Chapter 8 Detection and Measurement of Radiation Ionizing radiation is rarely detected directly. Instead, detectors usually measure the secondary products arising from the interactions of the radiation with the detector material. For example, as an alpha or beta particle traverses a detector's sensitive volume, electron-ion pairs or electron-hole pairs are created and the subsequent movement and collection of charges gives rise to an electrical pulse or current. Indirectly ionizing radiation such as gamma photons and neutrons must first undergo interactions in the detector material that produce secondary charged particles, recoil atoms or electrons that, in turn, produce charge pairs as they slow down. The collection of the ionization created by radiation in a detector volume can be used simply to detect the passage of a radiation particle. The rate of generation of radiation-induced pulses can then be used to measure the rate at which radiation particles traverse the detector. Such detectors are termed radiation counters. In some detectors, the magnitude of the radiation induced pulse is related to the type of radiation particle and its energy. By measuring both the number of pulses and the distribution of pulse sizes produced by a given type of radiation, both the number and energy distribution of the incident radiation can be determined. These detectors can then be used as energy spectrometers. In some detectors the average current can be used as a measure of the amount of ionization or energy deposition, per unit mass of detector material, caused by incident radiation. These detectors can then be calibrated to measure radiation absorbed doses and are thus called dosimeters. In this chapter, the properties of some of the most common radiation detectors are reviewed. An important aspect of radiation detection is an assessment of the uncertainties associated with ionization measurements. Both the release of radiation by radioactive decay and the interactions of radiation with matter are stochastic in nature. Thus, repeated measurements of radiation emitted by a source of constant activity, in a given detector volume in a given time interval, exhibit random statistical fluctuations. The quantification of such statistical fluctuations is a necessity in radiation measurements. This topic also is addressed briefly in this chapter.
8.1 Gas-Filled Radiation Detectors The idea of measuring the radiationinduced ionization in a gas volume dates to the nineteenth-century. These early gas-filled detectors became known as ionization chambers. As ionizing radiation passes through a chamber, the motion within an electric field of the ion pairs formed inside the chamber produces an electrical current. The magnitude of the current is measured and correlated (calibrated) to the intensity of the radiation field. A very common geometry is a coaxial detector that consists of a thin, positivelycharged, center wire anode (held in Cathode Figure 8.1. Basic elements of a gas-filled radiation-detector tube. The cathode is often used to seal the gas cavity from the ambient environment. The output voltage pulse is produced across the load resistor RL. After Knoll [2000]. place by insulators) surrounded by an outer, negatively-charged, cathode tube. The outer tube contains the gas and defines the active volume of the chamber. Air filled chambers may or may not be sealed from the ambient environment. A gas-filled chamber is illustrated in Fig. 8.1. Radiation either interacts in the wall of the chamber or directly in the filling-gas. For incident electromagnetic radiation the dominant interactions, photoelectric effect, Compton scatter, and pair production, occur primarily in the chamber wall material. If the electrons released from the atoms in the wall material escape from the wall and enter the active gas volume of the chamber, then these charged particles (secondary radiation) produce ionization as they pass through the gas. The potential difference between the anode wire and the cathode establishes an electric field that causes positive and negative charges to move in opposite directions. Electrons rapidly drift toward the anode and positive ions migrate more slowly toward the cathode. The motion of these ion pairs causes a flow of current in the external circuit and establishes a voltage across the load resistance. If the incident radiation field contains beta particles, for example, then the chamber must be constructed so that these particles can enter the volume. This is achieved by placing a very thin "window" on one end of the tube. There are three basic types of gas-filled radiation detectors: ionization chambers, proportional counters and Geiger-Mueller counters. All three are known as ionization chambers, but they each have a unique process for forming the total number of ion pairs that are collected at the electrodes. All three operate by forming initial ion pairs from the incident radiation. Once these ion pairs are formed, it is important that they do not recombine and thereby fail to contribute to the electrical signal. Figure 8.2 shows the various operational regions of gas-filled chambers. Region I represents the recombination region where the potential difference between the anode and cathode is not sufficient to collect all the initial ion pairs. Ion chambers operating in this region are not useful radiation detectors.
recombination before collection I I 250 500 Voltage (volts) 750 1000 Figure 8.2. Operational regions for gas-filled radiation detectors. As the potential difference between the anode and cathode is increased, the ionization chamber Region II is entered. In this region the resulting output current is referred to as the ionization chamber saturation current. Region III represents the proportional counter region. In this region, electrons acquire sufficient energy to induce secondary ionization, and hence multiplication. This internal ion-pair multiplication increases the total number of ion pairs in the active volume and, hence, the output current increases by a multiplicative factor M > 1. This region is called the proportional counter region since the output current, or total collected charge per interaction, is proportional to the initial number of ion pairs created by the incident radiation. Next is entered a limited proportionality region where the number of ion-electron pairs collected is relatively independent of the initial number of ion-electron pairs created by the incident radiation. This region generally is not useful operationally. The region of Geiger-Mueller (GM) counter operation, Region IV, exhibits a "plateau" over which M reaches a nearly constant value. In this region one avalanche produces secondary photons whose interactions produce other local
avalanches until the entire anode is surrounded by ion-electron pairs. In this region, the charge collected per interaction is no longer proportional to the initial number of ion pairs created. Therefore, it is very difficult to distinguish between different types of incident radiation or to gain knowledge about the energy of the incident radiation. If the tube voltage is too large, the tube undergoes continuous avalanches around the central anode, one leading to another. This region V, is known as the continuous discharge region. Its entry normally leads to failure of the ion chamber. 8.1.1 lonization Chambers lonization chambers or ion chambers are widely used as radiation monitors. They can be designed to respond to alpha particles, beta particles, gamma rays, x rays, and neutrons. The ion chamber is the most basic type of gas-filled radiation detector since it operates without gas multiplication. These chambers are frequently referred to as current chambers since they are often operated by directly measuring the output current. When operated at the saturation current level (region II of Fig. 8.2), the output current level is proportional to the intensity of the incident radiation and permits a direct measurement of the exposure rate. In an ion chamber with parallel-plate electrodes and a uniform electric field, current pulses from individual events can be registered, thereby allowing energy spectroscopy of the incident radiation. In their most common cylindrical geometry, however, ionization chambers cannot be used for this purpose since the pulse amplitude is dependent upon where the ion-electron pairs are formed in the detector. Gamma-Ray Ion Chambers Gamma-ray ion chambers, operated in the current mode, are very stable and have a long life. They can be fabricated in a variety of sizes and shapes. Large ion chambers are used as area monitors for ionizing radiation and high-pressure chambers offer a very high sensitivity, permitting measurement of exposure rates as low as 1 /ir/h. Small chambers with low gas pressures can be operated in radiation fields with exposure rates as great as 10 7 R/h. Air-filled ionization chambers vented to atmospheric pressure are often used to measure radiation exposure. After the formation of an electron-ion pair, the electrons and ions begin to move through the gas towards their collecting electrodes. However, the electronegative oxygen quickly captures the free electron to produce a heavy negative ion. Because of this attachment most of the free electrons disappear and the chamber current is produced by the motion of negative and positive ions. This is not a problem when a detector is operated in the current mode as long as the potential difference is sufficient to prevent significant recombination. However, such a chamber is not suited for pulse-mode operation, since the positive and negative ions move very slowly compared to electron speeds in the same electric field, and very broad and indistinct pulses are produced. Neutron-Sensitive Ion Chambers Because neutrons do not directly produce ion-electron pairs, an ion chamber sensitive to neutrons must incorporate some isotope with a large neutron cross section for a nuclear reaction that produces charged particles, which, in turn, create ionelectron pairs in the chamber gas. The most common gases in use are 3 He and BFs.
Alternatively, the inside of the tube may be lined with a neutron-sensitive material such as boron or 235 U which, upon absorbing a neutron, emits charged particles such as alpha particles or fission fragments into the chamber gas. Chambers lined with a fissile isotope are called fission chambers. Such chambers are often operated in the proportional-counter mode. Fission chambers are often used where there is a mixed radiation field containing a large component of gamrna rays. Since these detectors are operated in the pulse mode, the large voltage pulse formed by the large charge produced by the ionization of the heavy charged fission products or their resulting radioactivity makes it possible to discriminate against the much smaller gamma-ray voltage pulse. Fission chambers containing 235 U or 239 Pu are efficient thermal-neutron detectors often used for in-core nuclear reactor startup channels. A compensated ton chamber is a neutron detector that effectively discriminates against the gamma ray induced current. Typically the chamber has three concentric electrodes, one coated with a neutron sensitive material such as 23 U or a compound containing 10 B. Compensated ion chambers are widely used in nuclear reactors because of their ability to respond to neutron fields that vary up to ten orders of magnitude; i.e., these detectors have a very large "dynamic range." 8.1.2 Proportional Counters Proportional counters are gas-filled chambers operated in region III of Fig. 8.2. In this region, there is an internal multiplication in the chamber gas of the original ionelectron pairs created by the incident radiation. The physical mechanism for this multiplication process is discussed in the next section. This internal multiplication M is typically from 10 to 10,000. Because of the location of the gas multiplication avalanche, near the central wire, the output voltage pulse is not only amplified by a factor of M but the pulse amplitude is proportional to the total ionization energy deposited inside the active volume of the detector. This feature allows the proportional counter to be used as a spectrometer by using the output pulse amplitude to infer the energy of the incident radiation. Also, since the number of ion pairs is proportional to the initial ionization caused by the incident radiation, proportional counters can be used to distinguish between different types of charged particles. For example, one common application is to use a gas-flow counter to distinguish between beta particles and alpha particles. Gas Multiplication In a proportional counter, the electric field strength is large enough to increase the energy of the electrons generated in the active volume to a level sufficient to cause additional ionization. of the. cham- 25 im diameter ber gas. In a cylindrical counter, the elec- anode wire trie field strength varies inversely with the distance from the axis of the central Figure 8.3. An orthogonal view of an anode wire and essentially all of the mul- Blanche triggered by a single electron. I he density ot the shading indicates the tiplication occurs very near the central concentration of electrons formed in the electrode. This avalanche multiplication avalanche. After Knoll [2000].
process is illustrated in Fig. 8.3. Electrons produced by this avalanche are quickly collected at the anode leaving a sheath of positive-ion space charge. Regardless of where the initial ion pairs are formed in the active volume, most of the current during charge collection results from the motion of the positive ions moving from the anode wire to the cathode. Applications The proton recoil proportional counter is one of the few instruments available to measure fast neutron spectra in the energy range from 10 kev to about 800 kev. When a high energy neutron scatters from a hydrogen atom in the chamber gas, the recoil energy of the hydrogen nucleus (a proton) depends on the kinetic energy of the incident neutron. The recoil proton's energy is then registered as a current pulse in the external circuit of the proportional counter. From the measured distribution of the pulse magnitudes, the energy distribution of the incident neutrons can be inferred. Gas fillings for neutron energy spectroscopy are usually hydrogen or methane. Techniques have been developed to discriminate against gamma rays. Techniques are also available to infer tissue dose from neutrons by using counters filled with a tissue equivalent gas. Thin-window proportional counters are frequently used to measure low energy x-ray spectra. The efficiency is about 10% at 5 kev. To enhance the photoelectric response, the counters are filled with a counting gas mixture in which one component has a high atomic number, e.g., a mixture of xenon and CE^. Beryllium is a common window material, and, because of its low atomic number, it has a low photoelectric cross section and is relatively transparent to incident low energy photons. 3 He and boron trifluoride (BFs) gas-filled counters are mainly sensitive to slow neutrons based upon the two nuclear reactions: 3 He(n,p) 3 H and 10 B(n,a) 7 Li. Because of the large output voltage pulse caused by the alpha particles and the protons, these counters can be made essentially insensitive to gamma rays using pulse height discrimination. These detectors have many applications including thermal neutron diffraction, moisture content measurements in soil and concrete, nuclear material assay, and nuclear fuel fabrication monitoring. Many applications require that high-energy neutrons be moderated to increase the detector efficiency. 8.1.3 Geiger-Mueller Counters Geiger and Mueller developed the Geiger-Mueller (GM) detector in 1928. These radiation detectors are simple and robust devices that continue to be an important tool for sensing the presence of ionizing radiation. These gas filled detectors operate in region IV of Fig. 8.2, and have the remarkable property that the size of the pulse, or total charge produced in the active volume is independent of the ionization energy deposited by the initial ion pairs. Consequently, these detectors do not have the inherent capability of distinguishing between different types of radiation or measuring radiation energy. They can, however, be configured so that they are sensitive to both charged particles (alpha particles and beta particles) and electromagnetic radiation (x rays and gamma rays). A GM detector's sensitivity to charged particles is limited only by the thickness of the entrance window. The entrance window must be just thick enough to ensure that the chamber's filling gas does not escape. Once charged particles enter the
active volume of the detector, they are detected with almost 100% efficiency since it takes only one ion-electron pair to initiate the pulse formation process. Typical counter characteristics are shown in Table 8.1. Table 8.1. Typical GM counter specifications and characteristics. primary gas quench gas gas multiplication cathode (outer case) material anode wire diameter window thickness (mg cm" 2 ) window material operating voltage plateau slope dead time energy response for j3 particles energy response for x and gamma rays sensitivity for low energy /3 particles sensitivity for high energy beta particles sensitivity for x and gamma rays mode of operation spectroscopy applications advantages helium, argon or neon halogen or organic 10 4-10 8 stainless steel 0.1 mm 1.5-40 mica 500-1600 V 2-10 %/100 V 50-200 MS poor poor poor due to window thickness limitation excellent for thin window detectors good pulse none low cost and large output voltage pulse Pulse Formation in a GM Counter The initial ion pairs produced inside a GM tube are quickly separated in the electrical field between the anode and the cathode. As electrons move toward the anode, they gain sufficient energy to produce additional ionization. At this stage, the pulse formation follows a pattern similar to the operation of a proportional counter. Very quickly a dense collection of ion pairs forms near the anode, terminated when all of the electrons are collected. This initial avalanche is followed by a succession of other avalanches caused by the emission of light photons that are emitted from gas molecules excited by electron impact. These photons, in turn, interact with the walls of the detector and with the chamber gas to produce tertiary electrons, which spread the discharge over the whole volume near the positive electrode. This cascade of avalanches in a GM tube is shown in Fig. 8.4. The operation of the GM counter, thus, depends upon the probability that an electron in the gas produces a tertiary photoelectron. As the positive ion sheath grows near the anode, a space charge builds up. This causes the electric field near the wire to weaken and the multiplication to decrease. The discharge becomes convergent and ion pair production stops. During the collection of the electrons, the positive-ion sheath is essentially stationary because ion mobility is about 1000 times less than that for electrons.
Photon' f Y / r* '-v '- v ^_x~v_x-x_x-v_x*"* Y W Photon Anode wire Cathode Figure 8.4. The mechanism by which additional avalanches are triggered in a GM discharge. After Knoll [2000]. After collection of the electrons at the central anode, the positive ions slowly move toward the chamber's cathode. If they strike the wall, there is a finite probability that these ions will cause new electrons to be released into the active volume of the counter. If this happens, then the cascading avalanches begin anew. If the probability of this happening is high, the counter can enter a continuous discharge mode and destroy itself. These secondary avalanches caused by the positive ions are reduced to an acceptable level by adding a small quantity of a quench gas, normally a halogen or organic compound. The final result of this complex process is an internal gas multiplication of 10 4 to 10 8, which produces a very large charge and a correspondingly large output voltage pulse. Operating Voltage One of the first steps in characterizing the operation of any gas-filled radiation detector is establishing the correct operating voltage. GM counters have a fairly wide range over which they can be operated. However, each counter should be operated near the center of its plateau within the GM region shown as region IV in Fig. 8.2. This assures the best long-term performance because, at this voltage, small changes in the applied voltage have an insignificant effect on the amplitude of the output pulse (see Fig. 8.5). If the applied voltage is too low, then the gas multiplication is less than desired and the counting rate is very sensitive to small changes in voltage. Dead Time Corrections in GM Counters Because of the relatively long time required for the ionization process to take place in a GM counter, the chamber becomes insensitive to additional incident radiation during the ionization process. For large incident radiation intensities, the GM counter is unresponsive or dead for a large fraction of the counting time. These slow detectors require that dead time corrections be made even when the count rate
2000 1500-1000 500-750 850 950 1050 1150 Tube voltage (V) Figure 8.5. Operational plateau curve for a Geiger-Mueller counter. is relatively low. For any GM counter application, consideration must be given to systematic errors introduced by this count rate dependent phenomenon. Dead time corrections can be made to determine the true counting rate using the procedure discussed in Section 8.5.3. Electronics These detectors do not require expensive electronic components since the output signal is very large. Simple analog meters and basic pulse counting circuits are quite adequate. Applications GM counters are often the detectors of choice for applications requiring information about only the magnitude or intensity of the radiation field. As such they find wide application in hand-held survey meters used to detector radiation fields. Other types of radiation detectors are more suitable if information is needed on the type or energy of the radiation. GM counters can be made in essentially any size and shape. Their low cost and high efficiency make them suitable as sensors of beta particles, x-rays and gamma rays. This is especially true if the radiation level is low enough that dead time losses are not a concern. These detectors are seldom used to detect neutrons although, with a cadmium cover surrounding the tube, the GM detector is sensitive to thermal neutrons and may be calibrated for thermal neutron flux density. Because of their poor energy response, inability to distinguish among different types of radiation, arid their large resolving times, care must be taken to assure that some knowledge of the radiation field is available before making measurements with these counters.
8.2 Scintillation Detectors There are two types of scintillation detectors, (1) solid crystals of inorganic material, and (2) plastics and liquids consisting of organic molecules. Their modes of excitation differ but the final result is the same. As charged particles pass through the material the energy that they lose is transferred into excitation energy of the inorganic crystals or molecular excitation of the organic molecules. The excitation energy is released in fluorescence, i.e., scintillation. The number of light photons emitted in any one event is proportional to the energy lost by the initial charged particle in that event. The time dependence of the fluorescence emission, and hence the output pulse shape, is dependent upon the specific type of material. Although scintillator material has been used for almost 100 years to detect ionizing radiation, their widespread application dates from the development of the photomultiplier tube (PMT) some 50 years ago. This vacuum tube device allows the measurement of extremely low levels of light. In a PMT the incident photons strike a photocathode thereby liberating photoelectrons. These photoelectrons are then accelerated towards another electrode at a higher potential where the energetic impinging electrons cause more electrons to be emitted. This electron multiplication process continues along a series of electrodes, each at higher potential than the previous and, at each, the electron population is increased. The number of electrons finally collected at the last electrode may be millions of times greater than the number of electrons that began the cascade. In essence, the PMT is a photon to electron amplifier. A typical scintillator detector assembly consists of a hermetically sealed scintillation material optically mounted to the PMT's photocathode, a voltage divider string (YDS) in the PMT, and a preamplifier to produce a voltage pulse from the electrons collected at the last PMT electrode. These components are usually bound together in a single assembly By using this common configuration, the detector assembly is a stand-alone device that only requires an external voltage for the PMT and an external power supply for the preamplifier. Such an assembly is often called a scintillation detector. The amplitudes of the output voltage pulses are proportional to the energy deposited by charged particles produced in the scintillation material. A gamma ray penetrating the scintillator material may give up its energy to the scintillator material through photoelectric interactions, Compton scattering and pair production reactions. If all of the incident gamma-ray energy is deposited in the scintillator material, the number of scintillation photons produced is proportional to the incident gamma-ray energy. Thus, by measuring the distribution of pulse sizes or the pulse height distribution (PHD) produced by the scintillation detector, the energy distribution of the incident gamma rays can be determined. Thus, one of the most important applications of scintillation detectors is gamma-ray spectroscopy. Characteristics of some common scintillation detectors are shown in Table 8.2. Nal Scintillation Detectors The most popular inorganic scintillation material is Nal(Tl). These detectors are available in a variety of sizes and shapes. Because the maximum wavelength of light emitted by this material is 415 nm, it is easy to find commercially available PMTs whose maximum sensitivity matches the fluorescence emission spectrum. The
Table 8.2. Common scintillation materials. Material Decay constant Primary applications Comments Inorganic: Nal(Tl) 230 (1) x- and 7-ray detectors (2) x- and 7-ray spectrometers (1) hygroscopic (2) high Z material (3) 3x3 inch most common for 7 (4) 7-ray resolution of 6-8% (5) right cylindrical shape most common (6) well-type have outstanding efficiency Organic: BC-400 NE-102 plastic 2.4 (1) x- and 7-ray detectors (2) used in portal and large surface area detectors (3) a. (3 and fast n detectors (4) j3 spectrometers (1) non-hydroscopic (2) inexpensive (3) low Z; minimal photoelectric effect (4) high efficiency for charge particles (5) available in many sizes and shapes relatively large decay time constant is normally not a problem since a very high efficiency for x- rays and gamma rays dominates their radiation response. Of all the different Nal(Tl) detectors available to characterize gamma-ray radiation fields, the 3x3 inch right circular cylindrical detector, has historically been the favorite. This is the most extensively characterized Nal(Tl) detector and extensive efficiency data are available in the literature. A typical Nal(Tl) pulse height distribution is shown in Fig. 8.6. Because of its very high efficiency for electromagnetic radiation (see Fig. 8.7), Nal(Tl) is widely used to measure x rays and gamma rays. X-ray detectors with a thin entrance window containing a very thin Nal(Tl) detector are often used to measure the intensity and/or spectrum of low energy electromagnetic radiation. Because Nal(Tl) detectors do not require cooling, they can be used in a great variety of applications. Field applications are possible since they can operate over a long time period in warm and humid environments, resist a reasonable level of mechanical shock, and are resistant to radiation damage. Basically, for any application requiring a detector with a high gamma-ray efficiency and a modest resolution, the Nal(Tl) detector is clearly a good choice. Comparison of Inorganic Scintillators Advances in x-ray computed tomography and positron emission tomography have placed new demands on speed, efficiency, and effective atomic number of scintillators used for x-ray and annihilation photon detection. Table 8.3 compares with Nal several new inorganic scintillators. Organic Scintillation Detectors Organic scintillation detectors are also very popular. They provide an excellent option when a larger less expensive detector is needed. They can't be used for highenergy electromagnetic radiation spectroscopy since their low atomic number limits
Single Double escape escape r peak peak \ 500 1000 1500 2000 Energy (kev) 2500 3000 3500 Figure 8.6. Pulse height distribution of the gamma rays emitted by the radioactive decay of 24 Na as measured by a Nal(Tl) scintillation detector. In addition to the two photopeaks, corresponding to the complete absorption of the 24 Na 2.754 and 1.369-MeV gamma rays, several other peaks are also apparent. The single and double escape peaks arise from the escape of one and both 0.511-MeV annihilation photons generated in the Nal crystal by the annihilation of positrons created in pair-production interactions. The annihilation peak is caused by 0.511-MeV annihilation photons produced outside the Nal crystal and subsequently depositing their energy in the crystal. The backscatter peak arises from photons scattered from the source material into the Nal detector. 0.7 0.6 0.5 «u 0.4 0.3 3x1 Solid Crystal 2x2 I source 9.3cm 3x2 [ on axis j 0.2 0.1 0.0 0.20 0.40 0.60 0.80 1.0 1.2 Photon energy (MeV) 1.4 Figure 8.7. Intrinsic peak efficiency for Nal(Tl) detectors.
Table 8.3. Comparison of inorganic scintillation materials." Scintillator Nal(Tl) BGO Bi 4 Ge 3 Oi 2 LSO Lu 2 (SiO 4 )O GSO Gd 2 SiO 5 Relative light yield Peak wavelength (nm) Decay constant (ns) Density (g/crn'^) Effective Z Index of refraction 100 410 230 3.67 51 1.85 15-20 480 300 7.13 75 2.15 75 420 12, 42 7.40 66 1.82 20-25 440 30-60 6.71 59 1.85 "Compiled by Dr. Frederic H. Fahey. Wake Forest University Medical School. their photoelectric sensitivity. Thus, Compton scattering dominates from about 60 kev to 10 MeV. Moreover, since these detectors are composed of only hydrogen and oxygen with a density of 1.032 g cm~ 3 they make a u near tissue equivalent" detector. Typical applications include portable survey instruments, portal monitors, and beta-particle spectrometers. 8.3 Semiconductor lonizing-radiation Detectors Functioning of these radiation detectors is based upon the newest technologies. The impact on radiation detection and measurement has been revolutionary because of unique semiconductor properties, especially outstanding energy resolution. New semiconductor detectors continue to be introduced into the market place. The three most common types are described in Table 8.4. Germanium Semiconductor Detectors There are two main types of germanium semiconductor detectors: (1) Ge(Li) a germanium crystal doped with interstitially drifted lithium ions to cancel the effect of natural impurities in the germanium crystal, and (2) the more recent HPGe high purity germanium crystal in which impurity atom concentration are less than 10 10 cm~ 3. The more expensive HPGe detectors have replaced the older Ge(Li) technology since they can be kept at room temperature when not in use whereas Ge(Li) crystals must always be kept at liquid nitrogen temperatures. Germanium detectors, besides having exceptional energy resolution, are very efficient for detecting photons. Their efficiency ranges from excellent for low energy x-rays to good for medium to high-energy gamma rays over an energy range of 1 kev to 10 MeV. The performance of these detectors is often compared to Nal(Tl) and Cd/Zn telluride (CZT) detectors. Because of the higher atomic number arid larger size, Nal(Tl) detectors often have a higher efficiency for high energy gamma rays than do germanium detectors, but a much poorer energy resolution. The dramatic difference in the energy resolution between Na(Tl) and Ge(Li) spectrometers is shown in Fig. 8.8.
0) c 0) c o -(-> o JC. 0. o i i Iniii i i in o n CM O O O O o uiq jad sq.unoo S w) o < 00 O s 5 M) X
Table 8.4. Common semiconductor ionizing-radiation detectors. Material Resolution (FWHM) Principal application Comments HPGe 2.0 kev at 1.33 MeV 1.0 kev at 122 kev 0.15 kev at 5.9 kev x- and 7-ray energy measurements (1) excellent resolution (2) intermediate atomic number (3) high cost-varies with 7 eff. (4) cooling needed when used (5) small volume Si(Li) 0.14 kev at 5.9 kev from Fe x-ray (low kev) and charged particles (especially a particles) energy measurements (1) excellent resolution (2) low atomic number (3) low cost (4) doesn't require cooling (5) small volume (6) thin entrance windows Cd/Zn/ Telluride (CZT) 1.5 kev at 122 kev from 5T Co x-ray and low energy gamma-ray energy measurements (1) good resolution (2) high atomic number (3) expensive (4) minimum cooling needs (5) available only in small sizes Silicon Semiconductor Detectors Si (Li) detectors with thin entrance windows are commonly used in alpha- and betaparticle spectrometers. They can be configured to achieve essentially 100% intrinsic efficiency and have excellent resolution. They also offer an inexpensive option for x-ray spectroscopy. Since Si(Li) detectors have a much lower atomic number than CZT. Nal(Tl), and HPGe, their relative efficiency per unit thickness is significantly lower for electromagnetic radiation. However, for x-ray or gamma ray energies less than about 30 kev, commercially available Si(Li) detectors are thick enough to provide performance which is superior to CZT, Nal(Tl), and HPGe and detector materials. For example, as shown in Fig. 8.9, a 3 to 5 mm-thick detector with a thin entrance window has an efficiency of 100% near 10 kev. Based upon the fact that a majority of the applications require a thin window, Si(Li) detectors are often manufactured with very thin beryllium windows. Cadmium Zinc Telluride Detectors Cadmium zinc telluride (CZT) is a new high-resolution and high-atomic number semiconductor detector material. A reasonable degree of cooling for the detector and the directly coupled preamplifier enhances detector system performance. CZT detectors offer an excellent option for low energy x-ray spectroscopy where cooling is not possible. Keeping the detector and preamplifier at about -30 C, is adequate to achieve optimum energy resolution. By contrast, HPGe detectors must be cooled at liquid nitrogen temperatures to achieve optimum resolution. These detectors are not available in large sizes. Their small size diminishes the possibility of making detectors with large efficiencies for high-energy electromagnetic radiation (see Fig. 8.10). Therefore, the major application is low energy x- or gamma-ray spectroscopy.
100 80 60 40 20 0 10" 10' Photon energy (kev) 10' Figure 8.9. Full energy peak efficiency of Si(Li) detectors. hotoelectric + Compton 0 200 400 600 800 1000 Gamma-ray energy (kev) Figure 8.10. Gamma-ray efficiency for a 2 mm thick CZT detector.
When x-rays or gamma rays interact in a CZT detector, an average of one electron/hole pair is produced for every 5 ev of energy lost by the photoelectron or Compton electron. This is greater than in Ge or Si, so the resolution of these detectors is not as good as HPGe or Si(Li) detectors. However, they are the detector of choice for many applications requiring good energy resolution. For example, they perform better than HPGe when it is not feasible to provide liquid nitrogen cooling. Likewise, when detector sensitivity is the dominant consideration, they are much better than Si(Li). 8.4 Personal Dosimeters 8.4.1 The Pocket Ion Chamber Very familiar to radiation workers over many, many years are the self-reading pocket ionization chamber (PIC) and the film badge dosimeter. The pocket ionization chamber is an ion chamber, as described in Section 8.1, in the form of a cylinder about the size of a fountain pen. A charge is placed on the electrodes of the ion chamber and the corresponding voltage is displayed through an eyepiece using an electroscope. As the ion chamber receives radiation exposure, the electrodes are discharged and the voltage change of the electroscope is presented in a reticule scaled to radiation dose or exposure. More commonly, the PIC is sensitive only to gamma radiation. However, neutron sensitive ionization chambers are also used, calibrated in dose equivalent. 8.4.2 The Film Badge The film badge consists of a packet of photographic film sealed in a holder with attenuating filters. Ionizing radiation darkens the film, as in the production of an x-ray image. The filtration is designed to render the degree of film darkening as nearly as possible a known function of gamma-ray exposure, independent of the energy of the incident gamma rays. After the badge is carried by a radiation worker for a period of time, the film is processed, along with calibration films with the same emulsion batch exposed to known radiation doses. The worker's radiation dose for the period is assessed and ordinarily maintained in a lifetime record of exposure. In some cases special attenuation filters are used to relate the darkening of portions of the film to beta-particle or even neutron dose. Special badge holders in the form of rings or bracelets are used to monitor the radiation exposure of hands, wrists, and ankles. 8.4.3 The Thermoluminescent Dosimeter The thermolumiriescent dosimeter (TLD) is a solid state radiation detector, whose radiation dose may be gauged by measurement of the release of light upon heating of the detector after exposure. TLDs are inorganic crystals such as LiF or CaSO4 to which impurity elements, or dopants, such as Mg. P, Mn, or Cu are added in small concentrations. The TLD matrix is an insulator but the dopant adds hole and electron traps with energy levels within the insulator band gap. Upon radiation exposure, traps are filled. They remain filled until the TLD is heated, thereby releasing the trapped charge carriers. Recombination of newly mobile charged carriers leads to light emission. Natural Li, in which the abundance of 6 Li is 7.4%, is
somewhat sensitive to thermal neutrons. TLD neutron sensitivity may be enhanced by increasing the abundance of the lighter isotope. Similarly, neutron insensitive LiF employs only 7 Li. Dosimetry of mixed neutron and gamma-ray radiation may be accomplished by using pairs of neutron-sensitive and neutron-insensitive TLDs. 8.5 Measurement Theory 8.5.1 Types of Measurement Uncertainties The analysis of any type of experimental data always requires that an assessment be made of the uncertainties associated with each measurement. Without such an estimate, the data have very limited value. There are several types of uncertainties associated with any measurement. These include stochastic and sampling uncertainties and errors as well as systematic errors. For example, the decay of radioactive atoms occurs randomly or stochastically so that a measurement of the number of decays in a fixed time interval has an inherent stochastic uncertainty. Repeated measurements would give slightly different results. Systematic errors are introduced by some constant bias or error in the measuring system and are often very difficult to assess since they arise from biases unknown to the experimenter. Sampling errors arise from making measurements on a different population from the one desired. These biases too are hard to detect, let alone quantify. Engineers and scientists must always be aware of the difference between accuracy and precision, even though popular usage often blurs or ignores the important distinction. Precision refers to the degree of measurement quantification as determined, for example, by the number of significant figures. Accuracy is a measure of how closely the measured value is to the true (and usually unknown) value. A very precise measurement may also be very inaccurate. 8.5.2 Uncertainty Assignment Based Upon Counting Statistics Data measured with an ionizing-radiation detection system embody both random uncertainties and systematic errors. Uncertainty assignment requires knowledge of both. Even with a perfect measurement system capable of operating over a long time period without introducing significant systematic error, one must always consider the random nature of the data caused by the stochastic decay of radioactive nuclei. This section deals only with one component of the total uncertainty the random or statistical uncertainty, which can be estimated using the binomial distribution since it is the basic statistical distribution describing the stochastic decay process of radionuclides. It is usually possible to approximate the binomial distribution by a Poisson distribution, which, for a sufficiently large number of decays or measured events (> 40), can be well approximated by a Gaussian distribution. This approximation simplifies the process of assigning the statistical uncertainty since, for the Poisson distribution, the variance cr 2 is equal to the mean value x, the value usually measured. Data are routinely reported as x ± s where x is the number of measured counts during a specified time period and s is the standard deviation of x. When an error is reported as one standard deviation (or "one sigma") it is called the standard error. For a single measurement x of a radioactive sample, it can be shown that
the standard deviation, for x greater than about 40, is s = v/x (8.1) Thus, for a single count measurement, the expected or mean value is estimated as x with an estimated standard deviation of s = ^/x, so the reported value is x±s. Don't arbitrarily use this concept to assign uncertainties to other types of experimental measurements such as pressure or temperature measurement, which have totally different statistical distributions from the Poisson distribution governing radioactive decay. Frequently, more than one count measurement is recorded. For this case, assuming the radiation source has constant activity, estimates of the mean and standard deviation can be obtained directly from the TV measured values, Xi, 2,, ^TV- The mean or expected count is the average value of the counts, i.e., N-l x ~j^2^xi - \^ 1=1 and the corresponding estimate of the standard deviation of the mean is s = \/x/n - I Note that for replicate measurements the error is reduced by the square root of N. This assumes, of course that the measurements are independent and that no changes occur in the radiation field or with the detector's response during the time that the N data points were being recorded. If a Gaussian distribution can approximate the data, then s has a specific meaning, namely, there is a 68% probability that the mean value falls within the range x ± s. This means that if the measurement were repeated a large number of times, 68% of the observed counts would fall within x±s. Table 8.5 shows the relationship between the percent standard deviation and the number of counts recorded. For example, a standard error of one percent or less would require recording at least 10,000 counts. Uncertainties are not always reported as one standard deviation. Sometimes a larger uncertainty e is reported in order to increase the probability that the mean will be included within the range between x e and x + e. As shown in Table 8.6, if the error is reported as one standard deviation, the probability is 68%. However for an error range of x 1.65s to x + 1.65s the probability increases to 90%. This would be reported as the 90% confidence interval. The convention is to report errors as one standard deviation. For anything other than one standard deviation, one should specify the number of standard deviations or the probability interval. 8.5.3 Dead Time All radiation detection systems operating in the pulse mode have a limit on the maximum rate at which data can be recorded. The limiting component may be either the time response of the radiation detector, which is the case for the Geiger counter, or the resolving capability of the electronics. The true counting rate for a
Table 8.5. Standard deviation (%) of count data measured with a radiation detector operating in the pulse mode. Table 8.6. Probability interval for different number of standard deviations based upon a Gaussian distribution. Observed counts x Percent standard error (1 sigma error) No. of standard deviations Probability 100 400 1100 2500 10000 10% 5% 3% 2% 1% 0.68 1.00 1.65 1.96 3.00 0.50 0.68 0.90 0.95 0.997 detector with zero dead time losses (n) is related to the recorded counting rate for a detector with significant dead time losses (m) by ra n = 1 mr (8.4) where T is the dead time of the detector. Note that the term mr is the fraction of the time that the detector is unable to respond to additional ionization in the active volume of the detector. When designing an experiment, it is advisable to keep these losses to a minimum. If possible, this means that mr should be less than 0.05. For example, for a GM counter with a typical dead time of r = 100 yus, the maximum count rate would be 500 counts/s. 8.5.4 Energy Resolution One important application for radiation detectors is radiation spectroscopy, measurement of the energy distribution of a given type of radiation. Such spectrometers requires measuring some quantity that is proportional to the energy of the incident radiation. Typically, that quantity is the amplitude of a voltage pulse produced from collection of the separated charge deposited in the detector. Raw spectroscopy data, known as pulse height distributions (PHD), display count rate vs. pulse height. They contain peaks whose relative positions are related to the energy of the incident radiation. By processing a PHD, it is possible to determine the energy distribution of the incident radiation. BIBLIOGRAPHY BEVINGTON P.R. AND D.K. ROBINSON, Data Reduction and Error Analysis for the Physical Sciences, 2nd ed., McGraw-Hill, New York, 1992. EiCHHOLZ, G.G. AND J.W. POSTON, Principles of Radiation Detection, Ann Arbor Science, Ann Arbor, MI, 1979. KLEINKNECHT, K., Detectors for Particle Radiation, Cambridge University Press, U.K., 1999. KNOLL, G.F., Radiation Detection and Measurement, 3rd ed., Wiley, New York, 2000. PRICE, W.J., Nuclear Radiation Detection, McGraw-Hill, New York, 1964. TSOULFANIDIS, N., Measurement and Detection of Radiation, Hemisphere, New York, 1983.
PROBLEMS 1. Should the quenching gas in a GM tube have a higher or lower ionization potential than the major tube gas? Why? 2. What effect does each of the following changes have on the performance of a proportional counter? (a) The diameter of the anode wire is increased, (b) The pressure of the fill gas is increased, (c) The atomic number of the wall material is increased. 3. For 1-MeV photons, sketch how the detector efficiency of a proportional or GM chamber varies with the thickness of the cylindrical wall. 4. A given GM tube has a dead time of 0.25 ms. If the measured count rate is 900 counts per second, what would be the count rate if there were no dead time? 5. The decay constant for Nal(Tl) fluorescence radiation is about 230 /its. How long must one wait to collect 90% of the scintillation photons? 6. In anthracene, scintillation photons have a wavelength of 447 nm. If 1-MeV of energy is deposited in an anthracene crystal and 20,000 scintillation photons are produced, what is the scintillation efficiency? 7. Why is air often used as a tube fill gas in an ionization chamber, but not in a proportional counter? 8. If the energy resolution of a Nal(Tl) detector is 8%, what is the FWHM of the full-energy peak for a 137 Cs source? 9. Which detector has the greatest full-energy peak efficiency: Nal(Tl), HPGe, Si(Li). or CZT for (a) 100-keV gamma rays, and (b) for 1-MeV gamma rays? Justify your answer.