Mini Review

ISSN: 2574 -1241 DOI: 10.26717/BJSTR.2021.33.005413

Comparison of Methods and Systems in Internal Radiation

Guy Yembi Goma and Muhammad Maqbool* Program, Department of Clinical & Diagnostic Sciences, the University of Alabama at Birmingham, USA *Corresponding author: Muhammad Maqbool, Health Physics Program, Department of Clinical & Diagnostic Sciences, the University of Alabama at Birmingham, USA

ARTICLE INFO ABSTRACT

Received: January 17, 2021 Exact dose delivery to cancer patients in their treatment by radiation is very Published: January 29, 2021 patients and radiations workers is performed before any dose delivery. A cancer patient canimportant. be exposed Radiation to radiation Dosimetry in two is a ways: specific external area in exposure which exact and doseinternal calculation exposure. to External exposure occurs when source of radiation is located and placed outside a Citation: Guy Yembi Goma, Muhammad Maqbool. Comparison of Methods and patient. Internal exposure is due to radiopharmaceuticals taken inside a patient’s body. Systems in Internal Radiation Dosimetry. The area of dosimetry dealing with radiation delivered by external sources of radiation is called external dosimetry and the area in which radiation is obtained from radioactive BJSTR. MS.ID.005413. sources within the body is called internal dosimetry. In this minireview, various methods Biomed J Sci & Tech Res 33(3)-2021. and systems used in the internal dosimetry are analyzed. Comparison of those methods Keywords: Dosimetry; Internal Dosimetry reveals that every method and system has its own advantages and priorities over others Methods; Internal Dosimetry Systems in various cases and circumstances. Abbreviations: S: Source Organ; CFR: Code of Federal Regulations; GI: Gastrointestinal Tract; ALI: Annual Limit of Intake, DAC: Derived Air Concentration; ICRP: International Commission on Radiology Protection

Mini Review parameters such as: the activity of the , its physical This different internal dosimetry methods were developed for and biological half-life T and T , the fractional abundance of the the purpose of radioprotection, and radiation safety to minimize p b radiation with energy E emitted per nuclear transition n , the the risk of the effects of on the people. The i i fraction of energy emitted that is absorbed in the target volume current method for calculating dose was proposed by the ICRP , the pathology or biodistribution of the radioactivity in the body [1-4]. Internal dosimetry modalities or methods are used when a i [5,6]. The equivalent and effective doses are proportional to the radioactive material enters the body by different routes inhalation, ϕ radiation and tissue weighting factors and the . The ingestion, absorption, injection on different circumstances either absorbed dose rate is given by the equation: by breathing, eating, or drinking respectively, contaminated air,

~~ ~ . Φ ∆Φ Φ −1 A∑∑ii NEiii A ii A ∑ i NEiii occupational or accidentally. In this course class paper, we describe D(..) hr orGy s= k = k = 2.13 fluids foods and wounds. These occur in different occasions, either mm m the various mathematical models for calculating the absorbed dose Ci or Bq], is the accumulated activity, (1) by diverse organs of the body. Where A ̃ [μ −1 −−11 ∆=total [rad . g .µ Ci orGy . kg . Bq . s ] 2.13Σi N i E i Internal Dosimetry Dose Calculation is the equilibrium absorbed dose constant, and , is the mass(2) of the target volume. This formula is applicable only for non- amount of energy deposited per unit mass of the body tissue. That The quantification of the absorbed dose to tissue requires the penetrating radiations ( -rays and energy comes from different type of radiations, the penetrating energy is absorbed in the target volume. For penetrating radiations and α β-rays), implying that all the (x and . Furthermore, the radiation absorbed dose is a function of lot of emissions (x and γ-rays), as well as non-penetrating ones α γ-rays), all the energy or portion of it will be absorbed by β Copyright@ Muhammad Maqbool | Biomed J Sci & Tech Res | BJSTR. MS.ID.005413. 25881 Volume 33- Issue 3 DOI: 10.26717/BJSTR.2021.33.005413

the target volume. When the source and the target are different, we 1 fraction, with , being basically the absorbed fraction and is an r 2 energy by the target tissue, then, the absorbed dose is given by the μ must insert a coefficient to account for the partial absorption of the following equation. Internationalabsorption coefficient. Commission on Radiology Protection

Φ← (ICRP)  −1 A* NEii i () T S D( rad . hr )= 2.13 m The ICRP introduced two methodologies ICRP II [10] and

Where Φ=iT()← S AF() T ← S is the absorbed fraction (3) of dose ICRP 30 [1] for internal dosimetry implementable in occupational settings, particularly in the nuclear fuel cycle with reference to organ T. For and particles, x- and -radiation of energies less coming from the source organ (S), that is absorbed by the target than 11 keV, all the energy emitted. By the radiopharmaceutical is α β γ (Figure 1). The ICRP II is the foundation of the absorbed in the volume greater than 1cm. So, will take the value i regulations in the US (Code of Federal Regulations (CFR), 10 CFR 0, unless the source S and target T are the same, . For and new ICRP 30. The ICRP II and ICRP 30 systems have to do with ϕ i 20), which was revised (10 CFR 20) in 1994 and gave birth to the particles, most non-penetrating radiation is usually absorbed, so occupational exposure and its calculation of the dose equivalent ϕ α β we set the absorption fraction . For x and -rays, penetrating 51.2*A *ξ i using the formula: H = , with ξ =∑ iinQ Φ i i m radiation with energy greater than , the value of i varies inversely ϕ γ Where n , is the quality with increasing energy and between 0 and 1, contingent on the i i (2)i ϕ factor of the radiation to get the result in . The energy. The data of i are computed by statistical Monte Carlo ϕ are defined in section 2., while Q methods based on the interaction radiation and matter [6]. constant 5.12 is the k constant that converts into rem per day, for ϕ activity A( Different Dosimetry Systems certain amount of radiopharmaceutical enters the body either by μCi), mass m(g), and energy E(MeV). At time t = 0, a inhalation or The previous formulas have been derived using lot of dose for radiopharmaceutical with complex emission spectra. respiratory system models are used to compute the transfer of simplifications and are the most commonly used to calculate ingestion. For the gastrointestinal (GI) tract as well as the These dosimetry systems that seem to look different, where some radionuclide from the GI or the respiratory system to the body parameters have been combined, may look different, but yield the same output given the same input and assumptions. fluids as well as its excretion. Then, by either pathway of intake, the enters the body through compartment a which is connected to Marinelli-Quimby Methods radionuclide enters the body fluid system [11]. The radionuclide different compartments, b, c, d, etc. representing type of tissues The equation for the dose of non-penetrating beta ( or organs of the body, where the radiopharmaceutical experiences that decays completely in a body tissue is given by the equation: β) emitter [7,8]. biological clearance and physical decay, and finally goes out of the Dββ= 73.8* CE * * T to be uniform in the reference man. The ICRP metabolic design body. The dose distribution in the first compartment is assumed uses mathematical models for the reference man to track the where: Dβ [] rad is the concentration of(4) the radionuclide, radiopharmaceutical as it moves from the original compartment C[µ Ci .] g −1 is the concentration of the nuclide, E[] MeV is the β to other tissues or organs. The different compartment models mean energy emitted per decay of the nuclide, is the half-life of are: ICRP-30 Dosimetric Model for Respiratory System (ICRP the nuclide in the tissue. By analogy with the cumulative activity, ~ 73.8 A=1.44* fAT * * ,we see that k = = 51.1 , and C[µ Ci .] g −1 is the the Gastrointestinal Tract, and ICRP-30 Dosimetric Model for 0 1.44 Human Respiratory Tract Model), ICRP-30 Dosimetric Model for activity per unit mass, and for the emitter, the absorbed fraction Submersion in a Radioactive Gas Cloud. The solutions of these =1. For penetrating radiations such as -rays, we use the geometric computations are the equivalent dose and effective dose rates of β factors of Brownell and Hine [9] for spheres and cylinders of set various tissues and organs as a function of time. Then, the committed ϕ γ shape to calculate the data for the fraction of energy emitted that equivalent dose and the committed effective dose subsequent from is absorbed in the target volume. The dose in the vicinity of the the original intake can be evaluated, as well as the ALI (Annual ϕ

γ-emitter is given by the formula: e−µr  rad LimitWe of calculateIntake), the the DAC committed (Derived effective Air Concentration) dose for a [11].radionuclide D=10−3τ * c dV y ∫ r2  hr in the body with the ICRP mathematical tools and data from the

− reference man. The committed effective dose is given for a period where C[µ Ci .] g 1 is the activity of the gamma emitter. Similarly, (5) = 50 years by the formula: E(ττ )=∑⇔=∑ wH ( ) E (50) wH -rate constant is the exposure rate per TTT TTT disintegration from the point source (same as k**Σ nE τ by analogy, the specific γ Γ ii i e−µr where w is the weighting factor for the tissue, and H () (6) T T ∫ r 2 plays a role of the absorbed) to an (τ) is the infinite medium, while the factor committed equivalent dose of tissue T given by equation (7) Copyright@ Muhammad Maqbool | Biomed J Sci & Tech Res | BJSTR. MS.ID.005413. 25882 Volume 33- Issue 3 DOI: 10.26717/BJSTR.2021.33.005413

. t0 +τ H()τ = HT dt Tt∫ 0 expression: of the activity-time curve for a source region) and is given by the where (7) k∑Φ nE = iii i members of the public [12]. For a target organ T, we can calculate DF τ= 50 years for occupational use, and τ= 70 years for the m and sum the respective committed equivalent dose from all organs, Mathematically speaking is similar to the(13) mean dose per known as source organ S.

AF() T← S R cumulated activity defined in the MIRD system by the equation

∧ (11) in 3.3 . The RADAR team put together collections of decay data, (8) occupational workers and nuclear medicine patients. Furthermore, where HT()← S dose conversion factors, and classified, identical dose models for per disintegration of the radionuclide in S. U is the number of they developed a computer code, OLINDA/EXM that works with is the equivalent dose in Sthe target T(mean) disintegration of the nuclide in the source organ S for the 50 years of the committed equivalent dose. The unique difference between Specificformulas (12), Absorbed (13) and Fraction,input data from Specific RADAR Effective site [15]. Energy, the dose calculated with ICRP II in one hand and ICRP 30 including and Committed Quantities MIRD on the other hand, is that the committee of ICRP II used a very simple phantom, the sphere ( all body organs and whole were radiation emitted per disintegration by a radionuclide in source The specific∧ equivalent dose in a target organ T caused by the organ S is HT()← S represented by spheres). The source and the target were the same. Medical Internal Radiation Dosimetry (MIRD) System ∧ S , target organ T and. For the akind specific of radiation, combination we derive of the the source value organ of H The MIRD system has been extensively studied in class, so, from the fraction of the energy emitted in the organ S and absorbed here we are going write down the main equations that govern that in organ T denoted , where R represents the type of radiation. The system [13]. The absorbed dose in the MIRD system is given by the set of equations below: specific absorbed fraction is the ratio of over the mass of the target ~ publishes by the ICRP. The values of for different type of radiations D= AS* , is . Tables with value of the is the same as in equation (3) are are the same as in section 2 [11]. The Medical Internal Radiation This equation means that the Average or Mean Dose (9)D is the product of the cumulative activity ~ by Mean Dose per activity S. fraction for different organs and targets in the Reference Man A Dosimetry (MIRD) Committee have computed the absorbed ~ using Monte Carlo techniques based of the transport of photons A∑∆ DS= ii* m t body. Very often, a radiopharmaceutical in source organ S emits ~ (monoenergetic and poly-energetic beam) through the human the average or mean dose is the ratio of the term A ∑∆ (10)ii by the several kind of radiation R, with a yield YR with a mean energy ER. mass of target mt [g]. ← The product of YRER by AR() T S R gives the mean absorbed dose in

~ the target T per disintegration in source S by the type of radiation R. −−1 1 A∑∆ii[ rad .. gµ Ci . hr ]is the product of the cumulative activity ~ ∑∆ µ −−11 A ii[rad .. g Ci . hr ], the sum of all the equilibrium absorbed dose constants for all radiation emissions i. The ICRP expresses the specific effective energy (SEE) transmitted [ μCi.hr] by the source organ S per disintegration as follows [12], per gram of tissue in T from the emission of a specific radiation R in ~ A∑ Φ∆ ← ()i ii AF() T S R −1 SEE() T←= S R YRRR E w,[ MeV . g ] m −−1 1 = t =∑ Φ∆ µ mT S ~ i ii[rad .. g Ci . hr ] A Where w is the weighting factor of radiation R(14) emitted per (11) R Φ disintegration of a radionuclide in source S to the equivalent dose in where Φ= Equationmt (11) gives the Mean dose per cumulated activity, target T. To get this expression in radiation protection unit, we time the SEE by the factor: Radiation isDose called Assessment specific absorbed Resource fraction. (RADAR) −13 −1 − 3 − 1 −10 −−1 1 The Radiation Dose Assessment Resource set up a website (1.60*10J . MeV /10 kg . g )= 1.6*10Sv ( MeV . g ) www.doseinfo-radar.com. The introduction of the internet gave Performing the discrete sum for all kind of radiation emitted by of publications on data and procedures used in the system. The the scientific community the impulse to disseminate a number organ T, given below. RADAR system is ruled by the equation [14]. the radiopharmaceutical gives a specific equivalent dose in target Λ HTS(←= ) 1.6*10−10 ∑SEE ( T ← S ) ,[ Sv ] D= N* DF RR The committed equivalent dose is given by multiplying (15) the where and are respectively the number (12) of transformations number of disintegration U going on in the organ (the number of disintegrations is the area S dose above: in 50 years by the specific equivalent

Copyright@ Muhammad Maqbool | Biomed J Sci & Tech Res | BJSTR. MS.ID.005413. 25883 Volume 33- Issue 3 DOI: 10.26717/BJSTR.2021.33.005413

H (50)= 1.6*1010 ∑∑U SEE(),[] T ← S Sv 3. T SRR R Investigation of Internal Dosimetry Methodologies. Journal of Nuclear Lehnert W, Schmidt K, Kimiaei S, Bronzel M, Kluge A (2016) Comparative The committed effective dose obtained by multiplying (16) the committed equivalent dose by the sum of all weighting factors for 4. Medicine 57(2): 307. different tissues or organs: Sgouros G (2005) Dosimetry of Internal Emitters. Journal of Nuclear 5. Medicine January 46(1): 18-27. −10 E(50)= 1.6*10 ∑∑∑TTw S U S R SEE()[] T ← SR, Sv York,Michael USA. G Stabin (2007) Radiation Protection and Dosimetry: An Suppose we have multiple radiopharmaceuticals in the(17) body, Introduction to Health Physics. In Michael G Stabin(Eds.) Springer, New 6. therefore, to get the total committed effective dose, we add the 3rd individual contributions [12], Saha GB (2006) Physics and of Nuclear Medicine. In Saha 7. GB (eds.) (edn.) Springer, New York, USA. −10 radioactive isotopes II, practical considerations in therapy and E(50)= 1.6*10 ∑∑TTSSRw [ U ∑ SEE ( T ← S )Rj ] ,[ Sv ] Marinelli L, Quimby E, Hine G (1948) Dosage determination with

nd Conclusion (18) 8. protection. Am J Roent Radium Ther2 59(2): 260-281. andQuimby Febiger, E, FeitelbergWashington S Square, (1963) USA. ( Edn.). Radioactive isotopes in It has been long journey exploring, learning about the various medicine and biology. In Quimby E, Feitelberg S (eds.) Philadelphia, Lea internal dosimetry systems that are used to monitor, evaluate 9. exposure for occupational workers as well as the public. All these Hine G, Brownell G (1956) Radiation dosimetry. In Hine G, Brownell G dosimetry systems are crucial for protecting radiation workers 10. (eds.) Academic Press, New York, USA. radiation. and the public in order to prevent acute, rare effect of radiation, as ICRP (1960) Report of committee II on permissible dose for internal 11. well as minimizing the risk for long-term effects. Finally, we protect the people, also, by verifying adequacy at workplace controls and James E Turner (2007) Atoms Radiation and Radiation Protection. In 12. James E Turner (eds.) Wiley, New York, USA. demonstrating regulatory compliance. Statement of the , Mc Lean, Virginia. Health Physics Society (2004) Radiation Risk in Perspective. Position 13. References Dose Calculations. Society of Nuclear Medicine . 1. Loevinger R, Budinger T, Watson E (1988) MIRD Primer for Absorbed the ICRP 2: 3-4. 14. IRCP (1979) Limits for Intakes of by Workers. Annals of 2. Stabin MG, Siegel JA (2003) Physical Models and Dose Factors for Use in European Journal of Nuclear Medicine 23: 491-493. 15. Internalwww.doseinfo-radar.com Dose Assessment. Health Physics 85(3): 294-310. Mountford PJ (1996) Internal dosimetry: developments and limitations.

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