Portal Monitor Characterization for Internally and Externally Deposited Radionuclides

Operational Topic Detection efficiencies for internally and externally deposited radionuclides have been evaluated using a pass-through and static mode portal monitor. Portal Monitor Characterization for Internally and Externally Deposited Radionuclides Matthew Carey, Adam Kryskow, Fred Straccia, and Mark Tries* Abstract: In this evaluation, both the internal and external radionuclide detection efficiencies for a portal monitor were evaluated as a function of photon energy using an anthropomorphic phantom. Pass-through and static measurements were completed using 241 Am, 57 Co, 133 Ba, 137 Cs, 60 Co, 109 Cd, and 54 Mn in various locations both external and internal to the phantom. Other parameters, such as single detector uniformity, total detector uniformity, background linearity, and activity linearity have been analyzed. It was found that the minimum detectable activity for internally deposited 137 Cs in the abdomen was approximately ten times higher for pass-through versus static measurements. Additionally, it was found that the minimum detectable activity for 137 Cs in the abdominal region for both internal and external pass-through scenarios are nearly equivalent. In general, if the expected radionuclide source term is primarily non-transuranic, the pass-through mode offers sufficient sensitivity to identify potential overexposures while providing much greater personnel throughput. However, minimum detectable committed effective doses for transuranics such as 241 Am, show potential for personnel over exposure if the radionuclide mixture contains a significant fraction of transuranics. It is therefore recommended that nuclear facilities evaluate their radionuclide source term in *University of Massachusetts, Lowell, Radiological Sciences, University Avenue, MA 01854. Theauthorsdeclarenoconflictsofinterest. order to bound potential personnel doses. Health Phys. 107(Supplement 3):S188 S197; 2014 Key words: operational topics; contamination, internal; counting efficiency; detection limits INTRODUCTION Photon radiation portal monitors are commonly used as an element of radiation protection programs at nuclear facilities worldwide. These radiation monitors are primarily used for personnel monitoring of external contamination, but also have the ability to detect internally deposited radionuclides. Understanding the sensitivity of the radiation portal monitor to both externally and internally deposited radioactivity of single radionuclides or of radionuclide mixtures is an important element in assessing the potential total effective dose equivalent (TEDE) to facility personnel. The Mirion, (5000 Highlands Parkway Suite 150 Smyrna, GA 30082 USA) Fast-Track-Fibre (FTF) portal monitor was characterized Matthew Glen Carey is a health physicist at Radiation Safety & Control Services, Inc. Previously, he was a graduate student at the University of Massachusetts Lowell completing his Master s thesis on portal monitor characterization. This research involved Monte Carlo N-Particle (MCNP) radiation transport simulations, point kernel efficiency calculations, physical source measurements, and the comparison thereof. He specializes in using MCNP for solving complex health physics radiation transport problems such as shielding design, activation analysis, and detector response. His email address is Matthew_Carey@student.uml.edu. for response to internal and externally deposited radionuclides using an anthropomorphic phantom, as suggested by the portal monitor requirements of the American Nuclear Insurers (ANI) Engineering Inspection Criteria (ANI-EIC) Section 8.5.9 (ANI 2008). Detection efficiencies were measured with sources placed at internal and external positions on a phantom, using both static (stationary) and pass-through portal monitor detection modes. The FTF has proprietary algorithms that were not investigated, such as false alarm probability filters, time resolved algorithms, detection of fluctuating background conditions, and handling of special scenarios by pattern recognition. The efficiencies of multiple radionuclides with widely varying photon energies were evaluated in order to determine the FTF s energy response curve. This response curve, in terms of efficiency as a function of energy, allowed for the calculation of the minimum detectable activity (MDA), minimum detectable intake (MDI), and minimum detectable committed effective dose equivalent (MD- CEDE) per alarm for any photon emitting radionuclide mixture as described in Section 8.5.9.7 and Section 8.5.9.10 in ANI-EIC. The FTF was also analyzed for activity linearity, background linearity, single detector uniformity, and total detector uniformity as S188 www.health-physics.com November 2014

The Radiation Safety Journal Vol. 107, suppl 3 November 2014 described in Sections 8.5.9.4 and 8.5.9.8 in ANI-EIC. MATERIALS AND METHODS Portal monitor The FTF portal monitor contains 14 polyvinyl toluene (PVT) detectors with fiber optic coupling to two multichannel photomultiplier units (Mirion Technologies 2011). These PVT detectors run along the sides, bottom, and top of the device. The12sidedetectorsareeach85cm in height, 15 cm in width, 5 cm deep, and include 2.54 cm of lead shielding towards the exterior of the portal monitor, to reduce the background and potential cross-talk between multiple portal monitors if positioned side by side. The bottom and top detectors are 40 cm in width and length and 5 cm deep. The response from these detectors is grouped into eight channels leading to two multichannel photomultiplier units. For all static measurements recorded, except for those contained in the Single Detector Uniformity section, the responses were summed into a single system or total response. Radioactive sources Efficiency versus energy functions were acquired through the use of seven nuclides; five of which, 241 Am, 57 Co, 133 Ba, 137 Cs, 60 Co, are photon test standards recommended in Section 4.6 of N42.35 (ANSI 2006). Additionally, 109 Cd and 54 Mn were used to fill in gaps in the energy spectrum. The sources, with their percentage uncertainty and sigma values, are summarized in Table 1. Anthropomorphic phantom The Lawrence Livermore National Labs (LLNL) phantom was used for the both the internal static and pass-through measurements, as well as the external static measurements. An image of the LLNL phantom positioned inside the FTF Table 1. NIST sources used for evaluation. Nuclide Serial Number Corrected Activity (kbq) Half-life (y) 241 Am 1559-34-1 380.4 ± 3.0% (2.3σ) 432.17 ± 0.68 (1σ) 109 Cd 1559-34-3 368.0 ± 3.0% (2.3σ) 1.266 ± 0.002 (1σ) 57 Co 1559-34-4 391.8 ± 3.0% (2.3σ) 0.744 ± 0.0002 (1σ) 133 Ba 1559-34-2 39.11 ± 3.0% (2.3σ) 10.57 ± 0.04 (1σ) 137 Cs 1489-26-5 39.00 ± 3.1% (2.3σ) 30.17 ± 0.16 (1σ) 137 Cs 1559-34-7 9.487 ± 3.1% (2.3σ) 30.17 ± 0.16 (1σ) 137 Cs 1559-34-8 3.796 ± 3.1% (2.3σ) 30.17 ± 0.16 (1σ) 54 Mn 1559-34-5 37.81 ± 3.0% (2.3σ) 0.855 ± 0.001 (1σ) 60 Co 1559-34-6 37.81 ± 3.1% (2.3σ) 5.272 ± 0.001 (1σ) is shown in Fig. 1. Each source position shown in Fig. 1 is described in Table 2, per ANI-EIC Section 8.5.12. The portal monitor was evaluated for internal sensitivity per ANI-EIC Section 8.5.9 as well as external sensitivity. Prior to taking measurements, a two point calibration was completed using the 39 kbq 137 Cs source. This process was necessary for the external passthrough analysis described in the Internal and External Efficiencies section below. Static and passthrough measurements were completed for scenarios with an empty portal monitor (i.e., in-air measurements), and with sources positioned inside and outside a LLNL phantom. Due to the LLNL torso phantom not having extremities, external sources were placed on a human during pass-through measurements. Static data were recorded using the FTF s Wait-in mode, which allows for timed static measurements. Pass-through data were recorded using FTF s Fast-Track mode. Detector measurements were used to create photon energy response curves and calculate minimum detectable quantities. Detector linearity Activity linearity response. Section 8.5.9.4 and Section 8.5.9.8 in ANI-EIC requires counting rate to increase linearly with increasing source activity. This linearity was evaluated using three 137 Cs sources with activities of 39, 9.5, and 3.8 kbq as described in Table 1. Background linearity. Background levels were artificially increased using multiple sources positioned around the outside of the portal monitor. Background exposure rates were measured in the center of the FTF using a calibrated Ludlum, (501 Oak St, Sweetwater, TX 79556) 3 97 detector. The total gross counting rate of the portal monitor was evaluated for >linearity as a function of background exposure rate and then used to determine a conversion factor between counts per second (cps) and μr h 1. Effects of background counting rate on lower level discriminator. The lower limit of detection was calculated as a function of background counting rate using a 100 s background count time, 10 ssource count time, and detection probabilities of zα = 1.645 and zβ =3.09. Single detector uniformity. In order to measure the uniformity of a single detector, a side detector was removed and shielded, such that a source placed on the detector face would not add net counts to any other detector. With this approach, net counts on the detector channel were measured to determine detector uniformity over the detector face. Because the top and bottom detectors are on a dedicated photomultiplier channel, uniformity measurements only required background subtraction. Total detector uniformity. The total response of the system was evaluated at multiple locations, at a 10 cm distance from the inner portal monitor side walls. These measurements were taken from 20 to +20 cm in 10 cm intervals, Operational Radiation Safety www.health-physics.com S189

M. Carey et al. Portal monitor characterization FIG. 1. LLNL phantom within the FTF portal monitor. where 0 cm was taken to be the middle of the portal monitor (i.e., where a person would stand during a static measurement). Vertical measurements were taken at 20 cm intervals from the floor of the portal monitor to a height of 180 cm. Internal and external efficiencies Source locations for internal and external static and passthrough measurements are described in Table 2. The depths shown are given with respect to the surface tissue at each individual location. The width gives the lateral displacement of the source from the phantoms centerline. Right and left are taken from the phantom s frame of reference. A human was used in place of the LLNL phantom for elbow, knee, and foot measurements, as the phantom did not include these extremities. For pass-through measurement, when the FTF alarms, the device displays personnel transit speed (m s 1 ) and maximum and minimum expected activity but no counting data. Therefore, a roundabout method was created to determine the pass-through efficiency of the portal monitor as a function of personnel speed. The initial calibration with 137 Cs was used as the average total detector efficiency. When a pass-through alarm was triggered using the same 137 Cs source, the average of the maximum and minimum expected activities was recorded along with the estimated personnel speed. The average activity was then multiplied by the average detector efficiency for 137 Cs to calculate the expected net counting rate. This methodology was repeated 30 times for each measurement location and fit to a linear regression for counting rate vs. personnel speed. Minimum detectable activities were calculated under the assumption that the average alarm was equal to the alarm threshold. This procedure may be applied to other nuclides with similar photon energies to 137 Cs for external photon efficiency by using a ratio of the efficiency of 137 Cs pass-through to the 137 Cs static measurement. Efficiency calculations The efficiency of a nuclide with a half-life much greater than the counting interval was calculated experimentally in units of counts per nuclear transformation by dividing the net counting S190 www.health-physics.com November 2014

The Radiation Safety Journal Vol. 107, suppl 3 November 2014 Table 2. Source Locations within and on the LLNL Phantom. Region Height (cm) Width (cm) Depth (cm) Position Index Static Internal Left Lung 134 3 5 1 Right Lung 128 4 8 2 Thyroid 145 0 2 3 GI Tract 117 0 10 4 Static External Chest 130 0 0 5 Head 170 0 0 6 Gluteus 110 8 0 8 Knee a 42 10 0 10 Pass-through Internal Abdomen 115 0 0 7 Pass-through External Chest 130 0 0 5 Knee a 42 10 0 10 Abdomen 115 0 0 7 Foot a 1 15 0 11 Elbow a 120 25 0 9 a Used human in place of LLNL phantom. rate by the decay-corrected source activity A as: ε nuc transðcounts=decayþ ¼ R s A : (1) Most photon emitting nuclides emit multiple photon emissions per transformation, the yieldweighted average nuclide energy (also termed effective energy) was calculated by: E nuc ¼ ie i Y i i Y i : (2) The total yield of a nuclide was calculated as the sum of the photons emitted per transformation: Y nuc ¼ Y i : (3) i The effective nuclide efficiency in counts per photon was calculated by: ε nuc photonðcounts=photonþ¼ iε i Y i : (4) i Y i Minimum detectable quantities The efficiency of the portal monitor for different nuclides allowed for the calculation of the MDA, MDI, MD-CEDE per alarm, and MD-DAC-hr per alarm for all nuclides tested. The critical level (L C ) used to evaluate measurements for activity above background was established using a type I (false positive) error probability of five percent. This equates to a normal deviate z α of 1.645, which is used in the equation below (Currie 1968): sffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi L C ¼ z α R b þ R b ; T g (5) T b where T g is the sample counting interval (in this case the personnel counting interval) and T b is the background counting interval. The lower limit of detection (L D ), the smallest net signal that can be detected reliably was established using a type II (false negative) error probability of 0.10%. This equates to a corresponding normal deviate z β of 3.09 which was used in the equation below: L D ¼ Z β 2 þ L C þ 1 2T g 2 vffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi Z 4 Z 2 Z 2 u β β β T 2 þ 4LC þ 4 L 2 T g C g Z 2 α t : (6) The personnel counting interval, T g was equal to 10 s for this evaluation and the background counting interval was 100 s. The MDA for a given nuclide was calculated experimentally as the lower limit of detection divided by the counting efficiency in units of counts per nuclear transformation: MDA nuc ¼ L D ε nuc trans : (7) The MDA for a nuclide mixture was given by: L D MDA mix ¼ ; (8) i f j ε nuc trans j where f j is the nuclide activity fraction for the j th nuclide in the mixture. To determine the MDI for a given nuclide, it was assumed that the detection of the intake will be discovered in less than a half day following an acute inhalation or ingestion intake. The relationship between MDA and MDI was expressed using an intake retention function (IRF)asshownineqn(9): MDI nuc ¼ MDA nuc IRF : (9) For purposes of this evaluation, the delay post intake was assumed to be 0.2 d (4.8 h) and an aerodynamic median activity diameter (AMAD) of1micrometerwasused. NUREG/CR 4884, which is based on ICRP 30 methodology, was used to determine the IRF values for nuclides in the lungs, nasal passages, andgitract(u.s.nrc1987;icrp 1978). The IRF for ingestion was taken as 1.00. For a mixture of nuclides the MDI was calculated by: MDI mix ¼ MDA mix IRF : (10) The MD-CEDE was calculated from the MDI mix and the most limiting annual limit on intake (ALI) from 10 CFR 20, Appendix B (U. S. NRC 2012). Solubility classes for nuclides 133 Ba and 137 Cs were taken as Type D; 241 Am, 109 Cd, and 54 Mn as Type W; and 57 Co and 60 Co as Type Y. The resultant MD-CEDE represents the minimum internal dose from an acute intake of a photon emitting nuclide mixture (in units of msv) that would correspond to a portal monitor alarm: Operational Radiation Safety www.health-physics.com S191

M. Carey et al. Portal monitor characterization Background linearity. A linear fit (with the y-intercept set to zero) was applied to the total detector counting rates versus background exposure rate data shown in Fig. 3. These data indicate some slight deviations from linearity however, these deviations were considered acceptable (correlation coefficient of 0.9891) and are most likely attributable to the nonuniformity of the artificial background field that was created. A conversion factor between total detector counting rate and a uniform background exposure rate was determined as: Background μr hr ¼ R b 1652 ; (13) FIG. 2. Portal monitor response linearity. D MD-CEDE mix ¼ MDI mix ALI fi ALI j j ; (11) where f j is the nuclide activity fraction of the mix, and D ALI is 50 msv for SALI and 500 msv for NALI. The minimum detectable derived air concentration (MD-DAC) in DAC-hrs was calculated as: 2000DAC-hr MD-DAC-hr mix ¼ MDI mix f j SALI j j : (12) The MD-CEDE and MDDAChr for single nuclides are easily inferred from eqn (11) and eqn (12) by setting j equal to one. RESULTS AND DISCUSSION The initial performance parameters of the FTF portal monitor are described in the Detector Linearity section. Detector efficiencies for static and pass-through modes are presented in the following sections for both internal and externally deposited radionuclides. Minimum detectable quantity results such as MDA, MDI, MD-CEDE and MD-DAC-hr are discussed below. rate versus the source activity, as shown in Fig. 2. An unweighted linear function (with the y-intercept set to zero) was fit to the data. The resultant correlation coefficient of 0.9998 indicated the detector was behaving as expected. where 1,652 is the slope of the linear regression line found in Fig. 3. This slope is only accurate for the sources and geometry used to generate the high background field in this experiment. Single detector uniformity. Detector uniformity for the side, top, and bottom type detector is shown in Fig. 4 and Fig. 5. As expected, the detector response was greatest in the center, were the solid angle subtended was the greatest. Total detector response. The measured total detection uniformity in Fig. 6 shows larger normalized values in the center of the horizontal axis, with multiple larger regions along the vertical axis. The increase in response towards the center of the horizontal axis was expected due to the increased total solid angle subtended. The reduced efficiency around 100 cm in height was also expected due to the existence of an air gap between the Detector linearity Activity linearity response. The linearity of the detector was evaluated by plotting the net counting FIG. 3. Detector counting rate as a function of background. S192 www.health-physics.com November 2014

The Radiation Safety Journal Vol. 107, suppl 3 November 2014 Internal static efficiencies Static measurement efficiencies for various internal locations are shown in Fig. 7. Regression parameters for the thyroid, right lung, GI tract and left lung static internal efficiencies have been fit to third order polynomials given in eqns (14) through (17); with correlation coefficients of 0.99909, 0.99695, 0.99747, and 0.99914, respectively: ε Thyroid ¼ 0:09349 E 3 0:3151 E 2 þ 0:3461E 0:01044 (14) ε Right Lung¼ 0:08009E 3 0:2613E 2 þ 0:2880E 0:009167 (15) ε GI Tract ¼ 0:005043E3 0:08066E 2 þ 0:1657E 0:006220 (16) ε Left Lung¼ 0:08286E 3 0:2820E 2 þ 0:3160E 0:01052: (17) The efficiency curves for the internal efficiency measurements show higher overall efficiencies for the thyroid region, with the GI tract having the lowest efficiency. This is expected as the thyroid has less attenuating material between the source and the detectors, while the GI tract has the most attenuating material, thus the exiting photon has a lower probability of being detected. It should be noted that the left lung efficiencies are slightly higher than the right lung due to the left lung source depth being 5 cm, while the right lung was positioned deeper, at 8 cm. FIG. 4. Normalized single side detector uniformity. upperand lowersidedetectors.it is assumed that Fig. 6 would appear vertically symmetric had measurements been taken at the top of the detector (roughly 200 cm). External static efficiencies Theefficiencyresultsareshown as a function of effective photon energy in Fig. 8. The polynomial regression results for the external static efficiencies are contained in eqns (18) through (21); with correlation coefficients of 0.99766, 0.99383, 0.97850, and 0.98443, respectively: Operational Radiation Safety www.health-physics.com S193

M. Carey et al. Portal monitor characterization lower efficiency. Two trials were completed for the source on the right knee. A higher efficiency is found when the source leg steps into the portal monitor first, as expected due to the temporary lack of motion as weight is shifted onto the leading leg. FIG. 5. Normalized bottom detector uniformity. Minimum detectable quantities Effects of background counting rate on lower level discriminator. L D as a function of background counting rate is shown in Table 4 and shown in Fig. 10. As one can see,theshapeofthecurveinfig.10 approximates a square root function, as expected from eqns (5) and (6). ε Chest ¼ 0:03021 E 3 0:1146E 2 þ 0:1684E 0:00312(18) ε Gluteus ¼ 0:00255 E 3 0:0822E 2 þ 0:1746E 0:00443 (19) ε Head ¼ 0:02791E 3 0:1473E 2 þ 0:2275E 0:00223 (20) ε Knee ¼ 0:04469 E 3 0:2357E 2 þ 0:3192E 0:01014: (21) Internal pass-through efficiencies Measurement of the internal pass-through efficiency was completed for a source positioned within the abdomen of the LLNL phantom, with the results shown in Fig. 9. Due to the variance in the efficiency results and the uncertainty in entry velocity, a linear regression of the data would result in a very low linear regression correlation coefficient of 0.0046. Therefore, there is no appreciable correlation found between internal pass-through efficiency and entrance velocity. However, the average of the data in Fig. 9 gives an average entry velocity of 0.9 m per second, and an average detector efficiency of 0.8%. External pass-through efficiencies External pass-through efficiency data for each body part from linear regressions of each data set are shown in Table 3. As expected, faster transit times led to a FIG. 6. Normalized measured detector uniformity from detector center. S194 www.health-physics.com November 2014

The Radiation Safety Journal Vol. 107, suppl 3 November 2014 The expected MD-DAC-hr for each nuclide is calculated using eqn (12). This calculation was performed for an inhalation scenario the results of which are shown in Table 8. FIG. 7. Internal static efficiencies. External static minimum detectable activity. The MDA for external source positions was calculated for various locations for a background counting rate of 22,000 cps. These results are contained in Table 9 for the various nuclides used. Internal pass-through minimum detectable activity. An estimate of the MDA for internal pass-through of 137 Cs in the abdomen can be calculated using theaverageofthedatashownin Fig. 9. Using a 1 m per second entry velocity and the minimum efficiency of 0.6% yields an MDA for 137 Cs in the abdomen is found to be 5 10 4 Bq. FIG. 8. External static efficiencies. Internal static minimum detectable activity. The average measured internal efficiency for each nuclide was used to calculate the MDA of each nuclide for an average background of 22,000 cps. These results are contained in Table 5 using a false positive probability of 5% and a false negative probability of 0.10% as stated previously. The expected MD-CEDE for the likely solubility class for each nuclide was calculated using eqn (11). For these calculations the left and right lung MDA values were averaged and the thyroid measurements assumed to be the nasal passage (due to proximity). MD-CEDE values are contained for either an inhalation or ingestion scenario in Table 6 and Table 7. The lowest MD-CEDE for each nuclide are bolded in Table 6. FIG. 9. Internal pass-through efficiency of abdomen. External pass-through minimum detectable activity. Measurement of the efficiency as a function of entrance velocity was used to calculate the MDA for 137 Cs for pass-through monitoring. These results are shown for various body locations at different velocities in Table 10. Operational Radiation Safety www.health-physics.com S195

M. Carey et al. Portal monitor characterization Table 3. External pass-through efficiency vs. entrance velocity. Efficiency at given velocity Source location 0.5 (m s 1 ) 1 (m s 1 ) 1.5 (m s 1 ) 2 (m s 1 ) Chest 1.3% 1.2% 1.0% 0.9% Source on right knee, left in first 1.4% 1.4% 1.3% 1.3% Source on right knee, right in first 1.8% 1.8% 1.8% 1.8% Abdomen 1.5% 1.3% 1.2% 1.0% Left foot 2.0% 1.8% 1.5% 1.2% Elbow 2.2% 2.0% 1.7% 1.5% Table 7. Minimum detectable committed effective dose (Sv) from ingestion. Nuclide GI tract 241 Am 3 10 2 109 Cd 4 10 4 57 Co 5 10 5 133 Ba 2 10 5 137 Cs 3 10 5 54 Mn 6 10 6 60 Co 1 10 4 Table 4. Background total counting rate vs. lower limit of detection. Background counting rate (cps) Expected background (μr hr 1 ) FIG. 10. Lower limit of detection vs. total background counting rate. Lower limit of detection (cps) 22,000 13 982 26,400 16 1,075 30,800 19 1,161 35,200 21 1,241 39,600 24 1,316 44,000 27 1,386 Table 5. Minimum detectable activity (Bq). Nuclide Right lung Left lung GI Tract Thyroid 241 Am 3.95 10 4 3.52 10 4 9.78 10 4 2.86 10 4 109 Cd 1.25 10 4 1.36 10 4 2.30 10 4 1.17 10 4 57 Co 1.17 10 4 9.79 10 3 2.03 10 4 8.58 10 3 133 Ba 4.49 10 3 4.09 10 3 6.80 10 3 3.70 10 3 137 Cs 2.55 10 3 2.35 10 3 3.40 10 3 2.16 10 3 54 Mn 2.39 10 3 2.16 10 3 2.85 10 3 2.01 10 3 60 Co 2.32 10 3 2.17 10 3 2.72 10 3 2.04 10 3 Table 6. Minimum detectable committed effective dose (Sv) from inhalation. Nuclide Lungs GI Tract Nasal 241 Am 1 10 3 2 10 3 3 10 2 109 Cd 2 10 3 3 10 3 8 10 4 57 Co 3 10 4 4 10 4 8 10 5 133 Ba 7 10 5 9 10 5 2 10 1 137 Cs 4 10 5 1 10 2 5 10 1 54 Mn 5 10 5 5 10 5 2 10 5 60 Co 2 10 3 1 10 3 5 10 4 CONCLUSION The FTF portal monitor was found to be operating as expected in terms of activity and background linearity. Uniformity measurements of both the single and total detector uniformity resulted in responses that were mostly due to distance and the solid angle subtended by the detectors, as expected. The FTF was found to have an average static inhalation MDI of approximately 2 10 4 Bq for most nuclides (excluding 241 Am) for a measurement of 10 s. This inhalation MDI would result in MD-CEDE values on the order of tens to hundreds of μsv, which is well below regulatory limits. However, for the case of 241 Am, the MD-CEDE for ingestion and especially inhalation is particularly high and therefore shows that nuclides 241 Am can t be sufficiently measured by this technique, which poses a risk for overexposure. Internal static mode measurements found MDA values on the order of 10 4 Bq. External static mode measurements found MDA values for 137 Cs around 3 10 3 Bq. The internal pass-through MDA for 137 Cs was found to be approximately 5 10 4 Bq. External pass-through MDA results for 137 Cs are on the order of 10 4 Bq and 10 5 Bq for velocities of 0.5 and 2 m s 1,respectively. While the static mode is appreciably more sensitive compared to the pass-through mode, pass-through monitoring may be an acceptable practice to employ for both internal and external source terms. This is, of course, highly dependent on the nuclide mixture and deposition location. S196 www.health-physics.com November 2014

The Radiation Safety Journal Vol. 107, suppl 3 November 2014 Table 8. Minimum DAC-hr from inhalation. Nuclide Right lung Left lung GI tract Thyroid 241 Am 2.92 10 6 2.60 10 6 7.23 10 6 2.11 10 6 109 Cd 1.85 10 2 2.01 10 2 3.41 10 2 1.72 10 2 57 Co 1.24 10 1 1.04 10 1 2.15 10 1 9.14 10 0 133 Ba 4.75 10 0 4.33 10 0 7.19 10 0 3.92 10 0 137 Cs 9.45 10 0 8.72 10 0 1.26 10 1 8.02 10 0 54 Mn 2.22 10 0 2.00 10 0 2.64 10 0 1.86 10 0 60 Co 5.74 10 1 5.40 10 1 6.74 10 1 5.08 10 1 Table 9. External static MDA (Bq) for various source positions. Nuclide Chest Head Gluteus Knee 241 Am 3.32 10 4 1.14 10 5 6.86 10 4 6.31 10 4 109 Cd 1.78 10 4 8.19 10 3 1.45 10 4 1.08 10 4 57 Co 1.77 10 4 1.04 10 4 1.72 10 4 9.21 10 3 133 Ba 6.57 10 3 4.71 10 3 6.51 10 3 3.81 10 3 137 Cs 3.41 10 3 2.59 10 3 2.94 10 3 2.29 10 3 54 Mn 3.07 10 3 2.20 10 3 2.73 10 3 2.74 10 3 60 Co 2.62 10 3 2.16 10 3 2.54 10 3 2.10 10 3 Table 10. Pass-through external surface MDA vs. velocity and source position. Pass-through external surface 0.5 (m s 1 ) 1 (m s 1 ) 1.5 (m s 1 ) 2 (m s 1 ) Chest 1.67 10 4 4.46 10 4 7.25 10 4 1.00 10 5 Source of right knee, left in 1st 2.09 10 4 4.18 10 4 6.27 10 4 8.36 10 4 Source on right knee, right in 1st 1.77 10 4 3.53 10 4 5.29 10 4 7.06 10 4 Abdominal 1.69 10 4 4.40 10 4 7.12 10 4 9.84 10 4 Left foot 9.69 10 3 4.43 10 4 7.88 10 4 1.13 10 5 Elbow 5.31 10 3 3.11 10 4 5.69 10 4 8.27 10 4 In conclusion, understanding the response of portal monitors to internally deposited radioactivity can be an important tool in allowing radiation protection personnel to bound potential internal 137 Cs MDA (Bq) at given velocity doses (Carey et al. 2012). This information could be applied in litigation support; or in evaluation of potential missed doses if unmonitored internal exposures occurred in a facility. Acknowledgments This study was funded in part by the University of Massachusetts Lowell, Radiation Safety & Control Services, Inc., and Mirion Technologies HP Division. REFERENCES American Nuclear Insurers. Engineering inspection criteria for nuclear liability insurance. Section 8.5 Revision 5. West Harford, CT: American Nuclear Insurers; 2008. CareyM,DaroisE,StracciaF.Thesensitivity of gamma portal monitors to personnel intakes of radioactivity.ouray:sandrikwadepublishing; Nuclear Decommissioning Report; 2012. Mirion Technologies. CheckPoint: Gate and FastTrack Fibre Pedestrian Monitor: Technical handbook. Hamburg, Germany: Mirion Technologies; 2011. American National Standards Institute. American National Standard for evaluation and performance of radiation detection portal monitors for use in homeland security; New York: IEEE; N42.35 2006; 2006. Currie LA. Limits for quantative detection and quantitative deternatino. Analytical Chem 40: 587 593; 1968. U.S. Nuclear Regulatory Commission. Interpretation of bioassay measurements. Washington, DC: U.S. NRC; NUREG/CR 4884; 1987. International Commission on Radiological Protection. 1978 limits for intakes of radionuclides by workers. Oxford: Pergamon Press; ICRP Publication 30; 1978. U.S. Nuclear Regulatory Commission. Annual limits on intake (ALIs) and derived air concentrations (DACs) of radionuclides for occupational exposures. Washington, DC: U.S. Government Printing Office; 2012. Operational Radiation Safety www.health-physics.com S197