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Characterization of a Thermo Scientific D711 D-T neutron generator located in a low-scatter facility
Nuclear Instruments and Methods in Physics Research A 741 (2014) 57–66
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Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
Characterization of a Thermo Scientific D711 D-T neutron generator located in a low-scatter facility John W. Hayes 1, Erin Finn n, Larry Greenwood 2, Rick Wittman 3 Pacific Northwest National Laboratory, PO Box 999, Richland, WA 99352, USA
art ic l e i nf o
a b s t r a c t
Article history: Received 2 July 2013 Received in revised form 4 November 2013 Accepted 6 November 2013 Available online 28 November 2013
A dosimetry experiment used to measure the neutron flux and spectrum of a D-T neutron generator is presented. The D-T generator at Pacific Northwest National Laboratory is installed in the middle of a large room to minimize scatter of neutrons back to the sample. The efficacy of maintaining a pure fast neutron field for the sample is investigated. Twenty-one positions within 13 cm of the neutron source contained foils or wires of Fe, Ni, and Al with additional Au, and in monitors at some locations. Spectral adjustment of the neutron flux at each position based on the measured reaction rates and theoretical Monte Carlo calculations show that at least 99.1% of the spectrum lies above 110 keV for all measured positions, and neutrons above 14 MeV can account for as much as 91% at locations along the axis of the generator and close to the source. The 14 MeV component drops to 77% in radial positions far from the source. The largest total flux observed was 8.29E þ08 n/cm2 s (7 1.4%) in the center of the cooling cap, although additional experiments have shown this value could be as high as 1.20E þ09 n/cm2 s. & 2014 Published by Elsevier B.V.
Keywords: 14 MeV D-T generator Neutron spectral adjustment MCNP STAYSL PNNL
1. Introduction Pacific Northwest National Laboratory (PNNL) purchased and installed a D711 model D-T neutron generator (“generator”) from Thermo Scientific in August 2011. The generator nominally produces 14 MeV neutrons isotropically as a product of the fusion event from an accelerated deuteron on a tritium target. Fast neutron sources are important in matters of national security and nuclear data research for several reasons. First, fast neutrons provide the capability of harnessing threshold reactions for the production of rare isotopes, which are of interest to radiochemistry groups at PNNL concerned with validating radioanalytical techniques for the separation and characterization of these isotopes. Fast fission reactions on 235U, 238U, and 239Pu using the D-T source are also capable of producing sufficiently high yields of rare fission product isotopes in order to further develop these techniques. Research in this area has progressed significantly by having the capability to produce rare isotopes, which may not be produced by other means or are prohibitively expensive or logistically difficult to acquire from sources outside of PNNL such as nuclear reactors or other accelerators. Experiments with 14 MeV neutrons are also of interest because
n
Corresponding author. Tel.: þ 1 509375-7374. E-mail addresses: [email protected] (J.W. Hayes), [email protected] (E. Finn), [email protected] (L. Greenwood), [email protected] (R. Wittman). 1 Permanent address: 891 Bridge Walk Ct, Ada, MI 49301, USA, Tel.: þ 1 616 217-7245. 2 Tel.: þ 1 509 375-5301. 3 Tel.: þ 1 509 375-5111. 0168-9002/$ - see front matter & 2014 Published by Elsevier B.V. http://dx.doi.org/10.1016/j.nima.2013.11.023
nuclear data for fast fission has not been studied as extensively as it has been for thermal fission. Groups at PNNL have already performed experiments studying 14 MeV fission of 235U with the intent to determine the fission product yield of short-lived isotopes more precisely. Research applications of the D-T generator require that the neutron spectrum be as “pure” as possible near 14 MeV. The presence of a significant slow neutron component in the spectrum introduces larger uncertainties in the measurement of 14 MeV fission yields because slow neutrons contribute to the production of the fission product isotopes and one must correct for the slow component when calculating the 14 MeV production rate. This is especially a problem if the spectrum has a significant thermal or epithermal neutron component since the fission cross-sections are much higher at these energies than at 14 MeV. Therefore, installation of the generator was designed to minimize the slow component, which is primarily due to scattering of the source neutrons back to the generator head. By housing the generator in the center of a low-scatter room where the closest wall is more than 4 m away, the room return effect was diminished to the extent possible. The initial experiments performed with the generator were intended to characterize the neutron spectrum and intensity so that the proportion of the slow component to the total flux could be measured.
2. Materials and methods The first section includes a discussion of prior attempts at characterizing the flux at similar facilities as well as alternative
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methods to foil irradiation considered at this one. In the next sections, the details of the experiment are presented, followed by an explanation of the analysis methods used. Dosimetry measurements of the real flux environment are used to validate theoretical calculations and adjust flux spectra within associated uncertainties and based on limiting fitting parameters. 2.1. Previous characterization attempts Indirect methods of measuring the flux from similar D-T neutron generators by the associated alpha-particle method have been performed by Robertson et al. [1], Shirato et al. [2], and Hertel et al [3]. The method in these 3 papers however can only provide total flux results and not spectral information. Kehayias et al. [4] determined the total flux for an A-325 model D-T generator using silicon foil activation but again provided no spectral information. Jonah et al. [5] used foil activation to determine the complete energy spectrum for an A-711 KAMAN neutron generator with 106 n/s output. Irradiation of 8 pure metals and the unfolding of the spectrum by the SULSA computer code closely matches the method presented here. SULSA is an unfolding code based on a least squares method of measured reaction rates only, while STAYSL PNNL is an adjustment code that uses reaction rates and an input spectrum. Other methods of measuring the neutron flux directly, such as the NE213 liquid scintillator used by Shirato and Hertel, or a Bonner spheres method, can provide confidence to the total neutron yield determined in this paper, but have not been performed at this facility to date. These methods cannot determine flux as a function of space around the source which was also a goal of this experiment. Therefore, foil irradiation provided the most effective way to determine the flux spectra at many points near the source. 2.2. Sample irradiation In this experiment, an irradiation of dosimetry foils and wires was conducted with the aim of characterizing the neutron spectrum at 21 points in a hemispherical array near the generator head. An aluminum apparatus was designed to fit on the cooling cap and act as a sample holder for the space in a hemisphere of radius 13 cm above the center of the D-T reaction. The D-T source that produces a fast neutron is assumed to take place in the shape of a circular disc just below the top of the cooling cap. The apparatus is shown in Fig. 1 in its experimental configuration on the generator. The relation of the generator to the walls is also implied in the figure: the distances to the side walls are 5.4 m and 6.9 m. The D-T source is 4.3 m above the floor, which is also the closest boundary of the room to the generator. Pure iron, aluminum, nickel, gold, and indium metal foils or wires were heat sealed in tightly conformed plastic to make 21 sample packets. The sample packets were affixed to various positions on the apparatus. Fig. 2 shows a schematic of the locations of each of the sample packets relative to the disc source. This schematic comes from the Monte Carlo N-Particle (MCNP) model [6,7] used for analysis described in Section 2.2. The positions are labeled 1–21 for easier reference. Packets 1, 2, 7, and 8 contained 0.5 in. diameter foils of Fe (31 mils thickness), Ni (10 mils), Al (30mils), Au (2 mils), and In (5 mils). Packets 3 and 4 only included Fe, Ni, and Al foils, but of the same diameter and thickness. Packets 5 and 6 also only included Fe, Ni, and Al foils, but the Fe foil was 50 mils thick so that reasonable activities could still be achieved. Packets 9–21 contained Fe, Ni, and Al wires of 62 mil diameter and lengths ranging from 1.0 to 2.5 cm. Each of the metal foils and wires was carefully weighed for purposes of gamma spectroscopy following irradiation.
Fig. 1. The aluminum sample holder placed on the cooling cap of the D-T generator.
Fig. 2 also shows a coordinate system to describe any point in space relative to the center of the disc source. A point P is given by coordinates (a, h), where a is the vertical distance perpendicular to the plane of the disc source and h is the horizontal offset from the disc center. Later analysis and discussion of the spatial flux will take advantage of this coordinate system. For each position analyzed during this experiment, the point P is taken to be the geometric center of the space occupied by the sample. The coordinates of the position numbers, the direct distance d to the center of the disc source, and the angle θ defined by the vectors a and d are given in Table 1. The sample packets were irradiated for 3 h and 36 min with a ramp to full power (160 kV and 3.0 mA) of 34 min and 40 s. The irradiated samples were then gamma counted with high-purity germanium (HPGe) detectors and gamma spectra were acquired in the software program GENIE 2000 offered by Canberra [8]. The detectors are calibrated with National Institute of Standards and Technology; traceable standards and control counts for the detectors are performed daily, while background counts are performed at least weekly. Gamma spectroscopy analysis was performed using the GENIE 2000 software to determine the product activity decay-corrected to the end of irradiation. Nuclear data for analysis was taken from the National Nuclear Data Center at Brookhaven National Laboratory [9]. Figs. 3 and 4 show examples of spectra acquired in GENIE from the Fe wire in Position 1. The activation products 56Mn and 54Mn are created from (n, p) reactions on 56Fe and 54Fe, respectively. Fig. 3 is an immediate count, therefore the activity of 56Mn (2.6 h half-life) dominates the spectrum as evident by the peak at 846 keV. Fig. 4 shows that when the sample was recounted the following day, the 56Mn activity had decayed away appreciably. The ability to resolve the peak at 834 keV from 54Mn then allowed for a determination of its activity. As described in Section 2.3, the 56Mn and 54Mn activities (as well as all of the sample product activities) are then used as input into the STAYSL PNNL code. The complete list of reactions studied and the gamma lines from spectra acquired in GENIE used to calculate the product activity are summarized in Table 2. The table also provides a range of energies that were expected to account for 90% of the reaction rate based on preliminary calculations. The majority are purely fast reactions leading to high resolution in the region of the spectrum
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Fig. 2. MCNP model of the sample apparatus and experimental configuration for 21 sample positions. The two views are shown in the XZ (left) and YZ (right) planes.
Table 1 Sample positions relative to neutron source in centimeters. Position
a
h
d
θ
Position
a
h
d
θ
1 2 3 4 5 6 7 8 9 10 11
4.0 5.8 4.0 5.8 7.8 11.5 7.8 11.5 0.6 0.6 0.6
4.1 2.4 4.1 2.4 8.9 5.1 8.9 5.1 4.8 5.8 7.2
5.8 6.3 5.8 6.3 11.8 12.6 11.8 12.6 4.8 5.8 7.2
45.7 22.4 45.7 22.4 48.8 24.0 48.8 24.0 82.4 83.7 84.9
12 13 14 15 16 17 18 19 20 21 –
0.6 0.6 0.6 1.1 2.2 3.2 4.5 6.9 8.4 10.7 –
8.8 4.8 6.8 0.2 0.2 0.2 0.2 0.7 0.7 0.7 –
8.8 4.8 6.8 1.2 2.2 3.3 4.6 6.9 8.5 10.8 –
85.8 82.3 84.6 10.4 5.5 3.7 2.6 6.0 4.9 3.8 –
near 14 MeV. In locations that included Au and In foils, better resolution of the thermal region could also be achieved because of the capture reactions that extend to thermal energies. Both 58Co and 196Au have isomers with half-lives of about 9 h. Additional gamma counting was performed in both cases such that the reaction rates reported for both of these reactions include the complete in-growth of the isomeric states as well as direct production of the ground states.
neutron was tallied among 200 directional cosine bins from 01 to 1801 (0.01 rad¼ 0.5731 / bin.) The neutron energy was also tallied among 100 energy bins within each cosine bin. The energy bin boundaries varied depending on the direction but remained within several hundred kiloelectron volt of the average energy in order to obtain high resolution of the peak in that cosine bin. The neutron energy spectra as a function of angle are shown in Fig. 6. The total yield at each angle is normalized to 1. Neutrons are emitted with the greatest energy in the forward direction (θ ¼0) and are emitted with lower energies as the polar angle increases. The validity of this modeling method was checked by demonstrating that when 360 kV deuterons are used it gives consistent results to the neutron energy spectra calculated by Greenwood et al [11] for a beam of 360 kV deuterons incident on a tritium target. The neutron energy spectra with angular dependence from this initial model were then used in the full model5 of the generator present in the low-scatter room. The spatial distribution of the neutrons was modeled as a disc source of radius 1.65 cm near the top of the cooling cap (Fig. 2). The flux in each sample was calculated with an f4 tally in 129 groups including 100 keV bins from 12.5 to 16 MeV in order to achieve high resolution of the peak near 14 MeV. The spectra calculated by MCNP for each position could then be used as the first approximation for input into the STAYSL PNNL code described in the next section.
2.3. Monte Carlo calculation 2.4. Spectral adjustment with STAYSL PNNL The MCNP code was used to model the experimental setup and calculate the neutron spectrum in all 21 locations in 129 energy groups. The model was adapted from a previous model of the lowscatter room developed by PNNL to calculate dose from a 252Cf source also used in the facility [10]. A separate model of the generator received from Thermo Scientific was updated with asbuilt dimensions and then inserted into the low-scatter room model in its location in the center of the room (Fig. 5). The model of the neutron source in MCNP incorporated a sophisticated energy-angle dependent source representative of the kinematics of the D-T fusion event. First, MCNP4 was used to calculate the neutron energy spectrum in a model in which a deuterium particle was accelerated at 160 kV onto a tritium target adsorbed in a zinc hydride matrix inside a vacuum. The product 4
Monte Carlo N-Particle eXtended (MCNPX) was used for this calculation.
The STAYSL PNNL code [12] is comprised of 6 code modules designed to yield adjusted neutron flux spectra based on measured neutron activation rates and theoretical flux calculations as input. First the code was used to calculate saturated reaction rates by taking the activities of the irradiated samples calculated in GENIE and correcting for atomic weight, isotopic abundance, gamma self-absorption, decay during irradiation, and neutron self-absorption (neutron burnup of the target or product atoms was considered negligible due to the short time of irradiation). The first 3 corrections were calculated in the module SigPhi Calculator and are shown in Table 3. Gamma self-absorption corrections, which account for photon energy and sample geometry, averaged about 1% in the foils and about 3% in the wires. 5
MCNP5 was used in this test.
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Fig. 3. GENIE gamma spectrum for Fe wire in Position 1. The 846 keV peak is used to calculate the activity of
Fig. 4. GENIE gamma spectrum for Fe wire in Position 1 recounted 20 h after the end of irradiation. The
The decay during irradiation correction, which is also shown in Table 3, was determined in the module BCF, or Beam Correction Factor code, by calculating the production and decay of each activation product over the entire irradiation history. The control console of the generator outputs the measured values of high voltage (HV) and beam current (B) to a log file every 10 s so that a model of the flux can be fit based on these values following the experiment. Because the dependence of flux on the operating conditions was not known, the initial model assumed flux was linear with B but independent of HV. The beam profile as a
56
Mn.
54
Mn activity can be determined using the peak at 834 keV.
function of time is shown in Fig. 7. The correction is necessary because of the lower neutron intensities during the ramp-up period of the experiment. Neutron self-absorption is not a unilateral correction applied to the reaction rate but rather a modification of the cross-sections in all 129 energy groups performed by the module SHIELD. Additionally, the effect is only important for the thermal capture reactions in Au and In. Therefore, the magnitude of the neutron self-absorption effect was calculated by comparing the adjusted reaction rates, with and without SHIELD applied. The result was
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Table 2 Reactions used for dosimetry calculations. Reaction
T1/2
56
2.6 312.1 27.7 35.6 70.9 15.0 6.2 2.7 49.5 4.9 54.3
Fe (n, p) 56Mn Fe (n, p) 54Mn 54 Fe (n, α) 51Cr 58 Ni (n, 2n) 57Ni 58 Ni (n, p) 58Co 27 Al (n, α) 24Na 197 Au (n, 2n) 196Au 197 Au (n, γ) 198Au 115 In (n, 2n) 114mIn 115 In (n, n′) 115mIn 115 In (n, γ) 116mIn 54
h day day h day h day day day h min
γ [keV]
Range [MeV]
846 834 320 1377 810 1368 355 411 190 336 1293
13.5 6.7 13.7 14.2 6.0 13.8 13.8 2.3E 08 14.2 1.8 1.0E 08
14.9 14.9 14.9 14.8 14.9 14.8 14.8 1.2 14.7 14.7 2.0
Fig. 6. The neutron energy spectra as a function of angles with 160 kV deuterons incident on a tritium target. The curve for 901 is scaled by a factor of 0.2.
Fig. 5. MCNP model of the D-T generator with respect to its location within the low-scatter facility.
typically less than 0.1% for both the reactions 197Au (n, γ) 198Au and 115 In (n, γ) 116mIn, though in some cases it was as high as 2.7%. Following these corrections, the saturated reaction rates and the flux calculated by MCNP were put into the final module of STAYSL PNNL (the final module is also called STAYSL PNNL). STAYSL PNNL first calculates the theoretical spectrum averaged reaction rate from the 129 group MCNP flux and neutron cross-section libraries from the International Reactor Dosimetry File (IRDF) [13]. Because the default libraries in STAYSL PNNL (one each for cross-section data, crosssection covariances data, and neutron self-shielding data) only include 100 groups with a resolution of just 1 MeV at 14 MeV, it was necessary to reprocess these files to match the 129 group energy structure that was used to calculate the spectrum in MCNP and achieve a resolution of 100 keV from 12.5 to 16 MeV. The code subsequently adjusts the MCNP neutron spectrum in every group by performing a least-squares fit of the experimental reaction rates and the theoretical reaction rates. The least squares method accounts for uncertainty in the MCNP flux groups, uncertainty in the corrected saturated reaction rates, covariances between each reaction rate, covariances between each reaction's cross-sections, and covariances between flux groups. In addition to the adjusted flux spectrum, STAYSL PNNL outputs chi-squared values for the fit and 1 s uncertainties for the total flux as well as in each group.
3. Results Several presentations of the data analysis are discussed. The first section shows the flux spectrum at one of the positions and
also demonstrates how the peak neutron energy varies with a position's angle relative to the source. The next section describes the distribution of energy in the other locations where samples were placed. In the final section, discussion is extended to all the space near the source and how well the experimental data fits with the analytical solution for the flux as a function of space. 3.1. Flux spectra Fig. 8 shows the simulated MCNP spectrum and the adjusted spectrum from STAYSL PNNL (hereby referred to as the measured spectrum) for Position 1. The spectra are plotted as differential flux times the mid-point energy of the group versus energy. The magnitude of the adjustment is plotted on the secondary axis. Most of the adjustment occurs in the epithermal region, where the uncertainties in the MCNP flux calculation are high anyway, about 10%–25%. The adjustment at 14.5 MeV, where the differential flux is a maximum of 2.24Eþ08 n/MeV cm2 s (71.7%), was only 0.15% because the MCNP calculation is well known here. This can be seen in Fig. 9 where the same data is plotted on linear scales around the peak at 14.5 MeV. The simulated and measured spectra are indistinguishable on this scale. The total flux for Position 1 is 7.58E þ07 n/cm2 s ( 71.4%). The shape of the spectrum and the region of maximum adjustment is similar for the remaining positions (plots not shown), the main difference being the flux intensity in each position. Since the generator produces neutrons as a disc source and there is some degree of anisotropy, it was important to run STAYSL PNNL with at least 100 keV resolution in the region around 14 MeV (as opposed to the original STAYSL PNNL libraries which have 1 MeV resolution at 14 MeV). The benefit of finer resolution is evident by the nice peak shape in Fig. 9. Additionally, 100 keV resolution is necessary in order to analyze the spatial dependence of the peak neutron energy like the example shown in Fig. 10. The average energy differential flux times are plotted for Positions 1, 12, and 18 which represent samples at 3 different angles relative to
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Table 3 Correction factors used to determine saturated reaction rates. Reaction
56 54
Fe (n, p) Fe (n, p)
Atomic weight of target (g/mol)
56 54
Mn Mn
In (n, γ)
115
116m
In
Gamma absorption in wire
0.9801 0.9800 0.9678
0.9665 0.9663 0.9463
58.693 58.693 26.982
0.680769 0.680769 1.0
0.067842 0.0014663 0.15451
0.9941 0.9923 0.9946
0.9695 0.9604 0.9906
196.967 196.967
1.0 1.0
0.016727 0.037882
0.9868 0.9898
NA NA
114.818 114.818 114.818
0.9571 0.9571 0.9571
0.0020979 0.43356 0.98798
0.9838 0.9935 0.9977
NA NA NA
27
Au (n, γ) 198Au 115 In (n, 2n) 114mIn 115 In (n, n′) 115mIn
Gamma absorption in foil
0.63641 0.00033306 0.0037466
Fe (n, α) 51Cr Ni (n, 2n) 57Ni 58 Ni (n, p) 58Co
197
Irradiation history 0.91754 0.05845 0.05845
58
Al (n, α) 24Na Au (n, 2n) 196Au
Correction factors (unitless)
55.845 55.845 55.845
54
197
Isotopic abundance of target (mol fraction)
Fig. 7. The irradiation history for the experiment is plotted as relative beam current versus time.
the neutron source. The center of the peak appears to be dependent on the angle θ (see Fig. 2). The peak energy is greatest for an angle of 2.61 (Position 18) at 14.8 MeV, decreases to 14.5 MeV at 45.71 (Position 1), and is a minimum of 14.1 MeV at 85.81 (Position 12). The decrease in peak energy at the sample position with increasing polar angle θ closely matches the source energies shown in Fig. 6, suggesting that most of the neutrons that reach the sample do not undergo any collisions. 3.2. Energy distribution The experiment also validates that the D-T source is very nearly “pure” 14 MeV neutrons and that samples see mostly fast neutrons. In all 21 positions the flux above 110 keV relative to the total flux is at least 99.1%. The relative group flux is shown in Table 4 for Positions 15, 1, and 5, which generally lie along the line at approximately 451 relative to the disc source (although their centers given by θ in Table 1 vary). For these positions, the peak in the spectrum occurs at 14.5 MeV. Table 4 shows that there is
appreciable scattering of neutrons to lower energies as distance increases but that thermal energies are still negligible. This is to be expected as neutrons undergo more scattering before reaching samples further away. Positions extending radially from the source, such as Positions 9–12, show increased scatter compared to off-radial positions. The peak in each of these positions occurs at 14.1 MeV. Table 5 shows that neutrons scatter out of the 14 MeV range more often than for locations at shallower angles. This is likely caused by the anisotropy of the D-T reaction that emits the source neutron. The fact that neutrons must pass through more water present in the cooling cap when emitted in the radial direction also contributes to this effect. In order to minimize scattering in the neutron spectrum, samples should be placed as close as possible to the source. This position would be on the center of the cooling cap surface in Position 15, where 91.4% of the neutrons have energies between 14 and 15 MeV. The maximum total flux is also available at Position 15 with a measured value of 8.29E þ08 n/cm2 s (71.4%). However,
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Fig. 8. The simulated and measured flux spectra for Position 1. Flux is plotted on the left axis, and percent difference between the two on the right.
Fig. 9. The two spectra for Position 1, zoomed in on the peak at 14.5 MeV, are indistinguishable.
additional experiments have shown that the total flux is even larger if the sample lies flat on the center of the cap, suggesting that there is a steep axial flux gradient near the source. The same experimental method and data analysis applied for flat foils in this
position produced a total flux of 1.20E þ09 n/cm2 s (71.9%) with a group flux of 1.10E þ09 n/cm2 s (72.1%) in the 14–15 MeV range. This represents 91.2% for this component which is still similar to the 14–15 MeV fraction in Position 15.
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Fig. 10. Peak fluxes in three samples at different angles.
Table 4 Flux distribution for three positions at varying distances and along the line of 451 relative to the source. Energy [MeV]
Position, d (cm)
0–0.11 0.11–1 1–10 10–14 14–15
15 (1.2)
1 (5.8)
5 (11.8)
0.3% 1.6% 5.0% 1.6% 91.4%
0.7% 2.8% 7.7% 2.4% 86.5%
0.8% 3.3% 8.8% 2.7% 84.4%
Table 5 Flux distribution for four radial positions. Energy [MeV]
0–0.11 0.11–1 1–10 10–14 14–15
Position, d (cm) 9 (4.8)
10 (5.8)
11 (7.2)
12 (8.8)
0.7% 3.1% 9.2% 4.3% 82.7%
0.7% 3.4% 9.8% 4.9% 81.2%
0.8% 3.6% 10.8% 5.5% 79.3%
0.9% 4.0% 11.6% 6.1% 77.5%
A summary of the total flux and several group fluxes for every position is given in Table 6. The second value listed in each group in the table is the 1 s error in %. The relative error can be as high as a factor of 2 at thermal energies because of the high uncertainty in the MCNP calculation and the lower reaction rates at these energies during the actual experiment. The total and fast fluxes, however, are reported with only a 1%–2% error.
3.3. Uncollided flux As described earlier and shown in Fig. 2, the experiment can be approximated by a model of a disc source and a sample position at point P given by the coordinates (a, h). Additionally, if R is the radius of the disc source and Sa is the source intensity averaged over the area of the disc, the uncollided flux, Φun, at point P is given by the analytical equation [14] qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi3 2 2 2 2 2 2 4 2 2 Sa 4a þR h þ R þ 2R ða2 h Þ þ ða2 þ h Þ 5 ð1Þ Φun ¼ ln 4 2a2 Eq. (1) assumes the source is isotropic and there is no attenuation. This is a first order approximation of the experiment since the real system is anisotropic and some attenuation between the D-T source and the sample exists because of the water and stainless steel in the cooling cap. Nevertheless, the only means of achieving an easy analytical solution for all space like Eq. (1) is by ignoring these effects. Since the peak energy of the source neutron is anywhere from 14.1 to 14.8 MeV, the 14–15 MeV bin of the flux determined by STAYSL PNNL is assumed to represent uncollided neutrons arriving at the sample. Thus, the flux in this bin can be compared to the flux predicted by Eq. (1). Fig. 11 shows the result when the 14–15 MeV bin of the 21 experimental points (see Table 6) are superimposed onto the surface Φun. The value for R was taken to be 1.65 cm, the measured radius of the vacuum tube in the generator where deuterons are accelerated, which is based on the assumption that neutrons are produced uniformly in the tritium target. The value for Sa was then determined to be 3.14E þ09 n/s by normalizing the 21 experimental 14 MeV fluxes to the value for phi (if Sa were equal to 1) at that coordinate, and taking an average value. Multiplying Sa by πR2 yields the total source intensity of 2.69E þ 10 n/s. This is
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Table 6 Group fluxes and relative error for every position. Position
Total flux
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21
7.58E þ 07 6.89E þ 07 8.55E þ 07 7.28E þ 07 2.14E þ 07 1.94E þ 07 2.07Eþ07 1.97Eþ07 1.12E þ08 7.16E þ 07 4.73E þ 07 2.69E þ 07 1.23E þ 08 5.15E þ 07 8.29E þ 08 3.72E þ 08 1.94E þ 08 1.02E þ 08 5.02E þ 07 3.14E þ 07 1.62E þ 07
1.4 1.3 2.4 1.4 1.4 1.5 1.4 1.4 1.6 1.6 1.5 1.7 1.4 1.6 1.4 1.7 1.2 1.3 1.8 1.7 1.7
Thermal 1E 4 eV to 0.55 eV
Epithermal 0.55 eV to 110 keV
1.56E þ04 1.43E þ04 1.03E þ04 1.01E þ 04 9.72Eþ03 9.32E þ 03 1.26E þ04 1.16E þ04 1.38E þ04 7.58E þ03 1.18E þ 04 9.70Eþ 03 1.17E þ 04 1.06E þ04 1.70Eþ 04 1.57E þ 04 1.41E þ 04 1.07Eþ 04 7.71E þ 03 8.98E þ 03 5.48E þ03
5.09E þ05 4.33Eþ 05 5.12E þ 05 4.31E þ05 1.58E þ 05 1.36E þ 05 1.63E þ 05 1.49E þ 05 7.88E þ 05 4.88E þ05 3.70Eþ 05 2.21E þ05 7.58E þ 05 3.95E þ05 2.68E þ06 1.62E þ 06 9.63E þ05 5.55E þ05 3.11E þ 05 1.81E þ 05 1.04E þ 05
101.7 97.4 121.6 120.2 124.6 127.5 59.2 88.1 125.2 129.2 115.7 126.4 115.9 128.5 110.9 111.3 111.6 107.6 118.6 119.7 128.3
Fast 110 keV to 15 MeV 14.7 14.4 12.4 12.3 12.7 12.8 14.7 16.2 12.4 12.6 12.7 12.3 12.5 12.9 12.3 12.5 12.4 12.8 12.2 12.7 12.5
7.53E þ07 6.85E þ07 8.49E þ07 7.23E þ07 2.12E þ 07 1.92E þ07 2.05E þ07 1.95E þ07 1.12E þ 08 7.11E þ 07 4.69E þ07 2.67Eþ 07 1.22E þ08 5.11E þ07 8.27E þ 08 3.70Eþ 08 1.93E þ08 1.01E þ 08 4.99E þ07 3.12E þ 07 1.61E þ 07
Fast 14–15 MeV 1.4 1.3 2.4 1.4 1.4 1.5 1.4 1.4 1.6 1.6 1.5 1.7 1.4 1.6 1.4 1.7 1.2 1.3 1.8 1.7 1.7
6.55E þ07 6.01E þ 07 7.40E þ 07 6.35Eþ 07 1.81E þ 07 1.65E þ 07 1.75E þ 07 1.66E þ 07 9.30E þ07 5.81E þ07 3.75E þ07 2.08E þ07 1.02E þ 08 4.13E þ07 7.58E þ 08 3.35Eþ 08 1.73E þ08 8.94E þ07 4.35Eþ 07 2.71E þ 07 1.40E þ 07
1.4 1.1 2.7 1.4 1.3 1.5 1.2 1.2 1.6 1.6 1.5 1.8 1.4 1.7 1.5 1.9 1.2 1.3 2.0 1.9 1.9
placing samples on the center of the cooling cap. A generalized formula for the uncollided flux has been extended to all space near the D-T source.
Acknowledgments
Fig. 11. 14 MeV flux with experimental data points superimposed onto the surface Φ. Dark blue points lie above the surface, light blue below. For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.
34% higher than the value of 2E þ 10 n/s reported by the vendor [15]. Experimental data still fits the shape of the curve in Fig. 11 quite well. Absolute differences between the measured flux and the analytical flux range from 1.6% to 27%. In some cases the surface is an overestimate of the measured value by as much as 27%, and in other cases it is an underestimate by as much as 24%. These differences are likely due to the fact that this is a first-order approximation neglecting anisotropy and attenuation. Further investigation including a least-squares fit of R and Sa to the data is required in order to resolve the value for source intensity. It may be that there is an effective radius where the D-T reaction occurs.
4. Conclusions The neutron flux has been well characterized in the space near the source of the D-T generator located in a low-scatter facility at PNNL. The STAYSL PNNL code was used to adjust MCNP spectra based on measured neutron activation rates in dosimetry foils and appropriate corrections. In all 21 positions studied, the peak energy lies in the range of 14.1–14.8 MeV and the intensity of this peak is 7 orders of magnitude larger than the intensity at thermal energies. At least 99.1% of the spectrum is comprised of fast neutrons, and one can minimize the scatter in the spectrum by
The research described in this paper was conducted under the Laboratory Directed Research and Development Program at Pacific Northwest National Laboratory, a Multiprogram National Laboratory operated by Battelle for the U.S. Department of Energy under contract DE-AC05-76RL01830. The authors would like to thank Shannon Morley for her analysis in GENIE 2000, Laura Thierolf for editing and formatting, and Zachary Hilliard for making some of the figures. This document is PNNL-SA-99038.
Appendix A. Supporting information Supplementary data associated with this article can be found in the online version at http://dx.doi.org/10.1016/j.nima.2013.11.023. References [1] J.C. Robertson, K.J. Zieba, Nuclear Instruments and Methods 45 (1) (1966) 179. [2] S. Shirato, S. Shibuya, Y. Ando, T. Kokubu, K. Hata, Nuclear Instruments and Methods in Physics Research A 278 (2) (1989) 477. [3] N.E. Hertel, B.W. Wehring, Nuclear Instruments and Methods 172 (3) (1979) 501. [4] P.M. Kehayias, J.J. Keyahias, Nuclear Instruments and Methods in Physics Research B 261 (1–2) (2007) 827. [5] S.A. Jonah, K. Ibikunle, Nuclear Instruments and Methods in Physics Research A 501 (2–3) (2003) 514. [6] X-5 Monte Carlo Team, MCNP—A General Monte Carlo N-Particle Transport Code, Version 5, vol. I: Overview and Theory, LA-UR-03-1987, Los Alamos National Laboratory, Los Alamos, New Mexico, 2008. [7] D. Pelowitz (Ed.) , MCNPX User's Manual Version 2.7.0, LA-CP-11-00438, Los Alamos National Laboratory, Los Alamos, New Mexico, 2011. [8] Canberra Industries, Inc., GENIE 2000 Spectroscopy Software: Operations. 9233652G V3.2, 2009. [9] National Nuclear Data Center, Brookhaven National Laboratory. Information extracted from the NuDat 2 database. Available from: 〈http://www.nndc.bnl. gov/nudat2〉 (accessed 2013).
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[10] R. Traub, Calculation of Ambient (H*(10)) and Personal (Hp(10)) Dose Equivalent From a 252Cf Neutron Source. PNNL-19273, Pacific Northwest National Laboratory, Richland, Washington, 2010. [11] L.R. Greenwood, D.G. Doran, H.L. Heinisch, Physical Review C 35 (1) (1987). [12] L.R. Greenwood, C.D. Johnson, User Guide for the STAYSL PNNL Suite of Software Tools. PNNL-22253, Pacific Northwest National Laboratory, Richland, Washington, 2013. [13] IAEA, International Reactor Dosimetry File 2002 (IRDF-2002), Technical Reports Series No. 452, International Atomic Energy Agency, Vienna, Austria, 2006.
[14] O.J. Wallace. Semi-Analytic Flux Formulas for Shielding Calculations. WAPDTM-1197, Bettis Atomic Power Laboratory, West Mifflin, Pennsylvania, 1976, p. 75. [15] Thermo Scientific D 711 Neutron Generators. Available from: 〈http://www. thermoscientific.com/ecomm/servlet/productsdetail_11152___11962782_-1〉 (accessed 01.05.13).
Contents lists available at ScienceDirect
Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
Characterization of a Thermo Scientific D711 D-T neutron generator located in a low-scatter facility John W. Hayes 1, Erin Finn n, Larry Greenwood 2, Rick Wittman 3 Pacific Northwest National Laboratory, PO Box 999, Richland, WA 99352, USA
art ic l e i nf o
a b s t r a c t
Article history: Received 2 July 2013 Received in revised form 4 November 2013 Accepted 6 November 2013 Available online 28 November 2013
A dosimetry experiment used to measure the neutron flux and spectrum of a D-T neutron generator is presented. The D-T generator at Pacific Northwest National Laboratory is installed in the middle of a large room to minimize scatter of neutrons back to the sample. The efficacy of maintaining a pure fast neutron field for the sample is investigated. Twenty-one positions within 13 cm of the neutron source contained foils or wires of Fe, Ni, and Al with additional Au, and in monitors at some locations. Spectral adjustment of the neutron flux at each position based on the measured reaction rates and theoretical Monte Carlo calculations show that at least 99.1% of the spectrum lies above 110 keV for all measured positions, and neutrons above 14 MeV can account for as much as 91% at locations along the axis of the generator and close to the source. The 14 MeV component drops to 77% in radial positions far from the source. The largest total flux observed was 8.29E þ08 n/cm2 s (7 1.4%) in the center of the cooling cap, although additional experiments have shown this value could be as high as 1.20E þ09 n/cm2 s. & 2014 Published by Elsevier B.V.
Keywords: 14 MeV D-T generator Neutron spectral adjustment MCNP STAYSL PNNL
1. Introduction Pacific Northwest National Laboratory (PNNL) purchased and installed a D711 model D-T neutron generator (“generator”) from Thermo Scientific in August 2011. The generator nominally produces 14 MeV neutrons isotropically as a product of the fusion event from an accelerated deuteron on a tritium target. Fast neutron sources are important in matters of national security and nuclear data research for several reasons. First, fast neutrons provide the capability of harnessing threshold reactions for the production of rare isotopes, which are of interest to radiochemistry groups at PNNL concerned with validating radioanalytical techniques for the separation and characterization of these isotopes. Fast fission reactions on 235U, 238U, and 239Pu using the D-T source are also capable of producing sufficiently high yields of rare fission product isotopes in order to further develop these techniques. Research in this area has progressed significantly by having the capability to produce rare isotopes, which may not be produced by other means or are prohibitively expensive or logistically difficult to acquire from sources outside of PNNL such as nuclear reactors or other accelerators. Experiments with 14 MeV neutrons are also of interest because
n
Corresponding author. Tel.: þ 1 509375-7374. E-mail addresses: [email protected] (J.W. Hayes), [email protected] (E. Finn), [email protected] (L. Greenwood), [email protected] (R. Wittman). 1 Permanent address: 891 Bridge Walk Ct, Ada, MI 49301, USA, Tel.: þ 1 616 217-7245. 2 Tel.: þ 1 509 375-5301. 3 Tel.: þ 1 509 375-5111. 0168-9002/$ - see front matter & 2014 Published by Elsevier B.V. http://dx.doi.org/10.1016/j.nima.2013.11.023
nuclear data for fast fission has not been studied as extensively as it has been for thermal fission. Groups at PNNL have already performed experiments studying 14 MeV fission of 235U with the intent to determine the fission product yield of short-lived isotopes more precisely. Research applications of the D-T generator require that the neutron spectrum be as “pure” as possible near 14 MeV. The presence of a significant slow neutron component in the spectrum introduces larger uncertainties in the measurement of 14 MeV fission yields because slow neutrons contribute to the production of the fission product isotopes and one must correct for the slow component when calculating the 14 MeV production rate. This is especially a problem if the spectrum has a significant thermal or epithermal neutron component since the fission cross-sections are much higher at these energies than at 14 MeV. Therefore, installation of the generator was designed to minimize the slow component, which is primarily due to scattering of the source neutrons back to the generator head. By housing the generator in the center of a low-scatter room where the closest wall is more than 4 m away, the room return effect was diminished to the extent possible. The initial experiments performed with the generator were intended to characterize the neutron spectrum and intensity so that the proportion of the slow component to the total flux could be measured.
2. Materials and methods The first section includes a discussion of prior attempts at characterizing the flux at similar facilities as well as alternative
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methods to foil irradiation considered at this one. In the next sections, the details of the experiment are presented, followed by an explanation of the analysis methods used. Dosimetry measurements of the real flux environment are used to validate theoretical calculations and adjust flux spectra within associated uncertainties and based on limiting fitting parameters. 2.1. Previous characterization attempts Indirect methods of measuring the flux from similar D-T neutron generators by the associated alpha-particle method have been performed by Robertson et al. [1], Shirato et al. [2], and Hertel et al [3]. The method in these 3 papers however can only provide total flux results and not spectral information. Kehayias et al. [4] determined the total flux for an A-325 model D-T generator using silicon foil activation but again provided no spectral information. Jonah et al. [5] used foil activation to determine the complete energy spectrum for an A-711 KAMAN neutron generator with 106 n/s output. Irradiation of 8 pure metals and the unfolding of the spectrum by the SULSA computer code closely matches the method presented here. SULSA is an unfolding code based on a least squares method of measured reaction rates only, while STAYSL PNNL is an adjustment code that uses reaction rates and an input spectrum. Other methods of measuring the neutron flux directly, such as the NE213 liquid scintillator used by Shirato and Hertel, or a Bonner spheres method, can provide confidence to the total neutron yield determined in this paper, but have not been performed at this facility to date. These methods cannot determine flux as a function of space around the source which was also a goal of this experiment. Therefore, foil irradiation provided the most effective way to determine the flux spectra at many points near the source. 2.2. Sample irradiation In this experiment, an irradiation of dosimetry foils and wires was conducted with the aim of characterizing the neutron spectrum at 21 points in a hemispherical array near the generator head. An aluminum apparatus was designed to fit on the cooling cap and act as a sample holder for the space in a hemisphere of radius 13 cm above the center of the D-T reaction. The D-T source that produces a fast neutron is assumed to take place in the shape of a circular disc just below the top of the cooling cap. The apparatus is shown in Fig. 1 in its experimental configuration on the generator. The relation of the generator to the walls is also implied in the figure: the distances to the side walls are 5.4 m and 6.9 m. The D-T source is 4.3 m above the floor, which is also the closest boundary of the room to the generator. Pure iron, aluminum, nickel, gold, and indium metal foils or wires were heat sealed in tightly conformed plastic to make 21 sample packets. The sample packets were affixed to various positions on the apparatus. Fig. 2 shows a schematic of the locations of each of the sample packets relative to the disc source. This schematic comes from the Monte Carlo N-Particle (MCNP) model [6,7] used for analysis described in Section 2.2. The positions are labeled 1–21 for easier reference. Packets 1, 2, 7, and 8 contained 0.5 in. diameter foils of Fe (31 mils thickness), Ni (10 mils), Al (30mils), Au (2 mils), and In (5 mils). Packets 3 and 4 only included Fe, Ni, and Al foils, but of the same diameter and thickness. Packets 5 and 6 also only included Fe, Ni, and Al foils, but the Fe foil was 50 mils thick so that reasonable activities could still be achieved. Packets 9–21 contained Fe, Ni, and Al wires of 62 mil diameter and lengths ranging from 1.0 to 2.5 cm. Each of the metal foils and wires was carefully weighed for purposes of gamma spectroscopy following irradiation.
Fig. 1. The aluminum sample holder placed on the cooling cap of the D-T generator.
Fig. 2 also shows a coordinate system to describe any point in space relative to the center of the disc source. A point P is given by coordinates (a, h), where a is the vertical distance perpendicular to the plane of the disc source and h is the horizontal offset from the disc center. Later analysis and discussion of the spatial flux will take advantage of this coordinate system. For each position analyzed during this experiment, the point P is taken to be the geometric center of the space occupied by the sample. The coordinates of the position numbers, the direct distance d to the center of the disc source, and the angle θ defined by the vectors a and d are given in Table 1. The sample packets were irradiated for 3 h and 36 min with a ramp to full power (160 kV and 3.0 mA) of 34 min and 40 s. The irradiated samples were then gamma counted with high-purity germanium (HPGe) detectors and gamma spectra were acquired in the software program GENIE 2000 offered by Canberra [8]. The detectors are calibrated with National Institute of Standards and Technology; traceable standards and control counts for the detectors are performed daily, while background counts are performed at least weekly. Gamma spectroscopy analysis was performed using the GENIE 2000 software to determine the product activity decay-corrected to the end of irradiation. Nuclear data for analysis was taken from the National Nuclear Data Center at Brookhaven National Laboratory [9]. Figs. 3 and 4 show examples of spectra acquired in GENIE from the Fe wire in Position 1. The activation products 56Mn and 54Mn are created from (n, p) reactions on 56Fe and 54Fe, respectively. Fig. 3 is an immediate count, therefore the activity of 56Mn (2.6 h half-life) dominates the spectrum as evident by the peak at 846 keV. Fig. 4 shows that when the sample was recounted the following day, the 56Mn activity had decayed away appreciably. The ability to resolve the peak at 834 keV from 54Mn then allowed for a determination of its activity. As described in Section 2.3, the 56Mn and 54Mn activities (as well as all of the sample product activities) are then used as input into the STAYSL PNNL code. The complete list of reactions studied and the gamma lines from spectra acquired in GENIE used to calculate the product activity are summarized in Table 2. The table also provides a range of energies that were expected to account for 90% of the reaction rate based on preliminary calculations. The majority are purely fast reactions leading to high resolution in the region of the spectrum
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Fig. 2. MCNP model of the sample apparatus and experimental configuration for 21 sample positions. The two views are shown in the XZ (left) and YZ (right) planes.
Table 1 Sample positions relative to neutron source in centimeters. Position
a
h
d
θ
Position
a
h
d
θ
1 2 3 4 5 6 7 8 9 10 11
4.0 5.8 4.0 5.8 7.8 11.5 7.8 11.5 0.6 0.6 0.6
4.1 2.4 4.1 2.4 8.9 5.1 8.9 5.1 4.8 5.8 7.2
5.8 6.3 5.8 6.3 11.8 12.6 11.8 12.6 4.8 5.8 7.2
45.7 22.4 45.7 22.4 48.8 24.0 48.8 24.0 82.4 83.7 84.9
12 13 14 15 16 17 18 19 20 21 –
0.6 0.6 0.6 1.1 2.2 3.2 4.5 6.9 8.4 10.7 –
8.8 4.8 6.8 0.2 0.2 0.2 0.2 0.7 0.7 0.7 –
8.8 4.8 6.8 1.2 2.2 3.3 4.6 6.9 8.5 10.8 –
85.8 82.3 84.6 10.4 5.5 3.7 2.6 6.0 4.9 3.8 –
near 14 MeV. In locations that included Au and In foils, better resolution of the thermal region could also be achieved because of the capture reactions that extend to thermal energies. Both 58Co and 196Au have isomers with half-lives of about 9 h. Additional gamma counting was performed in both cases such that the reaction rates reported for both of these reactions include the complete in-growth of the isomeric states as well as direct production of the ground states.
neutron was tallied among 200 directional cosine bins from 01 to 1801 (0.01 rad¼ 0.5731 / bin.) The neutron energy was also tallied among 100 energy bins within each cosine bin. The energy bin boundaries varied depending on the direction but remained within several hundred kiloelectron volt of the average energy in order to obtain high resolution of the peak in that cosine bin. The neutron energy spectra as a function of angle are shown in Fig. 6. The total yield at each angle is normalized to 1. Neutrons are emitted with the greatest energy in the forward direction (θ ¼0) and are emitted with lower energies as the polar angle increases. The validity of this modeling method was checked by demonstrating that when 360 kV deuterons are used it gives consistent results to the neutron energy spectra calculated by Greenwood et al [11] for a beam of 360 kV deuterons incident on a tritium target. The neutron energy spectra with angular dependence from this initial model were then used in the full model5 of the generator present in the low-scatter room. The spatial distribution of the neutrons was modeled as a disc source of radius 1.65 cm near the top of the cooling cap (Fig. 2). The flux in each sample was calculated with an f4 tally in 129 groups including 100 keV bins from 12.5 to 16 MeV in order to achieve high resolution of the peak near 14 MeV. The spectra calculated by MCNP for each position could then be used as the first approximation for input into the STAYSL PNNL code described in the next section.
2.3. Monte Carlo calculation 2.4. Spectral adjustment with STAYSL PNNL The MCNP code was used to model the experimental setup and calculate the neutron spectrum in all 21 locations in 129 energy groups. The model was adapted from a previous model of the lowscatter room developed by PNNL to calculate dose from a 252Cf source also used in the facility [10]. A separate model of the generator received from Thermo Scientific was updated with asbuilt dimensions and then inserted into the low-scatter room model in its location in the center of the room (Fig. 5). The model of the neutron source in MCNP incorporated a sophisticated energy-angle dependent source representative of the kinematics of the D-T fusion event. First, MCNP4 was used to calculate the neutron energy spectrum in a model in which a deuterium particle was accelerated at 160 kV onto a tritium target adsorbed in a zinc hydride matrix inside a vacuum. The product 4
Monte Carlo N-Particle eXtended (MCNPX) was used for this calculation.
The STAYSL PNNL code [12] is comprised of 6 code modules designed to yield adjusted neutron flux spectra based on measured neutron activation rates and theoretical flux calculations as input. First the code was used to calculate saturated reaction rates by taking the activities of the irradiated samples calculated in GENIE and correcting for atomic weight, isotopic abundance, gamma self-absorption, decay during irradiation, and neutron self-absorption (neutron burnup of the target or product atoms was considered negligible due to the short time of irradiation). The first 3 corrections were calculated in the module SigPhi Calculator and are shown in Table 3. Gamma self-absorption corrections, which account for photon energy and sample geometry, averaged about 1% in the foils and about 3% in the wires. 5
MCNP5 was used in this test.
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Fig. 3. GENIE gamma spectrum for Fe wire in Position 1. The 846 keV peak is used to calculate the activity of
Fig. 4. GENIE gamma spectrum for Fe wire in Position 1 recounted 20 h after the end of irradiation. The
The decay during irradiation correction, which is also shown in Table 3, was determined in the module BCF, or Beam Correction Factor code, by calculating the production and decay of each activation product over the entire irradiation history. The control console of the generator outputs the measured values of high voltage (HV) and beam current (B) to a log file every 10 s so that a model of the flux can be fit based on these values following the experiment. Because the dependence of flux on the operating conditions was not known, the initial model assumed flux was linear with B but independent of HV. The beam profile as a
56
Mn.
54
Mn activity can be determined using the peak at 834 keV.
function of time is shown in Fig. 7. The correction is necessary because of the lower neutron intensities during the ramp-up period of the experiment. Neutron self-absorption is not a unilateral correction applied to the reaction rate but rather a modification of the cross-sections in all 129 energy groups performed by the module SHIELD. Additionally, the effect is only important for the thermal capture reactions in Au and In. Therefore, the magnitude of the neutron self-absorption effect was calculated by comparing the adjusted reaction rates, with and without SHIELD applied. The result was
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61
Table 2 Reactions used for dosimetry calculations. Reaction
T1/2
56
2.6 312.1 27.7 35.6 70.9 15.0 6.2 2.7 49.5 4.9 54.3
Fe (n, p) 56Mn Fe (n, p) 54Mn 54 Fe (n, α) 51Cr 58 Ni (n, 2n) 57Ni 58 Ni (n, p) 58Co 27 Al (n, α) 24Na 197 Au (n, 2n) 196Au 197 Au (n, γ) 198Au 115 In (n, 2n) 114mIn 115 In (n, n′) 115mIn 115 In (n, γ) 116mIn 54
h day day h day h day day day h min
γ [keV]
Range [MeV]
846 834 320 1377 810 1368 355 411 190 336 1293
13.5 6.7 13.7 14.2 6.0 13.8 13.8 2.3E 08 14.2 1.8 1.0E 08
14.9 14.9 14.9 14.8 14.9 14.8 14.8 1.2 14.7 14.7 2.0
Fig. 6. The neutron energy spectra as a function of angles with 160 kV deuterons incident on a tritium target. The curve for 901 is scaled by a factor of 0.2.
Fig. 5. MCNP model of the D-T generator with respect to its location within the low-scatter facility.
typically less than 0.1% for both the reactions 197Au (n, γ) 198Au and 115 In (n, γ) 116mIn, though in some cases it was as high as 2.7%. Following these corrections, the saturated reaction rates and the flux calculated by MCNP were put into the final module of STAYSL PNNL (the final module is also called STAYSL PNNL). STAYSL PNNL first calculates the theoretical spectrum averaged reaction rate from the 129 group MCNP flux and neutron cross-section libraries from the International Reactor Dosimetry File (IRDF) [13]. Because the default libraries in STAYSL PNNL (one each for cross-section data, crosssection covariances data, and neutron self-shielding data) only include 100 groups with a resolution of just 1 MeV at 14 MeV, it was necessary to reprocess these files to match the 129 group energy structure that was used to calculate the spectrum in MCNP and achieve a resolution of 100 keV from 12.5 to 16 MeV. The code subsequently adjusts the MCNP neutron spectrum in every group by performing a least-squares fit of the experimental reaction rates and the theoretical reaction rates. The least squares method accounts for uncertainty in the MCNP flux groups, uncertainty in the corrected saturated reaction rates, covariances between each reaction rate, covariances between each reaction's cross-sections, and covariances between flux groups. In addition to the adjusted flux spectrum, STAYSL PNNL outputs chi-squared values for the fit and 1 s uncertainties for the total flux as well as in each group.
3. Results Several presentations of the data analysis are discussed. The first section shows the flux spectrum at one of the positions and
also demonstrates how the peak neutron energy varies with a position's angle relative to the source. The next section describes the distribution of energy in the other locations where samples were placed. In the final section, discussion is extended to all the space near the source and how well the experimental data fits with the analytical solution for the flux as a function of space. 3.1. Flux spectra Fig. 8 shows the simulated MCNP spectrum and the adjusted spectrum from STAYSL PNNL (hereby referred to as the measured spectrum) for Position 1. The spectra are plotted as differential flux times the mid-point energy of the group versus energy. The magnitude of the adjustment is plotted on the secondary axis. Most of the adjustment occurs in the epithermal region, where the uncertainties in the MCNP flux calculation are high anyway, about 10%–25%. The adjustment at 14.5 MeV, where the differential flux is a maximum of 2.24Eþ08 n/MeV cm2 s (71.7%), was only 0.15% because the MCNP calculation is well known here. This can be seen in Fig. 9 where the same data is plotted on linear scales around the peak at 14.5 MeV. The simulated and measured spectra are indistinguishable on this scale. The total flux for Position 1 is 7.58E þ07 n/cm2 s ( 71.4%). The shape of the spectrum and the region of maximum adjustment is similar for the remaining positions (plots not shown), the main difference being the flux intensity in each position. Since the generator produces neutrons as a disc source and there is some degree of anisotropy, it was important to run STAYSL PNNL with at least 100 keV resolution in the region around 14 MeV (as opposed to the original STAYSL PNNL libraries which have 1 MeV resolution at 14 MeV). The benefit of finer resolution is evident by the nice peak shape in Fig. 9. Additionally, 100 keV resolution is necessary in order to analyze the spatial dependence of the peak neutron energy like the example shown in Fig. 10. The average energy differential flux times are plotted for Positions 1, 12, and 18 which represent samples at 3 different angles relative to
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Table 3 Correction factors used to determine saturated reaction rates. Reaction
56 54
Fe (n, p) Fe (n, p)
Atomic weight of target (g/mol)
56 54
Mn Mn
In (n, γ)
115
116m
In
Gamma absorption in wire
0.9801 0.9800 0.9678
0.9665 0.9663 0.9463
58.693 58.693 26.982
0.680769 0.680769 1.0
0.067842 0.0014663 0.15451
0.9941 0.9923 0.9946
0.9695 0.9604 0.9906
196.967 196.967
1.0 1.0
0.016727 0.037882
0.9868 0.9898
NA NA
114.818 114.818 114.818
0.9571 0.9571 0.9571
0.0020979 0.43356 0.98798
0.9838 0.9935 0.9977
NA NA NA
27
Au (n, γ) 198Au 115 In (n, 2n) 114mIn 115 In (n, n′) 115mIn
Gamma absorption in foil
0.63641 0.00033306 0.0037466
Fe (n, α) 51Cr Ni (n, 2n) 57Ni 58 Ni (n, p) 58Co
197
Irradiation history 0.91754 0.05845 0.05845
58
Al (n, α) 24Na Au (n, 2n) 196Au
Correction factors (unitless)
55.845 55.845 55.845
54
197
Isotopic abundance of target (mol fraction)
Fig. 7. The irradiation history for the experiment is plotted as relative beam current versus time.
the neutron source. The center of the peak appears to be dependent on the angle θ (see Fig. 2). The peak energy is greatest for an angle of 2.61 (Position 18) at 14.8 MeV, decreases to 14.5 MeV at 45.71 (Position 1), and is a minimum of 14.1 MeV at 85.81 (Position 12). The decrease in peak energy at the sample position with increasing polar angle θ closely matches the source energies shown in Fig. 6, suggesting that most of the neutrons that reach the sample do not undergo any collisions. 3.2. Energy distribution The experiment also validates that the D-T source is very nearly “pure” 14 MeV neutrons and that samples see mostly fast neutrons. In all 21 positions the flux above 110 keV relative to the total flux is at least 99.1%. The relative group flux is shown in Table 4 for Positions 15, 1, and 5, which generally lie along the line at approximately 451 relative to the disc source (although their centers given by θ in Table 1 vary). For these positions, the peak in the spectrum occurs at 14.5 MeV. Table 4 shows that there is
appreciable scattering of neutrons to lower energies as distance increases but that thermal energies are still negligible. This is to be expected as neutrons undergo more scattering before reaching samples further away. Positions extending radially from the source, such as Positions 9–12, show increased scatter compared to off-radial positions. The peak in each of these positions occurs at 14.1 MeV. Table 5 shows that neutrons scatter out of the 14 MeV range more often than for locations at shallower angles. This is likely caused by the anisotropy of the D-T reaction that emits the source neutron. The fact that neutrons must pass through more water present in the cooling cap when emitted in the radial direction also contributes to this effect. In order to minimize scattering in the neutron spectrum, samples should be placed as close as possible to the source. This position would be on the center of the cooling cap surface in Position 15, where 91.4% of the neutrons have energies between 14 and 15 MeV. The maximum total flux is also available at Position 15 with a measured value of 8.29E þ08 n/cm2 s (71.4%). However,
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Fig. 8. The simulated and measured flux spectra for Position 1. Flux is plotted on the left axis, and percent difference between the two on the right.
Fig. 9. The two spectra for Position 1, zoomed in on the peak at 14.5 MeV, are indistinguishable.
additional experiments have shown that the total flux is even larger if the sample lies flat on the center of the cap, suggesting that there is a steep axial flux gradient near the source. The same experimental method and data analysis applied for flat foils in this
position produced a total flux of 1.20E þ09 n/cm2 s (71.9%) with a group flux of 1.10E þ09 n/cm2 s (72.1%) in the 14–15 MeV range. This represents 91.2% for this component which is still similar to the 14–15 MeV fraction in Position 15.
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Fig. 10. Peak fluxes in three samples at different angles.
Table 4 Flux distribution for three positions at varying distances and along the line of 451 relative to the source. Energy [MeV]
Position, d (cm)
0–0.11 0.11–1 1–10 10–14 14–15
15 (1.2)
1 (5.8)
5 (11.8)
0.3% 1.6% 5.0% 1.6% 91.4%
0.7% 2.8% 7.7% 2.4% 86.5%
0.8% 3.3% 8.8% 2.7% 84.4%
Table 5 Flux distribution for four radial positions. Energy [MeV]
0–0.11 0.11–1 1–10 10–14 14–15
Position, d (cm) 9 (4.8)
10 (5.8)
11 (7.2)
12 (8.8)
0.7% 3.1% 9.2% 4.3% 82.7%
0.7% 3.4% 9.8% 4.9% 81.2%
0.8% 3.6% 10.8% 5.5% 79.3%
0.9% 4.0% 11.6% 6.1% 77.5%
A summary of the total flux and several group fluxes for every position is given in Table 6. The second value listed in each group in the table is the 1 s error in %. The relative error can be as high as a factor of 2 at thermal energies because of the high uncertainty in the MCNP calculation and the lower reaction rates at these energies during the actual experiment. The total and fast fluxes, however, are reported with only a 1%–2% error.
3.3. Uncollided flux As described earlier and shown in Fig. 2, the experiment can be approximated by a model of a disc source and a sample position at point P given by the coordinates (a, h). Additionally, if R is the radius of the disc source and Sa is the source intensity averaged over the area of the disc, the uncollided flux, Φun, at point P is given by the analytical equation [14] qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi3 2 2 2 2 2 2 4 2 2 Sa 4a þR h þ R þ 2R ða2 h Þ þ ða2 þ h Þ 5 ð1Þ Φun ¼ ln 4 2a2 Eq. (1) assumes the source is isotropic and there is no attenuation. This is a first order approximation of the experiment since the real system is anisotropic and some attenuation between the D-T source and the sample exists because of the water and stainless steel in the cooling cap. Nevertheless, the only means of achieving an easy analytical solution for all space like Eq. (1) is by ignoring these effects. Since the peak energy of the source neutron is anywhere from 14.1 to 14.8 MeV, the 14–15 MeV bin of the flux determined by STAYSL PNNL is assumed to represent uncollided neutrons arriving at the sample. Thus, the flux in this bin can be compared to the flux predicted by Eq. (1). Fig. 11 shows the result when the 14–15 MeV bin of the 21 experimental points (see Table 6) are superimposed onto the surface Φun. The value for R was taken to be 1.65 cm, the measured radius of the vacuum tube in the generator where deuterons are accelerated, which is based on the assumption that neutrons are produced uniformly in the tritium target. The value for Sa was then determined to be 3.14E þ09 n/s by normalizing the 21 experimental 14 MeV fluxes to the value for phi (if Sa were equal to 1) at that coordinate, and taking an average value. Multiplying Sa by πR2 yields the total source intensity of 2.69E þ 10 n/s. This is
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Table 6 Group fluxes and relative error for every position. Position
Total flux
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21
7.58E þ 07 6.89E þ 07 8.55E þ 07 7.28E þ 07 2.14E þ 07 1.94E þ 07 2.07Eþ07 1.97Eþ07 1.12E þ08 7.16E þ 07 4.73E þ 07 2.69E þ 07 1.23E þ 08 5.15E þ 07 8.29E þ 08 3.72E þ 08 1.94E þ 08 1.02E þ 08 5.02E þ 07 3.14E þ 07 1.62E þ 07
1.4 1.3 2.4 1.4 1.4 1.5 1.4 1.4 1.6 1.6 1.5 1.7 1.4 1.6 1.4 1.7 1.2 1.3 1.8 1.7 1.7
Thermal 1E 4 eV to 0.55 eV
Epithermal 0.55 eV to 110 keV
1.56E þ04 1.43E þ04 1.03E þ04 1.01E þ 04 9.72Eþ03 9.32E þ 03 1.26E þ04 1.16E þ04 1.38E þ04 7.58E þ03 1.18E þ 04 9.70Eþ 03 1.17E þ 04 1.06E þ04 1.70Eþ 04 1.57E þ 04 1.41E þ 04 1.07Eþ 04 7.71E þ 03 8.98E þ 03 5.48E þ03
5.09E þ05 4.33Eþ 05 5.12E þ 05 4.31E þ05 1.58E þ 05 1.36E þ 05 1.63E þ 05 1.49E þ 05 7.88E þ 05 4.88E þ05 3.70Eþ 05 2.21E þ05 7.58E þ 05 3.95E þ05 2.68E þ06 1.62E þ 06 9.63E þ05 5.55E þ05 3.11E þ 05 1.81E þ 05 1.04E þ 05
101.7 97.4 121.6 120.2 124.6 127.5 59.2 88.1 125.2 129.2 115.7 126.4 115.9 128.5 110.9 111.3 111.6 107.6 118.6 119.7 128.3
Fast 110 keV to 15 MeV 14.7 14.4 12.4 12.3 12.7 12.8 14.7 16.2 12.4 12.6 12.7 12.3 12.5 12.9 12.3 12.5 12.4 12.8 12.2 12.7 12.5
7.53E þ07 6.85E þ07 8.49E þ07 7.23E þ07 2.12E þ 07 1.92E þ07 2.05E þ07 1.95E þ07 1.12E þ 08 7.11E þ 07 4.69E þ07 2.67Eþ 07 1.22E þ08 5.11E þ07 8.27E þ 08 3.70Eþ 08 1.93E þ08 1.01E þ 08 4.99E þ07 3.12E þ 07 1.61E þ 07
Fast 14–15 MeV 1.4 1.3 2.4 1.4 1.4 1.5 1.4 1.4 1.6 1.6 1.5 1.7 1.4 1.6 1.4 1.7 1.2 1.3 1.8 1.7 1.7
6.55E þ07 6.01E þ 07 7.40E þ 07 6.35Eþ 07 1.81E þ 07 1.65E þ 07 1.75E þ 07 1.66E þ 07 9.30E þ07 5.81E þ07 3.75E þ07 2.08E þ07 1.02E þ 08 4.13E þ07 7.58E þ 08 3.35Eþ 08 1.73E þ08 8.94E þ07 4.35Eþ 07 2.71E þ 07 1.40E þ 07
1.4 1.1 2.7 1.4 1.3 1.5 1.2 1.2 1.6 1.6 1.5 1.8 1.4 1.7 1.5 1.9 1.2 1.3 2.0 1.9 1.9
placing samples on the center of the cooling cap. A generalized formula for the uncollided flux has been extended to all space near the D-T source.
Acknowledgments
Fig. 11. 14 MeV flux with experimental data points superimposed onto the surface Φ. Dark blue points lie above the surface, light blue below. For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.
34% higher than the value of 2E þ 10 n/s reported by the vendor [15]. Experimental data still fits the shape of the curve in Fig. 11 quite well. Absolute differences between the measured flux and the analytical flux range from 1.6% to 27%. In some cases the surface is an overestimate of the measured value by as much as 27%, and in other cases it is an underestimate by as much as 24%. These differences are likely due to the fact that this is a first-order approximation neglecting anisotropy and attenuation. Further investigation including a least-squares fit of R and Sa to the data is required in order to resolve the value for source intensity. It may be that there is an effective radius where the D-T reaction occurs.
4. Conclusions The neutron flux has been well characterized in the space near the source of the D-T generator located in a low-scatter facility at PNNL. The STAYSL PNNL code was used to adjust MCNP spectra based on measured neutron activation rates in dosimetry foils and appropriate corrections. In all 21 positions studied, the peak energy lies in the range of 14.1–14.8 MeV and the intensity of this peak is 7 orders of magnitude larger than the intensity at thermal energies. At least 99.1% of the spectrum is comprised of fast neutrons, and one can minimize the scatter in the spectrum by
The research described in this paper was conducted under the Laboratory Directed Research and Development Program at Pacific Northwest National Laboratory, a Multiprogram National Laboratory operated by Battelle for the U.S. Department of Energy under contract DE-AC05-76RL01830. The authors would like to thank Shannon Morley for her analysis in GENIE 2000, Laura Thierolf for editing and formatting, and Zachary Hilliard for making some of the figures. This document is PNNL-SA-99038.
Appendix A. Supporting information Supplementary data associated with this article can be found in the online version at http://dx.doi.org/10.1016/j.nima.2013.11.023. References [1] J.C. Robertson, K.J. Zieba, Nuclear Instruments and Methods 45 (1) (1966) 179. [2] S. Shirato, S. Shibuya, Y. Ando, T. Kokubu, K. Hata, Nuclear Instruments and Methods in Physics Research A 278 (2) (1989) 477. [3] N.E. Hertel, B.W. Wehring, Nuclear Instruments and Methods 172 (3) (1979) 501. [4] P.M. Kehayias, J.J. Keyahias, Nuclear Instruments and Methods in Physics Research B 261 (1–2) (2007) 827. [5] S.A. Jonah, K. Ibikunle, Nuclear Instruments and Methods in Physics Research A 501 (2–3) (2003) 514. [6] X-5 Monte Carlo Team, MCNP—A General Monte Carlo N-Particle Transport Code, Version 5, vol. I: Overview and Theory, LA-UR-03-1987, Los Alamos National Laboratory, Los Alamos, New Mexico, 2008. [7] D. Pelowitz (Ed.) , MCNPX User's Manual Version 2.7.0, LA-CP-11-00438, Los Alamos National Laboratory, Los Alamos, New Mexico, 2011. [8] Canberra Industries, Inc., GENIE 2000 Spectroscopy Software: Operations. 9233652G V3.2, 2009. [9] National Nuclear Data Center, Brookhaven National Laboratory. Information extracted from the NuDat 2 database. Available from: 〈http://www.nndc.bnl. gov/nudat2〉 (accessed 2013).
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[10] R. Traub, Calculation of Ambient (H*(10)) and Personal (Hp(10)) Dose Equivalent From a 252Cf Neutron Source. PNNL-19273, Pacific Northwest National Laboratory, Richland, Washington, 2010. [11] L.R. Greenwood, D.G. Doran, H.L. Heinisch, Physical Review C 35 (1) (1987). [12] L.R. Greenwood, C.D. Johnson, User Guide for the STAYSL PNNL Suite of Software Tools. PNNL-22253, Pacific Northwest National Laboratory, Richland, Washington, 2013. [13] IAEA, International Reactor Dosimetry File 2002 (IRDF-2002), Technical Reports Series No. 452, International Atomic Energy Agency, Vienna, Austria, 2006.
[14] O.J. Wallace. Semi-Analytic Flux Formulas for Shielding Calculations. WAPDTM-1197, Bettis Atomic Power Laboratory, West Mifflin, Pennsylvania, 1976, p. 75. [15] Thermo Scientific D 711 Neutron Generators. Available from: 〈http://www. thermoscientific.com/ecomm/servlet/productsdetail_11152___11962782_-1〉 (accessed 01.05.13).