Operational Topic DOSE EQUIVALENT RATE CONSTANTS AND BARRIER TRANSMISSION DATA FOR NUCLEAR MEDICINE FACILITY DOSE CALCULATIONS AND SHIELDING DESIGN Maggie Kusano* and Curtis B. Caldwell† Abstract—A primary goal of nuclear medicine facility design is to keep public and worker radiation doses As Low As Reasonably Achievable (ALARA). To estimate dose and shielding requirements, one needs to know both the dose equivalent rate constants for soft tissue and barrier transmission factors (TFs) for all radionuclides of interest. Dose equivalent rate constants are most commonly calculated using published air kerma or exposure rate constants, while transmission factors are most commonly calculated using published tenth-value layers (TVLs). Values can be calculated more accurately using the radionuclide's photon emission spectrum and the physical properties of lead, concrete, and/or tissue at these energies. These calculations may be non-trivial due to the polyenergetic nature of the radionuclides used in nuclear medicine. In this paper, the effects of dose equivalent rate constant and transmission factor on nuclear medicine dose and shielding calculations are investigated, and new values based on up-to-date nuclear data and thresholds specific to nuclear medicine are proposed. To facilitate practical use, transmission curves were fitted to the three-parameter Archer equation. Finally, the results of this work were applied to the design of a sample nuclear medicine facility and compared to doses calculated using common methods to investigate the effects of these values on dose estimates and shielding decisions. Dose equivalent rate constants generally agreed well with those derived from the literature with the exception of those from NCRP 124. Depending on the situation, Archer fit TFs could be significantly more accurate than TVL-based TFs. These results were reflected in the sample shielding problem, with unshielded dose estimates agreeing well, with the exception of those based on NCRP 124, and Archer fit TFs providing a more accurate alternative to TVL TFs and a simpler alternative to full spectral-based calculations. The data provided by this paper should assist in improving the accuracy and tractability of dose and shielding calculations for nuclear medicine facility design. Health Phys. 107(1):60–72; 2014

*Sunnybrook Research Institute, Toronto, Ontario, Canada; †Department of Medical Imaging, Sunnybrook Health Sciences Centre, Toronto, Ontario, Canada; †Departments of Medical Imaging and Medical Biophysics, University of Toronto, Toronto, Ontario, Canada. The authors declare no conflicts of interest. For correspondence contact: Curtis Caldwell, Medical Imaging, AG40 Sunnybrook Health Sciences Centre, 2075 Bayview Avenue, Toronto, Ontario, Canada M4N 3M5, or email at [email protected]. (Manuscript accepted 8 October 2013) 0017-9078/14/0 Copyright © 2014 Health Physics Society DOI: 10.1097/HP.0000000000000051

Key words: nuclear medicine; radiation, medical; radionuclide; shielding

INTRODUCTION FROM A radiation protection perspective, the primary goals of nuclear medicine facility design are to keep radiation doses to workers and members of the public both below regulatory limits and As Low As Reasonably Achievable (ALARA). This can be achieved, in part, by evaluating a facility's layout and protocols, estimating doses received by workers and members of the public due to these protocols, and adding structural shielding (e.g., lead or concrete barriers) to reduce doses where necessary. Assuming a point source geometry, the unshielded equivalent dose H (μSv) received by an individual at point q due to a nuclear medicine patient at point p can be given by: H¼

Dδst A0 F 1 t 2 Rt2 ; d2

 F 1 ¼ exp −0:693 t1 =T 1=2 ;   Rt2 ¼ 1:443 T 1=2 =t 2 1− exp −0:693 t2 =T 1=2 ;

(1) (2) (3)

where A0 is the activity administered (GBq) at time t0, t1 is the time between administration and arrival at point p (h), t2 is the time spent by the patient/radionuclide at p (h), T1/2 is the radiopharmaceutical effective half-life (h), d is the distance between p and q (m), F1 is the decay factor over time t1, Rt2 is the dose reduction factor due to decay over time t2, and Dδst is the dose equivalent rate constant for soft tissue for the administered radionuclide (μSv m2 GBq−1 h−1), calculated for photon emissions above a threshold of δ keV. The dose equivalent rate constant relates the activity of a radionuclide to the dose rate to a soft tissue target a given distance away and depends on the radionuclide’s decay spectrum and the response of tissue to photons at these energies. If one ignores bremsstrahlung, which is negligible for electrons generated by primary photon energies < 1 MeV in air (DIN 2006) and is expected to be even less

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Nuclear medicine facility dose calculations and shielding design c M. KUSANO AND C. B. CALDWELL

in soft tissue, the dose equivalent rate constants for soft tissue can be given by:   1 μ Dδst ¼ E i Y i en ; (4) 4π i ρ i

the shielded dose Hs at point q due to a nuclear medicine patient at location p is given by:



where Yi and (μen/ρ)i are the yield and mass energyabsorption coefficient of soft tissue (cm2 g−1) for each of the radionuclide's photon energies Ei > δ keV. A threshold, δ, of 20 keV is commonly used for practical health physics calculations (Sorenson and Phelps 1987; Wasserman and Groenewald 1988; Tschurlovits et al. 1992; Ninkovic et al. 2005; DIN 2006), since photons of lower energy are generally significantly attenuated by their containers (i.e., vial, syringe, or patient). Because of the polyenergetic nature of the radionuclides used in nuclear medicine (Firestone and Ekstrom 2004), calculation of Dδst using eqn (4) can be tedious due to the number of interpolations required to estimate mass energy-absorption values at untabulated values. Dose equivalent rate constants (i.e., Dδst) are more easily estimated by multiplying published air kerma rate constants or exposure rate constants by radionuclide-specific factors that relate the response of tissue to that of air over the range of photon energies emitted by the radionuclide. For example, Wasserman and Groenewald (1988) calculated their dose equivalent rate constants by multiplying published air kerma rate constants by 1.1, the approximate ratio of the mass energy-absorption coefficient of soft tissue to that of air between 20 keV and 1 MeV. Ninkovic et al. (2005) suggested a factor of 1.11 be used to convert air kerma rate constants. Spectrally averaged f-factors (cGy R−1) were used by Smith and Stabin (2012) to convert their exposure rate constants to dose to soft tissue. From Table 1, in can be seen that dose equivalent rate constants for 99mTc calculated using this “more convenient” technique can differ by up to 40%, implying that dose estimates can vary by up to 40% depending on choice of reference. Clearly, the choice of dose equivalent rate constant for soft tissue can significantly impact dose estimates and shielding requirements. Extending the point source/nuclear medicine patient scenario presented above to include the effects of shielding,

61

H s ¼ H  TF;

where H is given by eqn (1) and TF is the transmission factor of a barrier between points q and p for the photon spectrum of the radionuclide in question. One of the simplest ways to calculate a barrier’s transmission factor for a particular radionuclide is to use its published broad beam first tenth-value layer (TVL): T F TVL ðxÞ ¼ 10−ðx=TVLÞ ;

T F B;i ðxÞ ¼ expð−μi xÞBðxÞi ;

— 0.6 — 0.795

(7)

where μi is the material’s linear attenuation coefficient (cm−1) for photons of energy Ei, x is the barrier thickness (cm), and B(x)i is the material’s buildup factor for thickness x and energy Ei. Alone, linear attenuation coefficients underestimate dose and shielding requirements because they do not account for scatter. Buildup factors, derived using transport calculations (e.g., Takeuchi and Tanaka 1985; Tanaka and Takeuchi 1986; Subbaiah et al. 1982; Hirayama et al. 1990),

Exposure rate constant Air kerma rate constant (R cm2 mCi−1 h−1) (μGy m2 GBq−1 h−1)

NCRP 124 (1996) Wasserman and Groenewald (1998) Ninkovic et al. (2005) Smith and Stabin (2012)

(6)

where x is the thickness of the barrier in the same units as the TVL. Broad beam tenth-value layers for common nuclear medicine radioisotopes through lead can be found in a number of references such as NCRP Report 124 (1996), DIN 6844‐3 (DIN 2006), and Smith and Stabin (2012). Reliable references for concrete are more elusive and could only be found in DIN 6844‐3 at the time of writing. So, although easy to calculate, one drawback of the TVL method is that it is limited to the set of radionuclide/barrier material combinations for which TVLs have been calculated and published. An additional drawback is that TVLs generally cannot be used to extrapolate transmission values at smaller or larger shielding thicknesses due to the multienergetic nature of radionuclides in nuclear medicine. A more accurate estimate of a barrier’s attenuation can be calculated by applying linear attenuation coefficients and exposure buildup factors to each of the radionuclide’s photon energies Ei:

Table 1. The dose equivalent rate constant for 99mTc derived from reported exposure and air kerma rate constants. Reference

(5)

20 14.2 14.10 N/A

Conversion factor

Derived dose equivalent rate constant (μSv m2 GBq−1 h−1)

1.1a 1.1 1.11 0.959 cGy/R

22 15.6 15.7 20.6

a

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Monte Carlo modeling (e.g., Harima et al 1987; Hirayama et al. 1990; Chibani 2001; Kharrati et al. 2007, 2012), the moments method (e.g., Eisenhauer and Simmons 1975), or the method of invariant embedding (e.g., Shimizu et al. 2004) correct for this by relating the unscattered radiation at a point to the total radiation at that point. In practice, dose and shielding calculations are most commonly performed using simple calculation tools such as spreadsheets. As a number of different radionuclides, administration and uptake periods, and imaging procedures may be used for nuclear medicine imaging and treatment, these calculations can become complicated, laborious, and prone to human error, so the less computationally intensive methods for calculating Dδst and TF tend to be favored. Simplifications, such as ignoring procedures that are performed less frequently, involve low activities, or have short imaging times are also often exploited. More accurate estimates can be obtained by writing software programs to implement the spectrum-based equations (eqns 1–5 and 7) above and to interpolate tabulated values of mass energyabsorption coefficients, mass attenuation coefficients, and buildup factor parameters (Kusano and Caldwell 2008; McGinnis 2012). The purpose of this paper is to study the effects of different methods of deriving dose equivalent rate constant and transmission factors on nuclear medicine dose and shielding calculations. More specifically, the objectives of this paper are to: (1) evaluate whether there are significant differences in accuracy between different dose equivalent rate constant and shielding transmission factor calculation methods; (2) assess the effects of calculation methods on nuclear medicine facility shielding design; (3) if it appears that different methods lead to differences in these factors, determine when more simplistic calculation methods would provide acceptable accuracy and when more complex methods would be required; and (4) provide accurate dose equivalent rate constants and transmission factors for radionuclides commonly used in nuclear medicine. METHODS Calculation of dose equivalent rate constant Dose equivalent rate constants for 67Ga, 99mTc, 111In, 123 131 133 I, I, Xe, and 201Tl were calculated using eqn (4), nuclear decay data from ICRP Report 107 (ICRP 2008), and log-log cubic spline interpolation of mass energyabsorption coefficients for soft tissue from Hubbell and Seltzer (2004). Only gamma rays, prompt and delayed gamma rays of spontaneous fission, and x-rays with minimum yields of 10−4 and energies of at least 20 keV were considered. The dose equivalent rate constants determined by this work were compared to those published by Wasserman and Groenewald (1988), those converted from the exposure

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rate constants of Smith and Stabin (2012), and those derived from the air kerma rate constants published by NCRP 124 (NCRP 1996) and Ninkovic et al. (2005). The Smith and Stabin data were converted to dose equivalent rate constants by multiplying by the f-factors provided by that paper. The NCRP 124 air kerma rate constants were multiplied by 1.1, the approximate ratio of mass energyabsorption in tissue to that of air recommended by Wasserman and Groenewald (1988). The Ninkovic et al. (2005) data were multiplied by 1.11, the mass energy-absorption ratio recommended by that paper. Calculation of transmission factor Transmission factors for 67Ga, 123I, 131I, 111In, 99mTc, 201 Tl, and 133Xe in lead and concrete were calculated by combining eqns (1)–(5) and (7):

 

TF ðxÞ ¼

∑i E i Y i

μen ρ

expð−μi xÞBðxÞi   : μ ∑i Ei Y i en ρ i i

(8)

Nuclear decay data were taken from ICRP Report 107 (ICRP 2008), and mass attenuation coefficients and energyabsorption coefficients were calculated using log-log cubic spline interpolation of the data provided by Hubbell and Seltzer (2004). Buildup factors were drawn from ANSI/ANS‐6.4.3 (ANS 1991). ANSI/ANS‐6.4.3 was originally published in 1991 in response to the accident at Three Mile Island in 1979, and although withdrawn in 2001 due to a failure to update the document within 10 y of its initial publication, it continues to be one of the most complete and easy-to-use sources of buildup factor data. The document includes data for concrete, iron, water, air, and 15 low-Z elements, calculated using the method of moments, and for seven high-Z elements (including lead) calculated using the discrete ordinates-integral transport theory PALLAS code (Takeuchi and Tanaka 1981, 1984). All data were validated with Monte Carlo N-Particle (MCNP) transport code} or Electron Photon Shower Simulation 4 (EGS4) Monte Carlo simulations (Nelson et al. 1985) and Anisotropic Source-Flux Iteration Technique (ASFIT) discrete ordinates-integral transport theory code**. Depths range from 0.5–40 mean free paths (mfp), and energies range from 15 keV (30 keV for lead) and 15 MeV. Since the buildup factor data do not include the effects of coherent }

Radiation Shielding Information Center. Radiation Shielding Information Center code package CCC-200/MCNP, MCNP - a general Monte Carlo code for neutron and photon transport, contributed by Los Alamos National Laboratory. **Radiation Shielding Information Center. Radiation Shielding Information Center code package CCC-336/ASFIT-VARI, gamma-ray transport code for one-dimensional finite systems contributed by Indira Gandhi Cenre for Atomic Research, India.

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Nuclear medicine facility dose calculations and shielding design c M. KUSANO AND C. B. CALDWELL

scattering, attenuation coefficients calculated without coherent scatter are provided for calculation of the number of mean free paths. Coherent scatter correction factors for high-Z materials, including lead, are also included in the document. To improve the accuracy of interpolation between energies and thicknesses and facilitate use for dose and shielding calculations, a number of empirical formulae have been fit to tabulated buildup factor data. A good fit (Harima 1993) to the ANSI/ANS‐6.4.3 data is provided by the geometric-progression (GP) method (Harima et al. 1986):  BðxÞ ≅

1 þ ðb −1ÞðK x −1Þ=ðK −1Þ; K ≠ 1 1 þ ðb −1Þx; K ¼ 1′

K ðxÞ ¼ cx þ d a

  tanh ξx −2 − tanhð−2Þ 1− tanhð−2Þ

;

(9)

(10)

where a, b, c, d, and ξ are material and energy-dependent parameters of the GP formula. These values are also published in ANSI/ANS‐6.4.3 (ANS 1991) and can alternatively be found in Harima (1993) and Shultis and Faw (2000). For this study, buildup factors were calculated using the geometric-progression fit parameters and lead coherent scattering correction factors found in ANSI/ANS‐ 6.4.3. The geometric-progression formula was extended to 100 mfp using the work of Harima (1991, 1993). Lead barriers up to 100 mm thick at increments of 0.01 mm and concrete barriers up to 80 cm thick at increments of 0.01 cm were evaluated, encompassing the range of greatest interest in nuclear medicine shielding applications. To facilitate practical use of these data, transmission curves were fit to the mathematical model proposed by Archer et al. (1983):

63

   β β −1=γ expðαγ xÞ− TF ðxÞ ¼ 1þ ; α α

ð11Þ

where x is the thickness of lead (in mm) or concrete (in cm), and α, β, and γ are radionuclide- and materialdependent parameters. Fits were limited to 100 equally spaced transmission factor values between TF = 1 and TF = 10−4. Values of α, β, and γ were determined using MPFIT, a non-linear least squares fitting procedure written by Markwardt (2008) that was implemented in Interactive Data. Language (IDL) version 8.1 (Excelis Visual Information Solutions, Boulder, CO). First TVLs in lead were also determined from the original transmission curves. TVLs in lead were compared to those derived from NCRP 124 (NCRP 1996), DIN 6844‐3 (DIN 2006), and Smith and Stabin (2012) data. TVLs for concrete were compared to data from DIN 6844‐3. To derive TVLs from the transmission curves provided by DIN 6844‐3, graphs were digitized and discretized using DataThief III, version 1.6 (Tummers 2006). Tenth-value layers were then determined by linear interpolation of the discretized data. The estimated uncertainty in the digitization for the data in question was less than ±2% in all cases. Calculation of dose and shielding: a nuclear medicine scan room example The impact of the different methods of deriving dose equivalent rate constants and transmission factors on worker dose estimates and shielding decisions was evaluated by applying new and published values to a sample shielding problem: the calculation of unshielded and shielded doses outside a nuclear medicine scan room. The hypothetical single-camera nuclear medicine facility was assumed to be dominated by 99mTc-based procedures, performing primarily

Table 2. Hypothetical nuclear medicine scan room procedures used to test the effect of differences in dose rate constants and transmission factors in a practical scenario.

Scan room procedure

Nuclide

Procedures per year N

Bone scan Cardiac perfusion—rest scan Cardiac perfusion—stress scan Gallium oncology I-123 MIBG Whole body I-131 Octreotide

Tc-99m Tc-99m Tc-99m Ga-67 I-123 I-131 In-111

500 500 500 50 50 50 50

0.85 0.37 1.11 0.296 0.296 0.111 0.222

Cardiac viability

Tl-201

50

0.111

Lung Ventilation

Xe-133

50

0.555

Administered activity A0 (GBq)

Uptake time t1 (h) 3 1 1 72 24 72 4 (scan 1) 24 (scan 2) 0.17 (scan 1) 3 (scan 2) 0

Scan time t2 (h)

Percentage of scan room workloada

0.75 0.33 0.33 1.5 1.5 1.5 1.5 (scan 1) 1.5 (scan 2) 0.33 (scan 1) 0.33 (scan 2) 0.25

Approximate percentage of scan room workload = (NA0t2)procedure/Σ(NA0xt2)100%.

a

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56% 8% 24% 3% 3% 1% 4% 4% 0.2% 0.2% 1%

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bone scans and cardiac perfusion tests with a smaller number of studies representing each of the remaining nuclear medicine radioisotopes of interest. Details of these procedures are given in Table 2. The contribution of each procedure to total “workload” is also given in this table. Values were estimated based on procedure volume, activity, and scan time as suggested by the Canadian Nuclear Safety Commission (CNSC 2010). The facility was assumed to need to meet the ALARA dose design goals set by the Canadian Nuclear Safety Commission of 1 mSv y−1 for Nuclear Energy Workers (NEWs) and 50 μSv y−1 for members of the public (CNSC 2004). All workers adjacent to the scan room were classified as NEWs. All workers above the facility were considered members of the public, as is standard practice in Canada. The nearest neighboring workers and members of the public were taken to be 3 m from the center of the scanner bed, and a worst-case occupancy factor of 1 was assumed. To begin, all surrounding materials (e.g., scanner, walls, ceiling, etc.) were considered non-attenuating. From the procedures, occupancy, and layout described above, the unshielded dose received by a worker 3 m from the scanner bed was calculated using eqns (1)–(3) and the dose equivalent rate constants for soft tissue presented in Table 3 or drawn from the literature. Half lives were taken from ICRP 107 (2008). Unshielded doses due to each procedure, totals for 99mTc-only procedures, and totals for all procedures were determined for each dose equivalent rate constant reference. To evaluate the effects of differences in estimated barrier transmission factors, 1.59 mm (1/16”) of lead was assumed to have been added to the scan room walls, and 20 cm of solid concrete slab was assumed to have been added to the ceiling. An estimate of the “true” shielded dose received by a worker 3 m from the scanner bed through each of these barriers was obtained by analyzing the contributions of the component photon energies of each radionuclide after energy-dependent linear attenuation and buildup to dose at the point of interest. This Table 3. Dose equivalent rate constants for soft tissue and tenth-value layers in lead and concrete for seven radionuclides commonly used in nuclear medicine, derived using eqn (4).

Nuclide

Dose equivalent rate constant (μSv m2 GBq−1 h−1)

Tenth-value layer in lead (mm)

Tenth-value layer in concrete (cm)

Ga-67 I-123 I-131 In-111 Tc-99m Tl-201 Xe-133

20.85 41.57 57.44 83.14 16.10 11.29 13.87

4.89 1.23 10.28 2.33 1.08 0.89 0.40

17.74 12.58 21.23 15.17 14.32 10.50 5.58

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was achieved by implementing the spectrum-based eqns (1)–(5) and (7)–(10) using an in-house software program called ALARACAD (Kusano and Caldwell 2008) written in IDL 8.1 (Excelis Visual Information Solutions, Boulder, CO). Nuclide spectra were taken from ICRP 107 (2008); mass energy-absorption coefficients for soft tissue and mass attenuation coefficients for lead and concrete were taken from Hubbell and Seltzer (2004); geometric-progression fit coefficients and lead coherent scattering correction factors were taken from ANSI/ANS 6.4.3 (1998); and the geometric-progression formula was extended to 100 mfp using the work of Harima (1993). For comparison, lead-shielded doses were also calculated using: (1) dose equivalent rate constants from this work and transmission factors calculated using the Archer fit parameters of this work; (2) dose equivalent rate constants and first TVLs from this work; (3) air kerma rate constants and half-value layers from NCRP 124; and (4) exposure rate constants, f-factors, and TVLs from Smith and Stabin. Doses through concrete were compared with those calculated using: (1) dose equivalent rate constants from this work and transmission factors calculated using the Archer fit parameters of this work, (2) dose equivalent rate constants and TVLs from this work, and (3) exposure rate constants and f-factors from Smith and Stabin and TVLs from DIN 6844‐3, the only reference for TVLs in concrete that could be found at the time of writing. RESULTS Effect of calculation method on dose equivalent rate constant Dose equivalent rate constants for soft tissue for 67 Ga, 123I, 131I, 111In, 99mTc, 201Tl, and 133Xe are given in Table 3. Percent differences between dose equivalent rate constants found in or derived from the literature and those calculated by this work are given in Table 4. The dose equivalent rate constants provided by this work generally correspond well with the work of Wasserman and Groenewald (1988) (within 4%), Ninkovic et al. (2005) (within 15%), and Smith and Stabin (2012) (within 10%, with the exception of 99mTc, which is 28% greater than this work). There is poor agreement (up to 113% difference) between the dose equivalent rate constants calculated by this work and those derived from the air kerma rate constants published in NCRP 124. Effect of calculation method on transmission factor First tenth-value layers in lead and concrete for 67Ga, 123 131 111 I, I, In, 99mTc, 201Tl, and 133Xe are given in Table 3. Tables 5 and 6 quantify differences in the tenth-value layers for lead and concrete calculated by this work relative to published values. Tables 7 and 8 list Archer parameters and reduced chi-squared values for each radionuclide

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Nuclear medicine facility dose calculations and shielding design c M. KUSANO AND C. B. CALDWELL

Table 4. Percent difference between dose equivalent rate constants published by Wasserman and Groenewald (1988) and values derived from air kerma rate constants (NCRP, 1996; Ninkovic et al., 2005) and exposure rate constants (Smith and Stabin, 2012) and those published in this work. Values have been shaded to highlight more significant differences.

Ga-67

0%

113%

4%

0%

−4% −2% 1% −3% 4% −3%

1% 6% −55% 38% 18% −12%

−4% 1% 11% −3% 0% 15%

9% 0% 7% 28% 2% 3%

50%

a

Converted from air kerma rate constants. Converted from exposure rate constants.

b

through lead and concrete respectively, calculated by fitting the ANSI/ANS‐6.4.3‐based transmission data derived using eqn (8) to the Archer function in eqn (11). Similar to dose equivalent rate constant, there is poor agreement (up to 89% difference) between the first tenthvalue layers in lead presented here and those derived from NCRP 124 “broad-beam” half-value layers (Table 5). Values are more similar (up to 16% difference) to the tenth-value layers in lead given by DIN 6844‐3 and Smith and Stabin. The tenth-value layers in concrete calculated by this work are within 17% of those given by DIN 6844‐3 (Table 6). Figs. 1 and 2 compare transmission factors for the seven radionuclides studied in lead and concrete, respectively, calculated using ANSI/ANS‐6.4.3 buildup factor data and eqn (8), constant tenth-value layers derived from this work, constant tenth-value layers given by Smith and Stabin, and the Archer results derived from this work. TVLs, including those calculated by this work, do not Table 5. Percent difference between tenth-value layers in lead published in DIN 6844-3 (DIN 2006), NCRP 124 (NCRP 1996) and Smith and Stabin (2012) and those published in this work. Values have been shaded to highlight more significant differences. Difference from this work

a

Nuclide

NCRP 124

Ga-67 I-123 I-131 In-111 Tc-99m Tl-201 Xe-133

−55% −89% −3% 85% −8% −14% 66%

a

DIN 6844-3 −3% 11% 5% 15% −8% 5% (−25%)

Converted from half-value layers.

Difference from this work

50%

fully represent the non-linear attenuation of polyenergetic radionuclides. Attenuation is more accurately represented by the Archer fits of this work. Effect of dose equivalent rate constant and transmission factor on dose and shielding calculations: a nuclear medicine scan room example Unshielded doses due to each nuclear medicine procedure calculated using the dose equivalent rate constant references listed in Table 3 are summarized in Fig. 3. Shielded doses due to each nuclear medicine procedure calculated using various dose equivalent rate constant references and transmission factor calculation methods are compared in Fig. 4 (1.59 mm lead) and Fig. 5 (20 cm concrete). From Fig. 3, if one ignores procedures that contribute minimally to the total workload (e.g.,

Dose equivalent rate constants and barrier transmission data for nuclear medicine facility dose calculations and shielding design.

A primary goal of nuclear medicine facility design is to keep public and worker radiation doses As Low As Reasonably Achievable (ALARA). To estimate d...
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