This course meets the requirements for 3 hours of continuing education credit in jurisdictions which recognize NYS Dept. of Education approval; however participants should be aware that some boards have limitations on the number of hours accepted in certain categories and/or restrictions on certain methods of delivery of continuing education. A certificate will be emailed to you upon successful completion.
Listed below are some of the probable prompt and delayed effects of certain doses of radiation when the doses are received by an individual within a twenty-four hour period.
Dosages are in Roentgen Equivalent Man (Rem)
- 0-25 No injury evident. First detectable blood change at 5 rem.
- 25-50 Definite blood change at 25 rem. No serious injury.
- 50-100 Some injury possible.
- 100-200 Injury and possible disability.
- 200-400 Injury and disability likely, death possible.
- 400-500 Median Lethal Dose (MLD) 50% of exposures are fatal.
- 500-1,000 Up to 100% of exposures are fatal.
- 1,000-over 100% likely fatal.
The delayed effects of radiation may be due either to a single large overexposure or continuing low-level overexposure. Example dosages and resulting symptoms when an individual receives an exposure to the whole body within a twenty-four hour period.
|100 – 200 Rem|
|First Day||No definite symptoms|
|First Week||No definite symptoms|
|Second Week||No definite symptoms|
|Third Week||Loss of appetite, malaise, sore throat and diarrhea|
|Fourth Week||Recovery is likely in a few months unless complications develop because of poor health|
|400 – 500 Rem|
|First Day||Nausea, vomiting and diarrhea, usually in the first few hours|
|First Week||Symptoms may continue|
|Second Week||Epilation, loss off appetite|
|Third Week||Hemorrhage, nosebleeds, inflammation of mouth and throat, diarrhea, emaciation|
|Fourth Week||Rapid emaciation and mortality rate around 50%|
The U.S. Nuclear Regulatory Commission and the Code of Federal Regulations
Since working with ionizing radiation can present significant safety risks, its use is closely regulated. In the United States, the Nuclear Regulatory Commission (NRC) is responsible for protecting workers, the public and the environment from the effects of radiation. The NRC is an independent agency established by the Energy Reorganization Act of 1974 to regulate civilian use of nuclear materials. The NRC is headed by five commissioners appointed by the President and confirmed by the Senate for five-year terms.
The Code of Federal Regulations (CFR) is the system used by the US Federal Government to organize the rules published in the Federal Register by the executive departments and agencies. The CFR is divided into 50 titles that represent broad areas subject to Federal regulation. Title 10 of the code applies to energy and parts 0 through 50 of Title 10 apply to NRC rules. Part 19, Notices, instructions and reports to workers: inspection and investigations; Part 20, Standards for protection against radiation; and Part 34, Licenses for industrial radiography and radiation safety requirements for industrial radiographic operations, are areas of the Code that are of primary interest when addressing radiation safety in industrial radiography.
The NRC regulations can be accessed on the Internet at:
More than half of the states in the U.S. have entered into “agreement” with the NRC to assume regulatory control of radioactive material use within their borders. As part of the agreement process, the states must adopt and enforce regulations comparable to those found in Title 10 of the Code of Federal Regulations. Some of the requirement of the Code, such as exposure limits, responsibilities and procedures will be discussed in the following pages.
As discussed in the introduction, concern over the biological effect of ionizing radiation began shortly after the discovery of X-rays in 1895. Over the years, numerous recommendations regarding occupational exposure limits have been developed by the International Commission on Radiological Protection (ICRP) and other radiation protection groups. In general, the guidelines established for radiation exposure have had two principle objectives: 1) to prevent acute exposure; and 2) to limit chronic exposure to “acceptable” levels.
Current guidelines are based on the conservative assumption that there is no safe level of exposure. In other words, even the smallest exposure has some probability of causing a stochastic effect, such as cancer. This assumption has led to the general philosophy of not only keeping exposures below recommended levels or regulation limits but also maintaining all exposure “as low as reasonable achievable” (ALARA). ALARA is a basic requirement of current radiation safety practices. It means that every reasonable effort must be made to keep the dose to workers and the public as far below the required limits as possible.
Regulatory Limits for Occupational Exposure
Many of the recommendations from the ICRP and other groups have been incorporated into the regulatory requirements of countries around the world. In the United States, annual radiation exposure limits are found in Title 10, part 20 of the Code of Federal Regulations, and in equivalent state regulations. For industrial radiographers who generally are not concerned with an intake of radioactive material, the Code sets the annual limit of exposure at the following:
1) the more limiting of:
- A total effective dose equivalent of 5 rems (0.05 Sv) or
- The sum of the deep-dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 50 rems (0.5 Sv).
2) The annual limits to the lens of the eye, to the skin, and to the extremities, which are:
- A lens dose equivalent of 15 rems (0.15 Sv)
- A shallow-dose equivalent of 50 rems (0.50 Sv) to the skin or to any extremity.
The shallow-dose equivalent is the external dose to the skin of the whole-body or extremities from an external source of ionizing radiation. This value is the dose equivalent at a tissue depth of 0.007 cm averaged over and area of 10 cm2.
The lens dose equivalent is the dose equivalent to the lens of the eye from an external source of ionizing radiation. This value is the dose equivalent at a tissue depth of 0.3 cm.
The deep-dose equivalent is the whole-body dose from an external source of ionizing radiation. This value is the dose equivalent at a tissue depth of 1 cm.
The total effective dose equivalent is the dose equivalent to the whole-body.
Declared Pregnant Workers and Minors
Because of the increased health risks to the rapidly developing embryo and fetus, pregnant women can receive no more than 0.5 rem during the entire gestation period. This is 10% of the dose limit that normally applies to radiation workers. Persons under the age of 18 years are also limited to 0.5rem/year.
Non-radiation Workers and the Public
The dose limit to non-radiation workers and members of the public are two percent of the annual occupational dose limit. Therefore, a non-radiation worker can receive a whole body dose of no more that 0.1 rem/year from industrial ionizing radiation. This exposure would be in addition to the 0.3 rem/year from natural background radiation and the 0.05 rem/year from man-made sources such as medical x-rays.
Controlling Radiation Exposure
When working with radiation, there is a concern for two types of exposure: acute and chronic. An acute exposure is a single accidental exposure to a high dose of radiation during a short period of time. An acute exposure has the potential for producing both nonstochastic and stochastic effects. Chronic exposure, which is also sometimes called “continuous exposure,” is long-term, low level overexposure. Chronic exposure may result in stochastic health effects and is likely to be the result of improper or inadequate protective measures.
The three basic ways of controlling exposure to harmful radiation are: 1) limiting the time spent near a source of radiation, 2) increasing the distance away from the source, 3) and using shielding to stop or reduce the level of radiation.
The radiation dose is directly proportional to the time spent in the radiation. Therefore, a person should not stay near a source of radiation any longer than necessary. If a survey meter reads 4 mR/h at a particular location, a total dose of 4mr will be received if a person remains at that location for one hour. In a two hour span of time, a dose of 8 mR would be received. The following equation can be used to make a simple calculation to determine the dose that will be or has been received in a radiation area.
Dose = Dose Rate x Time
When using a gamma camera, it is important to get the source from the shielded camera to the collimator as quickly as possible to limit the time of exposure to the unshielded source. Devices that shield radiation in some directions but allow it pass in one or more other directions are known as collimators. This is illustrated in the images at the bottom of this page.
Increasing distance from the source of radiation will reduce the amount of radiation received. As radiation travels from the source, it spreads out becoming less intense. This is analogous to standing near a fire. The closer a person stands to the fire, the more intense the heat feels from the fire. This phenomenon can be expressed by an equation known as the inverse square law, which states that as the radiation travels out from the source, the dosage decreases inversely with the square of the distance.
Inverse Square Law: I1 / I2 = D22 / D12
The third way to reduce exposure to radiation is to place something between the radiographer and the source of radiation. In general, the more dense the material the more shielding it will provide. The most effective shielding is provided by depleted uranium metal. It is used primarily in gamma ray cameras like the one shown below. The circle of dark material in the plastic see-through camera (below right) would actually be a sphere of depleted uranium in a real gamma ray camera. Depleted uranium and other heavy metals, like tungsten, are very effective in shielding radiation because their tightly packed atoms make it hard for radiation to move through the material without interacting with the atoms. Lead and concrete are the most commonly used radiation shielding materials primarily because they are easy to work with and are readily available materials. Concrete is commonly used in the construction of radiation vaults. Some vaults will also be lined with lead sheeting to help reduce the radiation to acceptable levels on the outside.
Half-Value Layer (Shielding)
As was discussed in the radiation theory section, the depth of penetration for a given photon energy is dependent upon the material density (atomic structure). The more subatomic particles in a material (higher Z number), the greater the likelihood that interactions will occur and the radiation will lose its energy. Therefore, the more dense a material is the smaller the depth of radiation penetration will be. Materials such as depleted uranium, tungsten and lead have high Z numbers, and are therefore very effective in shielding radiation. Concrete is not as effective in shielding radiation but it is a very common building material and so it is commonly used in the construction of radiation vaults.
Since different materials attenuate radiation to different degrees, a convenient method of comparing the shielding performance of materials was needed. The half-value layer (HVL) is commonly used for this purpose and to determine what thickness of a given material is necessary to reduce the exposure rate from a source to some level. At some point in the material, there is a level at which the radiation intensity becomes one half that at the surface of the material. This depth is known as the half-value layer for that material. Another way of looking at this is that the HVL is the amount of material necessary to the reduce the exposure rate from a source to one-half its unshielded value.
Sometimes shielding is specified as some number of HVL. For example, if a Gamma source is producing 369 R/h at one foot and a four HVL shield is placed around it, the intensity would be reduced to 23.0 R/h.
Each material has its own specific HVL thickness. Not only is the HVL material dependent, but it is also radiation energy dependent. This means that for a given material, if the radiation energy changes, the point at which the intensity decreases to half its original value will also change. Below are some HVL values for various materials commonly used in industrial radiography. As can be seen from reviewing the values, as the energy of the radiation increases the HVL value also increases.
Approximate HVL for Various Materials when Radiation is from a Gamma Source
|Half-Value Layer, mm (inch)|
|Iridium-192||44.5 (1.75)||12.7 (0.5)||4.8 (0.19)||3.3 (0.13)||2.8 (0.11)|
|Cobalt-60||60.5 (2.38)||21.6 (0.85)||12.5 (0.49)||7.9 (0.31)||6.9 (0.27)|
Approximate Half-Value Layer for Various Materials when Radiation is from an X-ray Source
|Half-Value Layer, mm (inch)|
|Peak Voltage (kVp)||Lead||Concrete|
|50||0.06 (0.002)||4.32 (0.170)|
|100||0.27 (0.010)||15.10 (0.595)|
|150||0.30 (0.012)||22.32 (0.879)|
|200||0.52 (0.021)||25.0 (0.984)|
|250||0.88 (0.035)||28.0 (1.102)|
|300||1.47 (0.055)||31.21 (1.229)|
|400||2.5 (0.098)||33.0 (1.299)|
|1000||7.9 (0.311)||44.45 (1.75)|
Note: The values presented on this page are intended for educational purposes. Other sources of information should be consulted when designing shielding for radiation sources.
Since X-ray and gamma radiation are not detectable by the human senses and the resulting damage to the body is not immediately apparent, a variety of safety controls are used to limit exposure. The two basic types of radiation safety controls used to provide a safe working environment are engineered and administrative controls. Engineered controls include shielding, interlocks, alarms, warning signals, and material containment. Administrative controls include postings, procedures, dosimetry, and training.
Engineered controls such as shielding and door interlocks are used to contain the radiation in a cabinet or a “radiation vault.” Fixed shielding materials are commonly high density concrete and/or lead. Door interlocks are used to immediately cut the power to X-ray generating equipment if a door is accidentally opened when X-rays are being produced. Warning lights are used to alert workers and the public that radiation is being used. Sensors and warning alarms are often used to signal that a predetermined amount of radiation is present. Safety controls should never be tampered with or bypassed.
When portable radiography is performed, it is most often not practical to place alarms or warning lights in the exposure area. Ropes and signs are used to block the entrance to radiation areas and to alert the public to the presence of radiation. Occasionally, radiographers will use battery operated flashing lights to alert the public to the presence of radiation. Portable or temporary shielding devices may be fabricated from materials or equipment located in the area of the inspection. Sheets of steel, steel beams, or other equipment may be used for temporary shielding. It is the responsibility of the radiographer to know and understand the absorption value of various materials. More information on absorption values and material properties can be found in the radiography section of this site.
As mentioned above, administrative controls supplement the engineered controls. These controls include postings, procedures, dosimetry, and training. It is commonly required that all areas containing X-ray producing equipment or radioactive materials have signs posted bearing the radiation symbol and a notice explaining the dangers of radiation. Normal operating procedures and emergency procedures must also be prepared and followed. In the US, federal law requires that any individual who is likely to receive more than 10% of any annual occupational dose limit be monitored for radiation exposure. This monitoring is accomplished with the use of dosimeters, which are discussed in the radiation safety equipment section of this material. Proper training with accompanying documentation is also a very important administrative control.
Working safely with radiation is the responsibility of everyone involved in the use and management of radiation producing equipment and materials. Depending on the size of the organization, specific responsibilities may be assigned to various individuals and/or committees.
Radiation Safety Officer
All organizations that are licensed to use ionizing radiation must have a Radiation Safety Officer. The RSO is the individual authorized by the company to serve as point of contact for all activities conducted under the scope of the authorization. The RSO ensures that radiation safety activities are being performed in accordance with approved procedures and regulatory requirements. Some of the common responsible for the RSO include:
- Ensuring that all individuals using radiation equipment are appropriately trained and supervised.
- Ensuring that all individuals using the equipment have been formally authorized to use the equipment.
- Ensuring that all rules, regulations, and procedures for the safe use of radioactive sources and X-ray systems are observed.
- Ensuring that proper operating, emergency, and ALARA procedures have been developed and are available to all system users.
- Ensuring that accurate records of the use of the sources and equipment are maintained.
- Ensuring that required radiation surveys and leak tests are performed and documented.
- Ensuring that systems and equipment are protected from unauthorized access or removal.
The minimum qualifications, training, and experience for RSOs for industrial radiography are as follows: (1) Completion of the training and testing requirements of Sec. 34.43(a) of Part 10 of the Federal Code of Regulations, (2) 2000 hours of hands-on experience as a qualified radiographer in industrial radiographic operations, and (3) Formal training in the establishment and maintenance of a radiation protection program.
Radiation Safety Committee
Some organizations may have a Radiation Safety Committee (RSC) to assist the RSO. The RSC often provides oversight of the policies, procedures and responsibilities of an organizations radiation safety program.
The individuals authorized to use the X-ray producing system or gamma sources are responsible for ensuring that:
- All rules, regulations, and procedures for the safe use of the X-ray system are followed.
- An accurate record of the use of the system is maintained.
- All safety problems with the system are reported to the RSO and corrected before further use.
- The system is protected from unauthorized access or removal.
Standard Operating Procedures
As a minimum, operating procedures must include instructions for the following:
- Appropriate handling and use of licensed sealed sources and radiographic exposure devices so that no person is likely to be exposed to radiation doses in excess of the established exposure limits.
- Methods and occasions for conducting radiation surveys.
- Methods for controlling access to radiographic areas.
- Methods and occasions for locking and securing radiographic exposure devices, transport and storage containers and sealed sources.
- Personnel monitoring and the use of personnel monitoring equipment.
- Transporting sealed sources to field locations, including packing of radiographic exposure devices and storage containers in the vehicles, placarding of vehicles when needed, and control of the sealed sources during transportation.
- The inspection, maintenance, and operability checks of radiographic exposure devices, survey instruments, transport containers, and storage containers.
- The procedure(s) for identifying and reporting defects and noncompliance.
- Maintenance of records.
Written operating procedures must be developed and made available to anyone that will be working with radiation sources or X-ray producing equipment. These procedures must be specific to the equipment and its use in a particular application. Simply making the equipment manufacturers operating instructions available to workers does not satisfy this requirement. The operating procedure must be followed at all times unless written permission to deviate is received from the Radiation Safety Officer.
Procedures must also be developed that guide workers in the event of an emergency. A few of the items that could be covered include:
- Steps that must be taken immediately by radiography personnel in the event a pocket dosimeter is found to be off-scale or an alarm ratemeter alarms unexpectedly.
- Steps for minimizing exposure of persons in the event of an accident.
- The procedure for notifying proper persons in the event of an accident.
- Radioactive source recovery procedure if licensee will perform the recovery.
Transporting the Exposure Device
When transporting the exposure device, it must be stowed securely in the vehicle. A lockable metal box is often bolted in the rear of the vehicle. A survey of the over pack, the outside of the vehicle, and the drivers compartment is then conducted and documented.
Preparing for an Exposure
Once on the job site, the exposure area will be assessed, distance calculations made for restricted area boundaries, and ropes and signs placed appropriately. Once this is complete, the radiographer is ready to remove the exposure device from its storage compartment in the vehicle. The survey meter should be monitored as the storage compartment is approached and when removing the exposure device from the compartment. Daily safety checks should then be made. Once these checks are completed, the radiographer and assistant may then move the exposure device to the exposure location. As the cranks and guide tubes are attached in preparation for the first exposure, the survey meter should be monitored. Before the source is exposed, the assistant should check the area for persons who may have crossed into the restricted area, and then move outside the rope boundary.
The majority of over exposures in industrial radiography are the result of the radiographer not knowing the location of a gamma emitter and failing to conduct a proper radiation survey. Exposure vaults are equipped with warning lights and safety interlock switches which provide a margin of safety for workers. A survey must be performed occasionally to verify that vaults are not “leaking” radiation and that the safety devices are performing properly. However, when conducting radiography with gamma emitters in the field, the radiographer must rely heavily on measurements with a survey meter since other safety devices are uncommon. A series of surveys must be taken and some of the results from these surveys must be documented when transporting and working with gamma emitters in the field.
Approaching the Exposure Device
A technician should be thoroughly familiar with the operation of a survey meter since proper use of the device is essential. Before removing the exposure device (camera) from storage, the calibration of the survey meter must be verified and the battery level must be checked. When approaching the exposure device to remove it from the storage location, the survey meter should be in hand and operational. The survey meter should be placed next to the exposure device to verify that the source is contained inside the projector, and to verify that the survey meter is working properly. Survey meter readings should be compared to previous readings and recorded.
Making an Exposure
The radiographer should be at the maximum distance from the exposure device that the guide tube will allow as he or she quickly cracks the source out of the exposure device and into place. As the source moves out of the exposure device, the survey meter will increase to a very high level and then reduce once the source is inside the collimator. During the exposure, the assistant will survey the established boundary to determine the levels of radiation present. If the survey meter indicates levels are higher than calculated, the boundary must be extended.
Retracting the Source
On retraction of the source, the radiographers will see a rise in readings as the source moves from the collimator and is retracted into the projector. When the source is inside the exposure device, the radiographer should approach it while monitoring the survey meter. If the source is properly retracted, no increase in the survey meter reading should be seen when approaching the exposure device. The exposure device should be surveyed on all sides, paying special attention to the front of the device. The entire length of the guide tube must then be surveyed.
This process is repeated for each exposure. The survey results must be documented when the exposure device is returned to the vehicle for transportation, and when it is placed back into its storage location.
Instruments used for radiation measurement fall into two broad categories:
– rate measuring instruments and
– personal dose measuring instruments.
Rate measuring instruments measure the rate at which exposure is received (more commonly called the radiation intensity). Survey meters, audible alarms and area monitors fall into this category. These instruments present a radiation intensity reading relative to time, such as R/hr or mR/hr. An analogy can be made between these instruments and the speedometer of a car because both are measuring units relative to time.
Dose measuring instruments are those that measure the total amount of exposure received during a measuring period. The dose measuring instruments, or dosimeters, that are commonly used in industrial radiography are small devices which are designed to be worn by an individual to measure the exposure received by the individual. An analogy can be made between these instruments and the odometer of a car because both are measuring accumulated units.
The radiation measuring instruments commonly used in industrial radiography are described in more detail in the following pages.
The survey meter is the most important resource a radiographer has to determine the presence and intensity of radiation. A review of incident and overexposure reports indicate that a majority of these type of events occurred when a technician did not have or did not use a survey meter.
There are many different models of survey meters available to measure radiation in the field. They all basically consist of a detector and a readout display. Analog and digital displays are available. Most of the survey meters used for industrial radiography use a gas filled detector.
Gas filled detectors consists of a gas filled cylinder with two electrodes. Sometimes, the cylinder itself acts as one electrode, and a needle or thin taut wire along the axis of the cylinder acts as the other electrode. A voltage is applied to the device so that the central needle or wire become an anode (+ charge) and the other electrode or cylinder wall becomes the cathode (- charge). The gas becomes ionized whenever the counter is brought near radioactive substances. The electric field created by the potential difference between the anode and cathode causes the electrons of each ion pair to move to the anode while the positively charged gas atom is drawn to the cathode. This results in an electrical signal that is amplified, correlated to exposure and displayed as a value.
Depending on the voltage applied between the anode and the cathode, the detector may be considered an ion chamber, a proportional counter, or a Geiger-Müller (GM) detector. Each of these types of detectors have their advantages and disadvantages. A brief summary of each of these detectors follows.
Ion Chamber Counter
Ion chambers have a relatively low voltage between the anode and cathode, which results in a collection of only the charges produced in the initial ionization event. This type of detector produces a weak output signal that corresponds to the number of ionization events. Higher energies and intensities of radiation will produce more ionization, which will result in a stronger output voltage.
Collection of only primary ions provides information on true radiation exposure (energy and intensity). However, the meters require sensitive electronics to amplify the signal, which makes them fairly expensive and delicate. The additional expense and required care is justified when it is necessary to make accurate radiation exposure measurements over a range of radiation energies. This might be necessary when measuring the Bremsstrahlung radiation produced by an X-ray generator. An ion chamber survey meter is sometimes used in the field when performing gamma radiography because it will provide accurate exposure measurements regardless of the radioactive isotope being used.
Proportional counter detectors use a slightly higher voltage between the anode and cathode. Due to the strong electrical field, the charges produced in the initial ionization are accelerated fast enough to ionize other electrons in the gas. The electrons produced in these secondary ion pairs, along with the primary electrons, continue to gain energy as they move towards the anode, and as they do, they produce more and more ionizations. The result is that each electron from a primary ion pair produces a cascade of ion pairs. This effect is known as gas multiplication or amplification. In this voltage regime, the number of particles liberated by secondary interactions is proportional to the number of ions produced by the passing ionizing particle. Hence, these gas ionization detectors are called proportional counters.
Like ion chamber detectors, proportional detectors discriminate between types of radiation. However, they require very stable electronics which are expensive and fragile. Proportional detectors are usually only used in a laboratory setting.
Geiger-Müller (GM) Counter
Geiger-Müller counters operate under even higher voltages between the anode and the cathode, usually in the 800 to 1200 volt range. Like the proportional counter, the high voltage accelerates the charges produced in the initial ionization to where they have enough energy to ionize other electrons in the gas. However, this cascading of ion pairs occurs to a much larger degree and continues until the counter is saturated with ions. This all happens in a fraction of a second and results in an electrical current pulse of constant voltage. The collection of the large number of secondary ions in the GM region is known as an avalanche and produces a large voltage pulse. In other words, the size of the current pulse is independent of the size of the ionization event that produced it.
|The GM counter was named for Hans Geiger who invented the device in 1908, and Walther Müller who collaborated with Geiger in developing it further in 1928.|
The electronic circuit of a GM counters counts and records the number of pulses and the information is often displayed in counts per minute. If the instrument has a speaker, the pulses can also produce an audible click. When the volume of gas in the chamber is completely ionized, ion collection stops until the electrical pulse discharges. Again, this only takes a fraction of a second, but this process slightly limits the rate at which individual events can be detected.
Because they can display individual ionizing events, GM counters are generally more sensitive to low levels of radiation than ion chamber instruments. By means of calibration, the count rate can be displayed as the exposure rate over a specified energy range. When used for gamma radiography, GM meters are typically calibrated for the energy of the gamma radiation being used. Most often, gamma radiation from Cs-137 at 0.662 MeV provides the calibration. Only small errors occur when the radiographer uses Ir-192 (average energy about 0.34 MeV) or Co-60 (average energy about 1.25 MeV).
Since the Geiger-Müller counter produces many more electrons than a ion chamber counter or a proportional counter, it does not require the same level of electronic sophistication as other survey meters. This results in a meter that is relatively low cost and rugged. The disadvantages of GM survey meters are the lack of ability to account for different amounts of ionization caused by different energy photons and noncontinuous measurement (need to discharge).
Comparison of Gas Filled Detectors
The graph to the right shows the relationship of ion collection in a gas filled detector versus the applied voltage. In the ion chamber region, the voltage between the anode and cathode is relatively low and only primary ions are collected. In the proportional region ,the voltage is higher, and primary ions and a number of secondary ions (proportional to the primary ions originally formed) are collected. In the GM region, a maximum number of secondary ions are collected when the gas around the anode is completely ionized. Note that discrimination between kinds of radiation (E1 and E2) is possible in the ion chamber and proportional regions. Radiation at different energy levels forms different numbers of primary ions in the detector. However in the GM region, the number of secondary ions collected per event remains the same no matter what the energy of the radiation that initiated the event. The GM counter gives up the ability to accurately measure the exposure due to different energies of radiation in exchange for a large signal pulse. This large signal pulse simplifies the electronics that are necessary for instruments such as survey meters.
Pocket dosimeters are used to provide the wearer with an immediate reading of his or her exposure to x-rays and gamma rays. As the name implies, they are commonly worn in the pocket. The two types commonly used in industrial radiography are the Direct Read Pocket Dosimeter and the Digital Electronic Dosimeter.
Direct Read Pocket Dosimeter
A direct reading pocket ionization dosimeter is generally of the size and shape of a fountain pen. The dosimeter contains a small ionization chamber with a volume of approximately two milliliters. Inside the ionization chamber is a central wire anode, and attached to this wire anode is a metal coated quartz fiber. When the anode is charged to a positive potential, the charge is distributed between the wire anode and quartz fiber. Electrostatic repulsion deflects the quartz fiber, and the greater the charge, the greater the deflection of the quartz fiber. Radiation incident on the chamber produces ionization inside the active volume of the chamber. The electrons produced by ionization are attracted to, and collected by, the positively charged central anode. This collection of electrons reduces the net positive charge and allows the quartz fiber to return in the direction of the original position. The amount of movement is directly proportional to the amount of ionization which occurs.
By pointing the instrument at a light source, the position of the fiber may be observed through a system of built-in lenses. The fiber is viewed on a translucent scale which is graduated in units of exposure. Typical industrial radiography pocket dosimeters have a full scale reading of 200 milliroentgens but there are designs that will record higher amounts. During the shift, the dosimeter reading should be checked frequently. The measured exposure should be recorded at the end of each shift.
The principal advantage of a pocket dosimeter is its ability to provide the wearer an immediate reading of his or her radiation exposure. It also has the advantage of being reusable. The limited range, inability to provide a permanent record, and the potential for discharging and reading loss due to dropping or bumping are a few of the main disadvantages of a pocket dosimeter. The dosimeters must be recharged and recorded at the start of each working shift. Charge leakage, or drift, can also affect the reading of a dosimeter. Leakage should be no greater than 2 percent of full scale in a 24 hour period.
Digital Electronic Dosimeter
Another type of pocket dosimeter is the Digital Electronic Dosimeter. These dosimeters record dose information and dose rate. These dosimeters most often use Geiger-Müller counters. The output of the radiation detector is collected and, when a predetermined exposure has been reached, the collected charge is discharged to trigger an electronic counter. The counter then displays the accumulated exposure and dose rate in digital form.
Some Digital Electronic Dosimeters include an audible alarm feature which emits an audible signal or chirp with each recorded increment of exposure. Some models can also be set to provide a continuous audible signal when a preset exposure has been reached. This format helps to minimize the reading errors associated with direct reading pocket ionization chamber dosimeters and allows the instrument to achieve a higher maximum readout before resetting is necessary.
Audible Alarm Rate Meters and Digital Electronic Dosimeters
Audible alarms are devices that emit a short “beep” or “chirp” when a predetermined exposure has been received. It is required that these electronic devices be worn by an individual working with gamma emitters. These devices reduce the likelihood of accidental exposures in industrial radiography by alerting the radiographer to dosages of radiation above a preset amount. Typical alarm rate meters will begin sounding in areas of 450-500 mR/h. It is important to note that audible alarms are not intended to be and should not be used as replacements for survey meters.
Most audible alarms use a Geiger-Müller detector. The output of the detector is collected, and when a predetermined exposure has been reached, this collected charge is discharged through a speaker. Hence, an audible “chirp” is emitted. Consequently, the frequency or chirp rate of the alarm is proportional to the radiation intensity. The chirp rate varies among different alarms from one chirp per milliroentgen to more than 100 chirps per milliroentgen.
Personnel dosimetry film badges are commonly used to measure and record radiation exposure due to gamma rays, X-rays and beta particles. The detector is, as the name implies, a piece of radiation sensitive film. The film is packaged in a light proof, vapor proof envelope preventing light, moisture or chemical vapors from affecting the film.
A special film is used which is coated with two different emulsions. One side is coated with a large grain, fast emulsion that is sensitive to low levels of exposure. The other side of the film is coated with a fine grain, slow emulsion that is less sensitive to exposure. If the radiation exposure causes the fast emulsion in the processed film to be darkened to a degree that it cannot be interpreted, the fast emulsion is removed and the dose is computed using the slow emulsion.
The film is contained inside a film holder or badge. The badge incorporates a series of filters to determine the quality of the radiation. Radiation of a given energy is attenuated to a different extent by various types of absorbers. Therefore, the same quantity of radiation incident on the badge will produce a different degree of darkening under each filter. By comparing these results, the energy of the radiation can be determined and the dose can be calculated knowing the film response for that energy. The badge holder also contains an open window to determine radiation exposure due to beta particles. Beta particles are effectively shielded by a thin amount of material.
The major advantages of a film badge as a personnel monitoring device are that it provides a permanent record, it is able to distinguish between different energies of photons, and it can measure doses due to different types of radiation. It is quite accurate for exposures greater than 100 millirem. The major disadvantages are that it must be developed and read by a processor (which is time consuming), prolonged heat exposure can affect the film, and exposures of less than 20 millirem of gamma radiation cannot be accurately measured.
Film badges need to be worn correctly so that the dose they receive accurately represents the dose the wearer receives. Whole body badges are worn on the body between the neck and the waist, often on the belt or a shirt pocket. The clip-on badge is worn most often when performing X-ray or gamma radiography. The film badge may also be worn when working around a low curie source. Ring badges are worn on a finger of the hand most likely to be exposed to ionizing radiation. A LIXI system with its culminated and directional beam would be one example where monitoring the hands would be more important than the whole body.
Thermoluminescent dosimeters (TLD) are often used instead of the film badge. Like a film badge, it is worn for a period of time (usually 3 months or less) and then must be processed to determine the dose received, if any. Thermoluminescent dosimeters can measure doses as low as 1 millirem, but under routine conditions their low-dose capability is approximately the same as for film badges. TLDs have a precision of approximately 15% for low doses. This precision improves to approximately 3% for high doses. The advantages of a TLD over other personnel monitors is its linearity of response to dose, its relative energy independence, and its sensitivity to low doses. It is also reusable, which is an advantage over film badges. However, no permanent record or re-readability is provided and an immediate, on the job readout is not possible.
How it works
A TLD is a phosphor, such as lithium fluoride (LiF) or calcium fluoride (CaF), in a solid crystal structure. When a TLD is exposed to ionizing radiation at ambient temperatures, the radiation interacts with the phosphor crystal and deposits all or part of the incident energy in that material. Some of the atoms in the material that absorb that energy become ionized, producing free electrons and areas lacking one or more electrons, called holes. Imperfections in the crystal lattice structure act as sites where free electrons can become trapped and locked into place.
Heating the crystal causes the crystal lattice to vibrate, releasing the trapped electrons in the process. Released electrons return to the original ground state, releasing the captured energy from ionization as light, hence the name thermoluminescent. Released light is counted using photomultiplier tubes and the number of photons counted is proportional to the quantity of radiation striking the phosphor.
Instead of reading the optical density (blackness) of a film, as is done with film badges, the amount of light released versus the heating of the individual pieces of thermoluminescent material is measured. The “glow curve” produced by this process is then related to the radiation exposure. The process can be repeated many times.
Radiation Safety References
Andrews, H., Radiation Biophysics, Englewood Cliffs, New Jersey: Prentice Hall, Inc., 1974
Iddings, F., “Radiation Detection for Radiography”, Materials Evaluation, American Society for Nondestructive Testing, Columbus, OH, August 2001
National Research Council, “Health Exposure to Low Levels of Ionizating Radiation”, BEIR V, Washington D.C., 1990
Shapiro, J., Radiation Protection – A Guide for Scientists and Physicians, 4th Ed., Cambridge, MA: Harvard University Press, 2002
United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR), Ionizing Radiation Sources and Biological Effects, New York, NY, 2000
“Ionizing Radiation Exposure of the Population of the United States”, NCRP Report No. 93, 1987
“Radiation Safety Guide for Users of X-ray Systems”, Iowa State University, Environmental Health and Safety, Ames IA, 2004
“Radiation Safety Study Guide for Users of Analytical X-ray Systems”, Environment, Safety, Health and Assurance, Ames Laboratory, Ames, IA, April 1996
Code of Federal Regulations, Title 10, Energy, GPO Access web site @
NDT Education Resource Center Developed by the Collaboration for NDT Education
This course meets the requirements for 3 hours of continuing education credit in jurisdictions which recognize NYS Dept. of Education approval; however participants should be aware that some boards have limitations on the number of hours accepted in certain categories and/or restrictions on certain methods of delivery of continuing education. A certificate will be emailed to you upon successful completion.