Understanding neutron radiography post exam reading viii part 1 of 2a

Understanding Neutron Radiography Reading VIII Part 1of 2

13th 2016 August Post Exam Reading

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Reactor

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Reactor

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The Magical Book of Neutron Radiography

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数字签名者：Fion Zhang DN：cn=Fion Zhang, o=Technical, ou=Academic, email=fion_zhang@ qq.com, c=CN 日期：2016.08.30 20:15:39 +08'00' Charlie Chong/ Fion Zhang

ASNT Certification Guide NDT Level III / PdM Level III NR - Neutron Radiographic Testing Length: 4 hours Questions: 135 1. Principles/Theory • Nature of penetrating radiation • Interaction between penetrating radiation and matter • Neutron radiography imaging • Radiometry 2. Equipment/Materials • Sources of neutrons • Radiation detectors • Non-imaging devices

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3. Techniques/Calibrations

• Electron emission radiography

• Blocking and filtering

• Micro-radiography

• Multifilm technique

• Laminography (tomography)

• Enlargement and projection

• Control of diffraction effects

• Stereoradiography

• Panoramic exposures

• Triangulation methods

• Gaging

• Autoradiography

• Real time imaging

• Flash Radiography

• Image analysis techniques

• In-motion radiography • Fluoroscopy

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4. Interpretation/Evaluation • Image-object relationships • Material considerations • Codes, standards, and specifications 5. Procedures • Imaging considerations • Film processing • Viewing of radiographs • Judging radiographic quality 6. Safety and Health • Exposure hazards • Methods of controlling radiation exposure • Operation and emergency procedures Reference Catalog Number NDT Handbook, Third Edition: Volume 4, Radiographic Testing 144 ASM Handbook Vol. 17, NDE and QC 105 Charlie Chong/ Fion Zhang

Fion Zhang at Copenhagen Harbor 4th August 2016

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SME- Subject Matter Expert http://cn.bing.com/videos/search?q=Walter+Lewin&FORM=HDRSC3 https://www.youtube.com/channel/UCiEHVhv0SBMpP75JbzJShqw

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Gamma- Radiography TABLE 1. Characteristics of three isotope sources commonly used for radiography.

Source

T½

Energy

HVL Pb

HVL Fe

Specific Activity

Dose rate*

Co60

5.3 year

1.17, 1.33 MeV

12.5mm

22.1mm

50 Cig-1

1.37011

Cs137

30 years

0.66 MeV

6.4mm

17.2mm

25 Cig-1

0.38184

Ir192

75 days

0.14 ~ 1.2 MeV (Aver. 0.34 MeV)

4.8mm

?

350 Cig-1

0.59163

Th232

Dose rate* Rem/hr at one meter per curie

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0.068376

八千里路云和月

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闭门练功

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http://greekhouseoffonts.com/

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COLLIMATED NEUTRON BEAM FOR NEUTRON RADIOGRAPHY M. DINCA1, M. PAVELESCU2, C. IORGULIS1 1 Institute for Nuclear Research, P.O. Box 078, Pitesti, Romania, dinca@scn.ro 2 Romanian Scientist Academy, Bucharest, Romania, mpavelescu@pcnet.ro Received October 21, 2005

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Pitești

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The obtaining of a collimated neutron beam on the tangential channel of the ACPR reactor from INR Pitesti that to satisfy the requests of a neutron radiography facility it is presented. The collimation of neutrons means the elimination from the neutron beam of those neutrons that have trajectories that are not inside the space defined by walls or successive apertures that are made of neutron absorbent materials. The assembly that assures the collimation of neutrons, named collimator, is optimized using MCNP 4B code based on Monte Carlo method for neutrons and gamma radiation.

Key words: neutron radiography, collimator for neutrons, collimation ratio, MCNP 4B code.

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1. INTRODUCTION A tangential channel of a nuclear reactor has some peculiarities regarding intensity and energetic spectrum of neutrons in comparison with a radial channel of a nuclear reactor or tubes used to extract neutrons from other neutron sources.

On a tangential channel the neutron beam has a bigger cadmium ratio and a lower gamma contamination than on a radial channel and is more suited to be used for thermal neutron radiography. For neutron radiography, different of other nuclear physics applications that use neutron beams, are necessary large neutron beams to obtain images of a large area of the investigated objects. An ideal neutron beam should be parallel, monoenergetic, with big intensity, free of other contaminant radiation and uniform on its cross section.

In practice it is intended to have experimental arrangements that to accomplish neutron beam parameters as closely as possible to ideal ones.

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For this purpose it is used a collimator. The neutrons pass through a collimator from the entrance aperture placed nearby neutron source to the exit window where are used for neutron radiography investigations. The inner space of a collimator is evacuated or filled with air, or better filled with helium.

A characteristic parameter of a collimator that defines the degree of divergence of the neutron beam is the L/D ratio, where L is the length of the collimator and D is the diameter (or generally the opening) of the entrance aperture.

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The place from where thermal neutrons start (the source of neutrons) is a moderator that contains neutrons moving in all directions. In order to have a neutron beam on a direction, nearby the moderator it is placed a collimator.

The neutrons entering in the collimator must have the direction of the exit window to be useful otherwise they are captured by walls or apertures to avoid the scattering. The entrance aperture must be big enough to permit a larger number of neutrons to go inside the collimator but small enough to have a bigger L/D ratio. The L/D ratio depends also by the length of the collimator (or otherwise by the distance from entrance aperture to object plane if the object is put far away from collimator), a bigger L means a better resolution.

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Because the moderator emits neutrons in all directions, their intensity is proportionally with 1/r2. To have a bigger intensity the object must be placed closer to neutron source but for a better geometrical resolution it must be placed farther. Bigger neutron intensity determines a better statistics, therefore a bigger contrast of the image that is able to differentiate between different materials. But for dimensional measurements it is necessary to have precise separation lines, therefore a big geometrical resolution.

A compromise must be made between the two parameters, L and D. A transmission method for neutron radiography it is involved because are detected the neutrons that pass through investigated object. If the neutrons come to investigated object more scattered, then the projection of a detail is larger in the plane of the detector and the geometrical resolution of the image is poorer.

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There are known different types of collimators, more important are: ■ pin-hole, ■ Soller and ■ divergent collimators.

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Soller Collimator

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Sรณller Collimator

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http://pd.chem.ucl.ac.uk/pdnn/inst3/soller.htm

The photograph shows the front opening of the 10′ Soller collimators of the detector bank of D1A taken before its rebuild in the late 1990's. The protective shielding has been removed so that thin vertical mylar sheets covered in a white gadolinium oxide paint are visible. The ones shown here were designed to collimate neutrons to 10 arc minutes (0.17°). (The little hole seen on the detector bank was a large "pin-hole" collimator, positioned in front of the normally unused 11th detector.)

The Soller collimators designed for use on neutron diffractometers have large dimensions as illustrated by the figure below for 5′ collimation. The figure shows just two foils, but in practice many parallel foils are required since the diameter of the detectors is 2 to 5 cm.

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http://pd.chem.ucl.ac.uk/pdnn/inst3/soller.htm

The town of Sóller in the northwest of Mallorca became wealthy because of the valley’s abundant citrus groves. In the 19th century, when the area was isolated from the rest of Mallorca by mountains, the oranges were shipped to France from the nearby west coast Port de Sóller (or Puerto de Sóller). Many locals went to work in France and returned – their fortunes duly made – to build some of the handsome Modernista properties that grace this town today.

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The pin-hole collimator The pin-hole collimator has a simple construction. An aperture is placed at a distance from neutron source in order to establish a L/D ratio of the collimator.

For a pin-hole collimator it is necessary a large neutron source that to have an equal neutron flux on its surface in order to expose uniformly the object to neutrons.

The Soller collimators At Soller collimators appear on image the network of absorber walls that delimits inner minicollimators. This type of collimator requires a large uniform neutron source.

The divergent collimator The most used is the divergent collimator because it permits the investigation of large objects, every point of the object being exposed to a neutron beam with approximately the same L/D (this means an intrinsic geometrical resolution uniform in the exit window of the collimator). A divergence collimator has the neutron source in its entrance aperture.

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Based on dimensional constrains of the tangential channel of ACPR, previous experimental determinations of the thermal neutron flux and intensity (8 â‹… 1011 n/cm2/s near core and 1.12 x 106 n/cm2/s at the exit of tangential beam tube, at 100 kW operating power of ACPR) and working methods involved, were established the parameters of the divergent thermal neutron beam. Some of them are:

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the thermal neutron beam intensity at least 5 ⋅105 n/cm2/s; the collimation ratio, L/D, at least 90; the exit window, 250 mm in diameter; the n/γ ratio at least 1⋅106 n/cm2/mrem (that determines used investigation methods); • the divergent angle under 40°; • the cadmium ratio above 17.

• • • •

Note: Cadmium Ratio The ratio of the response of an uncovered neutron detector to that of the same detector under identical conditions when it is covered with cadmium of a specified thickness.

Hint: the larger the cadmium ratio, the more the thermal neutron with energy less than 0.5Mev?

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To obtain a thermal neutron beam with such parameters were used: 1. a graphite illuminator placed on channel nearby reactor core to scatter neutrons towards exit of the channel; 2. a mobile monocrystaline bismuth filter for the attenuation of the gamma radiationand scattering of fast neutrons that will allow performing direct neutron radiography investigations and also Îł radiography investigations; 3. a set of successive apertures from boron, indium and lead for the formation of the divergent collimator.

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The position and dimensions of these components were optimized by calculus made with MCNP 4B code based on Monte Carlo method both for thermal neutrons and both for gamma radiation.

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ACPR- Annular Core Pulse Reactor

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2. CALCULUS WITH WIMS 4D AND MCNP 4B CODES The tangential beam port has an overall length of 5644 mm (Figure 1) and has two sections. First with the length of about 2984 mm and the diameter of 219 x 6.5 mm, and second with the length of about 2660 mm and the diameter of 273 x 6.5 mm.

2660 mm

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The distance between the center of the reactor and the beam port axis is 575 mm. The beam port exceeds with 508 mm, to the axis of the pool, the perpendicular right line on its own axis that passes through the center of the core. The beam port contains a mobile lead shutter with the thickness of 381 mm and 406 mm in diameter placed at 1015 mm from beam port exit.

Between the edge of the core and the tangential beam port is a distance of 157.8 mm. The space between core and beam port is filled with regular demineralised water. A better transmission of the neutrons from core to channel will be assured placing aluminum in free locations of the reactor grid.

In this way the reduction of the initial thermal neutron flux of the channel through divergent collimator construction is compensated. (?) mobile lead shutter

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Fig. 1. Sketch of the collimator of the neutron radiography facility at the tangential beam port of the ACPR.

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The optimization of the transmission of neutrons to channel and the optimization of the dimensions and positions of the collimator components is done using WIMS D4 and MCNP 4B codes. To establish the spectrum of the neutron flux at the edge of the ACPR reactor, the transport program WIMS D4 was involved. Because of cylindrical shape of the reactor, it is suitable to be modeled by WIMS program. The model consists of cylindrical rings that cover the central hole, ACPR fuel, water etc. The neutron flux calculated for 69 broad groups in a thin volume at the edge of the core has been collapsed in 3 or 23 groups. Their weights, after the renormalization to unit and upper boundaries for energy groups were used in the inputs of the MCNPprogram.

The WIMS D4 code was used to study the effect of the replacement of the water between core and beam port with an aluminum block, aluminum pins placed in grid’s holes or air in aluminum box. The replacement of the water leads to an improvement of the transfer of the neutrons towards beam port.

The results of the calculations are shown in Table 1. It can be seen that the increase of the thermal neutron flux is maximum using a box filled with air. To disturb not other experiments for irradiation tests, a bell box is put in place from where the water is pushed out and replaced by air.

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Table 1: Relative units of the thermal flux in the graphite illuminator for some materials between core and tangential beam port Water Aluminum pins Aluminum block

Air

4.92

19.5

14.3

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16

In order to assure a maximum thermal neutron beam at the exit of the collimator and a suitable established collimation ratio were performed Monte Carlo calculation based on MCNP code. Two models were prepared for Monte Carlo calculations. The first model aimed to establish the thickness and position of the graphite illuminator for the maximum increase of the thermal neutron beam at the exit of the collimator. This model contains the source of neutrons offered by WIMS code for 3 and 23 groups, the box with air and the illuminator placed on beam port. The relative values obtained in a plane at 100 cm from illuminator, for different thicknesses of the illuminator are shown in Figure 2. The illuminator is placed near centerline. If the illuminator is placed in a centered position the thermal neutron flux is a little improved, but epithermal and fast neutrons increase more and it is not desirable. The maximum neutron beam is obtained for 6.5 cm and 7 cm illuminator thicknesses.

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Fig. 2. Neutron beam intensity.

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The second model targeted to establish the position and thickness of the single-crystal Bi filter, to obtain the maximum thermal neutron beam at the exit of the collimator. This model is based on the geometry of the collimator shown in Figure 1 and the source of neutrons is placed on the face of the illuminator. We consider the maximization of neutron flux below 1.E-06 MeV. On the geometry of the second model calculations were done for gamma radiation also. Based on previous flux measurements and the results estimated from first MCNP model, it was established a value of 4.5 cm for the diameter of the collimator main aperture. Preliminary results obtained with the second model established a value of 3 cm for the thickness of the Bi singlerystal. The main aperture will be built by 13 mm of boral, 1 mm indium and 200 mm lead.

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To optimize the position of the aperture and Bi filter, MCNP calculation were done for different positions of the filter. The results are shown in Figure 3.

Supplementary, it was used the condition to have a uniform intensity of the neutron beam in the exit window of the collimator. This was precisely established with AutoCAD program that drawn the extreme lines of the neutron beam. In this way every point in the exit window is seen by the same area from the surface of the illuminator. In these conditions it was established the maximum distance between illuminator and aperture to be 152.5 cm, although the maximum of the neutron beam is obtained for the distance of 190-200 mm. The calculations for the distance of 152.5 cm, the main aperture of 4.5 cm and 3 cm of Bi indicates a decrease of the gamma radiation of 65.19 times, and for neutrons of 16.15 times (the Bi filter itself decreases the beam intensities 8.22 and 2.22 times, respectively). The calculations with MCNP code were done with polycrystalline Bi. In the real case, for singlecrystal Bi with cross-section 3 times smaller at room temperature [1], it is expected a reduction of the beam intensity with 41% instead of 2.22 times reduction.

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Fig. 3. Intensity of the thermal neutron beam at the exit of the collimator.

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The minimum distance between illuminator and main aperture is considered to be 125 cm. For this distance the intensity of neutron beam decreases with 17%, but the resolution increases. The lead ring (20 cm) should be positioned at less 125 cm were is the edge of the concrete wall of the pool, otherwise the direct gamma radiation from reactor core cannot be properly stopped. The secondary apertures are positioned to avoid any trajectory of the neutron directly from illuminator to reach the wall of the beam tube. The secondary apertures are boral plates and lead rings. To increase the neutron beam for the direct method and to perform gamma radiographs it is designed to remove vertically the Bi filter with the help of a steel cable. The Bi filter is inside of a box, which contains lead ballast to fall back on position when cable is released.

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3. CONCLUSIONS The collimation of the neutrons on the tangential beam port of the ACPR reactor is done, in fact, with a pin-hole collimator with an aperture of 45 mm placed at the distance of 125-152.5 cm from the surface of the illuminator that has a thickness of 6.5 cm and the diameter of 18 cm. The estimated beam intensity for thermal neutrons with bismuth filter is 3.96⋅105 – 4.65⋅105 n/cm2/s and 4.85⋅105 – 5.70⋅105 n/cm2 /s without Bi filter. The estimated values for gamma debit doses (for 152.5 cm illuminator-main aperture distance) are 1.75 rem/h without bismuth and 213 mrem/h with bismuth. The estimated n/gamma ratio is 1.03⋅106 n/cm2/mrem and 8.44⋅106 n/cm2 /mrem, respectively. The divergent angle of the collimator is 3o- 3.3o and the collimation ratio 100-92.8 for the domain of distances 125-152.5 cm betweenilluminator and main aperture. These values of beam intensity, n/gamma ratio and collimation ratio are in concordance with that from other facilities built at TRIGA reactors and offer the base to use with good results the direct and the transfer methods for neutron radiography.

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Sandia’s Annular Core Research Reactor conducts 10,000th operation

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https://share.sandia.gov/news/resources/news_releases/acrr/#.V65mc-Qkpdg

ACPR- Annular Core Pulse Reactor

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ACRR- Annular Core (Pulse) Research Reactor

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https://share.sandia.gov/news/resources/news_releases/acrr/

ACRR- Annular Core (Pulse) Research Reactor

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https://share.sandia.gov/news/resources/news_releases/acrr/

ALBUQUERQUE, N.M. – With a muffled “pop,” a flash of blue light and a few ripples through 14,000 gallons of deionized water, Sandia National Laboratories’ Annular Core Research Reactor (ACRR) recently conducted its 10,000th operation.

“The ACRR has been a real workhorse for Sandia, and labs leadership and the nation rely on these experiments and other weapons component testing done at Sandia to support certification of the nuclear weapon stockpile,” said Lonnie Martin, an ACRR operator.

In its 32-year history, the ACRR time and again has proved itself a valuable resource for a wide variety of experiments in nearly every branch of nuclear science, especially the testing of radiation-hardened electronic components.

With a dry, 9-inch diameter cavity in the core’s center, and a 20-inch diameter external cavity, the ACRR subjects electronics to high-intensity neutron irradiation and conducts reactor safety research. The ACRR also has done testing for semiconductor manufacturers, NASA, the Large Hadron Collider in Switzerland and dozens of other users.

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Sandia’s ACRR is a water-moderated, pool-type research reactor capable of steady-state, pulsed and tailored transient operations and, in the past, has been configured for medical isotope production. Other duties for ACRR include: reactor-driven laser experiments; space reactor fuels development; pulse reactor kinetics; reactor heat transfer and fluid flow; electronic component hardening; and explosive component testing. It is also routinely used for education and training programs.

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At peak power in its steady state mode, the ACRR produces up to four megawatts of power. But during a maximum pulse, it generates a whopping 35,000 megawatts of power for seven milliseconds. Nuclear engineer and former University of New Mexico professor Ron Knief compares its power output to that of the Palo Verde Nuclear Generating Station, outside of Phoenix. “For that very short time, we produce three times more power than the nation’s largest nuclear power station. They have three big reactors, and yet, for a fraction of a second, we produce three times more power than they do,” Knief said.

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The ACRR is a descendent of the Sandia Annular Core Pulse Reactor (ACPR), which was replaced in 1978 and is part of a large family of Training, Research Isotope Production, General Atomics (TRIGA) reactors. The TRIGA concept is credited to Manhattan Project physicist Edward Teller and a group of distinguished scientists who assembled the first model in a “Little Red Schoolhouse” in San Diego in 1956. Teller’s mandate to the team was to “design a reactor so safe … that if it was started from its shut-down condition and all its control rods instantaneously removed, it would settle down to a steady level of operation without melting any of its fuel,” according to Freeman Dyson’s, “Disturbing the Universe.” Essentially, even if all the engineered safety mechanisms failed, the reactor would operate safely, based on the laws of physics.

In 1978, the original ACPR TRIGA fuel was replaced with a new ACRR ceramic-metal, uranium dioxide/beryllium oxide (UO2/BeO) fuel, which is designed to allow steady state and pulsed operation at fuel temperatures up to 2,552 degrees (1,400 degrees C). The reactor underwent extensive upgrades in 2002, including upgrades to reactivity control circuitry.

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Key!

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Reactor

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Reactor

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Radiography may be considered the most effective nondestructive testing method merely because of its universal use and acceptance in industry.

Radiography can be used to test most types of solid material. Exceptions include materials of very high or very low density. Neutron radiography, however, can often be used in such cases. There is wide latitude both of material thickness that can be tested and in the techniques that can be used.

Usually conditions that result in a two percent or greater difference in through- section thickness can usually be detected. (≼2% subject sensitivity

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Read The following article: • E 748-95, Standard Practices for Thermal Neutron Radiography of Materials • E 803, Standard Test Method for Determining the L/D Ratio of Neutron Radiography Beams • E 1496-97, Standard Test Method for Neutron Radiographic Dimensional Measurements

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Chapter 5 Radiation Measurement PART 6. Neutron Detection 1.0 Characteristics The neutron is a part of the nucleus, has no charge and is somewhat larger in mass than the proton. It is similar to the photon in that it has no charge and produces ionization indirectly; it is different from the photon because it is a nuclear particle and not a unit of electromagnetic energy. (for photon, E=hʋ) Because the neutron is an uncharged particle, its interactions with matter are different from those of charged particles or photons. Ionization by neutrons is indirect: as a result of neutron interactions with matter, recoil (1) nuclei, (2) photons or (3) charged particles are produced and then interact with matter by various mechanisms that cause ionization.

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Recoil - Measurement of Hydrogen Depth Profile Using Fast NeutronsMaterials Analysis with Deuterium and Tritium Fusion Neutrons

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http://jolisfukyu.tokai-sc.jaea.go.jp/fukyu/mirai-en/2006/3_13.html

2.0 Neutron Sources Neutrons are classified according to their energies as shown in Table 4.

Some radionuclides (such as californium-252) may decay by spontaneous fission and emit neutrons with fission fragments, photons and electrons.

Induced fission reactions, such as those occurring in a nuclear reactor with uranium, emit about 2.5 neutrons per fission.

Fission neutrons range in energy from 0.025 eV to about 16 MeV. Other neutron sources are the result of various nuclear reactions and produce either a spectrum of neutron energies or monoenergetic neutrons. Common neutron producing nuclear reactions are the (γ, n), (α, n), (p, n), (d, n) and (α, 2n) reactions and may use radionuclide emissions or accelerated particles to initiate the reaction.

Neutron radiography usually uses radionuclides that emit alpha or gamma photons and produce neutrons by (α, n) and (γ, n) reactions with various target materials.

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TABLE 4. Neutron classification.

Class Thermal Epithermal Slow Intermediate Fast Relativistic

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Energy < 0.3 meV >1 eV 30 meV to 100 eV 100 eV to 10 keV 10 keV to 10 MeV greater than 10 MeV

3.0 Neutron Detectors There are several mechanisms and devices used to detect neutrons of various energies. Ionization chambers, proportional counters, scintillators, activation foils, track etch detectors, film emulsions, nuclear emulsions and thermoluminescent phosphors are some of the many devices used to detect neutrons. The main mechanisms used to detect neutrons in these devices are the (n, Îą), (n, p), (n, d), (n, f ) and (n, Îł) nuclear reactions.

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3.1 Proportional Neutron Detectors Many fast and slow neutron counters use proportional counting chambers filled with boron trifluoride (BF3) gas, often enriched in boron-10.

The interaction of thermal (slow) neutrons with boron gas releases an alpha particle of several megaelectronvolts that is easily detected in the proportional mode. 105B(n,Îą)73Li Fast neutrons are detected by a similar counter, in which thermal neutrons are absorbed in an external cadmium shield (113Cd(n,Îł)114Cd ; the fast neutrons that pass through the shield are thermalized in hydrogen rich material and counted in the proportional chambers.

Îł

hydrogen rich material

external Pb Shield?

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boron trifluoride (BF3) gas

â&#x2013; http://minerals.usgs.gov/minerals/pubs/commodity/ Charlie Chong/ Fion Zhang

The Cross Section (barns) of commonly used conversion screen Careful on differential cross section for isotopes of same element.

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TABLE 6. Properties of Some Thermal Neutron Radiography Conversion Materials Material

Useful Reactions

Lithium

6Li(n,α) 3H

910

prompt

Boron

10B(n,α) 7Li

3,830

prompt

Rhodium

103Rh(n)104mRh

11

45 min

103Rh(n)104Rh

139

42 s

107Ag(n)108Ag

35

2.3 min

109Ag

91

24 s

Silver

(n)110 Ag

Cadmium

113Cd((n,γ)114Cd

Indium

115

In(n)116n

115 In(n)116mln

Samarium

149Sm(n,γ) 150Sm I52 Sm(n)153Sm

Cross Section for Thermal Neutrons (barns)

Life

20,000

prompt

157

54 min

42

14 s

41,000

prompt

210

47 h

Europium

151

Eu(n)152Eu

3,000

9.2 h

Gadolinium

155 Gd(n,γ) I56Gd

61,000

prompt

157 Gd(n,γ)158Gd

254,000

prompt

Dyprosium

164 Dy(n)165mDy 164 Dy(n)165Dy

Gold

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197

Au(n)198Au

2,200

1.25 min

800

140 min

99

2.7 days

Neutron Cross Section of the elements

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http://periodictable.com/Properties/A/NeutronCrossSection.html

Neutron Cross Section Ď&#x192;total for Gd =50000 barn

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http://periodictable.com/Properties/A/NeutronCrossSection.html

Neutron Cross Section Ď&#x192;total for Dy =1010 barn

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http://periodictable.com/Properties/A/NeutronCrossSection.html

TABLE X1.1 Thermal Neutron Linear Attenuation Coefficients Using Average Scattering and Thermal Absorption Cross Sections for the Naturally Occurring Elements

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E748-02 Standard Practices for Thermal Neutron Radiography of Materials

Material

Useful Reactions

Lithium

6Li(n,α) 3H

910

prompt

Boron

10B(n,α) 7Li

3,830

prompt

Rhodium

103Rh(n)104mRh

11

45 min

103Rh(n)104Rh

139

42 s

107Ag(n)108Ag

35

2.3 min

109Ag

91

24 s

Silver

(n)110 Ag

Cadmium

113Cd((n,γ)114Cd

Indium

115

In(n)116n

115 In(n)116mln

Samarium

149Sm(n,γ) 150Sm I52 Sm(n)153Sm

Cross Section for Thermal Neutrons (barns)

Life

20,000

prompt

157

54 min

42

14 s

41,000

prompt

210

47 h

Europium

151

Eu(n)152Eu

3,000

9.2 h

Gadolinium

155 Gd(n,γ) I56Gd

61,000

prompt

157 Gd(n.γ)158Gd

254,000

prompt

49,000

prompt

2,200

1.25 min

800

140 min

99

2.7 days

Avg

Dyprosium

64Gd(n.γ)

?Gd

164 Dy(n)165mDy 164 Dy(n)165Dy

Gold

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197

Au(n)198Au

FIG. X1.1 Approximate Mass Attenuation Coefficients as a Function of Atomic Number

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TABLE 6. Properties of Some Thermal Neutron Radiography Conversion Materials Material

Useful Reactions

Lithium

6Li(n,α) 3H

910

prompt

Boron

10B(n,α) 7Li

3,830

prompt

Rhodium

103Rh(n)104mRh

11

45 min

103Rh(n)104Rh

139

42 s

107Ag(n)108Ag

35

2.3 min

109Ag

91

24 s

Silver

(n)110 Ag

Cadmium

113Cd((n,γ)114Cd

Indium

115

In(n)116n

115 In(n)116mln

Samarium

149Sm(n,γ) 150Sm I52 Sm(n)153Sm

Cross Section for Thermal Neutrons (barns)

Life

20,000

prompt

157

54 min

42

14 s

41,000

prompt

210

47 h

Europium

151

Eu(n)152Eu

3,000

9.2 h

Gadolinium

155 Gd(n,γ) I56Gd

61,000

prompt

157 Gd(n.γ)158Gd

254,000

prompt

164 Dy(n)165mDy

2,200

1.25 min

800

140 min

99

2.7 days

Dyprosium

164 Dy(n)165Dy

Gold Charlie Chong/ Fion Zhang

197

Au(n)198Au

TABLE 6. Capture cross sections σ of strongly absorbing elements for neutrons in approximate thermal equilibrium at 300 K (27 °C = 80 °F).

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TABLE 4. Average Characteristics of Thermal-Sources Type of Source

Typical Radiographic Intensity*

Resolution**

Exposure

Characteristics

Time

Radioisotope

101 to 104

Poor to Medium

Long

Stable operation. medium investment cost. possibly portable.

Accelerator

103 to 106

Medium

Average

On-off operation. medium cost. possibly mobile.

Subcritical Assembly

104 to 106

Good

Average

Stable operation, medium to high investment cost, mobility difficult

Nuclear reactor

105 to 108

Excellent

Short

Stable operation, medium to high investment cost. mobility difficult

*Neutrons per square centimeter per second. n/cm2â&#x2C6;&#x2122;s **These classifications are relative Charlie Chong/ Fion Zhang

More Reading on Gadolinum Gadolinium is a silvery-white malleable and ductile rare-earth metal. It crystallizes in hexagonal, close-packed α-form at room temperature, but, when heated to temperatures above 1235 °C, it transforms into its β-form, which has a body-centered cubic structure.

Gadolinium-157 has the highest thermal neutron capture cross-section among any stable nuclides: 259,000 barns. Only xenon-135 has a higher cross section, 2 million barns, but that isotope is unstable.

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Gadolinium is generally believed to be ferromagnetic at temperatures below 20 °C (68 °F) and is strongly paramagnetic above this temperature. There is some evidence that gadolinium may be a helical antiferromagnet, rather than a ferromagnet, below 20 °C (68 °F). Gadolinium demonstrates a magnetocaloric effect whereby its temperature increases when it enters a magnetic field and decreases when it leaves the magnetic field. The temperature is lowered to 5 °C (41 °F) for the gadolinium alloy Gd85Er15, and the effect is considerably stronger for the alloy Gd5(Si2Ge2), but at a much lower temperature (<85 K (−188.2 °C; −306.7 °F)).[6] A significant magnetocaloric effect is observed at higher temperatures, up to 300 K, in the Gd5(SixGe1-x)4 compounds.

Individual gadolinium atoms have been isolated by encapsulating them into fullerene molecules and visualized with transmission electron microscope.

Individual Gd atoms and small Gd clusters have also been incorporated into carbon nanotubes.

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Gadolinum

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Gadolinum

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isotope

NA

148Gd 150Gd 152Gd 154Gd 155Gd 156Gd 157Gd 158Gd 160 Gd 64

syn 75y syn 1.8×106 y 0.20% 1.08×1014y 2.18% – 14.80% – 20.47% – 15.65% – 24.84% – 21.86% >1.3×1021y

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half-life

DM

DE (MeV)

Decay Product

α α α (α) (α) (SF) (SF) (SF) (β−β−)

3.271 2.808 2.205 0.0812 0.0812 <71.541 <70.531 <70.965 1.729 1

144Sm 146Sm 148Sm 150Sm 151Sm

Prompt Prompt Prompt 160 Dy 66

https://en.wikipedia.org/wiki/Gadolinium

σ,Cross Section of Gadolinium = 49000 barn prompt GdAverage 155 Gd(n,γ) 156Gd = 61,000 barn prompt 157 Gd(n.γ) 158Gd = 254,000 barn prompt

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Neutron For many years after the proton and electron became comfortable concepts for building models of the atoms of the elements but explanations eluded researchers for the existence of isotopes and the extremely penetrating radiation emitted by the bombardment of light elements with alpha particles.

In 1932, Chadwick described a neutral particle with a mass equal to a proton that he called a neutron. The neutron explained many observations concerning radiation and particle physics and the concept was rapidly accepted. Neutron characteristics are given in Table 3.

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TABLE 3. Neutron characteristics.

Quantity

Measurement

Charge Rest mass Classical radius Magnetic moment Compton wavelength

neutral 1.675 × 10–27 kg 1.532 × 10–18 m –9.662 × 10–27 J·T–1 1.320 × 10–15 m

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Hydrogen, Deuterium, Tritium Radioactive materials have existed since the earth was created. All elements with atomic numbers greater than 83, bismuth, exist only as radioactive elements and many elements below atomic number 83 have radioactive isotopes that exist in nature.

The difference between a stable or nonradioactive atom of an element and an unstable or radioactive atom is in the energy content of the nucleus. Most often an excess or deficiency in the number of neutrons in the nucleus provides the excess energy or instability.

As an example: most hydrogen in nature exists as atoms with only 1 proton and 1 electron. About 15 of every 100 000 atoms of hydrogen have a neutron plus the proton in the nucleus, giving the atom a mass of 2 or twice the mass of most hydrogen atoms. Mass 2 hydrogen is called deuterium or heavy hydrogen and is stable. When a second neutron is added to the nucleus of hydrogen, the atom has a mass of 3, is called tritium and is radioactive.

The tritium atom is produced in nature by cosmic bombardment to produce a pre- 1952 concentration in nature of between 1 ~ 10 tritium atoms per 1018 hydrogen atoms.

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Neutrons Neutrons produced by fission, accelerator nuclear reactions or radioisotope sources have considerable kinetic energy. This kinetic energy is most often lost by scattering interactions with or absorption in the nuclei of the atoms in their path. Absorption of the neutron is followed by release of electromagnetic radiation or large particles such as protons, multiple neutrons, deuterons or alpha particles. Interactions with the orbital electrons contribute negligibly to the absorption of neutrons by matter. The nucleus is much smaller than the electron orbits, so neutron interactions are less frequent than those of alpha or beta particles. And because the neutron has no charge, ionization and excitation are not major absorption processes.

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FIGURE 4. Ionization by alpha particle that ejects orbital electron from atom.

Specific ionization is number of ion pairs generated by particle per unit path.

Total ionization designates number of ion pairs produced by particle along its entire path.

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Neutron- Elastic Scattering For elastic scattering, the neutron collides with the nucleus and bounces off, leaving the nucleus unchanged. This type of collision can be treated straightforwardly as a mechanical billiard ball type of collision. In the collision the energy of the neutron is shared by the nucleus, thus each collision reduces the energy of the neutron. After a number of collisions with the nuclei, the energy is reduced to the same average kinetic energy as that of the absorbing medium. This energy is often referred to as the thermal energy because it depends primarily on the temperature. Neutrons at thermal equilibrium with their surroundings are thermal neutrons. At 20°C (68°F), a thermal neutron would have a kinetic energy of about 0.025 eV and a velocity of 2200 m·s–1 (4900 mi·h–1).

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The transfer of energy from the neutron to the nucleus is greater for light nuclei. Therefore, low atomic nuclei containing materials such as water, hydrocarbons, graphite and beryllium are used to reduce neutron energies.

Such materials are called moderators. Hydrogen nuclei have essentially the same mass as neutrons and can undergo nearly complete kinetic energy transfer in a single collision. Energy transfer to larger nuclei require many collisions.

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Neutron- Inelastic Scattering Here the neutron collides with the nucleus leaving the nucleus in an excited state. In this process, the nucleus may either stay in the excited state (n,n’) as a metastable isomer or will immediately emit gamma radiation (n,γn) and return to the ground or original state.

Keywords: Excited state - (n,n’) Emit gamma and return to stable state - (n,n’)

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Nuclear Neutron Absorption As the neutron has no charge, it can approach the nucleus until the close range attractive forces of the nucleus begin to operate. In this process, the neutron is captured, forming a compound nucleus. Because there is no charge barrier, even the slowest neutron can be readily captured.

As the binding energy of a neutron into a compound nucleus is nearly 8MeV, even the capture of thermal neutrons can result in a highly excited state for the nucleus. (the thermal neutron has 0.025~0.1 MeV of energy, how this relate to the 8MeV?) This excited nucleus can attain relative stability by: ■ ejecting a proton, ■ ejecting an alpha particle, or ■ emitting the excess energy as gamma radiation.

When a particle is ejected, the nucleus becomes a new element; then the process is also known as nuclear transmutation. The discovery of transmutation by slow neutrons led to the realization of nuclear fission.

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As the binding energy of a neutron into a compound nucleus is nearly 8MeV, even the capture of thermal neutrons can result in a highly excited state for the nucleus. (the thermal neutron has 0.025~0.1 MeV of energy, how this relate to the 8MeV?)

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The simplest capture reaction is that of capture of slow neutrons with emission of gamma rays (n,γ). Thermal neutron reaction with cobalt is an example: 59Co

+ n → 60Co + γ → 60Ni + β- + γ

In heavy nuclei, the capture of a slow neutron, followed by the emission of gamma radiation, increases the neutron-to-proton ratio — usually making the nucleus radioactive with decay by electron emission likely. More information on production of radioactive material by neutron capture may be found in the discussion of radioactive materials.

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As the energy of the impinging neutron is made larger, a charged particle can be ejected. However, a charged particle, because of the short range attractive forces of the nuclei, is hindered from leaving the nucleus and processes such as (n,p), (n,Îą) and (n,d) can only take place when the incident neutron supplies sufficient energy to overcome the binding energies of the particles in the nucleus. For heavy nuclei these forces are appreciable and the requisite neutron energy becomes greater. Thus, for example, a particle ejection is possible only if the neutron has sufficient energy to overcome the binding energy of the alpha particle; that is, the neutron must be a fast neutron.

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In the (n,Îą) reaction, the product nucleus contains one neutron and two protons less than the original nucleus. The neutron-to-proton ratio is increased and the transmutation usually produces a radioactive nucleus that decays by the emission of an electron (beta disintegration).

As the energy of the incident neutron approaches 30 MeV, the compound nucleus can eject three neutrons (n, 3n) or two neutrons and a proton (n, 2np) as well as other combinations of particles. At even higher energies, more particles may be ejected until the nucleus essentially disappears (spallation).

Finally, nuclear fission (n,f), where the nucleus breaks up with the release of several larger particles and several neutrons, can be induced in certain large nuclides, such as uranium-235, by neutrons of almost any energy, whereas in other nuclides, fast or energetic neutrons are required.

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Nuclear Cross Sections Because of many reactions possible for absorbing neutrons and their complicated energy and mass dependencies, there is no simple way to present the total absorption effect. However, the probability of any interaction between neutrons and matter can be made qualitative by means of the concept of cross sections. The cross section σ is the effective target area of the nucleus as seen by the impinging neutron of a given energy. The number of interactions per unit time will be nvNσ, where n is the number of neutrons per unit volume moving with velocity v towards the target of N nuclei. The quantity nv is the neutron flux density (neutrons per square centimeter second). The cross section σ is usually expressed in square meters (m2) or barns (b), where 1 b = 10–24 cm2 = 10–28 m2.

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In discussing the variation of nuclear cross section with energy of the incident neutrons, certain generalizations of a broad character can be made. In general, there are three regions that can be distinguished.

■ First is the low energy region, which includes the thermal range, where the cross section decreases steadily with increasing neutron energy. The total cross section is the sum of two terms, one due to neutron scattering is quite small and almost constant, the other representing absorption by the nucleus is inversely proportional to the velocity (energy) . This low energy range is termed the v–1 region, where the time spent by the neutron near the nucleus is proportional to v–1.

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Second, following the somewhat indefinite v–1 region, many elements exhibit peaks called resonance peaks, where the neutron cross sections rise sharply to high values for certain energies, then fall to lower values again. Depending on the element, the number of such peaks may number three or more. These peaks may be found mostly in the energy range 0.1 to 1 eV, although in a few elements like uranium-238, they may be found up to energies of 10 eV. These reactions are of the (η,γ) (Eta, gamma) type.

And third, with neutrons of high energy in the MeV range, the cross sections are very low, less than 10–27 m2 (10 b), compared to the very high cross sections of 4 × 10–25 m2 (several thousand barns, ~4000 b) at low energies.

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A simple example of the total absorption cross section is that of cadmium, shown in Fig. 5. The v–1 region is shown up to about 0.03 eV, the resonance at 0.176 eV and the low cross section region for energies greater than about 2 MeV. The dramatic increase in cross sections at the resonance have been worked out by the theory of G. Breit and E.P. Wigner. In its simplicity, if the energy of the neutron is such that a compound nucleus can be formed at or near one of its energy levels, then the probability of capture of these neutrons will be exceptionally high.

All elements do not show the resonant absorption effect; for example, boron has no measurable resonance and the cross section follows the v–1 law from 0.01 eV to over 1000 eV. However, its cross section for (n,α) is so large for neutrons of low energy that this reaction is often used for neutron detectors.

Table 6 shows the dramatic variation of cross section for absorbing thermal neutrons of some of the better neutron absorbers.

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FIGURE 5. Absorption of neutrons by cadmium, showing resonance peak at 0.176 eV.

0.176 eV

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TABLE 6. Capture cross sections σ of strongly absorbing elements for neutrons in approximate thermal equilibrium at 300 K (27 °C = 80 °F).

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Neutron Activation In the section on neutron interactions with materials, neutron capture was briefly discussed. This technique, coupled with the large fields of neutrons available in nuclear reactors, produces most of the radioisotopes used in radiography. cobalt-60 and iridium-192 come from thermal neutron bombardment of the stable isotopes (cobalt-59 and iridium-191) of these two elements. Production of the radioactivity can be predicted by Eq. 21:

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in which A is the activity produced in disintegrations per second, N is number of target atoms being bombarded, f is the neutron flux (in neutrons per centimeter second), Ď&#x192; is the cross section for neutron capture (in square centimeter), ti is the irradiation time in the same units as the half life and T is the half life of the radioisotope produced.

The exponential portion of the equation corrects the production of the radioactive material for the amount that decays away while more is being made. This leads to the point of diminishing returns for production in that after about five half lives, almost as much of the radioactive material is decaying as is being produced per each increment of neutron bombardment time.

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Also, the equation is correct only for thin samples of the bombarded material.

Absorption of neutrons in the outer layers of the sample (usually a metal pellet) reduces the number of neutrons incident on the interior atoms. This self-shielding of neutrons coupled with a self-absorption of gamma rays released by radioactive atoms inside of the sample gives a gamma output considerably lower than calculated.

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Fission Fragments When uranium-235 or other fissionable atom undergoes fission, multiple neutrons and two major fragments of the nucleus are released. The two fragments are called fission fragments and are a source of radioactive materials for industrial, medical and research use. The fragments are usually of unequal size and are grouped in two distributions around mass numbers 96 and 138. One of the major products is cesium-137, which can be chemically separated from the other fission fragments for use as a gamma ray source in radiography, medical therapy and large irradiation facilities for preservation of food and for sterilization of medical supplies.

Fission Fragment: cesium-137

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Accelerator Production Large particle accelerators such as linatrons, van de graaff generators and cyclotrons can provide appreciable neutron fluxes or streams of high energy particles including protons, deuterons and helium nuclei. When appropriate target materials are bombarded by these particles, radioactive nuclei can be produced. Although radioactive materials for medical use are being produced in this fashion, generally radiographic sources are not commercially produced in this fashion.

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The Measurement Units

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Rolf Maximilian Sievert

Sievert - (Sv) is an unit for health effect of ionizing radiation, numerically it is the energy absorbed by human body (by matters?) of 1J·Kg-1 The sievert (symbol: Sv is a derived unit of ionizing radiation dose in the International System of Units (SI). It is a measure of the health effect of low levels of ionizing radiation on the human body.

Quantities that are measured in sieverts are intended to represent the stochastic 随机的 health risk, which for radiation dose assessment is defined as the probability of cancer induction and genetic damage.

To enable consideration of stochastic health risk, calculations are performed to convert the physical quantity absorbed dose into equivalent and effective doses, the details of which depend on the radiation type and biological context. For applications in radiation protection and dosimetry assessment the International Commission on Radiological Protection (ICRP) and International Commission on Radiation Units and Measurements (ICRU) have published recommendations and data which are used to calculate these.

These are under continual review, and changes are advised in the formal "Reports" of those bodies.

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https://en.wikipedia.org/wiki/Sievert

The sievert is used for radiation dose quantities such as equivalent dose, effective dose, and committed dose.

It is used to represent both the risk of the effect of external radiation from sources outside the body and the effect of internal irradiation due to inhaled or ingested radioactive substances.

Conventionally, the sievert is not used for high dose rates of radiation that produce deterministic effects, which is the severity of acute tissue damage that is certain to happen. Such effects are compared to the physical quantity absorbed dose measured by the unit gray (Gy).

The sievert is of fundamental importance in dosimetry and radiation protection, and is named after Rolf Maximilian Sievert, a Swedish medical physicist renowned for work on radiation dosage measurement and research into the biological effects of radiation. One sievert carries with it a 5.5% chance of eventually developing cancer.

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https://en.wikipedia.org/wiki/Sievert

One Sievert equals 100 rem. The rem is an older, non-SI unit of measurement.

To enable a comprehensive view of the Sievert this article deals with the definition of the Sievert as an SI unit, summarises the recommendations of the ICRU and ICRP on how the Sievert is calculated, includes a guide to the effects of ionizing radiation as measured in Sievert, and gives examples of approximate figures of dose uptake in certain situations.

The gray - quantity "D" 1 Gy = 1 joule/kilogram - a physical quantity. 1 Gy is the deposit of a joule of radiation energy in a kg of matter or tissue.

The sievert - quantity "H" 1 Sv = 1 joule/kilogram - a biological effect. The Sievert represents the equivalent biological effect of the deposit of a joule of radiation energy in a kilogram of human tissue. The equivalence to absorbed dose is denoted by Q. H=QĂ&#x2014;D

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https://en.wikipedia.org/wiki/Sievert

REM - Roentgen equivalent man The roentgen equivalent in man (abbreviated rem; symbol rem, or often but incorrectly R) is an older, CGS unit of equivalent dose, effective dose, and committed dose. Quantities measured in rem are designed to represent the stochastic biological effects of ionizing radiation, primarily radiation-induced cancer. These quantities are a complex weighted average of absorbed dose, which is a clear physical quantity measured in rads. There is no universally applicable conversion constant from rad to rem; the conversion depends on relative biological effectiveness (RBE).

The rem is defined since 1976 as equal to 0.01 sievert, which is the more commonly used SI unit outside of the United States. A number of earlier definitions going back to 1945 were derived from the roentgen unit, which was named after Wilhelm Rรถntgen, a German scientist who discovered Xrays. The acronym is now a misleading historical artifact, since 1 roentgen actually deposits about 0.96 rem in soft biological tissue, when all weighting factors equal unity. Older units of rem following other definitions are up to 17% smaller than the modern rem.

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https://en.wikipedia.org/wiki/Roentgen_equivalent_man

One rem carries with it a 0.055% chance of eventually developing cancer.

Doses greater than 100 rem received over a short time period are likely to cause acute radiation syndrome (ARS), possibly leading to death within weeks if left untreated. Note that the quantities that are measured in rem were not designed to be correlated to ARS symptoms. The absorbed dose, measured in rad, is the best indicator of ARS. A rem is a large dose of radiation, so the millirem (mrem), which is one thousandth of a rem, is often used for the dosages commonly encountered, such as the amount of radiation received from medical x-rays and background sources.

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Rem Usage The rem and millirem are CGS units in widest use among the American public, industry, and government. SI units are the norm outside of the United States, and they are increasingly encountered within the US in academic, scientific, and engineering environments.

The conventional units for dose rate is mrem/h. Regulatory limits and chronic doses are often given in units of mrem/yr or rem/yr, where they are understood to represent the total amount of radiation allowed (or received) over the entire year. In many occupational scenarios, the hourly dose rate might fluctuate to levels thousands of times higher for a brief period of time, without infringing on the annual total exposure limits.

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There is no exact conversion from hours to years because of leap years, but approximate conversions are: 1 mrem/h = 8766 mrem/yr 0.1141 mrem/h = 1000 mrem/yr The ICRP once adopted fixed conversion for occupational exposure, although these have not appeared in recent documents: 8 h = 1 day 40 h = 1 week 50 week = 1 yr Therefore, for occupation exposures of that time period, 1 mrem/h = 2000 mrem/yr 0.5 mrem/h = 1000 mrem/yr

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The US National Institute of Standards and Technology (NIST) strongly discourages Americans from expressing doses in rem, in favor of recommending the SI unit. The NIST recommends defining the rem in relation to the SI in every document where this unit is used. For US industries and US firms that do not require the sole use of SI, however, the unit rem is often preferred.

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Health Effects Ionizing radiation has deterministic and stochastic effects on human health.

The deterministic effects that can lead to acute radiation syndrome only occur in the case of high doses (> ~10 rad or > 0.1 Gy) and high dose rates (> ~10 rad/h or > 0.1 Gy/h). A model of deterministic risk would require different weighting factors (not yet established) than are used in the calculation of equivalent and effective dose.

To avoid confusion, deterministic effects (either chronic & acute?) are normally compared to absorbed dose in units of rad, not rem.

Stochastic effects are those that occur randomly, such as radiation-induced cancer. The consensus of the nuclear industry, nuclear regulators, and governments, is that the incidence of cancers due to ionizing radiation (not including excessive high dose rate?) can be modeled as increasing linearly with effective dose at a rate of 0.055% per rem (5.5%/Sv). Individual studies, alternate models, and earlier versions of the industry consensus have produced other risk estimates scattered around this consensus model.

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There is general agreement that the risk is much higher for infants and fetuses than adults, higher for the middle-aged than for seniors, and higher for women than for men, though there is no quantitative consensus about this.

There is much less data, and much more controversy, regarding the possibility of cardiac and teratogenic引起畸型的 effects, and the modelling of internal dose The International Commission on Radiological Protection (ICRP) recommends limiting artificial irradiation of the public to an average of 100 mrem (1 mSv) (0.1Rem for public and 5Rem for radiation worker?) of effective dose per year, not including medical and occupational exposures.

For comparison, radiation levels inside the US United States Capitol are 85 mrem/yr (0.85 mSv/yr), close to the regulatory limit, because of the uranium content of the granite structure. According to the ICRP model, someone who spent 20 years inside the capitol building would have an extra one in a thousand chance of getting cancer, over and above any other existing risk.

(20 yr × 85 mrem/yr × 0.001 rem/mrem × 0.055%/rem = ~0.1%) That "existing risk" is much higher; an average person would have a one in ten chance of getting cancer during this same 20-year period, even without any exposure to artificial radiation.

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Radiation-related Quantities

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Radiation-related Quantities 1 gray = 100 rad J·Kg-1 = 100,000 erg·Kg-1 Joule = 1 x 10-5 erg 1 Rongent ≠ 1 Rad

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The Rรถntgen equivalent physical or rep (symbol rep) is a unit of absorbed dose first introduced by Herbert Parker in 1945 to replace an improper application of the roentgen unit to biological tissue. It is the absorbed energetic dose before the biological efficiency of the radiation is factored in. The rep has variously been defined as 83 or 93 ergs per gram of tissue (8.3/9.3 mGy)[2] or per cm3 of tissue. At the time, this was thought to be the amount of energy deposited by 1 roentgen.

Improved measurements have since found that one roentgen of air kerma deposits 8.77 mGy in dry air, or 9.6 mGy in soft tissue, but the rep was defined as a fixed number of ergs per unit gram. A 1952 handbook from the US National Bureau of Standards affirms that "The numerical coefficient of the rep has been deliberately changed to 93, instead of the earlier 83, to agree with L. H. Gray's 'energy-unit'." It is unclear what was meant by Gray's 'energy unit', since the gray was not defined until the 1970s; perhaps the gram-roentgen he introduced in 1940? The rep was commonly used until the 1960s, but was gradually displaced by the rad starting in 1954 and later the gray starting in 1977.

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Air Kerma means kerma in a given mass of air. The unit used to measure the quantity of air kerma is the Gray (Gy). For X-rays with energies less than 300 kiloelectronvolts (keV), 1 Gy = 100 rad. In air, 1 Gy of absorbed dose is delivered by 114 roentgens (R) of exposure.

100 rad = 114 R

Kerma - kinetic energy released in the medium (Kinetic Energy Released per Unit Mass) 1 abbreviation for kinetic energy released in the medium, a quantity that describes the transfer of energy from a photon to a medium as the ratio of energy transferred per unit mass at each point of interaction.

2 abbreviation for kinetic energy released in matter, a unit of quantity referring to the kinetic energy transferred from photons to charged particles, such as electrons in Compton interactions, per unit mass. The SI unit for the KERMA is the gray, and the special unit is the rad.

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http://medical-dictionary.thefreedictionary.com/KERMA

Kerma Kinetic Energy Released per Unit Mass.

Kerma is a dose variable. Kerma K is the quotient of dEtr and dm; whereby dEtr is the sum of the starting values of kinetic energies of all charged particles released by indirectly ionizing radiation from the material in a volume element dV, and dm is the mass of the material in this volume element. All indications for a Kerma must mention the reference material (i.e. the material dm). The SI unit of the kerma is gray (Gy).

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http://www.euronuclear.org/info/encyclopedia/k/kerma.htm

Teratogenic 引起畸型的

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Deterministic Effects

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Chronic Effect

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Deterministic Effects

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ionization chambers

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ionization chambers

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ionization chambers

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ionization chambers

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ionization chambers

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More Reading on ionization chambers Basically, an ionization chamber consists of two electrodes kept at a potential difference and a gas that fills the space between the electrodes. The detection process occurs when an X-ray photon interacts with the gas inside the chamber, forming â&#x20AC;&#x153;Nâ&#x20AC;? number of electron-ion pairs. The electrons and ions are separated due to the direction and sense of the electric field.

Continuous streams of photons originate a continuous production of electronhole pairs and consequently an electric current between two electrodes. This current, typically in the order of a few pico-ampere (pA) (10-12) , is proportional to the photon flux of X-rays.

The electrode measuring the current generated by the camera is called the collector electrode. This electrode (anode?) is typically maintained at a potential close to the ground. The other electrode (cathode?) is called the high-voltage electrode, and must be maintained at a positive (?) voltage (to collect positive charges) or negative (to collect negative charges). These electrodes are fixed inside the chamber through electrical insulators.

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http://lnls.cnpem.br/beamlines/xafs/equipments/ion-chambers/

The chamber can be sealed to use different gases at different pressures. The figure below represents a parallel plate chamber where radiation passes between the electrodes. There are two windows, one where the beam reaches the sensitive volume and the other where the beam exits the chamber. These windows should be composed of a material of low atomic number and should be thin, so not to reduce the intensity of radiation. Note that this type of chamber is not stopping the X-ray beam.

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http://lnls.cnpem.br/beamlines/xafs/equipments/ion-chambers/

Schematic diagram of an ionization chamber with parallel plates.

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http://lnls.cnpem.br/beamlines/xafs/equipments/ion-chambers/

The Ionization Chamber is the simplest of all gas-filled radiation detectors, and is widely used for the detection and measurement of certain types of ionizing radiation; X-rays, gamma rays and beta particles.

Conventionally, the term "ionization chamber" is used exclusively to describe those detectors which collect all the charges created by direct ionization within the gas through the application of an electric field.

It only uses the discrete charges created by each interaction between the incident radiation and the gas, and does not involve the gas multiplication mechanisms used by other radiation instruments, such as the Geiger-MĂźller counter or the proportional counter.

Ion chambers have a good uniform response to radiation over a wide range of energies and are the preferred means of measuring high levels of gamma radiation. They are widely used in the nuclear power industry, research labs, radiography, radiobiology, and environmental monitoring.

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https://en.wikipedia.org/wiki/Ionization_chamber

The Ionization Chamber  Detectors which collect all the charges created by direct ionization within the gas through the application of an electric field.  It only uses the discrete charges created by each interaction between the incident radiation and the gas,  It does not involve the gas multiplication mechanisms used by other radiation instruments, such as the Geiger-Müller counter or the proportional counter.  Ion chambers have a good uniform response to radiation over a wide range of energies

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The Ionization Chamber ď Ž It does not involve the gas multiplication mechanisms used by other radiation instruments, such as the Geiger-MĂźller counter or the proportional counter.

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Schematic diagram of parallel plate ion chamber, showing drift of ions.

Electrons typically drift 1000 times faster than positive ions due to their much smaller mass.

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Principle of operation An ionization chamber measures the charge from the number of ion pairs created within a gas caused by incident radiation. It consists of a gas-filled chamber with two electrodes; known as anode and cathode. The electrodes may be in the form of parallel plates (Parallel Plate Ionization Chambers: PPIC), or a cylinder arrangement with a coaxially located internal anode wire.

coaxially located

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Parallel Plate

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A voltage potential is applied between the electrodes to create an electric field in the fill gas. When gas between the electrodes is ionized by incident ionizing radiation, ion-pairs are created and the resultant positive ions and dissociated electrons move to the electrodes of the opposite polarity under the influence of the electric field. This generates an ionization current which is measured by an electrometer circuit. The electrometer must be capable of measuring the very small output current which is in the region of femto amperes (10-15) to pico amperes (10-12) , depending on the chamber design, radiation dose and applied voltage.

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Each ion pair created deposits or removes a small electric charge to or from an electrode, such that the accumulated charge is proportional to the number of ion pairs created, and hence the radiation dose. This continual generation of charge produces an ionization current, which is a measure of the total ionizing dose entering the chamber. However, the chamber cannot discriminate between radiation types (beta or gamma) and cannot produce an energy spectrum of radiation.

The electric field also enables the device to work continuously by mopping up electrons, which prevents the fill gas from becoming saturated, where no more ions could be collected, and by preventing the recombination of ion pairs, which would diminish the ion current.

Current Mode This mode of operation is referred to as "current" mode, meaning that the output signal is a continuous current, and not a pulse output as in the cases of the Geiger-MĂźller tube or the proportional counter.

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Referring to the accompanying ion pair collection graph, it can be seen that in the "ion chamber" operating region the collection of ion pairs is effectively constant over a range of applied voltage, as due to its relatively low electric field strength the ion chamber does not have any "multiplication effect". This is in distinction to the Geiger-MĂźller tube or the proportional counter whereby secondary electrons, and ultimately multiple avalanches, greatly amplify the original ion-current charge.

Plot of ion current against voltage for a wire cylinder gaseous radiation detector. The ion chamber uses the lowest usable detection region.

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Chamber types and construction The following chamber types are commonly used.

Free-air chamber This is a chamber freely open to atmosphere, where the fill gas is ambient air.

The domestic smoke detector is a good example of this, where a natural flow of air through the chamber is necessary so that smoke particles can be detected by the change in ion current. Other examples are applications where the ions are created outside the chamber but are carried in by a forced flow of air or gas.

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Chamber pressure â&#x2013; Vented chamber These chambers are normally cylindrical and operate at atmospheric pressure, but to prevent ingress of moisture a filter containing a desiccant is installed in the vent line. This is to stop moisture building up in the interior of the chamber, which would otherwise be introduced by the "pump" effect of changing atmospheric air pressure. These chambers have a cylindrical body made of aluminium or plastic a few millimetres thick. The material is selected to have an atomic number similar to that of air so that the wall is said to be "air equivalent" over a range of radiation beam energies. This has the effect of ensuring the gas in the chamber is acting as though it were a portion of an infinitely large gas volume, and increases the accuracy by reducing interactions of gamma with the wall material. The higher the atomic number of the wall material, the greater the chance of interaction. The wall thickness is a trade-off between maintaining the air effect with a thicker wall, and increasing sensitivity by using a thinner wall.

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These chambers often have an end window made of material thin enough, such as mylar, so that beta particles can enter the gas volume. Gamma radiation enters both through the end window and the side walls. For handheld instruments the wall thickness is made as uniform as possible to reduce photon directionality though any beta window response is obviously highly directional. Vented chambers are susceptible to small changes in efficiency with air pressure and correction factors can be applied for very accurate measurement applications.

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â&#x2013; Sealed low pressure chamber These are similar in construction to the vented chamber but are sealed and operate at or around atmospheric pressure. They contain a special fill gas to improve detection efficiency as free electrons are easily captured in air-filled vented chambers by neutral oxygen which is electronegative, to form negative ions (?) . These chambers also have the advantage of not requiring a vent and desiccant. The beta end window limits the differential pressure from atmospheric pressure that can be tolerated, and common materials are stainless steel or titanium with a typical thickness of 25 Âľm. â&#x2013; High pressure chamber The efficiency of the chamber can be further increased by the use of a high pressure gas. Typically a pressure of 8-10 atmospheres can be used, and various noble gases are employed. The higher pressure results in a greater gas density and thereby a greater chance of collision with the fill gas and ion pair creation by incident radiation. Because of the increased wall thickness required to withstand this high pressure, only gamma radiation can be detected. These detectors are used in survey meters and for environmental monitoring.

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Instrument types Ion chambers are widely used in hand held radiation survey meters to measure beta and gamma radiation. They are particularly preferred for high dose rate measurements and for gamma radiation they give good accuracy for energies above 50-100 keV. There are two basic configurations; the "integral" unit with the chamber and electronics in the same case, and the "two-piece" instrument which has a separate ion chamber probe attached to the electronics module by a flexible cable.

The chamber of the integral instrument is generally at the front of the case facing downwards, and for beta/gamma instruments there is a window in the bottom of the casing. This usually has a sliding shield which enables discrimination between gamma and beta radiation. The operator closes the shield to exclude beta, and can thereby calculate the rate of each radiation type.

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Some hand held instruments generate audible clicks similar to that produced by a G-M counter to assist operators, who use the audio feedback in radiation survey and contamination checks. As the ion chamber works in current mode, not pulse mode, this is synthesised from the radiation rate.

Installed For industrial process measurements and interlocks with sustained high radiation levels, the ion chamber is the preferred detector. In these applications only the chamber is situated in the measurement area, and the electronics are remotely situated to protect them from radiation and connected by a cable. Installed instruments can be used for measuring ambient gamma for personnel protection and normally sound an alarm above a preset rate, though the Geiger-MĂźller tube instrument is generally preferred where high levels of accuracy are not required.

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Hand-held integral ion chamber survey meter in use

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View of sliding beta shield on integral hand held instrument

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General precautions in use Moisture is the main problem that affects the accuracy of ion chambers. The chamber's internal volume must be kept completely dry, and the vented type uses a desiccant to help with this.[3] Because of the very low currents generated, any stray leakage current must be kept to a minimum in order to preserve accuracy. Invisible hygroscopic moisture on the surface of cable dielectrics and connectors can be sufficient to cause a leakage current which will swamp any radiation-induced ion current. This requires scrupulous cleaning of the chamber, its terminations and cables, and subsequent drying in an oven. "Guard rings" are generally used as a design feature on higher voltage tubes to reduce leakage through or along the surface of tube connection insulators which can require a resistance in the order of 1013 Ί.

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For industrial applications with remote electronics, the ion chamber is housed in a separate enclosure which provides mechanical protection and contains a desiccant to remove moisture which could affect the termination resistance.

In installations where the chamber is a long distance from the measuring electronics, readings can be affected by external electromagnetic radiation acting on the cable. To overcome this a local converter module is often used to translate the very low ion chamber currents to a pulse train or data signal related to the incident radiation. These are immune to electromagnetic effects.

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Applications Nuclear industry Ionization chambers are widely used in the nuclear industry as they provide an output that is proportional to radiation dose. They find wide use in situations where a constant high dose rate is being measured as they have a greater operating lifetime than standard Geiger-MĂźller tubes, which suffer from gas break down and are generally limited to a life of about 1011 count events. Additionally, the Geiger-MĂźller tube cannot operate above about 104 counts per second, due to dead time effects, whereas there is no similar limitation on the ion chamber.

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Keywords:  Ionization Chamber provides an output that is proportional to radiation dose.  Geiger-Müller tubes, which suffer from gas break down and are generally limited to a life of about 1011 count events.  Geiger-Müller tube cannot operate above about 104 counts per second (104 N·s-1) , due to dead time effects, whereas there is no similar limitation on the ion chamber.

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Smoke detectors The ionization chamber has found wide and beneficial use in smoke detectors.

In a smoke detector, ambient air is allowed to freely enter the ionization chamber. The chamber contains a small amount of americium-241, which is an emitter of alpha particles which produce a constant ion current. If smoke enters the detector, it disrupts this current because ions strike smoke particles and are neutralized. This drop in current triggers the alarm. The detector also has a reference chamber which is sealed but is ionized in the same way.

Comparison of the ion currents in the two chambers allows compensation for changes due to air pressure, temperature, or the ageing of the source.

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Medical radiation measurement In medical physics and radiotherapy, ionization chambers are used to ensure that the dose delivered from a therapy unit or radiopharmaceutical is what is intended. The devices used for radiotherapy are called "reference dosimeters", while those used for radiopharmaceuticals are called radioisotope dose calibrators. A chamber will have a calibration factor established by a national standards laboratory such as ARPANSA in Australia or the NPL in the UK, or will have a factor determined by comparison against a transfer standard chamber traceable to national standards at the user's site.

Guidance on application use In the United Kingdom the HSE has issued a user guide on selecting the correct radiation measurement instrument for the particular application concerned. This covers all radiation instrument technologies, and is a useful comparative guide to the use of ion chamber instruments.

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The Geiger counter is an instrument used for measuring ionizing radiation used widely in such applications as radiation dosimetry, radiological protection, experimental physics and the nuclear industry.

It detects ionizing radiation such as alpha particles, beta particles and gamma rays using the ionization effect produced in a Geiger–Müller tube; which gives its name to the instrument. In wide and prominent use as a hand-held radiation survey instrument, it is perhaps one of the world's best-known radiation detection instruments.

The original detection principle was discovered in 1908, but it was not until the development of the Geiger-Müller tube in 1928 that the Geiger-Müller counter became a practical instrument. Since then it has been very popular due to its robust sensing element and relatively low cost.

However, there are limitations in measuring high radiation rates and the energy of incident radiation.

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A "two-piece" bench type Geigerâ&#x20AC;&#x201C;MĂźller counter with end-window detector

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Schematic of a Geiger counter using an "end window" tube for low penetration radiation. A loudspeaker is also used for indication

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Principle of operation A Geiger counter consists of a Geiger-Müller tube, the sensing element which detects the radiation, and the processing electronics, which displays the result.

The Geiger-Müller tube is filled with an inert gas such as helium, neon, or argon at low pressure, to which a high voltage is applied.

The tube briefly conducts electrical charge when a particle or photon of incident radiation makes the gas conductive by ionization. The ionization is considerably amplified within the tube by the Townsend discharge effect to produce an easily measured detection pulse, which is fed to the processing and display electronics. This large pulse from the tube makes the G-M counter relatively cheap to manufacture, as the subsequent electronics is greatly simplified. The electronics also generates the high voltage, typically 400–600 volts, that has to be applied to the Geiger-Müller tube to enable its operation.

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Readout There are two types of radiation readout; counts or radiation dose.

The counts display is the simplest and is the number of ionizing events displayed either as a count rate, commonly "counts per second", or as a total over a set time period (an integrated total). The counts readout is normally used when alpha or beta particles are being detected. More complex to achieve is a display of radiation dose rate, displayed in a unit such as the sievert which is normally used for measuring gamma or X-ray dose rates. A G-M tube can detect the presence of radiation, but not its energy which influences the radiation's ionising effect. Consequently, instruments measuring dose rate require the use of an energy compensated G-M tube, so that the dose displayed relates to the counts detected. The electronics will apply known factors to make this conversion, which is specific to each instrument and is determined by design and calibration.

The readout can be analog or digital, and increasingly, modern instruments are offering serial communications with a host computer or network.

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There is usually an option to produce audible clicks representing the number of ionization events detected. This is the distinctive sound normally associated with hand held or portable Geiger counters. The purpose of this is to allow the user to concentrate on manipulation of the instrument whilst retaining auditory feedback on the radiation rate.

Limitations There are two main limitations of the Geiger counter. Because the output pulse from a Geiger-MĂźller tube is always the same magnitude regardless of the energy of the incident radiation, the tube cannot differentiate between radiation types. A further limitation is the inability to measure high radiation rates due to the "dead time" of the tube. This is an insensitive period after each ionization of the gas during which any further incident radiation will not result in a count, and the indicated rate is therefore lower than actual.

Typically the dead time will reduce indicated count rates above about 104 to 105 counts per second depending on the characteristic of the tube being used.

Whilst some counters have circuitry which can compensate for this, for accurate measurements ion chamber instruments are preferred for high radiation rates.

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For accurate measurements ion chamber instruments are preferred for high radiation rates.

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Ionization Chamber

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Geiger Muller Counter

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G-M counter with pancake type probe

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Laboratory use of a G-M counter with end window probe to measure beta radiation from a radioactive source

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Types and applications The application and use of a Geiger counter is dictated entirely by the design of the tube, of which there are a great many, but they can be generally categorised as "end-window", or windowless "thin-walled" or "thick-walled", and sometimes hybrids of these types.

Particle detection The first historical uses of the Geiger principle were for the detection of alpha and beta particles, and the instrument is still used for this purpose today. For alpha particles and low energy beta particles the "end-window" type of G-M tube has to be used as these particles have a limited range even in free air, and are easily stopped by a solid material. Therefore the tube requires a window which is thin enough to allow as many as possible of these particles through to the fill gas. The window is usually made of mica with a density of about 1.5 - 2.0 mg/cm2.

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Alpha particles have the shortest range, and to detect these the window should ideally be within 10mm of the radiation source due to alpha particle attenuation in free air. However, the G-M tube produces a pulse output which is the same magnitude for all detected radiation, so a Geiger counter with an end window tube cannot distinguish between alpha and beta particles. A skilled operator can use distance to differentiate alpha and high energy beta, but with the detector in close contact with the radiation source the types are indistinguishable. The "pancake" Geiger-Muller detector is a variant of the end window probe, but designed with a larger detection area to make checking quicker. However the pressure of the atmosphere against the low pressure of the fill gas limits the window size due to the limited strength of the window membrane.

High energy beta particles can also be detected by a thin-walled "windowless" G-M tube, which has no end window. Although the tube walls have a greater stopping power than a thin end window, they still allow these more energetic particles to reach the fill gas.

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End-window G-M detectors are still used as a general purpose portable radioactive contamination measurement and detection instrument, owing to their relatively low cost, robustness and their relatively high detection efficiency; particularly with high energy beta particles.

However for discrimination between alpha and beta particles or provision of particle energy information, (1) scintillation counters or (2) proportional counters should be used. Those instrument types are manufactured with much larger detector areas, which means that checking for surface contamination is quicker than with a G-M instrument.

Keywords: However for discrimination between alpha and beta particles or provision of particle energy information, (1) scintillation counters or (2) proportional counters should be used.

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Gamma and X-ray detection Geiger counters are widely used to detect gamma radiation, and for this the windowless tube is used. However, efficiency is generally low due to the poor interaction of gamma rays compared with alpha and beta particles. For instance, a chrome steel G-M tube is only about 1% efficient over a wide range of energies.

The article on the Geiger-Muller tube carries a more detailed account of the techniques used to detect photon radiation. For high energy gamma it largely relies on interaction of the photon radiation with the tube wall material, usually 1â&#x20AC;&#x201C;2 mm of chrome steel on a "thick-walled" tube, to produce electrons within the wall which can enter and ionize the fill gas. This is necessary as the low pressure gas in the tube has little interaction with high energy gamma photons. However, for low energy photons there is greater gas interaction and the direct gas ionisation effect increases. With decreasing energy the wall effect gives way to a combination of wall effect and direct ionisation, until direct gas ionisation dominates. Due to the variance in response to different photon energies, windowless tubes employ what is known as "energy compensation" which attempts to compensate for these variations over a large energy range.

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Low energy photon radiation such as low energy X rays or gamma rays interacts better with the fill gas. Consequently a typical design for low energy photon detection for these is a long tube with a thin wall or with an end window. The tube has a larger gas volume than a steel walled tube to give an increased chance of particle interaction.

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Neutron detection A variation of the Geiger tube is used to measure neutrons, where the gas used is boron trifluoride (BF3) or helium-3 (3He) and a plastic moderator is used to slow the neutrons. This creates an alpha particle inside the detector and thus neutrons can be counted.

Geiger tube filled with BF3 for detection of thermal neutrons 10 B + n → 7 Li + 4 α https://www.orau.org/PTP/collection/proportional%20counters/bf3info.htm 5 3 2 3 He + n → 3 Li + 1 P http://large.stanford.edu/courses/2012/ph241/lam1/ 2 1 1

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Boron trifluoride Detector

This neutron detector was produced by 20th Century Electronics in England. The company began manufacturing BF3 counters in the early 1950s. It is approximately 16 1/2 inches long, 2 inches in diameter, copper walled and filled with BF3. One end (towards the right in the above photo) has a threaded cap to protect the fragile glass insulator.

The model number, marked on the wall of the tube, is 15EB70/50/G/UA0539. The EB refers to "enriched boron trifluoride." The 50 refers to the tube diameter, i.e., 50 mm.

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Gamma measurementâ&#x20AC;&#x201D;personnel protection and process control The term "Geiger counter" is commonly used to mean a hand-held survey type meter, however the Geiger principle is in wide use in installed "area gamma" alarms for personnel protection, and in process measurement and interlock applications. A Geiger tube is still the sensing device, but the processing electronics will have a higher degree of sophistication and reliability than that used in a hand held survey meter.

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Physical design For hand-held units there are two fundamental physical configurations: the "integral" unit with both detector and electronics in the same unit, and the "two-piece" design which has a separate detector probe and an electronics module connected by a short cable.

The integral unit allows single-handed operation, so the operator can use the other hand for personal security in challenging monitoring positions, but the two piece design allows easier manipulation of the detector, and is commonly used for alpha and beta surface contamination monitoring where careful manipulation of the probe is required or the weight of the electronics module would make operation unwieldy. A number of different sized detectors are available to suit particular situations, such as placing the probe in small apertures or confined spaces.

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Gamma and X-Ray detectors generally use an "integral" design so the Geigerâ&#x20AC;&#x201C;MĂźller tube is conveniently within the electronics enclosure. This can easily be achieved because the casing usually has little attentuation, and is employed in ambient gamma measurements where distance from the source of radiation is not a significant factor. However, to facilitate more localised measurements such as "surface dose", the position of the tube in the enclosure is sometimes indicated by targets on the enclosure so an accurate measurement can be made with the tube at the correct orientation and a known distance from the surface.

There is a particular type of gamma instrument known as a "hot spot" detector which has the detector tube on the end of a long pole or flexible conduit.

These are used to measure high radiation gamma locations whilst protecting the operator by means of distance shielding.

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Particle detection of alpha and beta can used in both integral and two-piece designs. A pancake probe (for alpha/beta) is generally used to increase the area of detection in two-piece instruments whilst being relatively light weight.

In integral instruments using an end window tube there is a window in the body of the casing to prevent shielding of particles. There are also hybrid instruments which have a separate probe for particle detection and a gamma detection tube within the electronics module. The detectors are switchable by the operator, depending the radiation type that is being measured.

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Pancake G-M tube used for alpha and beta detection; the delicate mica window is usually protected by a mesh when fitted in an instrument.

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Guidance on application use[edit] In the United Kingdom the HSE has issued a user guidance note on selecting the best portable instrument type for the radiation measurement application concerned.[4][3] This covers all radiation protection instrument technologies and is a useful comparative guide to the use of G-M detectors. The guide does not recommend the G-M detector for mixed alpha and beta contamination detection, and they are only recommended as "satisfactory" for beta-only contamination. However for gamma and low-voltage X-rays they are recommended as the best instrument type.

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Use of a "hot spot" detector on a long pole to survey waste casks

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G-M pancake detector feeding a microcontroller data-logger sending data to a PC via bluetooth. A radioactive rock was placed on top the G-M causing the graph to rise.

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G-M counters being used as gamma survey monitors, seeking radioactive satellite debris

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Geiger–Müller tube The Geiger–Müller tube or G–M tube is the sensing element of the Geiger counter instrument used for the detection of ionizing radiation. It was named after Hans Geiger, who invented the principle in 1908, and Walther Müller, who collaborated with Geiger in developing the technique further in 1928 to produce a practical tube that could detect a number of different radiation types.

It is a gaseous ionization detector and uses the Townsend avalanche phenomenon to produce an easily detectable electronic pulse from as little as a single ionising event due to a radiation particle. It is used for the detection of gamma radiation, X-rays, and alpha and beta particles. It can also be adapted to detect neutrons. The tube operates in the "Geiger" region of ion pair generation. This is shown on the accompanying plot for gaseous detectors showing ion current against applied voltage.

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Whilst it is a robust and inexpensive detector, the G–M is:  unable to measure high radiation rates efficiently,  has a finite life in high radiation areas and  is unable to measure incident radiation energy, so no spectral information can be generated and there is no discrimination between radiation type.

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Plot of ion pair current against applied voltage for a cylindrical gaseous radiation detector with a central wire anode.

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Principle of operation The tube consists of a chamber filled with an inert gas at low-pressure (about 0.1 atmosphere). The chamber contains two electrodes, between which there is a potential difference of several hundred volts. The walls of the tube are either metal or have their inside surface coated with a conductor to form the cathode, while the anode is a wire in the center of the chamber. several hundred volts

0.1 atmosphere

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When ionizing radiation strikes the tube, some molecules of the gas are ionized, either directly by the incident radiation or indirectly by means of secondary electrons produced in the walls of the tube. This creates positively charged ions and electrons, known as â&#x20AC;&#x153;ion pairsâ&#x20AC;?, in the fill gas.

The strong electric field created by the tube's electrodes accelerates the positive ions towards the cathode and the electrons towards the anode. Close to the anode in the "avalanche region" the electrons gain sufficient energy to ionize additional gas molecules and create a large number of electron avalanches which spread along the anode and effectively throughout the avalanche region. This is the "gas multiplication" effect which gives the tube its key characteristic of being able to produce a significant output pulse from a single ionising event

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If there were to be only one avalanche per original ionising event, then the number of excited molecules would be in the order of 106 to 108. However the production of multiple avalanches results in an increased multiplication factor which can produce 109 to 1010 ion pairs.

Keypoint: Single event multiplication factor: 106 to 108 Multiple event multiplication factor: 109 to 1010

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The creation of multiple avalanches is due to the production of UV photons in the original avalanche, which are not affected by the electric field and move laterally to the axis of the anode to instigate further ionising events by collision with gas molecules. These collisions produce further avalanches, which in turn produce more photons, and thereby more avalanches in a chain reaction which spreads laterally through the fill gas, and envelops the anode wire.

The accompanying diagram shows this graphically. The speed of propagation of the avalanches is typically 2â&#x20AC;&#x201C;4 cm per microsecond, so that for common sizes of tubes the complete ionisation of the gas around the anode takes just a few microseconds. This short, intense pulse of current can be measured as a count event in the form of a voltage pulse developed across an external electrical resistor. This can be in the order of volts, thus making further electronic processing simple.

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Visualisation of the spread of Townsend avalanches by means of UV photons.

This mechanism allows a single ionising event to ionise all the gas surrounding the anode by triggering multiple avalanches.

Spread of avalanches In a GM Tube

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The discharge is terminated by the collective effect of the positive ions created by the avalanches. These ions have lower mobility than the free electrons due to their higher mass and remain in the area of the anode wire.

This creates a "space charge" which counteracts the electric field which is necessary for continued avalanche generation.

For a particular tube geometry and operating voltage this termination always occurs when a certain number of avalanches have been created, therefore the pulses from the tube are always of the same magnitude regardless of the energy of the initiating particle. Consequently, there is no radiation energy information in the pulses which means the Geiger–Muller tube cannot be used to generate spectral information about the incident radiation.

Pressure of the fill gas is important in the generation of avalanches. Too low a pressure and the efficiency of interaction with incident radiation is reduced.

Too high a pressure, and the “mean free path” for collisions between accelerated electrons and the fill gas is too small, and the electrons cannot gather enough energy between each collision to cause ionisation of the gas.

The energy gained by electrons is proportional to the ratio “e/p”, where “e” is the electric field strength at that point in the gas, and “p” is the gas pressure.

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Detection of higher energy gamma in a thick-walled tube. Secondary electrons generated in the wall can reach the fill gas to produce avalanches.

Multiple avalanches omitted for clarity

Interaction of gamma radiation with GM tube wall

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Types of tube Broadly, there are two main types of Geiger tube construction.

End window type For alpha particles, low energy beta particles, and low energy X-rays, the usual form is a cylindrical end-window tube. This type has a window at one end covered in a thin material through which low-penetrating radiation can easily pass. Mica is a commonly used material due to its low mass per unit area. The other end houses the electrical connection to the anode.

Pancake tube Pancake Gâ&#x20AC;&#x201C;M tube, the circular concentric anode can clearly be seen.

The pancake tube is a variant of the end window tube, but which is designed for use for beta and gamma contamination monitoring. It has roughly the same sensitivity to particles as the end window type, but has a flat annular shape so the largest window area can be utilised with a minimum of gas space. Like the cylindrical end window tube, mica is a commonly used window material due to its low mass per unit area. The anode is normally multi-wired in concentric circles so it extends fully throughout the gas space.

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Windowless type This general type is distinct from the dedicated end window type, but has two main sub-types, which use different radiation interaction mechanisms to obtain a count.

Thick walled A selection of thick walled Gâ&#x20AC;&#x201C;M tubes for gamma detection. The largest has an energy compensation ring; the others are not energy compensated Used for high energy gamma detection, this type generally has an overall wall thickness of about 1-2 mm of chrome steel. Because most high energy gamma photons will pass through the low density fill gas without interacting, the tube uses the interaction of photons on the molecules of the wall material to produce high energy secondary electrons within the wall. Some of these electrons are produced close enough to the inner wall of the tube to escape into the fill gas. As soon as this happens the electron drifts to the anode and an electron avalanche occurs as though the free electron had been created within the gas. The avalanche is a secondary effect of a process that starts within the tube wall; the avalanche is not the effect of radiation directly on the gas itself.

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Thin walled Thin walled tubes are used for: high energy beta detection, where the beta enters via the side of the tube and interacts directly with the gas, but the radiation has to be energetic enough to penetrate the tube wall. Low energy beta, which would penetrate an end window, would be stopped by the tube wall.

Low energy gamma and X-ray detection.

The lower energy photons interact better with the fill gas so this design concentrates on increasing the volume of the fill gas by using a long thin walled tube and does not use the interaction of photons in the tube wall. The transition from thin walled to thick walled design takes place at the 300â&#x20AC;&#x201C;400 KeV energy levels. Above these levels thick walled designs are used, and beneath these levels the direct gas ionisation effect is predominant.

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Schematic of a Geiger counter using an "end window" tube for lowpenetrating radiation. A loudspeaker is also used for indication

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Pancake Gâ&#x20AC;&#x201C;M tube, the circular concentric anode can clearly be seen.

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A selection of thick walled Gâ&#x20AC;&#x201C;M tubes for gamma detection. The largest has an energy compensation ring; the others are not energy compensated

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Neutron G–M tubes will not detect neutrons since these do not ionise the gas.

However, neutron-sensitive tubes can be produced which either have the (1) inside of the tube coated with boron, or (2) the tube contains boron trifluoride or (3) helium-3 as the fill gas.

The neutrons interact with the boron nuclei, producing alpha particles, or directly with the helium-3 nuclei producing hydrogen (proton) and tritium ions and electrons. These charged particles then trigger the normal avalanche process. + n → 73Li + 42α + 2e3 He + n → 1 H+ + 3 H + e1 1 2 3 He + n → 1 H + 3 H+ + e1 1 2

10

5B

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Gas mixtures The main component of the gas fill mixture is an inert gas such as helium, argon or neon, in some cases in a Penning mixture, and a "quench" gas of 5– 10% of an organic vapor or a halogen gas to prevent multiple pulsing. The halogen-filled G–M tube was invented by Sidney H. Liebson in 1947 and has several advantages over the tubes with older organic mixtures. The halogen tube discharge takes advantage of a metastable state of the inert gas atom to more-readily ionize a halogen molecule than an organic vapor, enabling the tube to operate at much lower voltages, typically 400–600 volts instead of 900–1200 volts. It also has a longer life than tubes quenched with organic compounds, because the halogen ions can recombine while the organic vapor is gradually destroyed by the discharge process (giving the latter a life of around 108 events). For these reasons, the halogen-filled tube is now the most common.

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Geiger plateau The Geiger plateau is the voltage range in which the GM tube operates in its correct mode. If a Gâ&#x20AC;&#x201C;M tube is exposed to a steady radiation source and the applied voltage is increased from zero, it follows the plot of ion current shown in this article. In the "Geiger region" the gradient flattens; this is the Geiger plateau.

Depending on the characteristics of the specific tube (manufacturer, size, gas type, etc.) the voltage range of the plateau will vary. In this region, the potential difference in the counter is strong enough to allow the creation of multiple avalanches.

A lower voltage is not sufficient to cause a complete discharge along the anode, and individual Townsend avalanches are the result, and the tube tries to act as a proportional counter. If the applied voltage is higher than the plateau, a continuous glow discharge is formed and the tube cannot detect radiation.

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The plateau has a slight slope caused by increasing sensitivity to low energy radiation as the voltage increases. Normally when a particle ionizes gas atoms, complete ionization of the gas occurs. But for a low energy particle, it is possible that the kinetic energy in addition to the potential energy of the voltage are insufficient for the avalanche to occur and the ion recombines. As applied voltage rises, the threshold for the minimum radiation response falls, thus the counter's sensitivity rises; giving rise to the slope.

The counting rate for a given radiation source varies slightly as the applied voltage is varied and to prevent this, a regulated voltage is used. However, it is normal to operate the tube in the middle of the plateau to allow for variations in the tube supply voltage.

Keypoints: In Geiger Muller counter the ionization is by Multiple Avalanches

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Quenching and dead time The ideal Gâ&#x20AC;&#x201C;M tube should produce a single pulse on entry of a single ionising particle. It must not give any spurious pulses, and must recover quickly to the passive state.

Unfortunately for these requirements, when positive argon ions reach the cathode and become neutral argon atoms again by obtaining electrons from it, the atoms can acquire their electrons in enhanced energy levels. These atoms then return to their ground state by emitting photons which can in turn produce further ionisation and hence cause spurious secondary pulse discharges. If nothing were done to counteract it, ionisation could even escalate, causing a so-called current "avalanche" which if prolonged could damage the tube. Some form of quenching of the ionisation is therefore essential.

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The disadvantage of quenching is that for a short time after a discharge pulse has occurred (the so-called dead time, which is typically 50â&#x20AC;&#x201C;100 microseconds), the tube is rendered insensitive and is thus temporarily unable to detect the arrival of any new ionising particle. This effectively causes a loss of counts at sufficiently high count rates and limits the Gâ&#x20AC;&#x201C;M tube to a count rate of between 104 to 105 counts per second, depending on its characteristic. A consequence of this is that ion chamber instruments were sometimes preferred for higher count rates, however the modern application of "electronic quenching" (see below) can extend this upper limit considerably.

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Chemical quenching Self-quenching or internal-quenching tubes stop the discharge without external assistance, by means of the addition of a small amount of a polyatomic organic vapor such as butane or ethanol, or alternatively a halogen such as bromine or chlorine.

If a poor diatomic gas quencher is introduced to the tube, the positive argon ions, during their motion toward the cathode, would have multiple collisions with the quencher gas molecules and transfer their charge and some energy to them. Thus, neutral argon atoms would be produced and the quencher gas ions in their turn would reach the cathode, gain electrons there from, and move into excited states which would decay by photon emission, producing tube discharge. However, effective quencher molecules, when excited, lose their energy not by photon emission, but by dissociation into neutral quencher molecules. No spurious pulses are thus produced.

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External quenching, sometimes also called "active quenching" or "electronic quenching", uses high speed control electronics to rapidly remove and reapply the high voltage between the electrodes after each discharge peak.

This results in faster quenching of the tube than using the effect of gas alone, and allows for greatly increased tube lifetimes.

A technique known as "time-to-first-count" is sometimes used in conjunction with this to greatly increase the maximum count rate of the tube.

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Dead time and recovery time in a Geiger Muller tube.[4] The tube can produce no further pulses during the dead time, and is able to produce only pulses of limited height until the recovery time elapses.

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Fold-back One consequence of the dead time effect is the possibility of a high count rate continually triggering the tube before the recovery time has elapsed. This may produce pulses too small for the counting electronics to detect and lead to the very undesirable situation whereby a Gâ&#x20AC;&#x201C;M counter in a very high radiation field is falsely indicating a low level. This phenomenon is known as "foldback". An industry rule of thumb is that the discriminator circuit receiving the output from the tube should detect down to 1/10 of the magnitude of a normal pulse to guard against this. Additionally the circuit should detect when "pulse pile-up " has occurred, where the apparent anode voltage has moved to a new dc level through the combination of high pulse count and noise. The electronic design of Geigerâ&#x20AC;&#x201C;Muller counters must be able to detect this situation and give an alarm; it is normally done by setting a threshold for excessive tube current.

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Detection efficiency The efficiency of detection of a Gâ&#x20AC;&#x201C;M tube varies with the type of incident radiation. Tubes with thin end windows have very high efficiencies (can be nearly 100%) for high energy beta, though this drops off as the beta energy decreases due to attenuation by the window material. Alpha particles are also attenuated by the window. As alpha particles have a maximum range of less than 50 mm in air, the detection window should be as close as possible to the source of radiation. The attenuation of the window adds to the attenuation of air, so the window should have a density as low as 1.5 to 2.0 mg/cm2 to give an acceptable level of detection efficiency. The article on stopping power explains in more detail the ranges for particles types of various energies. The counting efficiency of photon radiation (gamma and X-rays above 25 keV) depends on the efficiency of radiation interaction in the tube wall, which increases with the atomic number of the wall material. Chromium iron is a commonly used material, which gives an efficiency of about 1% over a wide range of energies.

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Energy compensation If a Gâ&#x20AC;&#x201C;M tube is to be used for gamma or X-ray dosimetry measurements the energy of incident radiation, which affects the ionising effect, must be taken into account. However individual pulses from a Gâ&#x20AC;&#x201C;M tube do not carry any energy information. A solution is to assign a radiation dose to each counting event, so the tube characteristic relates the number of counts to the intensity of incident radiation.

At low photon energy levels the response increases as low energy photons have a greater interaction with the fill gas than high energy photons. The tube therefore has an increased response for radiation which has a lower dose rate, and a correction must be applied to prevent an incorrect high reading for low energy photons. This discrepancy can be 2â&#x20AC;&#x201C;3 times greater or more, and for a thick-walled tube usually peaks at about 60 keV, where radiation interactions with the gas are still large, but the shielding effect of the wall has not become dominant.

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This correction is achieved by 'energy compensation' of the tube, which modifies the number of count events in accordance with the energy of the incident radiation by using an external filter collar of energy absorbing material. The collar has an increased attenuation of low energy gamma, and so compensates for the increased energy response of the naked tube at those levels. The aim is that sensitivity/energy characteristic of the tube should be matched by the absorption/energy characteristic of the filter. This results in a more uniform response over the stated range of detection energies for the tube.

Lead and tin are commonly used materials, and a simple filter effective above 150 keV can be made using a continuous collar along the length of the tube.

However, at lower energy levels this attenuation can become too great, so air gaps are left in the collar to allow low energy radiation to have a greater effect.

In practice, compensation filter design is an empirical compromise to produce an acceptably uniform response, and a number of different materials and geometries are used to obtain the required correction.

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Comparative response curves for GM tube with and without radiation energy compensation

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Thin-walled glass Gâ&#x20AC;&#x201C;M tube showing a spiral wire cathode. The tape bands are for fixing compensating rings

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Thin-walled glass Gâ&#x20AC;&#x201C;M tube showing a spiral wire cathode. The tape bands are for fixing compensating rings

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Thin-walled glass Gâ&#x20AC;&#x201C;M tube showing a spiral wire cathode. The tape bands are for fixing compensating rings

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Thin-walled glass Gâ&#x20AC;&#x201C;M tube with energy compensating rings fitted. The complete assembly fits into the aluminium housing.

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Thin-walled glass Gâ&#x20AC;&#x201C;M tube with energy compensating rings fitted. The complete assembly fits into the aluminium housing.

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Thin-walled glass Gâ&#x20AC;&#x201C;M tube with energy compensating rings fitted. The complete assembly fits into the aluminium housing.

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Proportional counter The proportional counter is a type of gaseous ionization detector device used to measure particles of ionizing radiation.

The key feature is its ability to measure the energy of incident radiation, by producing a detector output that is proportional to the radiation energy; hence the detector's name. It is widely used where energy levels of incident radiation must be known, such as in the discrimination between alpha and beta particles, or accurate measurement of X-ray radiation dose. A proportional counter uses a combination of the mechanisms of a Geigerâ&#x20AC;&#x201C; MĂźller tube and an ionization chamber, and operates in an intermediate voltage region between these. The accompanying plot shows the proportional counter operating voltage region for a co-axial cylinder arrangement.

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Plot of variation of ion pair current against applied voltage for a wire cylinder gaseous radiation detector.

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Operation In a proportional counter the fill gas of the chamber is an inert gas which is ionised by incident radiation, and a quench gas to ensure each pulse discharge terminates; a common mixture is 90% argon, 10% methane, known as P-10.

What is P10: Proportional counter with 90% Argon (inert gas) with 10% quenched gas (10% methane) An ionising particle entering the gas collides with an atom of the inert gas and ionises it to produce an electron and a positively charged ion, commonly known as an "ion pair". As the charged particle travels through the chamber it leaves a trail of ion pairs along its trajectory, the number of which is proportional to the energy of the particle if it is fully stopped within the gas.

Typically a 1 MeV stopped particle will create about 30,000 ion pairs.

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What is P10: Proportional counter with 90% Argon (inert gas) with 10% quenched gas (10% methane)

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The chamber geometry and the applied voltage is such that in most of the chamber the electric field strength is low and the chamber acts as an ion chamber. However, the field is strong enough to prevent re-combination of the ion pairs and causes positive ions to drift towards the cathode and electrons towards the anode. This is the "ion drift" region. In the immediate vicinity of the anode wire, the field strength becomes large enough to produce Townsend avalanches. This avalanche region occurs only fractions of a millimeter from the anode wire, which itself is of a very small diameter. The purpose of this is to use the multiplication effect of the avalanche produced by each ion pair. This is the "avalanche" region.

A key design goal is that each original ionising event due to incident radiation produces only one avalanche. This is to ensure proportionality between the number of original events and the total ion current. For this reason the applied voltage, the geometry of the chamber and the diameter of the anode wire are critical to ensure proportional operation. If avalanches start to self-multiply due to UV photons as they do in a Geigerâ&#x20AC;&#x201C;Muller tube, then the counter enters a region of "limited proportionality" until at a higher applied voltage the Geiger discharge mechanism occurs with complete ionisation of the gas enveloping the anode wire and consequent loss of particle energy information.

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Plot of variation of ion pair current against applied voltage for a wire cylinder gaseous radiation detector.

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Therefore, it can be said that the proportional counter has the key design feature of two distinct ionisation regions: 1. Ion drift region: in the outer volume of the chamber â&#x20AC;&#x201C; creation of number ion pairs proportional to incident radiation energy.

2. Avalanche region: in the immediate vicinity of the anode â&#x20AC;&#x201C; Charge amplification of ion pair currents, while maintaining localised avalanches.

The process of charge amplification greatly improves the signal-to-noise ratio of the detector and reduces the subsequent electronic amplification required.

In summary, the proportional counter is an ingenious combination of two ionisation mechanisms in one chamber which finds wide practical use.

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The generation of discrete Townsend avalanches in a proportional counter.

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Plot of electric field strength at the anode, showing boundary of avalanche region.

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The gas flow proportional counter for alpha counting was invented in 1943 by John Simpson at the University of Chicago Metallurgical Laboratory. Its purpose was to measure plutonium (an alpha emitter) in the presence of beta-emitting fission products.

The key feature of this instrument that allowed it to reject beta pulses was its use of methane as the counting gas. Simpson would later invent P-10 gas (10% methane, 90% argon), the most widely employed gas in proportional counters. The instrument also featured a short time constant which reduced pulse pile up and assisted in rejecting the beta pulses.

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RCL Mark 2, Model 201 Fast Neutron Proportional Counter (1950s) This is a fast neutron detector produced by Radiation Counter Laboratories (RCL) of Skokie, Illinois.

The tube is approximately 8 1/4 inches long and 1 7/8 inches in diameter. A brass evacuation tube can be seen projecting to the right from the brass chamber. The actual proportional counter chamber is 1.2 inches long, lined with 1/16 inch of polyethylene, and filled with methane at a pressure of 150 cm. It operated at 2100 volts.

Fast neutrons knock protons off the polyethylene lining. The protons then ionize the methane fill gas to produce the signal. The RCL detector designation is the Mark 2, Model 201, Serial 127. The "1" at the end of the model number (201) refers to the number of chambers housed in the unit. The Models 202 and 203 used two and three chambers respectively.

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Proportional Counter

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Proportional Counter

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Applications Spectroscopy The proportionality between the energy of the charged particle travelling through the chamber and the total charge created makes proportional counters useful for charged particle spectroscopy. By measuring the total charge (time integral of the electric current) between the electrodes, we can determine the particle's kinetic energy because the number of ion pairs created by the incident ionizing charged particle is proportional to its energy.

The energy resolution of a proportional counter, however, is limited because both the initial ionization event and the subsequent 'multiplication' event are subject to statistical fluctuations characterised by a standard deviation equal to the square root of the average number formed. However, in practice these are not as great as would be predicted due to the effect of the empirical Fano factor which reduces these fluctuations. In the case of argon, this is experimentally about 0.2.

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Photon detection Proportional counters are also useful for detection of high energy photons, such as gamma-rays, provided these can penetrate the entrance window.

They are also used for the detection of X-rays to below 1 Kev energy levels, using thin walled tubes operating at or around atmospheric pressure.

Radioactive contamination detection Proportional counters in the form of large area planar detectors are used extensively to check for radioactive contamination on personnel, flat surfaces, tools and items of clothing. This is normally in the form of installed instrumentation because of the difficulties of providing portable gas supplies for hand-held devices. They are constructed with a large area detection window made from such as metallised mylar which forms one wall of the detection chamber and is part of the cathode. The anode wire is routed in a convoluted manner within the detector chamber to optimise the detection efficiency.

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They are normally used to detect alpha and beta particles, and can enable discrimination between them by providing a pulse output proportional to the energy deposited in the chamber by each particle. They have a high efficiency for beta, but lower for alpha. The efficiency reduction for alpha is due to the attenuation effect of the entry window, though distance from the surface being checked also has a significant effect, and ideally a source of alpha radiation should be less than 10mm from the detector due to attenuation in air.

These chambers operate at very slight positive pressure above ambient atmospheric pressure. The gas can be sealed in the chamber, or can be changed continuously, in which case they are known as "gas-flow proportional counters". Gas flow types have the advantage that they will tolerate small holes in the mylar screen which can occur in use, but they do require a continuous gas supply.

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Guidance on application use In the United Kingdom the HSE has issued a user guidance note on selecting the correct radiation measurement instrument for the application concerned.

This covers all radiation instrument technologies, and is a useful comparative guide to the use of proportional counters.

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■ ωσμ∙Ωπ∆ ∇ º≠δ≤>ηθφФρ|β≠Ɛ∠ ʋ λ α ρτ√ ≠≥ѵФε ≠≥ѵФdsssa

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Biological Half-life

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Biological Half-life The radioactive half-life for a given radioisotope is physically determined and unaffected by the physical or chemical conditions around it. However, if that radioisotope is in a living organism it may be excreted so that it no longer is a source of radiation exposure to the organism. For a number of radioisotopes of particular medical interest, the rate of excretion has been cast in the form of an effective biological half-life. The rate of decrease of radiation exposure is then affected by both the physical and biological half-life, giving an effective half-life for the isotope in the body. Though the biological half-life cannot be expected to be as precise as the physical half-life, it is useful compute an effective half-life fromďź&#x161;

1/TEffective = 1/TPhysical + 1/TBiological

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Examples of the half-lives show that biological clearing is sometimes dominant and sometimes physical decay is the dominant influence.

Half Life in day Isotopes

T Physical

T Biological

T Effective

3H

4.5 x 103

12

12

22P

14.3

1155

14.1

90Sr

1.1 x 104

1.8 x 104

6.8 x 103

99mTc

0.25

1

0.20

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Tritium, 3H, has a fairly long physical half life but clears from the body quickly, lessening the exposure. Phosphorous, 32P, is used for some kinds of bone scans. The phosphorous tends to be held in the bones, leading to a long biological half-life, but its physical half-life is short enough to minimize exposure. Strontium, 90Sr, is very bad news in the environment. It mimics calcium and therefore gets trapped in bone. This gives it a long biological half-life to go with its long physical half-life, making it doubly dangerous.

Technetium, 99mTc, is one of the favorites for diagnostic scans because of short physical and biological half-lives. It clears from the body very quickly after the imaging procedures.

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The biological half-life or terminal half-life of a substance is the time it takes for a substance (for example a metabolite, drug, signalling molecule, radioactive nuclide, or other substance) to lose half of its pharmacologic, physiologic, or radiologic activity, according to the Medical Subject Headings (MeSH) definition. Typically, this refers to the body's cleansing through the function of kidneys and liver in addition to excretion functions to eliminate a substance from the body. In a medical context, half-life may also describe the time it takes for the blood plasma concentration of a substance to halve (plasma half-life) its steady-state. The relationship between the biological and plasma half-lives of a substance can be complex depending on the substance in question, due to factors including accumulation in tissues (protein binding), active metabolites, and receptor interactions.

Biological half-life is an important pharmacokinetic parameter and is usually denoted by the abbreviation t ½ While a radioactive isotope decays perfectly according to first order kinetics where the rate constant is fixed, the elimination of a substance from a living organism follows more complex chemical kinetics.

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More Reading http://www.euronuclear.org/info/encyclopedia.htm

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■ ωσμ∙Ωπ∆ ∇ º≠δ≤>ηθφФρ|β≠Ɛ∠ ʋ λ α ρτ√ ≠≥ѵФε ≠≥ѵФdsssa

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Good Luck!

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