Published on October 21, 2007
Radiation Protection: Principles, Methods and Monitoring Moderator: Mr Anoop Bharadwaj
Retrospective • “..the surgeons of Vienna and Berlin believe that the Roentgen photograph is destined to render inestimable services to surgery.” – JAMA, 1896 • “(Radiation is) the most serious agent of pollution of the environment and the greatest threat to man’s survival on earth.” – E.F. Shumacher, in Small is Beautiful, 1973
Introduction • The acceptance by society of the risks associated with radiation is conditional on the benefits to be gained from the use of radiation. • Nonetheless, the risks must be restricted and protected against by the application of radiation safety standards. • It is therefore essential that activities involving radiation exposure be subject to certain standards of safety in order to protect the individuals who are exposed to radiation: – Occupationally, – For medical diagnostic or therapeutic purposes – As members of the public.
Effects of Radiation Exposure Radiation Exposure Stochastic Effects / Non stochastic effects/ Probablistic Deterministic ➢ Have no threshold levels of ➢ Have definite threshold levels radiation dose. of radiation dose. ➢ The probablity of the effects ➢ The probablity of the effects is proportional to the dose. is proportional to the dose. ➢ A latent period is seen ➢ A latent period is seen between the time of between the time of exposure and the events to exposure and the events to manifest manifest ➢ Severity independant of dose ➢ Severity may be proportional received to the dose received. ➢ Seen when the cells are ➢ Seen when the cells are killed modified rather than killed. or loose capability to divide.
Effects if Radiation Exposure Deterministic Stochastic, Quadratic Stochastic, Linear
Stochastic Effects • These are primarily of two types: – Carcinogenesis – Hereditary effects • Both have: – A random nature of appearance. – No threshold dose for appearance. – Definite latent period for appearance after exposure. – Probablity of induction increases with the dose received. – Severity of the effect is independent of the dose received. – A risk that can be defined on epidemiological studies only.
Radiation Induced Carcinogenesis • Data from BIER V (US population) and the UNSCEAR (Japanese Atomic Bomb survivors) show that: – Relative risk of radiation induced cancers is a linear function of doses upto 2 Sv – At lower dose range of 0 -0.5 Sv risks are slightly higher than the extrapolated risk. – Risk of radiation induced cancers varies with age with patients at younger age being more susceptible. – Females < 15 years are most susceptible.
Risk Estimates for Carcinogenesis • Cancer accounts for ~ 25% of all deaths in developed nations • The annual risk of fatality from cancer is shown in the table below. • ICRP estimates that for each radiation induced cancer 13 -15 years of life will be lost (but most will occur at ages of 68 – 70 yrs) Population High Dose, High Dose Low Dose, Low Dose Rate Rate Working Population 8 x 10 per Sv 4 x 10 -2 per Sv -2 Whole Population 10 x 10 -2 per Sv 5 x 10 -2 per Sv
Radiation induced Hereditary effects • Radiation induced hereditary effects are secondary to mutations which are passed on to the progeny • Radiation doesnot cause new types of mutations but increases the frequency of naturally occuring mutations • Three classes of hereditary effects are known: – Gene mutations – Chromosomal abberations ( Down's Syndrome) – Multifactorial (Neural tube defects) • ICRP estimates that risk of hereditary effects due to radiation exposure is: – 0.2% per Sv in general population – 0.1% per Sv in working population
Radiation induced Mutations • Radiosensitivity of different mutations vary widely – so an average mutation risk is considered. • Low dose rate radiation is less effective in inducing mutations • Time interval between exposure and conception plays an important protective role. – A period of 6 months is therefore recommended between radiation exposure and conception. • Radiation induced mutations can be transmitted across the generations. • Average “doubling dose” for humans is considered to be 1.56 Sv
Risk estimates of Hereditary effects Disease Class Base Frequency First genera- 2nd generation per million live tion risk per risk per million births million live live births births Mendelian mutations 16500 750 – 1500 1300 – 2500 Chronic multifactorial 65000 250 – 1200 250 -1200 diseases Congenital abnormal- 60000 2000 2400 – 3000 ities Total 738000 3000 – 4700 3950 – 6700 Total risk per Gy NA 0.41 – 0.64 % 0.53 – 0.91 %
Fetal Effects • Radiation risks to fetus are related to: – Exposure magnitude – Time of pregnancy • Radiation risks are most significant during organogenesis and in the early fetal period. • Threshold for malformations: – 100 – 200 mGy (Malformations) – 100 mGy (Mental Retardation): Risk coeffecient is 0.4 per Sv • A exposure 1mSv is safe for a fetus – normal exposure from background radiation • A dose > 0.1 Sv is considered the threshold beyond which an MTP should be considered.
Fetal Effects • Time of radiation vs effect: – 2 – 3 weeks: Most embryos are aborted – 4 – 11 weeks: Severe abnormalities in most organs – 11 – 16 weeks: Mental retardation and stunting more common – 16 – 25 weeks: Mild degree of mental retardation and microcephaly – > 30 weeks: Usually leads to functional disabilities in later life • Carcinogenesis: – Most exposures implicated occur in 3rd trimester – Doses more than 1 mSv will increase the risk – but no threshold apparent. – Excess absolute risk – 6% per Sv.
ICRP quantities for Stochastic effects • The concept of detriment as recommended by the ICRP for stochastic effects includes the following quantities: – The probability of fatal cancer attributable to radiation exposure; – The weighted probability of incurring a non-fatal cancer – The weighted probability of severe hereditary effects – The length of lifetime lost, if the harm occurs.
Safety Standards • These are defined based on the knowledge of the knowledge of effects of radiation and the principles of radiation protection. • The UNSCEAR (United Nations Scientific Committee on the Effects of Atomic Radiation, 1955) is the central body which collects the data regarding this. • The IAEA uses the data from ICRP in designing the safety standards. • Consensus guidelines on these safety standards are published as International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (referred to as BSS) • Latest version – 1996 (IAEA safety series 115)
Types of Radiation Exposures
ICRP Reference Man • The concept of a ‘reference human’ to help manage the many different situations in which human beings would or could be exposed to ionising radiations. • ICRP published comprehensive report on Reference Man in 1975 (ICRP, 1975) – recent revision is in ICRP 2002. • The purpose of Reference Man was: – Create points of reference for the procedure of dose estimations to humans. – Derivation of relevant quantities and units for their interpretation – Considering the relationships between doses to different parts of the human body and their effects. • Most countries have modified the concept of reference man – e.g. Indian Reference Man
ICRP Reference Man • ICRP 89 has given definitions and data for reference humans of both sexes in 6 different age groups: – Newborn – 1 Yr – 5 Yr – 10 Yr – 15 Yr – Adult • Reference man is defined as being between 20-30 years of age, weighing 70 kg, is 170 cm in height, and lives in a climate with an average temperature of from 10 to 20 C. He isa Caucasian and is a Western European or North American in habitat and custom.” • Recently weight has been revised to 73 Kgs.
Indian Reference Man *Jain S.C., Mehta S.C., Kumar B., Reddy A.R. and Nagaratnam A . (1995) Formulation of the Reference Indian Adult: Anatomical and Physiological Data. Health Phys. 68, 509-522.
Types of exposure • 2 types of exposure have been defined by the ICRP: – Normal exposures – Potential exposures • Humans can be exposed to radiation while: – Persuing their normal occupation (Occupational exposure) – Undergoing some medical or dental procedures for diagnosis or evaluation (Medical exposure) – During normal daily life (Public exposure)
Normal Exposures • These are defined as the radiation exposure that will occur as a matter of course in the industry or in the medical procedures. • The exposure will have a predictable magnitude albiet with certain degree of uncertainity. • Controlling the doses delivered in these normal exposures is the method specified by IAEA.
Potential Exposures • These are exposures that are feasible but not certain to occur. • These can become actual exposures if there is equipment malfunction, design flaws of operating failures. • The primary means for controlling potential exposures is by optimizing the design of installations, equipment and operating procedures with the following aims: – To restrict the probability of occurrence of events that could lead to unplanned exposures – To restrict the magnitudes of the exposures that could result if such events were to occur.
Occupational Exposures • Occupational exposure which is defined as all exposures of workers incurred in the course of their work (with the exception of exposures excluded from the BSS and exposures from practices or sources exempted by the BSS).
Medical Exposures • Medical exposure which is defined as exposure incurred: – By patients as part of their own medical or dental diagnosis or treatment – By persons, other than those occupationally exposed, knowingly while voluntarily helping in the support and comfort of patients – By volunteers in a programme of biomedical research involving their exposure. • Three broad classes: – Patients – Caregivers – Medical test subjects
Public Exposure • Public exposure, which is defined as exposure incurred by members of the public from radiation sources, – Excludes any occupational or medical exposure and the normal local natural background radiation – But includes exposure to authorized sources and practices and from intervention situations.
Radiation Protection Quantities
Quantities for radiation protection • The absorbed dose is the basic physical dosimetry quantity, but it is not entirely satisfactory for radiation protection purposes because the effectiveness in damaging human tissue differs for different types of ionizing radiation. • So other qunatities that have been proposed to quantify the biological and physical effect of the radiation dose are: – Organ dose (Mean Physical Dose, DT) – Equivalent dose (Modifier for radiation type, HT) – Effective dose (Modifier for organ type, E) – Committed dose (Modifier for the time of exposure) – Collective dose (Modifier for the population exposed)
Organ Dose • The organ dose (DT) is defined as the mean dose in a specified tissue or organ of the human body. • Unit is cGy / Kg or joules / Kg
Equivalent Dose • The extent of biological effect produced by radiation depends on: – The physical organ dose – The characteristics of the radiation beam employed. – The pattern of dose deposition. • The equivalent dose (HT) is the organ dose multiplied by a radiation weighting factor wR to account for the effectiveness of the given radiation in inducing health effects. If more than one type of radiation is used.
Equivalent Dose • The SI unit for equivalent dose is joules/kg or sievert • One Sievert = 100 rem (radiation equivalent man) • Older ICRP recommendations used the quantity of dose equivalent (H) • This was the dose to the a point in the organ multiplied by the radiation weighting factor (wr). • Radiation Weighting factor is a somewhat arbitrarily chosen conservative value based on a range of RBEs related to the linear energy transfer (LET) of the radiation.
Radiation Weighting Factors Radiation Type Weighting Factor Photons all energies 1 Electrons all energies 1 Protons, other than fission fragments, energy more than 2 MeV 5 Neutrons energy < 10 KeV 5 10 KeV to 100 KeV 10 100 KeV to 2 MeV 20 > 2 MeV to 20 MeV 10 > 20 MeV 5 Alpha particles, fission fragments, heavy nuclei 20 1) Quality factors for various radiation types; Source: ICRP publication 60 2) Excludes Auger electrons 3) All values relate to radiation incident on the body for external radiation and for internal sources, emitted from the source
Effective Dose • The probablity of stochastic effect in a tissue depends on the type of tissue irradiated. • For the same equivalent dose the probablity of stochastic effects in two organs will vary – so the concept of effective dose. • Previously the concept of Effective dose equivalent was used which defined the effective dose to a point (as in Dose Equivalent)
Effective Dose: Definition • The Effective dose (E) is defined as the summation of tissue equivalent doses, each multiplied by the appropriate tissue weighting factor (wT), to indicate the combination of different doses to several different tissues in a way that correlates well with all stochastic effects combined (ICRP Publication 60) • The unit for effective dose is also Joules / Kg or Seivert (Sv)
Effective Dose: Tissue Weighting Factors • Tissue weighting factors wT should represent the relative contribution of an organ or tissue to the total detriment due to the effects resulting from a uniform irradiation of the whole body. • They represent the likely proportionate risks of stochastic events when tissues are irradiated. • For low doses, individual organ or tissue detriments can be treated as additive and the total detriment to the whole body is the summation of individual detriments. • For the whole body wT = 1 ; so a uniform equivalent whole body radiation dose has a effective dose E = 1 x HT = HT
Tissue Weighting Factors Tissue Weighting Factors For purposes of calculation, the remainder is composed of the following Gonads 0.20 additional tissues and organs: Breast 0.12 Adrenals • Red Bone Marrow 0.12 Brain, • Colon 0.12 Upper large intestine, • Lung 0.12 Small intestine, • Kidney, Stomach 0.05 • Muscle, • Breast 0.05 Pancreas, • Bladder 0.05 Spleen, • Liver 0.05 Thymus and • Esophagus 0.05 Uterus • These organs are choosen as they are Thyroid 0.05 likely to be selectively irradiated. Skin 0.01 Bone Surface 0.01 Remainder 0.05 Total 1.00 Source ICRP Publication 60
Tissue Weighting factors Those exceptional cases in which a single one of the remainder tissues or organs receives an equivalent dose in excess of the highest dose in any of the twelve organs for which a weighting factor is specified, a weighting factor of 0.025 should be applied to that tissue or organ and a weighting factor of 0.025 to the average dose in the rest of the remainder as defined above.
Effective dose: Use and Limitations • Effective dose is a useful concept as: – It measures the degree of harm from a given dose of radiation – It can be used to compared different types of radiation – Can be used to compare the dose from various types of exposures. • Limitation: – Tissue weighting factors are defined based on flawed estimates of stochastic effects • For the whole body the ICRP recommends the use of Effective doses for defining safety standards e.g. annual dose limits. – For individual organs and extremities, however equivalent doses are used.
Committed Dose • Useful concept for nuclear medicine • Relates to the time which a radionuclide will remain inside the body. • It is defined as the specific time integral of the dose received by the body after exposure to a specific radionuclide. • A committed effective dose and equivalent dose may also be defined by multiplying the tissue and radiation weighting factors with the committed dose.
Collective Dose • The collective dose relates to exposed groups or populations • Is defined as the summation of the products of the mean dose in the various groups of exposed people and the number of individuals in each group. • Can be defined for effective (E) or equivalent doses (H). • The unit of collective dose is the man-sievert (man-Sv). + + E1 x 6 E2 x 4 E3 x 9
Operational Quantities • The organ dose (DT), equivalent dose (H) and effective dose (E) are not directly measurable and there are no laboratory standards to obtain traceable calibrations for the radiation monitors using these quantities. • For this reason, the ICRU has defined three measurable operational quantities for protection purposes: – Ambient Dose equivalent – Directional Dose equivalent – Personal Dose equivalent
Ambient Dose Equivalent • The ambient dose equivalent at a point in a radiation field is the dose equivalent that would be produced by the corresponding aligned and expaned radiation field in the ICRU sphere at the depth d at the radius opposite to the direction of the radiation field. – The ICRU sphere is a 30 cm diameter tissue equivalent sphere composed of: • 76.2% Oxygen • 11.1% Carbon • 10.1% Hydrogen • 2.6% Nitrogen – depth d = 10 mm is recommended for penetrating Radiation.
Directional Dose Equivalent • The directional dose equivalent at a point in a radiation field H (d,Ω) is defined as the dose equivalent that would be produced by the corresponding expanded field in the ICRU sphere at depth d on a radius in a specified direction Ω. • For weakly penetrating radiation the depth of measurement recommended is 0.07 mm.
Ambient and Directional Dose equivalent a Radiation Field d ICRU Sphere
Personal Dose Equivalent • The personal dose equivalent Hp(d) is defined for both strongly and weakly penetrating radiations as the equivalent dose in soft tissue below a specified point on the body at an appropriate depth d. • The relevant depth is: – 10 mm for photon energies above 15 KeV – 0.07 mm for photon energies below 15 KeV and β particles in skin – 3 mm for photon energies below 15 KeV and β particles in the lens • The personal dose equivalent from exposure to penetrating radiation during the year is the radiation quantity to be compared with the annual dose limits (for effective dose).
Background Radiation Exposure
Background Radiation • The total effective dose equivalent for a member of the population in the United States from various sources of natural background radiation is approximately 3.0 mSv/year (300 mrem/year). • The sources of background radiation include: – Cosmic Radiation – Terrestrial Radiation – Internal Radioactive nuclei • The estimated total annual exposure is estimated to be 3 mSv.
Natural Radiation: Estimated Doses Lung Gonads Bone Surfaces Bone Marrow Other Tissue Total W 0.12 0.25 0.03 0.12 0.48 1 T Cosmic 0.03 0.07 0.01 0.03 0.13 0.27 Cosmogenic 0 0 NA 0 0 0.01 Terrestrial 0.03 0.07 0.01 0.03 0.14 0.28 Inhaled 2 NA NA NA NA 2 In the body 0.04 0.09 0.03 0.06 0.17 0.4 Estimated total effective dose equivalent rate for a member of the population in the United States and Canada from various sources of natural background radiation
Radiation Protection Principles
Basic Principles of Radiation Protection Radiation Protection Dose Justifiable ALARA Limits Exposure
Protetion Principles: Justifiable Use A practice that entails exposure to radiation should only be adopted if it yields sufficient benefit to the exposed individuals or to society to outweigh the radiation detriment it causes or could cause (i.e. the practice must be justified). • Example: – Benefit gained from the use of radiation therapy for treatment of Ptyrisasis versicolor or thyroid goitre doesnot allow the use of therapeutic radiation for these treatments – Use of x-ray pelvimetry has significant hazaards for the fetus so is no longer used routinely although may be useful for the rare mother with some pelvic deformity
Protection Principles: Dose Limits • Individual doses due to the combination of exposures from all relevant practices should not exceed specified dose limits for occupational and public exposure. • Different dose limits are specified for the radiation workers as the expected benefit from the work they do while handling radiation will outweigh the small increase in risk. • Pregnant radiation workers have to be protected so that the fetus or embryo is given the same radiation protection as given to that of the public. – Dose limits are not relevant for control of potential exposures – nor are they relevant for descisions on whether or how to undertake an intervention. – Dose limits are not applicable to medical exposure as the benefit gained from the exposure outweighs the harm.
Protection Principles: Dose Limits • Examples: – Individual dose for a person who is delivering a letter to the Co60 room would be counted in the quantum of radiation exposure he had but not the radiation exposure after an x-ray he had to have for the ankle after he slipped on the wet floor while going out. – If a pregnant female is given radiation for a therapeutic indication then the dose to her is not counted but dose to the fetus is counted. – When a radiation worker becomes pregnant care should be taken that during the period of gestation the exposure to the fetus doesnot exceed that of a general person – note the limits no longer apply after gestation unless she is working in nuclear medicine where radionuclides can be absorbed by the body – hence can be applicable for lactation.
Protection Principles: ALARA Radiation sources and installations should be provided with the best available protection and safety measures under the prevailing circumstances, so that the magnitudes and likelihood of exposures and the numbers of individuals exposed be As Low As Reasonably Achievable (ALARA), economic and social factors being taken into account, and the doses they deliver and the risk they entail be constrained (i.e. protection and safety should be optimized.)
Protection Principles: ALARA In diagnostic medical exposure, optimization of protection is • achieved by keeping the exposure of patients to the minimum necessary to achieve the required diagnostic objective – Thus in Obsteterics Xrays or CT Scans is better replaced with clinical examination, USG or MRI. – A skeletal survey is thus used for multiple myeloma but not for other conditions like TB. In therapeutic medical exposure, optimization is achieved by • keeping exposure of normal tissue ALARA consistent with delivering the required dose to the planning target volume (PTV). As a thumb rule a value of $ 1000 / 10mSv reduction in • exposure is considered reasonable. ( $ 10,000 in some special situations too.)
Radiation Protection History and Evolution
Radiation Protection: Dawn • 1904: William H. Rollins, a Harvard-trained physician and a practicing dentist, reported that the hazards of x-rays could be reduced by using simple shielding methods. • 1913: German Roentgen society developed 1st guidelines to shield x-ray technicians from excessive exposure. • 1928: The 2nd International Congress for radiology established the International X-ray and Radium protection committee. – 1929: American Medical Association condemned the use of radiation to remove hair. – 1932: American Medical Association withdrew radium from the list of remedies used for internal administration.
Radiation Protection: Tolerance Doses • 1934: The International X-ray and Radium Protection Committee along with it's American counterpart Advisory Committee on X-ray and Radium Protection gave a qunatitative tolerance dose of external radiation. – Based on research on time to cause erythema – Recommended limits: • 0.1 R per day for the whole body American Committee • 5 R per day for the fingers • 0.2 R per day whole body limit – International Committee – The discrepancy was the result of the difference in rounding off of similiar data by the two committes. • 1941: The American committee defined the tolerance dose from internal deposition of radium and radon.
The Birth of the Atomic Age
Radiation Protection in the Atomic Age • Became more important as: – Increasing use in the public, health and military domains made radiation exposure more common – Different aritficial radio-isotopes were generated with new characteristics. – Protection no longer was confined to radium and xrays. • The American Committee was changed to National Committee on Radiation Protection (NCRP) in 1946. • The international Committee was changed to International Committee on Radiation Protection (ICRP) similiarly. • Based on the work of the geneticist H.J.Muller who had indicated that the reproductive cells were vulnerable to even smallest doses of radiation the concept of tolerance doses were abandoned.
Permissible Dose • The concept of tolerance dose indicated that there was a level of radiation below which it was safe. • The concept of stochastic effects of radiation invalidated this dogma • Most scientists rejected that there was a threshold below which exposure to radiation was biologically innoncous. • The concept of permissible dose was therefore introduced.
Maximum Permissible Dose (MPD) • The NCRP defined the Maximum Permissible Dose as that dose which, in the light of present knowledge is not expected to cause appreciable bodily injury to the person at any point during his lifetime. • Advantages: – Explicit acknowledgement that doses below MPD have a risk of detrimental effects. – Acknowledged danger due to stochastic effects of radiation. – Introduced the concept of acceptable risk – probablity of the radiation induced injury was to be kept low to be easily acceptable to the individual • Allowed different levels for radiation workers and public – Allowed modifications in the advent of new knowledge.
Maximum Permissible Dose: NCRP • The NCRP initially defined a dose level of 0.3 R per six day work week for whole body external radiation exposure (agreed on in 1948; published in 1954). • Also proposed a detailed set of recommendations for permissible levels of internal exposure from radio-isotopes taken in the body. • To provide an adequate margin of safety, it proposed permissible levels as low as 1/10th of the numerical values derived from the sketchy data then available. • A seperate limit of 1/10th the adult dose (0.03 R/week) was proposed for the minors < 18 yrs age.
Maximum Permissible Dose: ICRP • The ICRP also used the proposal of MPD as given by the NCRP and used the same dose limits (0.3 R/week). • The one major area of difference was that it proposed the dose of 1/10th the level of exposure for the radiation worker, for the general public (0.03 R/week). • This level was first given in 1953, to account for the risk that the general population had from accidental exposure to radiation. • The ICRP's effort was the first attempt to produce radiation protection guidelines for the general population outside controlled areas where radiation workers worked.
Change in MPD levels • Several changes were introduced by the NCRP and ICRP after the 1960s to account for the increased exposure from the radioactive fallout from above ground nuclear bomb explosions. – Reduced the permissible doses for whole body exposures by 1/3rd. – Introduced the rad as the unit for dose rather than roentogen – Also introduced the term rem – radiation equivalent man – for chronic low level radiation exposure – The concept of RBE – Relative Biological Effectiveness – was introduced to account for the difference in the biological effects produced by different types of radiation.
Maximum Permissible Doses: ICRP • In 1956, the ICRP published the revised limits for MPD in terms of rem – Maximum permissible dose limit for occupational exposure was kept at 5 rem per year. – Defined the MPD for the general population as 0.5 rem per year. – In addition defined dose levels that could be accumulated at various ages: • 50 rems to age 30 years • 100 rems to age 40 years • 200 rems to age 60 years. – Thus the main thrust was providing additional protection for the younger population who were considered to be at greater risk for tranmiting genetic damage due to radiation.
Maximum Permissible Dose: NCRP • The revised NCRP recommendations were published in 1958. – Recommended average whole body radiation exposure from external sources to 5 rem per year. – Allowed a maximum limit of occupational exposure of 12 rem in a year if • Dose to person did not exceed the limit of 5 rems in the previous year. • Adequte past record of exposure existed. – Also gave a formula to calculate the MPD as per the age: • Accumulated MPD = 5 (Age in years - 18) – Recommended that doses to persons living close to radiation sources but outside controlled areas not to exceed 1/10th of those proposed for occupational exposure.
MPD – Radionuclides • In 1959 both ICRP and NCRP met and recommended the permissible body burden of radionuclides. • The recommended dose limits were: – 0.5 rem per year for an individual (1/10th of the occupational limit) – 0.17 rem per year for the entire population ( 1/30th of the occupational limit)
As Low as Practicable (ALAP) • 1st proposed by Forrest Western, Director of the Division of Radiation Protection Standards, AEC, USA (published in 1970) • Applied primarily to low level radiation exposure due to leakage from the Nuclear Reactors. • Basically gave the axiom that the radiation exposure was to be kept as low as possible practically from these reactors. • The changes were introduced in an effort to tighten the limits for effluent generated by the nuclear power plants – however concrete dose limits could not be specified.
Radiation Protection Modern Day
As Low As Reasonably Acheivable (ALARA) • Initally defined by the NRC (1975) for the nuclear power plants. • Defined as: – A level as low as reasonably achievable taking into account the state of technology, and the economics of improvements in relation to beneﬁts to the public health and safety, and other societal and socioeconomic considerations, and in relation to the utilization of atomic energy in the public interest. • Primarily designed to make radiation protection for the nuclear power plants economically acheivable – Example: It is practical to shield a reactor inside a shield of tungsten but not acheivable economically so concrete has to be used.
Negligible Individual Risk (NIRL) • Given by the NCRP only. • Defined as the dose level at which the risk of annual fatal health effects due radiation is so low that efforts to reduce the exposure to the individual is unwarranted. • The NCRP defined this limit as there was a need to define the threshold of the lower risk below which efforts to reduce the risk would not be warranted. • However doesnot contravene the principles of permissible dose limits and ALARA. • The risk level of radiation induced fatal events choosen for setting the NIRL is 10-7. • This limit is specified for members of public only.
Rationale behind dose limits • The risk for stochastic effects is taken into consideration while designing the dose limits for occupational and public exposure. • These limits represent the dose levels at which the risk for detriment are unacceptable (old reports had the dose levels which were harmful or fatal). • These represent the highest tolerable limits for continous exposure. • For the purpose of computations the risk coeffecients are defined based on the risk per unit equivalent dose: – Total risk coeffecient : 165 x 10-4 Sv-1 – Somatic risk component: 125 x 10-4 Sv-1 – Genetic risk component: 40 x 10-4 Sv-1 (Risk of severe genetic damage over first 2 generations)
Modern Day Levels: ICRP The maximum effective dose for occupational exposure is 50 mSv in 1 year and the total exposure for 5 years should not exceed 100 mSv (avg 20 mSv).
Dose Limits: ICRP The dose limits include: • – The effective dose from external sources in the specified period and – The 50 year committed dose for internal sources (for children 70 year commited doses) The 5 year limit is predefined – i.e. can't be defined after the • exposure has occured. – Seperate limits are specified for the skin as it may be subject to localized exposure – for skin, exposure averaged over an area of 1 cm2 is used irrespective of the area exposed. – Eye/lens are defined as they make minimal contribution to the total effective dose. For the pregnant female the ICRP recommends a dose limit of 2 • mSv to the women's abdomen.
Dose Limits: NCRP Exposure Type Dose Levels Occupational Exposure (Annual) 50 mSv Whole Body Effective Dose Limits (Stochastic Effects) Dose Equivalent Limits for organs (Non Stochastic Effects) 150 mSv Lens of Eye All others 500 mSv Public Exposure (annual) Effective dose euivalent limit, continous exposure 1 mSv Effective dose euivalent limit, infrequent exposure 5 mSv Dose Equivalent Limits for lens, skin and extremities 50 mSv Embryo Fetus Exposures 5 mSv Total dose equivalent limit Dose Equivalent in a month 0.5 mSv Annual effective education and training exposure 1 mSv 50 mSv Dose Equivalent Limits for lens, skin and extremities Source: The NCRP Report Number 91
Dose Limits : NCRP • The NCRP has got certain guidelines that are not there in the ICRP recommendations: – Cumulative Dose: The numerical value of the cumulative dose in terms of the tens of Sv exposure should not exceed the numerical age of the person – Example: For a 35 year old person the maximum cumulative exposure is 350 mSv. – For Students and apprentices below 18 yrs age, who are exposed to radiation as a part of their education, the annual effective dose limit is 1 mSv – Guidelines for remedial action for public exposures: • Effective dose limits > 5 mSv • Exposure to radon or decay products > 0.007 J/m3
Difference between NCRP and ICRP • Annual effective dose limits for occupational exposure: – ICRP: 20mSv recommended, 50 mSv maximum allowed. – NCRP: Allows 50 mSv annually. • Risk defined for annual effective dose limits: – ICRP: Both stochastic and nonstochastic events – NCRP: For only stochastic events • Cumulative Dose Limits for occupational exposure: – In terms of contigous, predefined 5 year periods in ICRP. – In NCRP the numerical calculation based on age is used. • Organ Equivalent dose limits: – Specified for lens, eye and skin in ICRP – Specified for many organs in addition in NCRP
Difference between NCRP and ICRP • Limits of Public exposure: – 2 limits given by NCRP alone (ICRP limits correspond to the limits of NCRP for continous or frequent exposure). – ICRP max exposure limit in 1 year corresponds to NCRP annual limits (5 mSv) for infrequent exposure. • Fetal Dose Limits: – In ICRP the total dose limit for fetus or embryo is 1 mSv (2 mSv to the lower abdomen of the pregnant lady). – In NCRP the total dose limit for fetus or embryo is 5 mSv. – In NCRP a monthly dose limit (1/10th the total limit) is also specified. • Threshold doses for remedial action and Negligible individual annual risk levels: – Defined by the NCRP only
Annual Limit on Intake (ALI) • Radiation protection quantity defined for radionuclides which can be ingested or inhaled. • ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent of 5 rems (0.05 Sv) or a committed dose equivalent of 50 rems (0.5 Sv) to any individual organ or tissue. – Thus in abscence of any external radiation exposure a radiation worker can have 1 ALI of exposure from internal isotopes • Derived Air Concentration: The concentration of a given radionuclide in air which, if breathed by the reference man for a working year of 2,000 hours under conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one ALI.
Comparision of risks Occupation Mean Rate (per million per year) Trade 40 Manufacture 60 Service 40 Government 90 Transport/ Public Utilities 240 Construction 320 Mines / Quarries 430 Radiation Worker* 20 Fatal Accident rates per million workers per year in choosen occupations * For radiation worker the 20 x 10-6 per year is the risk of any detriment not fatality
Organizations and Authorities • The basic responsibility for radiation protection rests with the authorized persons who will conduct the intervention causing radiation exposure. • The Goverment appoints a regulatory authority who will enforce the rules of radiation protection • In India, the regulatory authority is the AERB. – Authorization granted for practice of radiotherapy is done via liscences. – The authorized person for the radiotherapy department is therefore called a liscencee. – Also provides services like personal dosimetry, equipment calibration and maintainence of national standard dosimeter (primary standard dosimeter).
Design of Radiation Facilities: Principles
Principles • 3 basic parameters influence the exposure that an individual receives in a radiation field. – Time: Longer the time spent in the radiation field greater the exposure – Distance: The exposure falls as a function of square of the distance from the radiation source – Shielding: Exposure can be reduced due to attenuation of the priamry beam by shielding. • Design of radiation facilities is basically design of the shielding for a given set of time of exposure and distance. • The design and amount of shielding will be a function of the desired effective dose in a given area and the calculated dose at that place in the event of no shielding.
Terminology: Types of Radiation • Primary radiation: which is the radiation directly emitted from the treatment machine through the collimator opening in the case of external sources and from the radioactive source in the case of brachytherapy. • Scatter radiation: which is the radiation produced by the scattering of the primary radiation beam from various media struck by the primary beam, such as the patient, collimators, beam shaping accessories and air. • Leakage radiation: which is the radiation that escapes through the shielded head of the therapy unit (for accelerators leakage radiation only exists while the beam is on; for cobalt units leakage radiation is always present).
Terminology: Barriers • Primary Barrier: Where the primary radiation beam strikes the wall it becomes a primary barrier. It includes the roof and the floor if the radiation facility is situated above another room. • Secondary Barrier: These protect against the scattered and leakage radiation.
Terminology: Areas • Controlled areas: A controlled area is any area in which specific protection measures and safety provisions could be required for: – Controlling normal exposures and preventing spread of contamination during normal working hours – Preventing or limiting the extent of potential exposures. • Supervised area: An area for which occupational exposure conditions are kept under review though protective measures and safety provisions are not needed. – Number and extent of controlled areas should be as limited as possible as working in these areas require some form of personal protection measures and administrative control. – Controlled areas should be labelled appropiately and should have restricted access.
Terminology: Workload and Use factors • Workload: Refers to the typical radiation output from a source per week at a well designated point. • Use Factor: Refers to the number of times in a day the radiation beam is directed towards the barrier in question • Occupancy Factor: Refers to the amount of time the rooms adjacent to the treatment room are occupied. – The occupancy factor for an area should be considered as the fraction of time spent by a single person who is there the longest. – The occupancy factor is best defined as the fraction of an 8 h day or 2000 h year for which a single individual may occupy a particular area.
Terminology: IDR and TADR • When weekly dose limits are used instead of the instantaneous dose the shielding requirements are often less. – IDR refers to the direct reading of a dosimeter in dose per hour averaged over one minute – The IDR is an useful quantity for QA of the barrier after construction. – It is usually measured at a point 0.3 m from the barrier. • Another quantity in this senario is the TADR – Time averaged Dose Rate. – TADR = IDR x ( Daily Beam on time / Length of working day) – In UK an area of TADR2000 of ≤ 0.15 μSv/hr doesnot need supervision – US specifes a TADR limit of 20 μSv/hr in a public place.
General Design Guidelines • Usually located at periphery of the hospital complex – avoids the problem of therapy rooms being adjacent to high occupancy areas. • The ground level is preferred as the problem of shielding the floor is less. – If facility is located above the floor needs to be reinforced to bear the weight of the machine in addition to the barrier. • When ever possible the areas around a therapy machine should be designated as a controlled area (including roofs over the machines). • Mazes should be designed whenever possible as they reduce the need for a heavy shielded door. • Doors should be provided at the maze entrance so that casual entry of public can be avoided.
General Design Guidelines • If a shieding barrier is required to reduce dose at the entrance the door may be motorized. However in that event manual operating facilities and several interlocks are needed. • In ordinary dorros also interlocks are provided so that the beam may be turned off when the door is opened. • The control console should be provided with devices for keeping a watch on the patient at all times. – CCTV: Two cameras are recommended – 15 off and above the gantry rotation axis for optimum patient viewing. – Mirrors and door glass arrangement – Lead glass for direct viewing – expensive, only for low energy.
General Design Guidelines • Ducts for electrical materials should always run through a secondary barrier and at an angle through it. – No duct with diameter > 30 mm should penetrate the primary barrier. – The total length of the duct in the barrier should be greater than the thickness of the radiation barrier. – Ducts are best designed to follow the maze or reach the treatment room through the ground. • Warning signs, lights and audible alarms should be provided. – Illumination signs may be two stage or three stage: • 1st stage: When there is power in the treatment unit • 2nd stage: When the treatment unit is programmed • 3rd stage: When the beam is actually turned on.
Examples Radiation on indicator Gamma Zone Monitor Control Console
Examples Steel Door with Trefoil Lead Glass Window CCTV Mirror
Brachytherapy Facilities • Sources should be stored in a seperate lead lined chamber. • Storage area for radium should be ventilated by a direct filtered exhaust to the outdoors. • The storage rooms are usually provided with a sink of cleaning source applicators – a trap is provided to avoid source loss. • Source preperation bench should be provided with an L- Bench with a lead glass window. • Suitably long forceps should be used to provide as much distance as practical between sources and the operator. • Leak Test: A source is considered to be leaking if a presence of 0.005 µCi or more of removable contamination is measured
Design of Radiation Facilities: Method
Building Material: Concrete • Concrete density varies according to the mixture. • Concrete should be poured in one go to prevent seams from forming. • Also during the mixing with cement air may be trapped so continous vibration of the mix is required. TVL of different materials used for shielding • If preformed blocks are used, heavy mortar is needed to prevent shine paths. Typical Barrier Thickness for different beam energies
Workload • In EBRT, the typical number of patients treated in an eight hour day is 50. • NCRP Report 49 suggests a workload figure of 1000 Gy/week based on a dose of 4 Gy at 1 m per patient, assuming a five day week for megavoltage facilities. • For dual energy machines the same workload figure of 1000 Gy/week may be used. NCRP Report 51 suggests an assumed workload of 500 Gy week for the higher energy, with the remainder of the workload bein attributed to lower energy X rays or electrons. • If the workload is greater, the figures used for Gy/week should reflect this. So 50 petients per day , 5 days a week and 8 hours in a day is the basis for workload calculations
Use Factors • If conventional treatment techniques are to be used, NCRP Report 49 suggests – Use factors: • 1 for the floor with the beam pointing vertically down. • 0.25 for each wall and ceiling if specific values are not available. – Use factors for brachytherapy, secondary barriers and doors is considered as 1. • These use factors may depend on the particular use of the facility and also on the energy used. • For example, a facility performing a large number of total body irradiations may have a use factor greater than 0.25 for one wall, and lower for other walls.
Weekly Dose Rates • ICRP 33 states that for design of shielding in radiation facilities the actual dose values to individuals in the occupied areas should be only 1/10th of the effective dose values ( 1/30th for the equivalent dose) used for calculation of the shielding requirements. – Using the dose limit of 0.01 mSv/ week at a distance of 4 m for 20 MV photons can increase the shielding requirement for the barrier by 50 cm of RCC if designated as a public area as compared to what can be used if it was designated as a controlled area.
Design Effective Limits Design Limits USA UK Occupational exposure Fraction of 10mSv in a year 6 mSv in a year IDR is 7.5 μSv·h–1 Public exposure 1 mSv annually 0.3 mSv in a year 20μSv per hour IDR is <7.5 μSv·h–1 TADR is <0.5 μSv·h–1 TADR2000 <0.15 μSv·h–1
Weekly Dose Limits • When caclulating the dose limits for shielding design a number of conservative estimates are required: – The attenuation of the beam by the patient is not considered. – The maximum possible leakage radiation is assumed. – The workload, as well as the use and occupancy factors, is overestimated. – An assumption is made that staff are always in the most exposed place of the occupied area. • For linacs producing X rays and electrons an assumption is made that the linac always operates in the X ray mode. • For dual energy linacs an assumption is made that the linac always runs in the higher energy mode.
Thickness of Primary Barriers • The thickness is back calculated using the desired dose constraint. • The formula used is: P = Allowed dose per week outside the barrier – d = distance from the isocenter to the point outside barrier – SAD = Source to Axis distance (in m). – W = Workload (Gy / week at 1 m) – U = Use factor for that barrier – T = Occupancy factor or the fraction of time the area outside the barrier is likely to be – occupied.
Thickness of primary Barrier • The value B directly gives the desired attenuation. • This can be used to calulate the required thickness from the next formula: • An altenate method of calculation used is the use of IDR.
Secondary Barriers: Scattered Radiation • The formula for calculation of the scattered radiation from the patient is given as: dsca is the distance from the radiation source to the patient, in m. dsec is the distance from the patient to the point of interest, in m. a is the scatter fraction defined at dsca. F is the field area incident on the patient, in cm2. The scatter primary ratio (a) is dependent on the energy of the X ray beam and the scattering angle. These data are tabulated per 400 cm2 of irradiated field area for 60Co, 6, 10, 18 and 24 MV X ray beams.
Secondary Barrier: Scatterd Radiation • The formula for calculation of the scattered radiation from the walls is given as: dw is the distance from the radiation source to the scattering surface (wall), in m; dr is the distance from the scattering surface (wall) to the point of interest, in m; α is the wall reflection coefficient, which depends on the wall material, scattering angle, and beam energy (albedo factor) A is the field area projected on the scattering surface (wall), in m2.
Secondary Barrier: Scattered Radiation • For megavoltage beams, on the other hand, the maximum energy of the 90-degree scattered photons is 500 keV. • Therefore, the transmission of this scattered radiation through a barrier is estimated to be approximately the same as that for a 500-kVp useful beam. • The use factor for the secondary barrier is considered unity.
Secondary barrier: Leakage Radiation • The thickness of the secondary barrier to shield against leakage radiation is given by the relationship P is the design dose limit ds is the distance from the isocentre to the point of interest in m W is the workload T is the occupancy factor. The Use factor is always taken as 1. As the use factor is one the distance of the source from the point of concern is taken to lie at the isocenter as the average distance from all gantry angles.
Secondary barrier: Leakage Radiation • For a linear accelerator, national and international protocols state that the leakage from the treatment head must not exceed 0.5% of the primary beam, outside the useful beam at 1 m from the path of the electrons between the gun and target window and averaged over 100 cm2. • In the plane of the patient, the leakage must not exceed an average of 0.1% and a maximum of 0.2% over a 2 m radius measured from the beam central axis. • A simple quantity to keep in mind is the total dose from leakage radiation should not exceed 1 mrad at 1 m per hour from the gantry at any position.
Neutron Protection • Considered for LINACs operating above 10 MV. • Concrete has high water content – TVL for photoneutrons half that of the photons – additional shielding not needed. • Neutron capture produces secondary photons (capture photons): – Average capture photon energy is 3.6 MeV – Max energy is 8.0 MeV • Slow neutron capture is enhanced by use of materials like Boron. – Slow neutron capture by Boron results in low energy gamma ray production of 473 KeV. – For doors 5% Boron is included in the polythelene in treatment rooms.
Radiation Measurement and Monitoring
Types of Monitoring Radiation Monitoring Personal Monitoring Workplace Monitoring •For individual radiation •For the entire workplace and worker the radiation rooms •In form of radiation badges •Usually require some form of •Worn on the person of the radiation detectors worker •Integrated with the workplace •Allows estimation of individual •Allows estimation of exposure doses levels in the envioronment
Monitoring For Radiation Protection • Types of instruments for Dose Measurements – Dose rate meters used to measure the external exposure. – Dosimeters which indicate the cumulative external exposure. – Other types: • Surface contamination meters which indicate the potential internal exposure when a radioactive substance is distributed over a surface. • Airborne contamination meters and gas monitors which indicate the internal exposure when a radioactive substance is distributed within an atmosphere.
Dose Rate Meters • A dose rate meter absorbs energy from penetrating radiation. • Capable of providing direct readings of the dose equivalent rate in microsieverts per hour (mSv/h or mSv·h–1). • Disadvantages – These usually respond only to X, gamma and/or beta radiations. – Specialized instruments are necessary to measure neutron dose equivalent rates. – Dose rate meters may not be able to provide an accurate response to rapidly changing or pulsed radiation fields. Integrating dose rate meters and dosimeters are more appropriate in such circumstances.
Dosimeter • A dosimeter measures the cumulative energy absorbed as a consequence of exposure to ionizing radiation. – Personal dosimeters must be worn by radiation workers to measure their radiation exposure. – Passive dosimeters routinely monitor cumulative doses resulting from an external exposure. – Active dosimeters provide an immediate reading of the dose in microsieverts (mSv) and may also provide an immediate alarm signal when the measured dose approaches a value pre- set by the manufacturer or user.
Principle of construction • Detectors used: Detector – Gas filled detectors – Ionization chambers – Proportional counters – Geiger-Müller counters Amplifier – Scintillation counters – Solid state detectors. Processor Display
Gas Filled Ionization Chambers E1 and E2: Charged Electrodes A: Housing for the gas R : Radiation P : Power Supply D : Display
Practical Ionization Chambers • Components: – The cylindrical detector wall serves as the cathode (negative electrode) and is normally made of air-equivalent, carbon coated plastic or aluminium. – The axial anode (positive electrode). – Beta window made of thin foil (3–7 mg·cm–2). – Protective buildup cap (200–300 mg·cm–3) made of toughened plastic or aluminium – to improve detection effeciency for high energy photon radiation. • Detector volumes of a few hundred cc are needed to measure exposures in nanocoulombs per hour (nC·h–1).
Uses of Ionization chambers • Simple Ionization Chambers: Can be used as personal dosimeters and integrating dosimeters. They are most commonly used for personal protection. • Proportional Counters: Best used as surface contamination monitors. Are sensitive to low energy photons and beta particles. • Geiger Muller Counters: These are commonly used for dose and dose rate measurement. • Scintillation Chambers: Primarily used for gamma, electron and neutron spectrometry. • Solid state detectors: Primarily used for gamma and heavy particle spectrometry.
Neutron Detectors • These consist of a ionization chamber surrounded by a wall made of polyethylene to “thermalize” neutrons (M). • The dosimeter is filled with a gas like Boron trifluride (BF3) or He3. • A perforated cadmium or boronated plastic sheath (S) that modifies the neutron energy so that doses can be read in terms of dose equivalent (directly accounting for quality factor due to different neutron energies).
Individual Monitoring: Uses • The exposures of the individual radiation worker needs to be routinely monitored and records kept of their cumulative radiation doses. – They can also be used to retrospectively determine a dose received by a worker. – Individual monitoring is used to verify the effectiveness of radiation control practices in the workplace. – It is also used to detect changes in the workplace – Confirm or supplement static workplace monitoring, – Identify working practices that minimize doses a
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