Basic Ionization Dosimetry for Radiological Protection Management

Introduction

This paper analyzes the basic quantities along the dosimetric chain, the absorbed dose - the effective dose, in external and in-ternal exposure, to ionizing radiation, based on the nominal risk coefficients to avoid induced cancers and hereditary diseases. In this paper, we mention some aspects as follows. The paper begins with the presentation of the radiation characteristics, regarding the radiation field, particle fluence, Φ energy fluence Ψ and their rates (Φ, Ψ) necessary for defining and measuring the fundamental dosimetric quantities, exposure, X, kerma, K, and absorbed dose, D, under charged particle equilibrium and Bragg-Gray conditions. These dosimetric quantities support the management of radiation exposure after justification and optimization, in the application stage of the dose limits, ensuring that the dose (applied - limit) by calibration factors (N_X, N_K, N_D) is traceable to the national do- simetry standards PSDL. The determination of the absorbed dose in case of photon irradiation is done with the quantities resulting from the application of the Law of Attenuation and Absorption of Photons: the total linear attenuation coefficient, µ / ρ, the mass energy transfer coefficient, µtr / ρ, and the mass energy absorption coefficient, µab / ρ. For the absorbed dose from charged particles, the Law of Mass Stopping Power allows the definition of electron interaction coefficients by inelastic collisions with matter, the mass electronic (or collision) stopping power, Scol / ρ, and linear energy transfer, Lâ??, which are used to determine and measure the dose. Radiation Protection Quantities for External and Internal Exposure are also presented: Absorbed dose, DTR, in target (organ or tissue) T, multiplied by the radiation weighting factor, wR, determines the equivalent dose, HT, in target organ, T. Based on epidemiological data for each target organ, a nominal risk coefficient for cancer and hereditary effects at single or annual low dose,(≤ 100mSv), is estimated. This coefficient amplified by the equivalent dose, HT,determines the risk of cancer induction in the target. The nominal coefficient per organ divided by the sum of the nominal coeffi-cients of all the organs that compose the ensemble, represents the tissue weighting factor wT, normalized so ∑wT= 1. This, in turn, multiplied by the organ equivalent dose determines the effective dose of the considered organ, E. The paper further presents opera¬tional quantities that replace the quantities of radiation protection for external exposure i.e. when the radiation source is outside the body.Because the doses (HT, E) are not measurable in practice, the evaluation of the effective dose is made with the operational quantities, the equivalent of the area dose H* (10), and the equiv¬alent of the personal dose, HP (10).They represent, the monitoring of strongly penetrating radiation (d = 10 mm) and neutrons.The quantity H”(d, Ω),) is used for the monitoring of low penetration radiation, e.g. β particles, in the case of dose control in the lens of the eye (d = 3 mm) and dose control in the skin, hands and feet (d = 0.07 mm). For internal exposure, when the radiation source is in the body, the quantities of radiological protection are based on the law of nuclear decay. In the MIRD methodology, the internal dose, DÌ? = Ãâ?? S (rT ← rS), is the product of the physical parameter S (rT ← rS) , a tabulated “dose coefficient” and the cumulated activity, Ã =∫A(t)dt), for an infinite interval of time. In the ICRP meth¬odology, the organ equivalent dose HT and the internal effective dose E are replaced by the committed equivalent dose HT (50) in organ or tissue, T, and the committed effective dose E(50), result- ing from an incorporation of radionuclides followed by a period of 50 years old. In the formula of the equivalent dose, HT(50), enter the number of transformations in the source organ, SU(50), and the specific effective energy, SEE (T ← S). Derived amounts for the control of internal exposure to ingestion and inhalation (ALI), the activity concentration of radionuclide in breathed air by man(-DAC), the external exposure to tritium or noble gases (DAC) and the radon exposure in air (WL, WLM). Limits of the absorbed dose and the effective dose are presented. For occupational exposure, the effective dose, the value of which depends on the prevention of induction of stochastic and non-stochastic effects, is calculated by the formula, E=HP(10)+E(50). For the public, the equivalent of the annual effective dose is 2.4 mSv of which 1.6 mSv belongs to internal exposure and 0.8 mSv of external exposure due to the background of natural radiation.