 14 MeV neutron generator gives a signal which is not very sensitive to the 235U content. A measurement campaign [22] has also shown that for measurements times which are compatible with in the field measurements (typically 5 min), the estimated count rates for a large germanium detector are far below the requested accuracy for depleted uranium in 48Y containers [1]. Also the use of a more efficient detector such as a LaBr3 detector tested in the field [23] does not allow the requested accuracy to be reached. From the above, the 235U enrichment and the UF6 mass determination can be only verified with passive neutron counting provided that the ratio 234U/235U is known. MCNP calculations were carried out to confirm the relationship between the neutron output and the UF6 weight or the 235U enrichment given in Refs. [5,6].
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6.1. Calculation of the neutron detector response for passive measurements In Ref. [7], the response of 3He detectors due to emission of neutrons from UF6 containers was calculated with MCNP for four neutron detectors arranged around the steel container (Fig. 6). The detectors are in fixed positions for both containers at 60 cm distance and at mid height of the axis of the containers. Each detector comprises five 3He tubes of 2.54 diameter and 1 m active length with a pressure of 4 atm. The neutron count rate of this set-up can be used to determine the 235U, provided the 234U/235U ratio is known. If also the enrichment is known, the total UF6 mass can be calculated; alternatively, if the UF6 mass is known, the enrichment. The relation between the detector response and the fuel enrichment or the total UF6 mass was investigated for the containers types 30B and 48Y. Since the UF6 filling profile is not known a priori five different fuel filling profiles—numbered from 1 (x =0) to 5 (x =100) were taken into account (Fig. 7). They cover the two cases described in Figs. 5a and b of Section 3.1. The parameter x represents the percentage of UF6 covering all the inner cylinder walls with a layer of constant thickness. The rest of the material fills the remaining volume from below. The simulations were made for eight different isotopic compositions including that of reprocessed material (Table 3). For the container 30B all fuel compositions were taken into account whereas for the 48Y container, only the compositions 1, 2, 3 and 7 of Table 3 were considered. The neutron flux and the detector response were calculated with the Monte Carlo code MCNPX [24]. All neutron cross-sections were taken from the ENDF/B VI.1 and ENDF/B VI.O libraries. Table 4a and b give the detector response calculated for four neutron detectors for
315
both container types. The statistical uncertainty was always less than 1%. One can see that the detector response depends very much on the filling profile as it increases from filling profile x = 0 to 100 by a factor of 1.6. The calculations show that a minimum of 156,000 counts and 828,000 counts can be achieved in 2 min for the 30B container and for the 48Y, respectively, with the considered device. The potential background contribution of one container with the same load of UF6 at 300 cm distance behind the detector is 10% of the total neutrons signal, but it can be reduced substantially by appropriate shielding of the detector. Monitoring of background radiation in the environment of the measurement station may be used to make automatically background corrections. 6.2. Detector response as a function of enrichment The simulation suggests that the neutron source strength is a linear function of the 235U wt% for a given UF6 mass. Figs. 8a and b show the detector response as a function of 235U wt%, for different filling profiles and for both containers filled to 100% of the UF6 mass. Table 3 Isotopic composition used in MCNP calculations (in wt%). No.
Material type
232
1 2 3 4 5 6 7 8
Natural Dep03 LEU10 LEU30 LEU50 Repro 6.49% Sample3.25% Sample 5%
0 0 0 0 0 1.200E–08 4.800E–07 7.600E–07
U
234
235
0.0056 0.0023 0.0078 0.0234 0.0390 0.0670 0.0850 0.1330
0.7180 0.3000 1.0000 3.0000 5.0000 6.4710 3.2500 5.0000
U
U
236
U
0 0 0 0 0 0.05 1.09 1.55
238
234
99.2764 99.6977 98.9922 96.9766 94.9610 93.4620 96.6650 94.8670
0.007799 0.0078 0.0078 0.0078 0.0078 0.010354 0.026154 0.0266
U
U/235U
detector Table 4 No. Composition Detector response 1/s x=0
120
x =25
x= 50
x =75
x= 100
2.589E+ 02 1.796E+ 02 3.109E+ 02 7.049E+ 02 1.136E+ 03 1.924E+ 03 2.474E+ 03 3.822E+ 03
2.809E + 02 1.939E + 02 3.400E+ 02 7.722E + 02 1.211E + 03 2.029E + 03 2.640E + 03 4.101E + 03
2.941E +02 2.075E +02 3.576E +02 8.115E +02 1.292E +03 2.145E +03 2.817E +03 4.410E +03
(b) Detector response for the container 48Y 1 Natural 1.008E+ 03 1.268E + 02 1.420E+ 03 2 Dep03 6.909E+ 02 8.691E + 02 9.766E+ 02 3 LEU10 1.209E+ 03 1.550E + 03 1.723E+ 03 7 Sample3.25% 1.009E+ 03 1.255E + 04 1.413E+ 04
1.548E + 03 1.070E+ 03 1.888E + 03 1.524E + 04
1.625E +03 1.115E +03 1.984E +03 1.609E +04
(a) Detector response for the container 30B 1 Natural 1.872E + 02 2.316E + 02 2 Dep03 1.301E+ 02 1.620E + 02 3 LEU10 2.235E + 02 2.803E + 02 4 LEU30 5.152E + 02 6.469E + 02 5 LEU50 8.282E + 02 1.027E + 03 6 Repro 6.49% 1.410E+ 03 1.748E + 03 7 Sample3.25% 1.807E+ 03 2.239E + 03 8 Sample 5% 2.850E+ 03 3.507E + 03
30B 100
48Y 58 10
Fig. 6. Measurement geometry for passive neutron measurement (dimensions in cm).
X=0
X = 25
X = 50
Fig. 7. UF6 filling profiles.
X = 75
X = 100
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detector response 1/s
1400 1200 1000 800 x=0 x = 25 x = 50 x = 75 x = 100
600 400 200 0 0
1
2
3
4
5
6
U-235 wt%
2000
Fig. 10. Cross-section through the containers with position of detectors (in cm).
1500
2200
x=0 x = 25 x = 50 x = 75 x = 100
1000 500 0 0
0.2
0.4
0.6 U-235 wt%
0.8
1
2000
1.2
Fig. 8. (a) Dependence of the detector response on the 235U wt% for container 30B (2.2 tons of UF6). (b) Dependence of the detector response on the 235U wt% for container 48Y (12 tons of UF6).
detector response 1/s
detector response 1/s
2500
detector position 1 detector position 2 detector position 3
1600 1400 1200 1000 800
2500
0
25
50 filling profile number
75
100
2000
5000
1500 1000 0.7% enrichment 0.3% enrichment 1% enrichment
500 0 0
25
50
75
100
Filling profile number
detector response 1/s
Number of counts
1800
4500 4000 detector position 1 detector position 2 detector posiion 3
3500 3000
Fig. 9. Dependence of the detector response on the filling profile.
2500
Consequently, the relation between the count rate R, UF6 mass and the enrichment f35 can be approximately described by the formula R ¼ ZðMUF6 Þ ðc þd f35 Þ: The function Z and the parameters c and d depend on the container type and the specific set-up of the measurement station. The gradient d of the functions increases from x= 0 to 100. The equally thick layer of UF6 on all internal surfaces gives the highest detector response. The dependence of the detector response on the geometry parameter x (Fig. 7) is illustrated in Fig. 9 for the 48Y container. Because of the strong dependence of the detector response on the filling profile, the filling profile has to be known in order to determine the 235U wt% from the detector response. Otherwise, e.g., depleted uranium cannot be distinguished from 1% enriched U (48Y) or 3% from 5% (30B), in unfortunate cases.
0
25
50 filling profile number
75
100
Fig. 11. (a) Detector response for container 48Y. (b) Detector response for container 30B.
6.3. Influence of detector positions In order to reduce the strong dependence of the detector response on the filling profile, the detector positions were changed (Fig. 10). For the first case, the lateral distance of the detectors from the containers is kept but the detectors are shifted vertically by 120 and 120 cm, respectively (positions 2). In the next case the detectors are positioned 120 cm above and below of the containers (positions 3). The calculated detector responses for the three detector positions are given in Fig. 11a and b. The detector positions 2
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and 3 clearly reduce the dependence of the detector response on the filling profile whereas the position 3 always shows the highest detector response. Because of the rotation symmetry the detector positions 1 and 3 give the same values for the filling profile x =100. For the container 48Y, the ratio of the maximum to the minimum neutron signal is 1.63, 1.13 and 1.09 for the detector positions 1, 2 and 3, respectively. The corresponding values are 1.55, 1.14 and 1.03 for the container type 30B. Further optimization calculations can be made but they are of interest only if they take into consideration the local conditions of a possible measurement station. They have to include scattered radiation, the possible presence of other UF6 containers with different enrichments in vicinity of the measurement station and a shielding of the detectors.
Table 5 Contributions of volume layers to the detector response for geometry 1. Filling profile
x =100
relative contribution to the total signal (%)
the function Z(MUF6) has to be evaluated for a given cylinder type. For the calculation of the contributions of different UF6 volume layers to the detector response of a 100% filled container, the UF6 volume was divided into four volume layers numbered from 1 to 4 as shown in Fig. 12a and b. Only the filling profiles x= 100 and 50 of Fig. 7 were considered. All volume layers have the same UF6 mass, 3000 kg for container 48Y and 550 kg for container 30B. The results calculated with the MCNP code are represented in Table 5 for detector positions 1 of Fig. 10. For the filling profile x =100 all four volume elements are contributing in a similar way to the detector signal. None of them is ‘‘invisible’’. For the filling profile x= 50, the figures show that number 3 is the least visible volume element.
Detector response contribution % 48Y/LEU10
30B/Sample 5%
20.3 22.8 26.6 30.3 31.8 34.6 8.6 25
23.4 24.5 25.2 26.9 29.3 30.5 15.4 24.8
120.00 filling profile x = 100 filling profile x = 50
100.00 80.00
y = -8E-06x2 + 0.0622x + 0.2486
60.00 40.00
y = -4E-06x2 + 0.0536x - 0.275 R2 = 0.998
20.00 0.00 0
500
1500 1000 mass (kg)
2000
2500
120.00 relative contribution to the total signal (%)
R ¼ ZðMUF6 Þ ðc þd f35 Þ
Layer no.
1 2 3 4 1 2 3 4
x =50
6.4. Contributions of the volume elements inside the container If the detector response is written as
317
100.00 y = -8E-07x2 + 0.0181x - 7.8 80.00 60.00
y = -5E-07x2 + 0.0148x - 6.075
40.00 filling profile x = 100 filling profile x = 50
20.00 0.00 0
2000
4000
6000 8000 mass (kg)
10000
12000
14000
Fig. 13. (a) Relative contribution to the total signal for a 30B container, 5% enrichment. (b) Relative contribution to the total signal for a 48B container, 1% enrichment.
6.5. Detector response to partially filled drums
Fig. 12. (a) x–z cross-section of the UF6 volume layers for the filling profile x= 100 and (b) x–z cross-section of the UF6 volume layers for the filling profile x = 50.
The assumed filling processes were mentioned in Section 6.1. The filling by desublimation corresponds to the filling profile x= 100 whereas the filling at 80 1C corresponds to the filling profile x= 50. Figs. 13a and b show the relative detector response as a function the mass. This simulation function is referred as Z(MUF6) and the calculation were made for the detection position 1 of Fig. 10. The function Z(MUF6) can be fitted with a polynomial. It does not show a saturation effect. The very slight non-linearity is due to the self absorption in the material. To characterize once again the ‘‘visibility’’ of the UF6 and its dependence on the filling profile, the neutron leakage fraction was calculated in Fig. 14. Results were arbitrarily normalized to 1 for an UF6 weight equal to one fourth of the total load for both container types. Fig. 14a and b show that the leakage fraction
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1.1
The present MCNP calculations of the passive neutron emission carried out have shown that:
1
The results depend strongly on the UF6 filling profile, however,
0.9
0.8 filling profile x = 50 filling profile x = 100
0.7
0.6 500
0
1000 1500 weight (kg)
2000
2500
1.2
neutron leakage fraction
1
an effect which can be minimized by optimization of the detector position relative to the containers. The total number of counts is found to be a linear function of the 235U enrichment (for 100% filled drums) and is also rather a linear function of the UF6 mass (for partially filled drums). In a scenario whereby 10% of the UF6 mass would be missing from the cylinder, the total number of neutrons would also diminish by nearly 10% for both container types. At least 156,000 counts and 828,000 counts may be collected in 2 min, for the 30B container and for the 48Y, respectively, as a sum over four 3He detectors (without back shielding), thus allowing a very short measurement time. The potential background contribution of a drum at 300 cm distance is 10% of the total neutron signal.
However, the inspection systems have to take into consideration the physics limitations of the proposed method and MCNP calculations can be made to estimate which mass of high enriched uranium could be hidden in the centre of the container.
0.8 0.6 0.4
Acknowledgments
filling profile x = 50 filling profile x = 100
0.2 0 0
2000
4000
6000 8000 weight (kg)
10000
12000
14000
Fig. 14. (a) Neutron leakage fraction for a 30B container filled with 5% enriched UF6 for detectors placed in position 1 of Fig. 10. (b) Neutron leakage fraction for a 48Y container filled with 1% enriched UF6 for detectors placed in position 1 of Fig. 10.
remains quite high also for the large drums and demonstrate the clear effect of the filling profile for detector geometry 1. We observe small variations of the leakage fraction as a function of the UF6 mass and confirm in this way the results of Ref. [6]: in this paper, the authors calculated a leakage factor varying from 0.93 to 0.82 as the weight of the 30B container is increased from 2000 to 5000 lb for a filling profile which is not described. The ratio of the leakage fraction for 5000 lb over the leakage fraction for 2000 lb gives 1.13. Given the results of the present document, the corresponding ratio is equal to 1.09 and 1.23 for x =100 and 50. Since we can easily imagine that the filling profile of Ref. [6] is something between x= 100 and 50, respectively, there is a very good agreement between the results of the present simulation [7,21] and the calculations of Ref. [6].
7. Conclusion Of the several methods investigated, passive-neutron counting is the most appropriate to verify the 235U enrichment and UF6 weight in unattended mode provided that the ratio f34/f35 is known. As already mentioned in Ref. [6], all the volume elements inside the large containers contribute to the measured signal whereas the established inspection mode-gamma spectrometric enrichment measurements-see only a thin surface layer of the UF6.
The authors would like express their thanks to Mr. Olivier HALNA DU FRETAY, AREVA NC for very helpful discussions on the chemistry of cylinders heels as well as the cylinder filling profiles, to Robert ALVES, AREVA NC and Antoine RUIZ, DG TREN for their continuous support, Dr Bent PEDERSEN/JRC Ispra for helpful discussions. References [1] A. Lebrun, User requirements ‘‘state of the art NDA methods applicable to UF6 cylinders’’, SG-EQ-NDA-UR 0001, 2007. [2] S. Richter, A. Alonso, W. De Bolle, R. Wellum, P.D.P. Taylor, International Journal of Mass Spectrometry 193 (1999) 9. [3] R. Gunnink, W.D. Ruhter, P. Miller, J. Goerten, M. Swinhoe, H. Wagner, J. Verplancke, M. Bickel, S. Aboushal, MGAU, a new analysis code for measuring U-235 enrichments in arbitrary samples, IAEA symposium on International Safeguards, Vienna, Austria, March 8–14, 1994. [4] P. Mortreau, R. Berndt, Handbook of Gamma Spectrometry Methods for NonDestructive Assay of Nuclear Materials, EUR 19822 EN, version 3, 2006 /http://nuclearsafeguards.jrc.it/nda/handbook3-2006.pdfS. [5] P.E. Fehlau, W.H. Chambers, Perimeters safeguards techniques for uranium enrichment plants, LA-8997-MS, 1981. [6] R.B. Walton, T. Reilly, J.L. Parker, J.H. Menzel, E.D. Marshall, L.W. Fields, Measurements of UF6 cylinders with portable instruments, Nuclear Technology 21 February, 1974. [7] E. Franke, Calculation of the neutron response for passive measurements, Report-No UF6-01/08, 2008. [8] UF6, a manual of good handling practices, USEC-651, revision 7, January 1995. [9] R.S.T. Shaw, The routine non-destructive measurement of the 235U concentration of enriched uranium hexafluoride in transit cylinders using high resolution gamma spectrometry Proceeding ESARDA 1979, Brussels. [10] W. Ruhter, T.F. Wang, C.F. Hayden, Uranium enrichment measurements without calibration using gamma rays above 100 keV, UCRL-JC-142832, October 29–November, 2001. [11] E. Sampson, P.A. Hypes, D.T. Vo, FRAM isotopic analysis of uranium in thickwalled containers using high energy gamma rays and planar HPGe detectors Proceedings, Institute Nuclear Management, 2002. [12] T.D. Reilly, R.B. Walton, J.L. Parker, Nuclear safeguards research and development LA-4605-MS, Los Alamos Scientific Laboratory, 1970, pp. 19–21. [13] R. Berndt, P. Mortreau, E. Franke, Survey of state-of-art NDA methods applicable to UF6 cylinders—IAEA task n 07/TAU-04 JRC Scientific and Technical Reports, internal distribution. [14] A.R. Flynn, Field determination of uranium-235 enrichment by portable gamma ray spectrometry, Report K-1819-1972. [15] R.C. Hagenauer, H.Y. Rollen J.M. Whittaker, R.L. Mayer, T. Biro, Nondestructive measurement of UF6 cylinders at the Portsmouth gaseous diffusion plant, Nuclear material management—annual meeting, 1998.
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[16] J.T. Caldwell, Nuclear material management 2 (2) (1973). [17] W.L. Myers, C.A. Goulding, C.L. Hollas, Determination of the 235U enrichment of bulk Uranium samples using delayed neutrons, American nuclear society, 2006 Physor. [18] L. Lakosi, C. Tam Nguyen, J. Bagi, NIM B 266 (2008) 295. [19] F. Carrel, M. Gmar, F. Laine´, J. Loridon, J.L. Ma, C. Passard, IEEE Nuclear Science symposium NS-2 (2006) 909. [20] E.B. Norman, et al., NIM A 521 (2004) 608.
319
[21] E. Franke, Calculation for the neutron detector response for passive and active measurements Report-02/08. [22] P. Mortreau, R. Berndt, Measurement campaign on UF6 containers at Georges Besse 1 (1997) JRC Ispra-Internal report. [23] P. Mortreau, R. Berndt, 235U enrichment determination of UF6 containers with a LaBr3 detector (2009) JRC Ispra- Internal report. [24] J. F. Briesmeister, MCNP-A General Monte Carlo N-Particle Transport Code (Version 4B) LA-12625-M, 1997.
Contents lists available at ScienceDirect
Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
235
U enrichment or UF6 mass determination on UF6 cylinders with non-destructive analysis methods
R. Berndt a, E. Franke b, P. Mortreau a, a b
The European commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, TP 800, Via Fermi, Ispra, Italy Consultant, Dresden, Germany
a r t i c l e in fo
abstract
Article history: Received 10 July 2009 Received in revised form 22 September 2009 Accepted 2 October 2009 Available online 1 November 2009
The measurement of the 235U enrichment and the mass of UF6 in 30B and 48Y transit cylinders are important safeguards verification tasks of IAEA. In the framework of a project aiming to establish an unattended measurement station at isotope enrichment facilities (Lebrun, 2007) [1], a study was carried out to describe the state-of-the-art of non-destructive assay methods applicable to UF6 cylinders. The objective of the present work is to provide a feasibility assessment study of all known NDA techniques applicable to the quantitative verification of all uranium categories involved in an enrichment processing plant. Based on this investigation, the most appropriate techniques were then investigated for suitability of use in an unattended measurement station. & 2009 Elsevier B.V. All rights reserved.
Keywords: Gamma ray spectrometry Uranium enrichment Neutron assay MCNP calculations UF6 Non-destructive assay methods
1. Introduction
2. Radiation characteristics of uranium hexafluoride
Nuclear safeguards inspections have to verify the amounts of declared nuclear materials such as uranium and plutonium. The 235U enrichment as well as UF6 weight determination in 30B and 48Y transit containers used in enrichment plants represents a significant portion of this program of verification. In the framework of a project aiming to establish an unattended measurement station at an isotope enrichment facility [1], a study was carried out to describe the state-of-the-art of non-destructive assay (NDA) methods applicable to UF6 cylinders. Well-established NDA methods as well as techniques based on the analysis of delayed neutrons and delayed photons were first investigated. The technical requirements (such as operation in unattended mode, counting time compatible with the need of plant operation, and implementation timeframe limitations), have then orientated the work towards a more specific investigation. Previous results of in field measurements as well as MCNP calculations, lead to a proposal for an unattended measurement station of UF6 cylinders.
This section describes the emission of gamma rays and neutrons from UF6 and from the daughter isotopes for U of natural origin or U from reprocessing. Since alpha and beta particles emitted by the uranium isotopes as well as uranium X-ray cannot cross the thick wall of UF6 cylinders, the only visible signature comes from gamma and neutron radiation. 2.1. Description of uranium isotopes 2.1.1. Natural uranium In its natural state, uranium consists of three isotopes 234U, 235 U and 238U. Extremely low quantities of 236U due to activation processes such as cosmic ray activation can also be present [2]. Mass spectrometry analysis of U samples shows that the isotopic composition of natural uranium varies slightly as a function of its geographical origin. By measuring a set of different U ores, it was shown in Ref. [2] that the range of natural variation of the three natural isotopes expressed in at% is 234
U
0:0054 0:0051 at%;
0:7201 at%;
Corresponding author.
E-mail address: [email protected] (P. Mortreau). 0168-9002/$ - see front matter & 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.nima.2009.10.060
238
U
235
U
0:7207
99:2748 99:2739 at%
2.1.2. Reprocessed uranium In addition to the three naturally occurring uranium isotopes, the reprocessed uranium may contain all the isotopes from 232U
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Table 1 Nuclide concentrations (g/t) of a PWR fuel with a burn up of 33000 MWd and an initial enrichment of 3.2%.
Counts
3
4
4.248E–04 4.796E–03 1.177E + 02 7.516E + 03 4.437E + 03 2.951E–05 9.434E + 05 0.000E + 00 2.203E–21
6.138E–04 5.118E–03 1.203E +02 7.516E + 03 4.438E + 03 2.969E–05 9.434E + 05 0.000E + 00 2.203E–21
7.233E–04 5.440E–03 1.230E+ 02 7.516E+ 03 4.437E+ 03 2.450E–05 9.434E+ 05 0.000E +00 2.203E–21
102.31 keV - Th231
98.44 keV - UKα1
1
95.87 keV - Pa Kα2
100000
2.810E–04 4.643E–03 1.163E+ 02 7.516E+ 03 4.437E+ 03 1.021E+ 01 9.434E+ 05 0.000E +00 2.203E–21
94.66 keV - UKα2
1000000
Discharge cooling time (year)
93.35 keV - ThKα1
0.000E + 00 0.000E + 00 2.574E +02 3.201E+ 04 0.000E + 00 0.000E + 00 9.677E +05 0.000E + 00 2.203E–21
92.30 keV - Pa Kα2 92.367 keV - Th234 92.792 keV - Th234
69.8a 1.592 105a 2.457 105a 7.037 108a 2.342 108a 6.75d 4.469 109 23.5 min 14.1 h
Discharge
89.95 keV - Th231 89.96 keV -Th K α2
234
84.21 keV - Th231 84.8 keV - PbKβ1
U U U 235 U 236 U 237 U 238 U 239 U 240 U 233
Charge
82.09 keV - Th231
232
Half-life
81.23 keV - Th231
U isotope
10000 high enriched Uranium
1000 low enriched Uranium
100 80
85
90
95
100
105
Energy (keV) Fig. 1. Uranium spectra taken with a germanium detector—80–105 keV region [4].
to 240U. Table 1 shows the example of the U isotopic composition of a LWR spent fuel reactor (burn up= 33000 MWd, initial enrichment= 3.2%) at the end of the irradiation. The isotopes 237U, 239U and 240U have a very short half-life and 233 U is present in very small quantities. 2.2. Gamma ray spectra Typical features of U spectra as well as the difference between enriched natural UF6 and enriched reprocessed uranium are illustrated by Figs. 1–3. The low energy part of the U spectrum (80–105 keV region—Fig. 1) is characterized by an overlapping of many X and gamma rays which contains the signature from 235U, and from 231Th and 234Th, daughter products of 235U, and 238U, respectively. The most intense lines (U Ka at 98.43 and 94.65 keV) are due to self fluorescence of the material itself. This region is used to determine the 235U enrichment with the intrinsic calibration method [3]. The 100–210 keV region (Fig. 2) shows the typical lines of 235U, the most intense line of which (185.7 keV) is used to determine the enrichment by application of the enrichment meter principle. Fig. 3 shows spectra of depleted natural UF6 and depleted reprocessed uranium stored in transit containers in the 50–2500 keV range. The 235U gamma lines are strongly attenuated in the container walls but the 185.7 keV is still intense enough for
the measurement of the 235U enrichment. 238U is identified with the gamma rays of its daughter nuclides 234mPa (at 766.4 and 1001.0 keV) and 234Pa (880, 883, 925 and 926.7 keV). That isotope is in secular equilibrium with 238U after several periods of half-life (24.1 days of 234Th). Its signature can only be used if the history of the individual drum is known which is not compatible with the use of automated systems. The depleted reprocessed uranium presents several specific lines emitted by the decay products of 232U (212Pb at 238 keV, 208 Tl at 583, 860 and 2614.6 keV) or by contribution of the scattered radiation of the 2614.6 keV line as well as the single and double escape peak associated with this line. 2.2.1. Other possible features in UF6 spectrum The spectra shown in Fig. 3 were measured with depleted UF6. However, when measuring high enriched UF6, the 1274 keV line from 22Na may become visible in UF6 spectra [5]. This line is produced from the 19F (a, n) reaction and could be used to reveal the presence of HEU material hidden behind a thin layer of LEU in a UF6 drum since the half thickness at 1274 keV is about 2.5 cm in solid UF6. The presence of lines from 137Cs and 233Pa (a daughter of 237U) were also observed during the measurement of UF6 product from reprocessing of spent fuel [6]. In spite of the fact that the backscattered peak of 137Cs is located at
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194.94 keV - U235
185.715 keV- U235
182.61 keV - U235
163.33 keV - U235
150.93 keV - U235
143.76 keV - U235
140.76 keV - U235
10000 high enriched Uranium
311
198.9 keV - U235 202.11 keV - U235 205.311keV - U235
Counts
100000
120.91 keV - U234
114.4 keV - UKβ22 114.6 keV - UKβ23
108.99 keV - ThKα2 109.178 keV - U235 110.41 keV - UKβ3 111.30 keV - UKβ1
1000000
105.362keV - ThKβ1 106.581keV - Th231
R. Berndt et al. / Nuclear Instruments and Methods in Physics Research A 612 (2010) 309–319
1000 low enriched Uranium
112.8keV - Th234 / U238
100 100
120
140
160 Energy (keV)
180
200
Counts
766.4 keV234mPa
100
Tl (2614 keV) peak of
208
1592 keV Double escape
Tl
peak of 208Tl (2614.6 eV)
X
2103 keV Single escape
X
860.37
185.7 keV
keV208
X
1000
727.33 keV212Bi
10000
583.4 keV208Tl
238.6 keV212Pb
Fig. 2. Uranium spectra taken with a germanium detector—100–210 keV region [4].
X
1001 keV-234mPa
X X
10
X
X UF6, U from natural origin enr = 0.3%, 48Y container UF6, U from reprocessing, enr = 0.252%,48Y container
1 0
500
1000
1500 Energy (keV)
2000
2500
Fig. 3. UF6 spectra, from natural or reprocessed Uranium-The peaks marked with an X appear only in reprocessed Uranium.
184.2 keV, the determination of the 235U enrichment based on the 185.7 keV line analysis in Ref. [6] for these cylinders were very good.
are mainly emitted by 238U. 235U cannot be measured directly with passive neutron measurements. Fig. 4 shows the example of a 100 group neutron spectrums for natural UF6.
2.3. Neutron output 3. Characteristics of the UF6 cylinders In UF6, the neutron output arises primarily from spontaneous fission of 238U, and by 234U alpha decay and subsequent 19F (a, n) 22 Na reaction. Table 2 gives the relative contribution of the different U isotopes to the primary neutron production and for four typical enrichments. For natural uranium, the 234U and 238U contributions are almost equal. The 234U-produced neutrons dominate the (a, n) intensity of low enriched 235U. The spontaneous fission neutrons
The most commonly used cylinders are the 30B (2.5 tons UF6) cylinders for low-enriched product and 48Y (12 tons UF6) cylinders for natural feed and 48G for depleted uranium storage only. The nominal length is 206 cm for the 30B and 380 cm for the 48Y. The cylinders are constructed of ASTM A-516 steel [8]. The measurements performed in Ref. [9] showed that the 30B
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Table 2 Calculated relative contribution of the U isotopes to the total output neutron signal [7]. Isotopes
Amount (%) Neutron source 1/s/ton UF6
Relative yield (%)
Spontaneous fission UF6(a,n) Depleted U 232 U 234 U 235 U 236 U 238 U
0 0.00234 0.3 0 99.6977
0 0 0 0 9.96E + 03
0 7.85E+ 03 1.86E+ 03 0 7.53E+ 03
0 28.9 6.9 0 64.3
Natural U 232 U 234 U 235 U 236 U 238 U
0 0.0056 0.718 0 99.2764
0 0 0 0 9.92E + 03
0 1.88E+ 04 4.45E+ 02 0 7.49E+ 03
0 51.3 1.2 0 47.5
LEU 5% 232 U 234 U 235 U 236 U 238 U
0 0.039 5 0 94.961
0 0 0 0 9.50E +03
0 1.31E+ 05 3.10E+ 03 0 7.17E+ 03
0 86.9 2.1 0 11.1
0 0 0 0 9.50E +03
1.66E+ 04 4.46E+ 05 3.10E+ 03 2.76E+ 04 7.04E+ 03
3.26 87.48 0.61 5.41 3.24
neutron 100 group source 1/s/ton UF6
LEU 5% repr. 232 U 7.60E–07 234 U 0.133 235 U 5 236 U 1.55 238 U 93.317
3.0E+03 2.5E+03 2.0E+03 1.5E+03
(Alpha,n) spontaneous fission total
1.0E+03 5.0E+02 0.0E+00 1E+04
1E+05
1E+06
1E+07
Energy eV Fig. 4. Hundred group neutron spectra for natural UF6 [7].
container thickness varies from 12.5 to 13.8 mm. These deviations influence strongly the enrichment determination by application of the enrichment meter principle which requires a precise knowledge of the steel thickness at the measurement position. A variation of 70.5 mm of the wall thickness leads to a minimum error of 75% of the intensity of the 185.7 keV line of 235U. If there is no possibility to measure the wall thickness of the container with an ultrasonic gauge (contact measurement), the uncertainty on the 235U enrichment determination is at least 5%. 3.1. UF6 filling profile As it will be discussed later, the response of a neutron detector in the vicinity of a container depends strongly on the filling profile. UF6 is characterized by a high coefficient of expansion in the liquid phase. When heated in the containment autoclaves, the transformation of solid UF6 at 20 1C to liquid UF6 at 110 1C goes with a volume increase of 53%. Consequently, to prevent
Fig. 5. (a) Profile after filling at 80 1C. (b) Profile after desublimation.
deformation and rupture of the cylinders, the maximum UF6 mass must not exceed 2/3 of the maximum possible load of the cylinders. The distribution of the material within the container depends on how it was filled, on the last operation made on it (for instance, sampling in liquid phase after homogenization) or how long and under which conditions (temperature, sunshine) it was stored. When it is filled in liquid phase or after sampling, UF6 solidifies using all possible heat exchange surfaces. Most of UF6 remains in the lower part. However, there is a deposit of UF6 of several centimeters on all the upper part (Fig. 5a). When the container is filled by desublimation in a cylinder cooled at 25 1C, UF6 is deposited in a uniform way on all the
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inner surface of the cylinder (Fig. 5b) and forms an annular solid ring. When the container is filled with UF6 gas at 80 1C in a cylinder at 15 1C, first a layer of solid UF6 is deposited on the entire inner surface. Later, due to the increasing temperature of the solid UF6, additional amount of UF6 will condense inside the shell of solid UF6, flows by gravity in the cylinder and then solidifies. Consequently, this material will fill the lower part inside the solid UF6 shell (it is assumed that 75% of the gas is liquefied). As a result, we have the same filling profile as in Fig. 5a. With the changes of temperatures during the day, the filling profile changes. The coating sublimes from the cylinder wall to cooler cylinder areas, thus decreasing the thickness. However, no important modification due to the daily heat cycle is expected. 3.2. Chemical and radiochemical impurities When the cylinder is emptied by off gassing the UF6, the nonvolatile daughter products of 238U (234Th and 234mPa) as well as uranium deposit in the form of UF6, UO2F2, UF4 and UF5 and also volatile UF6 remain in the ‘‘heel’’. Heels in excess of 22.7 kg for 48Y or 48G and 11.3 kg for 30B require removal by cylinder cleaning. On filling a freshly emptied cylinder with UF6, the daughter products clearly remain mainly plated on the cylinder wall. The heel is responsible for significant radiation levels observed for empty cylinders. Since the non-volatile products are not self-shielded any more for an empty cylinder, the radiation dose increases by a factor of 12 for a freshly emptied 48Y cylinder and a factor of about 6 for a 30B cylinder [8]. The spatial distribution of the heel influences the amplitude and the spectral shape of the gamma radiation background. The contribution to the spectrum of the Compton scattering of the high energy gamma rays from 234mPa as well as Bremstrahlung production from the 2.3 MeV beta particle emitted in the decay of 234m Pa are sensitive to the location of the material. In addition, since the 185 keV gamma ray used for determining the 235U enrichment has a half thickness of 0.13 cm, the deposit can mask the signal from the actual UF6 filling. In the measurements carried out in Ref. [9], the authors mentioned that the residual activity was often clearly defined as forming a lateral band along the length of the cylinder. When measuring the 235U enrichment, the detector position has to be chosen in a way that the deposit does not influence the result. As already mentioned in Section 2.2, [5,6] also mentioned the presence of 137Cs and 233Pa (daughter of 237U) during their measurements on UF6 cylinders.
4. Classical NDA methods Typical NDA measurements are currently carried out by inspectors and the measurement devices are placed in contact with the cylinders. 4.1. Enrichment determination by gamma spectrometry Two basic NDA methods are routinely used to determine the U enrichment. One is based on a self-intrinsic calibration with unfolding of complex energy region 80–110 keV whereas the other makes use of the enrichment meter principle applied to the 185.7 keV line. In the case of measurement of UF6 containers, the thick shielding of the cylinder precludes the use of method based on the unfolding of low energy region. Several attempts were made to overcome the previous limitation when measuring U through thick container wall
235
313
[10,11]. These papers refer to the analysis of the 120–1200 keV region by intrinsic calibration using, the 143.8–163.3, 185.7 and 205.3 keV of 235U and the 258.3, 766.6 and 1001 keV from 234mPa, with an additional gamma ray from 234U at 120.9 keV for [10] only. Using either a standard empirical relative efficiency curve or a physical model for the sample detection system, these two applications were tested with various uranium enrichment standards using in particular 13 to 16 mm steel absorber and different germanium detectors. This approach has delivered good results but was tested on a limited number of items. Precise measurements would require very long counting times, especially when the detector has to be at 60 cm distance instead of being in contact with the drum surface. In addition, the presence of nonvolatile daughter product of 238U (paragraph 3.2) remaining in the heel may bias the results. The classical method to determine the 235U enrichment is the enrichment meter principle [12] making use of the proportionality between the net peak area of 185.7 keV line of 235U and the enrichment for infinitely thick items. This method requires the use of reference samples and since UF6 is very absorbent, the measurement results indicate only the enrichment at the cylinder wall. On the whole, the limitations encountered with the application of this method are mostly defined as a function of either the performance of the detectors used (NaI, LaBr3, Ge or CZT) in terms of efficiency or resolution or of the algorithms used in the analysis codes. For instance, the necessity of making wall thickness corrections accounting for the thickness and type of material placed between the source and the detector or the presence of interfering lines precludes the use of NaI detector with an analysis based on two ROIs of interest [13]. This method is used routinely to determine the 235U enrichment on UF6 containers either with low resolution detectors [5,6,14] or high resolution detectors [9,15]. The described characteristics of the gamma radiation measurements do not allow enrichment measurements of the required quality nor the direct or indirect determination of the 235U mass in a UF6 container. 4.2. UF6 mass and 235U enrichment determination with passive and active neutron assay 4.2.1. Passive-neutron assay The passive neutron detection consists in measuring the fast neutron flux emerging from a UF6 cylinder. The primary flux is due to F(a, n) reactions induced by the 234U alpha activity. If a constant 235U/234U ratio can be assumed, the fast neutron flux is then a linear function of 235U. The total neutron source term from a mass of UF6 can be written as [7] Q ¼ Mu ð3230000 f32 þ496 f34 þ0:0917 f35 þ 2:64 f36 þ ð0:0112 þSF38 Þ f38 Þ where Q is the neutron source strength in n/s, Mu is the uranium mass in g, f is the fractional isotopic abundance of the subscripted isotope and SF38 is the spontaneous fission coefficient of 238U. If the terms with f32, f35 and f36 can be neglected, the previous expression becomes Q ¼ Mu ð496f34 þ ðð0:0112 þ SF38 Þf38 ÞÞ: If the ratio f34/f35 is known and constant, the neutron source strength can be written as Q ¼ Mu ða þ bf35 Þ:
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Several experiments [5,6] were carried out to find the relationship between the neutron detector response, the 235U enrichment and the total UF6 mass. The passive-neutron measurements were performed with shielded neutron assay probes (SNAP) on the container surface and in a low lateral position. The conclusions of Refs. [5,6] were:
There is a linear relationship between the detector response
and the enrichment for a given UF6 mass for an enrichment ranging from depleted to 5%. The variations in the neutron leakage (fraction of the source neutron that escape from the cylinder) per source neutrons caused by large change in enrichment and cylinders loadings are small.
A drawback of the passive-neutron assay is that a falsification of the 235U enrichment declaration, which cannot be detected by gamma spectrometry, could lead to the expected neutron count rate (i.e. according to the declaration) by compensating for the missing neutrons with HEU material hidden in the cylinder.
a linear relationship between the delayed neutron signal and the uranium enrichment in Ref. [18], 235 U enrichment with a calibration curve using the 235U weight and the ratio 974 keV(132Sb)/1031 keV (89Rb) [19].
the
These methods demonstrated a high sensitivity well adapted to the detection of hidden nuclear material and appear as a powerful and promising tool for determining either the U enrichment or U mass given a value for the other. However, the transposition of such experiments to the UF6 containers would require an experimental study which up to now has not been done.
6. Specific investigations with respect to the design of an unattended measurement station for UF6 containers The previous section showed that it is possible to determine: 235 U enrichment with a gamma spectrometer when the wall thickness can be measured at the measurement point, when the contribution of interfering gamma lines to the 185.7 keV line can be evaluated, and when 234mPa deposits do not mask the signal, The 235U content and the UF6 mass with passive neutron counting provided that the 234f/235f ratio is known. Contrary to the gamma spectrometric measurements, neutron-based techniques are not sensitive to deposits of protactinium, to the container wall thickness variation and to the presence of potential interference, The 235U enrichment with an active neutron system making use of special geometry conditions (source and detector in contact with the container).
The 4.2.2. Active-neutron assay A usual active measurement technique for the enrichment is the irradiation of the UF6 with thermal neutrons and the observation of the fast neutron from induced fissions of 235U. Contrary to the passive measurement case, there is no need to know the f34/f35 ratio. The mean-free-path for thermal neutrons in solid UF6 is a strong function of 235U enrichment varying from 24 cm for 0.3% to 4 cm for 4%. This leads to a correspondingly strong dependence of the detected fast neutron flux on 235U enrichment. Several experiments [5,6,16] were performed in the 70’s with techniques based on thermal neutron induced fission of 235U in UF6 cylinders (30B and 48Y). For the three experiments, the neutron detector was placed in contact with the cylinders. The results showed that, up to 3% the net count rate is a linear function of the enrichment for the three experiments whereas for the 0.3–4% enrichment range, results of Ref. [6] show a non-linear behavior which was then confirmed with MCNP calculations.
5. Analysis of delayed neutrons and delayed photons Both classical gamma spectrometric and neutron assay described in the two previous sections cannot provide any information about the inner part of the large UF6 cylinders. For this reason, alternative methods having the possibility of overcoming this limitation are of interest. Applications such as systems for border security monitoring or for radioactive waste management for which highly sensitivity methods are required lead to recent investigations making use of the analysis of delayed neutrons and b-delayed gamma rays emitted by uranium fission products to determine the 235U enrichment or 235U mass. In these experiments, the interrogation of small samples was made with either a 14 MeV pulsed generator [17], or with a pulsed neutron beam produced by a LINAC [18,19] or with thermal neutron flux [20]. The combination of an irradiation with very penetrating 14 MeV neutrons associated to the detection of delayed gamma lines above 3 MeV (also very penetrating) could be used. Proof of principle measurements were performed in the previously cited papers on small samples for short counting time and it was possible to determine
the
235
U enrichment by fitting the temporal die away tail in Ref. [17],
For the unattended measurement, a minimum distance between the cylinders and the measurement device has to be respected to allow safe movement of the cylinders. Consequently, the wall thickness cannot be measured point with ultrasonic gauge (contact measurement). Typical variation of the wall thickness from the nominal value easily leads [13] to an enrichment error of 6% which is above the maximum uncertainty specified in Ref. [1] for low enriched products. The use of active neutron system for determining the 235U is also prohibited in the sense that this technique makes use of a special geometry and cannot be applied easily if the cylinder is not in contact with the detector. Additional MCNP calculations [21] showed that activation with a 14 MeV neutron generator gives a signal which is not very sensitive to the 235U content. A measurement campaign [22] has also shown that for measurements times which are compatible with in the field measurements (typically 5 min), the estimated count rates for a large germanium detector are far below the requested accuracy for depleted uranium in 48Y containers [1]. Also the use of a more efficient detector such as a LaBr3 detector tested in the field [23] does not allow the requested accuracy to be reached. From the above, the 235U enrichment and the UF6 mass determination can be only verified with passive neutron counting provided that the ratio 234U/235U is known. MCNP calculations were carried out to confirm the relationship between the neutron output and the UF6 weight or the 235U enrichment given in Refs. [5,6].
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6.1. Calculation of the neutron detector response for passive measurements In Ref. [7], the response of 3He detectors due to emission of neutrons from UF6 containers was calculated with MCNP for four neutron detectors arranged around the steel container (Fig. 6). The detectors are in fixed positions for both containers at 60 cm distance and at mid height of the axis of the containers. Each detector comprises five 3He tubes of 2.54 diameter and 1 m active length with a pressure of 4 atm. The neutron count rate of this set-up can be used to determine the 235U, provided the 234U/235U ratio is known. If also the enrichment is known, the total UF6 mass can be calculated; alternatively, if the UF6 mass is known, the enrichment. The relation between the detector response and the fuel enrichment or the total UF6 mass was investigated for the containers types 30B and 48Y. Since the UF6 filling profile is not known a priori five different fuel filling profiles—numbered from 1 (x =0) to 5 (x =100) were taken into account (Fig. 7). They cover the two cases described in Figs. 5a and b of Section 3.1. The parameter x represents the percentage of UF6 covering all the inner cylinder walls with a layer of constant thickness. The rest of the material fills the remaining volume from below. The simulations were made for eight different isotopic compositions including that of reprocessed material (Table 3). For the container 30B all fuel compositions were taken into account whereas for the 48Y container, only the compositions 1, 2, 3 and 7 of Table 3 were considered. The neutron flux and the detector response were calculated with the Monte Carlo code MCNPX [24]. All neutron cross-sections were taken from the ENDF/B VI.1 and ENDF/B VI.O libraries. Table 4a and b give the detector response calculated for four neutron detectors for
315
both container types. The statistical uncertainty was always less than 1%. One can see that the detector response depends very much on the filling profile as it increases from filling profile x = 0 to 100 by a factor of 1.6. The calculations show that a minimum of 156,000 counts and 828,000 counts can be achieved in 2 min for the 30B container and for the 48Y, respectively, with the considered device. The potential background contribution of one container with the same load of UF6 at 300 cm distance behind the detector is 10% of the total neutrons signal, but it can be reduced substantially by appropriate shielding of the detector. Monitoring of background radiation in the environment of the measurement station may be used to make automatically background corrections. 6.2. Detector response as a function of enrichment The simulation suggests that the neutron source strength is a linear function of the 235U wt% for a given UF6 mass. Figs. 8a and b show the detector response as a function of 235U wt%, for different filling profiles and for both containers filled to 100% of the UF6 mass. Table 3 Isotopic composition used in MCNP calculations (in wt%). No.
Material type
232
1 2 3 4 5 6 7 8
Natural Dep03 LEU10 LEU30 LEU50 Repro 6.49% Sample3.25% Sample 5%
0 0 0 0 0 1.200E–08 4.800E–07 7.600E–07
U
234
235
0.0056 0.0023 0.0078 0.0234 0.0390 0.0670 0.0850 0.1330
0.7180 0.3000 1.0000 3.0000 5.0000 6.4710 3.2500 5.0000
U
U
236
U
0 0 0 0 0 0.05 1.09 1.55
238
234
99.2764 99.6977 98.9922 96.9766 94.9610 93.4620 96.6650 94.8670
0.007799 0.0078 0.0078 0.0078 0.0078 0.010354 0.026154 0.0266
U
U/235U
detector Table 4 No. Composition Detector response 1/s x=0
120
x =25
x= 50
x =75
x= 100
2.589E+ 02 1.796E+ 02 3.109E+ 02 7.049E+ 02 1.136E+ 03 1.924E+ 03 2.474E+ 03 3.822E+ 03
2.809E + 02 1.939E + 02 3.400E+ 02 7.722E + 02 1.211E + 03 2.029E + 03 2.640E + 03 4.101E + 03
2.941E +02 2.075E +02 3.576E +02 8.115E +02 1.292E +03 2.145E +03 2.817E +03 4.410E +03
(b) Detector response for the container 48Y 1 Natural 1.008E+ 03 1.268E + 02 1.420E+ 03 2 Dep03 6.909E+ 02 8.691E + 02 9.766E+ 02 3 LEU10 1.209E+ 03 1.550E + 03 1.723E+ 03 7 Sample3.25% 1.009E+ 03 1.255E + 04 1.413E+ 04
1.548E + 03 1.070E+ 03 1.888E + 03 1.524E + 04
1.625E +03 1.115E +03 1.984E +03 1.609E +04
(a) Detector response for the container 30B 1 Natural 1.872E + 02 2.316E + 02 2 Dep03 1.301E+ 02 1.620E + 02 3 LEU10 2.235E + 02 2.803E + 02 4 LEU30 5.152E + 02 6.469E + 02 5 LEU50 8.282E + 02 1.027E + 03 6 Repro 6.49% 1.410E+ 03 1.748E + 03 7 Sample3.25% 1.807E+ 03 2.239E + 03 8 Sample 5% 2.850E+ 03 3.507E + 03
30B 100
48Y 58 10
Fig. 6. Measurement geometry for passive neutron measurement (dimensions in cm).
X=0
X = 25
X = 50
Fig. 7. UF6 filling profiles.
X = 75
X = 100
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detector response 1/s
1400 1200 1000 800 x=0 x = 25 x = 50 x = 75 x = 100
600 400 200 0 0
1
2
3
4
5
6
U-235 wt%
2000
Fig. 10. Cross-section through the containers with position of detectors (in cm).
1500
2200
x=0 x = 25 x = 50 x = 75 x = 100
1000 500 0 0
0.2
0.4
0.6 U-235 wt%
0.8
1
2000
1.2
Fig. 8. (a) Dependence of the detector response on the 235U wt% for container 30B (2.2 tons of UF6). (b) Dependence of the detector response on the 235U wt% for container 48Y (12 tons of UF6).
detector response 1/s
detector response 1/s
2500
detector position 1 detector position 2 detector position 3
1600 1400 1200 1000 800
2500
0
25
50 filling profile number
75
100
2000
5000
1500 1000 0.7% enrichment 0.3% enrichment 1% enrichment
500 0 0
25
50
75
100
Filling profile number
detector response 1/s
Number of counts
1800
4500 4000 detector position 1 detector position 2 detector posiion 3
3500 3000
Fig. 9. Dependence of the detector response on the filling profile.
2500
Consequently, the relation between the count rate R, UF6 mass and the enrichment f35 can be approximately described by the formula R ¼ ZðMUF6 Þ ðc þd f35 Þ: The function Z and the parameters c and d depend on the container type and the specific set-up of the measurement station. The gradient d of the functions increases from x= 0 to 100. The equally thick layer of UF6 on all internal surfaces gives the highest detector response. The dependence of the detector response on the geometry parameter x (Fig. 7) is illustrated in Fig. 9 for the 48Y container. Because of the strong dependence of the detector response on the filling profile, the filling profile has to be known in order to determine the 235U wt% from the detector response. Otherwise, e.g., depleted uranium cannot be distinguished from 1% enriched U (48Y) or 3% from 5% (30B), in unfortunate cases.
0
25
50 filling profile number
75
100
Fig. 11. (a) Detector response for container 48Y. (b) Detector response for container 30B.
6.3. Influence of detector positions In order to reduce the strong dependence of the detector response on the filling profile, the detector positions were changed (Fig. 10). For the first case, the lateral distance of the detectors from the containers is kept but the detectors are shifted vertically by 120 and 120 cm, respectively (positions 2). In the next case the detectors are positioned 120 cm above and below of the containers (positions 3). The calculated detector responses for the three detector positions are given in Fig. 11a and b. The detector positions 2
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and 3 clearly reduce the dependence of the detector response on the filling profile whereas the position 3 always shows the highest detector response. Because of the rotation symmetry the detector positions 1 and 3 give the same values for the filling profile x =100. For the container 48Y, the ratio of the maximum to the minimum neutron signal is 1.63, 1.13 and 1.09 for the detector positions 1, 2 and 3, respectively. The corresponding values are 1.55, 1.14 and 1.03 for the container type 30B. Further optimization calculations can be made but they are of interest only if they take into consideration the local conditions of a possible measurement station. They have to include scattered radiation, the possible presence of other UF6 containers with different enrichments in vicinity of the measurement station and a shielding of the detectors.
Table 5 Contributions of volume layers to the detector response for geometry 1. Filling profile
x =100
relative contribution to the total signal (%)
the function Z(MUF6) has to be evaluated for a given cylinder type. For the calculation of the contributions of different UF6 volume layers to the detector response of a 100% filled container, the UF6 volume was divided into four volume layers numbered from 1 to 4 as shown in Fig. 12a and b. Only the filling profiles x= 100 and 50 of Fig. 7 were considered. All volume layers have the same UF6 mass, 3000 kg for container 48Y and 550 kg for container 30B. The results calculated with the MCNP code are represented in Table 5 for detector positions 1 of Fig. 10. For the filling profile x =100 all four volume elements are contributing in a similar way to the detector signal. None of them is ‘‘invisible’’. For the filling profile x= 50, the figures show that number 3 is the least visible volume element.
Detector response contribution % 48Y/LEU10
30B/Sample 5%
20.3 22.8 26.6 30.3 31.8 34.6 8.6 25
23.4 24.5 25.2 26.9 29.3 30.5 15.4 24.8
120.00 filling profile x = 100 filling profile x = 50
100.00 80.00
y = -8E-06x2 + 0.0622x + 0.2486
60.00 40.00
y = -4E-06x2 + 0.0536x - 0.275 R2 = 0.998
20.00 0.00 0
500
1500 1000 mass (kg)
2000
2500
120.00 relative contribution to the total signal (%)
R ¼ ZðMUF6 Þ ðc þd f35 Þ
Layer no.
1 2 3 4 1 2 3 4
x =50
6.4. Contributions of the volume elements inside the container If the detector response is written as
317
100.00 y = -8E-07x2 + 0.0181x - 7.8 80.00 60.00
y = -5E-07x2 + 0.0148x - 6.075
40.00 filling profile x = 100 filling profile x = 50
20.00 0.00 0
2000
4000
6000 8000 mass (kg)
10000
12000
14000
Fig. 13. (a) Relative contribution to the total signal for a 30B container, 5% enrichment. (b) Relative contribution to the total signal for a 48B container, 1% enrichment.
6.5. Detector response to partially filled drums
Fig. 12. (a) x–z cross-section of the UF6 volume layers for the filling profile x= 100 and (b) x–z cross-section of the UF6 volume layers for the filling profile x = 50.
The assumed filling processes were mentioned in Section 6.1. The filling by desublimation corresponds to the filling profile x= 100 whereas the filling at 80 1C corresponds to the filling profile x= 50. Figs. 13a and b show the relative detector response as a function the mass. This simulation function is referred as Z(MUF6) and the calculation were made for the detection position 1 of Fig. 10. The function Z(MUF6) can be fitted with a polynomial. It does not show a saturation effect. The very slight non-linearity is due to the self absorption in the material. To characterize once again the ‘‘visibility’’ of the UF6 and its dependence on the filling profile, the neutron leakage fraction was calculated in Fig. 14. Results were arbitrarily normalized to 1 for an UF6 weight equal to one fourth of the total load for both container types. Fig. 14a and b show that the leakage fraction
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1.1
The present MCNP calculations of the passive neutron emission carried out have shown that:
1
The results depend strongly on the UF6 filling profile, however,
0.9
0.8 filling profile x = 50 filling profile x = 100
0.7
0.6 500
0
1000 1500 weight (kg)
2000
2500
1.2
neutron leakage fraction
1
an effect which can be minimized by optimization of the detector position relative to the containers. The total number of counts is found to be a linear function of the 235U enrichment (for 100% filled drums) and is also rather a linear function of the UF6 mass (for partially filled drums). In a scenario whereby 10% of the UF6 mass would be missing from the cylinder, the total number of neutrons would also diminish by nearly 10% for both container types. At least 156,000 counts and 828,000 counts may be collected in 2 min, for the 30B container and for the 48Y, respectively, as a sum over four 3He detectors (without back shielding), thus allowing a very short measurement time. The potential background contribution of a drum at 300 cm distance is 10% of the total neutron signal.
However, the inspection systems have to take into consideration the physics limitations of the proposed method and MCNP calculations can be made to estimate which mass of high enriched uranium could be hidden in the centre of the container.
0.8 0.6 0.4
Acknowledgments
filling profile x = 50 filling profile x = 100
0.2 0 0
2000
4000
6000 8000 weight (kg)
10000
12000
14000
Fig. 14. (a) Neutron leakage fraction for a 30B container filled with 5% enriched UF6 for detectors placed in position 1 of Fig. 10. (b) Neutron leakage fraction for a 48Y container filled with 1% enriched UF6 for detectors placed in position 1 of Fig. 10.
remains quite high also for the large drums and demonstrate the clear effect of the filling profile for detector geometry 1. We observe small variations of the leakage fraction as a function of the UF6 mass and confirm in this way the results of Ref. [6]: in this paper, the authors calculated a leakage factor varying from 0.93 to 0.82 as the weight of the 30B container is increased from 2000 to 5000 lb for a filling profile which is not described. The ratio of the leakage fraction for 5000 lb over the leakage fraction for 2000 lb gives 1.13. Given the results of the present document, the corresponding ratio is equal to 1.09 and 1.23 for x =100 and 50. Since we can easily imagine that the filling profile of Ref. [6] is something between x= 100 and 50, respectively, there is a very good agreement between the results of the present simulation [7,21] and the calculations of Ref. [6].
7. Conclusion Of the several methods investigated, passive-neutron counting is the most appropriate to verify the 235U enrichment and UF6 weight in unattended mode provided that the ratio f34/f35 is known. As already mentioned in Ref. [6], all the volume elements inside the large containers contribute to the measured signal whereas the established inspection mode-gamma spectrometric enrichment measurements-see only a thin surface layer of the UF6.
The authors would like express their thanks to Mr. Olivier HALNA DU FRETAY, AREVA NC for very helpful discussions on the chemistry of cylinders heels as well as the cylinder filling profiles, to Robert ALVES, AREVA NC and Antoine RUIZ, DG TREN for their continuous support, Dr Bent PEDERSEN/JRC Ispra for helpful discussions. References [1] A. Lebrun, User requirements ‘‘state of the art NDA methods applicable to UF6 cylinders’’, SG-EQ-NDA-UR 0001, 2007. [2] S. Richter, A. Alonso, W. De Bolle, R. Wellum, P.D.P. Taylor, International Journal of Mass Spectrometry 193 (1999) 9. [3] R. Gunnink, W.D. Ruhter, P. Miller, J. Goerten, M. Swinhoe, H. Wagner, J. Verplancke, M. Bickel, S. Aboushal, MGAU, a new analysis code for measuring U-235 enrichments in arbitrary samples, IAEA symposium on International Safeguards, Vienna, Austria, March 8–14, 1994. [4] P. Mortreau, R. Berndt, Handbook of Gamma Spectrometry Methods for NonDestructive Assay of Nuclear Materials, EUR 19822 EN, version 3, 2006 /http://nuclearsafeguards.jrc.it/nda/handbook3-2006.pdfS. [5] P.E. Fehlau, W.H. Chambers, Perimeters safeguards techniques for uranium enrichment plants, LA-8997-MS, 1981. [6] R.B. Walton, T. Reilly, J.L. Parker, J.H. Menzel, E.D. Marshall, L.W. Fields, Measurements of UF6 cylinders with portable instruments, Nuclear Technology 21 February, 1974. [7] E. Franke, Calculation of the neutron response for passive measurements, Report-No UF6-01/08, 2008. [8] UF6, a manual of good handling practices, USEC-651, revision 7, January 1995. [9] R.S.T. Shaw, The routine non-destructive measurement of the 235U concentration of enriched uranium hexafluoride in transit cylinders using high resolution gamma spectrometry Proceeding ESARDA 1979, Brussels. [10] W. Ruhter, T.F. Wang, C.F. Hayden, Uranium enrichment measurements without calibration using gamma rays above 100 keV, UCRL-JC-142832, October 29–November, 2001. [11] E. Sampson, P.A. Hypes, D.T. Vo, FRAM isotopic analysis of uranium in thickwalled containers using high energy gamma rays and planar HPGe detectors Proceedings, Institute Nuclear Management, 2002. [12] T.D. Reilly, R.B. Walton, J.L. Parker, Nuclear safeguards research and development LA-4605-MS, Los Alamos Scientific Laboratory, 1970, pp. 19–21. [13] R. Berndt, P. Mortreau, E. Franke, Survey of state-of-art NDA methods applicable to UF6 cylinders—IAEA task n 07/TAU-04 JRC Scientific and Technical Reports, internal distribution. [14] A.R. Flynn, Field determination of uranium-235 enrichment by portable gamma ray spectrometry, Report K-1819-1972. [15] R.C. Hagenauer, H.Y. Rollen J.M. Whittaker, R.L. Mayer, T. Biro, Nondestructive measurement of UF6 cylinders at the Portsmouth gaseous diffusion plant, Nuclear material management—annual meeting, 1998.
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