Understanding Neutron Radiography Post Exam Reading VIII-Part 1 of 2A

Understanding Neutron Radiography 

Reading VIII Part 1 of 2 

13 th 2016 August 

Post Exam Reading 

Charlie Chong/ Fion Zhang

Reactor 


Reactor 


The Magical Book of Neutron Radiography 



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 


3. Techniques/Calibrations 

• Blocking and filtering 

• Multifilm technique 

• Enlargement and projection 

• Stereoradiography 

• Triangulation methods 

• Autoradiography 

• Flash Radiography 

• In-motion radiography 

• Fluoroscopy 

• Electron emission radiography 

• Micro-radiography 

• Laminography (tomography) 

• Control of diffraction effects 

• Panoramic exposures 

•Gaging 

• Real time imaging 

• Image analysis techniques 


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 

4 th August 2016


SME- Subject Matter Expert 

http://cn.bing.com/videos/search?q=Walter+Lewin&FORM=HDRSC3 

https://www.youtube.com/channel/UCiEHVhv0SBMpP75JbzJShqw

Gamma- Radiography 

TABLE 1. Characteristics of three isotope sources commonly used for 

radiography. 

Source 

T½ 

Energy 

HVL 

HVL 

Specific 

Dose rate* 

Pb 

Fe 

Activity 

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 

4.8mm 

? 

350 Cig -1 

0.59163 

(Aver. 0.34 MeV) 

Th232 

0.068376 

Dose rate* Rem/hr at one meter per curie 


八千里路云和月 



闭门练功 



http://greekhouseoffonts.com/


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 


Pitești 


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. 


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. 


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. 


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. 


Because the moderator emits neutrons in all directions, their intensity is 

proportionally with 1/r 2 . 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. 


There are known different types of collimators, more important are: 

■ pin-hole, 

■ Soller and 

■ divergent collimators. 


Soller Collimator 


Sóller Collimator 


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. 


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. 


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. 


Based on dimensional constrains of the tangential channel of ACPR, previous 

experimental determinations of the thermal neutron flux and intensity 

(8 ⋅ 10 11 n/cm 2 /s near core and 1.12 x 10 6 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: 


• 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⋅10 6 n/cm 2 /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? 


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 γ radiography investigations; 

3. a set of successive apertures from boron, indium and lead for the 

formation of the divergent collimator. 


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. 


ACPR- Annular Core Pulse Reactor 



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 


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 


Fig. 1. Sketch of the collimator of the neutron radiography facility at the 

tangential beam port of the ACPR. 


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. 


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 14.3 16 19.5 


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. 


Fig. 2. Neutron beam intensity. 


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. 


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. 


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


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. 


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⋅10 5 n/cm 2 /s 

and 4.85⋅105 – 5.70⋅10 5 n/cm 2 /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⋅10 6 n/cm2/mrem and 8.44⋅10 6 n/cm 2 /mrem, 

respectively. The divergent angle of the collimator is 3 o -3.3 o 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. 


Sandia’s Annular Core Research Reactor 

conducts 10,000th operation 


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

ACPR- Annular Core Pulse Reactor 


ACRR- Annular Core (Pulse) Research Reactor 


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

ACRR- Annular Core (Pulse) Research Reactor 


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. 


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. 


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. 


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. 


Key! 



Reactor 


Reactor 


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) 


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 


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. 


Recoil - Measurement of Hydrogen Depth Profile Using Fast Neutrons- 

Materials Analysis with Deuterium and Tritium Fusion Neutrons 


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. 


TABLE 4. Neutron classification. 

Class 

Thermal 

Epithermal 

Slow 

Intermediate 

Fast 

Relativistic 

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, α), (n, p), (n, d), (n, f ) and (n, γ) nuclear reactions. 


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. 10 5 B(n,α)7 3 Li 

Fast neutrons are detected by a similar counter, in which thermal neutrons 

are absorbed in an external cadmium shield ( 113 Cd(n,γ) 114 Cd ; the fast 

neutrons that pass through the shield are thermalized in hydrogen rich 

material and counted in the proportional chambers. 

γ 

hydrogen 

rich material 

boron trifluoride (BF3) gas 

external 

Pb Shield? 


■ 

http://minerals.usgs.gov/minerals/pubs/commodity/ 


The Cross Section (barns) of 

commonly used conversion 

screen 

Careful on differential cross 

section for isotopes of same 

element. 


TABLE 6. Properties of Some Thermal Neutron Radiography Conversion Materials 

Material 

Useful Reactions 

Cross Section for 

Life 

Thermal Neutrons (barns) 

Lithium 

6 Li(n,α) 3 H 

910 

prompt 

Boron 

10 B(n,α) 7 Li 

3,830 

prompt 

Rhodium 

103 Rh(n) 104m Rh 

11 

45 min 

103 

Rh(n) 104 Rh 

139 

42 s 

Silver 

107 Ag(n) 108 Ag 

35 

2.3 min 

109 Ag (n) 110 Ag 

91 

24 s 

Cadmium 

113 Cd((n,γ) 114 Cd 

20,000 

prompt 

Indium 

115 In(n) 116 n 

157 

54 min 

115 

In(n) 116m ln 

42 

14 s 

Samarium 

149 Sm(n,γ) 150 Sm 

41,000 

prompt 

I52 

Sm(n) 153 Sm 

210 

47 h 

Europium 

151 Eu(n) 152 Eu 

3,000 

9.2 h 

Gadolinium 

155 

Gd(n,γ) I56 Gd 

61,000 

prompt 

157 

Gd(n,γ) 158 Gd 

254,000 

prompt 

Dyprosium 

164 

Dy(n) 165m Dy 

2,200 

1.25 min 

164 

Dy(n) 165 Dy 

800 

140 min 

Gold 

197 Au(n) 198 Au 

99 

2.7 days 


Neutron Cross Section of the elements 


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

Neutron Cross Section σtotal for Gd =50000 barn 



Neutron Cross Section σtotal for Dy =1010 barn 



TABLE X1.1 Thermal Neutron Linear Attenuation Coefficients Using 

Average Scattering and Thermal Absorption Cross Sections for the 

Naturally Occurring Elements 


E748-02 Standard Practices for Thermal Neutron Radiography of Materials

Material 



Life 


Lithium 

6 Li(n,α) 3 H 

910 

prompt 

Boron 

10 

B(n,α) 7 Li 

3,830 

prompt 

Rhodium 

103 Rh(n) 104m Rh 

11 

45 min 

103 Rh(n) 104 Rh 

139 

42 s 

Silver 

107 Ag(n) 108 Ag 

35 

2.3 min 

109 Ag (n) 110 Ag 

91 

24 s 

Cadmium 

113 Cd((n,γ) 114 Cd 

20,000 

prompt 

Indium 

115 

In(n) 116 n 

157 

54 min 

115 

In(n) 116m ln 

42 

14 s 

Samarium 

149 Sm(n,γ) 150 Sm 

41,000 

prompt 

I52 

Sm(n) 153 Sm 

210 

47 h 

Europium 

151 Eu(n) 152 Eu 

3,000 

9.2 h 

Gadolinium 

155 


61,000 

prompt 

157 

Gd(n.γ) 158 Gd 

254,000 

prompt 

Avg 64 Gd(n.γ) ? Gd 

49,000 

prompt 

Dyprosium 

164 

Dy(n) 165 mDy 

2,200 

1.25 min 

164 

Dy(n) 165 Dy 

800 

140 min 

Gold 

197 

Au(n) 198 Au 

99 

2.7 days 



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

Atomic Number

TABLE 6. Properties of Some Thermal Neutron Radiography Conversion 

Materials 

Material 



Life 


Lithium 

6 

Li(n,α) 3 H 

910 

prompt 

Boron 

10 B(n,α) 7 Li 

3,830 

prompt 

Rhodium 

103 Rh(n) 104m Rh 

11 

45 min 

103 

Rh(n) 104 Rh 

139 

42 s 

Silver 

107 

Ag(n) 108 Ag 

35 

2.3 min 

109 Ag (n) 110 Ag 

91 

24 s 

Cadmium 

113 

Cd((n,γ) 114 Cd 

20,000 

prompt 

Indium 

115 In(n) 116 n 

157 

54 min 

115 

In(n) 116m ln 

42 

14 s 

Samarium 

149 Sm(n,γ) 150 Sm 

41,000 

prompt 

I52 

Sm(n) 153 Sm 

210 

47 h 

Europium 

151 Eu(n) 152 Eu 

3,000 

9.2 h 

Gadolinium 

155 


61,000 

prompt 

157 

Gd(n.γ) 158 Gd 

254,000 

prompt 

Dyprosium 

164 

Dy(n) 165 mDy 

2,200 

1.25 min 

164 

Dy(n) 165 Dy 

800 

140 min 

Gold 

197 

Au(n) 198 Au 

99 

2.7 days 


TABLE 6. Capture cross sections σ of strongly absorbing elements for 

neutrons in approximate thermal equilibrium at 300 K (27 °C = 80 °F). 


TABLE 4. Average Characteristics of Thermal-Sources 

Type of Source 

Typical 

Resolution** 

Exposure 

Characteristics 

Radiographic 

Intensity* 

Time 

Radioisotope 

10 1 to 10 4 

Poor to Medium 

Long 

Stable operation. 

medium investment cost. 

possibly portable. 

Accelerator 

10 3 to 10 6 

Medium 

Average 

On-off operation. medium 

cost. possibly mobile. 

Subcritical 

10 4 to 10 6 

Good 

Average 

Stable operation, 

Assembly 

medium to high investment 

cost, mobility difficult 

Nuclear reactor 

10 5 to 10 8 

Excellent 

Short 

Stable operation, 

medium to high investment 

cost. mobility difficult 

*Neutrons per square centimeter per second. n/cm 2 ∙s 

**These classifications are relative 


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. 


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 (

Gadolinum 


Gadolinum 


isotope NA half-life DM DE (MeV) Decay Product 

148Gd syn 75y α 3.271 

144 

Sm 

150Gd syn 1.8×10 6 y α 2.808 

146 

Sm 

152Gd 0.20% 1.08×10 14 y α 2.205 

148 

Sm 

154Gd 2.18% – (α) 0.0812 

150 

Sm 

155Gd 14.80% – (α) 0.0812 

151 

Sm 

156Gd 20.47% – (SF)

σ,Cross Section of Gadolinium 

Gd Average = 49000 barn prompt 

155 

Gd(n,γ) 156 Gd = 61,000 barn prompt 

157 

Gd(n.γ) 158 Gd = 254,000 barn prompt 


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. 


TABLE 3. Neutron characteristics. 

Quantity 

Measurement 

Charge 

neutral 

Rest mass 

1.675 × 10 –27 kg 

Classical radius 1.532 × 10 –18 m 

Magnetic moment –9.662 × 10 –27 J·T –1 

Compton wavelength 1.320 × 10 –15 m 


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 10 18 

hydrogen atoms. 


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. 


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. 


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 ). 


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. 


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’) 


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. 


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?) 


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: 

59 

Co + n → 60 Co + γ → 60 Ni + 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. 


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,α) 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. 


In the (n,α) 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. 


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 (m 2 ) or 

barns (b), where 1 b = 10 –24 cm 2 =10 –28 m 2 . 


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. 


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 m 2 (10 b), compared to the very high cross 

sections of 4 × 10 –25 m 2 (several thousand barns, ~4000 b) at low energies. 


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. 


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

0.176 eV. 

0.176 eV 


TABLE 6. Capture cross sections σ of strongly absorbing elements for 

neutrons in approximate thermal equilibrium at 300 K (27 °C = 80 °F). 


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: 


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), 

σ is the cross section for neutron capture (in square centimeter), 

t i 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. 


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. 


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 


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. 



The Measurement Units 


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. 


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. 



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 = Q × D 



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 Röntgen, a German scientist who discovered X- 

rays. 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. 


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. 


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. 


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 


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. 


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. 


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. 


Radiation-related Quantities 


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 


The Röntgen 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 cm 3 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. 


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. 


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). 


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

Teratogenic 引起畸型的 


Deterministic Effects 


Chronic Effect 


Deterministic Effects 


ionization chambers 






ionization 

chambers 




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 “N” 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. 


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. 



Schematic diagram of an ionization chamber with parallel plates. 



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-Müller 

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. 


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 



The Ionization Chamber 

• It does not involve the gas multiplication mechanisms used by other 

radiation instruments, such as the Geiger-Müller counter or the 

proportional counter. 



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. 


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 

Parallel Plate 



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. 



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-Müller tube or the proportional counter. 



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-Müller 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. 









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. 



Chamber pressure 

■ 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. 



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. 



■ 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 µm. 

■ 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. 



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. 



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-Müller tube instrument is generally preferred 

where high levels of accuracy are not required. 



Hand-held integral ion chamber survey meter in use 



View of sliding beta shield on integral hand held instrument 



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 10 13 Ω. 



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. 



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-Müller tubes, which suffer 

from gas break down and are generally limited to a life of about 10 11 count 

events. Additionally, the Geiger-Müller tube cannot operate above about 10 4 

counts per second, due to dead time effects, whereas there is no similar 

limitation on the ion chamber. 



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 10 11 count events. 

• Geiger-Müller tube cannot operate above about 10 4 counts per second 

(10 4 N·s -1 ) , due to dead time effects, whereas there is no similar limitation 

on the ion chamber. 



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. 



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. 




http://www.world-nuclear-university.org/imis20/wnu/default.aspx


http://www.world-nuclear-university.org/imis20/wnu/default.aspx

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. 


A "two-piece" bench type Geiger–Müller counter with end-window detector 


Schematic of a Geiger counter using an "end window" tube for low 

penetration radiation. A loudspeaker is also used for indication 



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. 


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. 


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-Müller 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 10 4 to 

10 5 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. 


For accurate measurements ion chamber instruments are preferred for 

high radiation rates. 


Ionization Chamber 


Geiger Muller Counter 


G-M counter with pancake type probe 


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

Laboratory use of a G-M counter with end window probe to measure beta 

radiation from a radioactive source 



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/cm 2 . 



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. 



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. 



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 

1–2 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. 



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. 



Neutron detection 

A variation of the Geiger tube is used to measure neutrons, where the gas 

used is boron trifluoride (BF 3 ) or helium-3 ( 3 He) 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 BF 3 for detection of thermal neutrons 

10 

5 B + n → 7 3 Li + 4 2 α https://www.orau.org/PTP/collection/proportional%20counters/bf3info.htm 

3 

2 He + n → 3 1 Li + 1 1 P http://large.stanford.edu/courses/2012/ph241/lam1/ 



Boron trifluoride Detector 

This neutron detector was produced by 20th 

Century Electronics in England. The company 

began manufacturing BF 3 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. 


https://www.orau.org/PTP/collection/proportional%20counters/bf3twentiethlarge.htm

Gamma measurement—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. 



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. 



Gamma and X-Ray detectors generally use an "integral" design so the 

Geiger–Müller 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. 



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. 



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. 



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. 



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



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. 



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

satellite debris 


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. 


https://en.wikipedia.org/wiki/Geiger%E2%80%93M%C3%BCller_tube

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. 



Plot of ion pair current against applied voltage for a cylindrical gaseous 

radiation detector with a central wire anode. 










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 



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 “ion pairs”, 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 



If there were to be only one avalanche per original ionising event, then the 

number of excited molecules would be in the order of 10 6 to 10 8 . However the 

production of multiple avalanches results in an increased multiplication factor 

which can produce 10 9 to 10 10 ion pairs. 

Keypoint: 

Single event multiplication factor: 10 6 to 10 8 

Multiple event multiplication factor: 10 9 to 10 10 



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 2–4 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. 



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 



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. 



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 



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 G–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. 



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 G–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. 



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 300–400 

KeV energy levels. Above these levels thick walled designs are used, and 

beneath these levels the direct gas ionisation effect is predominant. 



Schematic of a Geiger counter using an "end window" tube for lowpenetrating 

radiation. A loudspeaker is also used for indication 



Pancake G–M tube, the circular concentric anode can clearly be seen. 



A selection of thick walled G–M tubes for gamma detection. The largest has 

an energy compensation ring; the others are not energy compensated 



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. 

10 

5 B + n → 7 3 Li + 4 2 α + 2e- 

3 

2 He + n → 1 1 H+ + 3 1 H + e- 

3 

2 He + n → 1 1 H + 3 1 H+ + e - 



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 10 8 events). For these reasons, the halogen-filled tube is now the 

most common. 



Geiger plateau 

The Geiger plateau is the voltage range in which the GM tube operates in its 

correct mode. If a G–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. 



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 



Quenching and dead time 

The ideal G–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. 



The disadvantage of quenching is that for a short time after a discharge pulse 

has occurred (the so-called dead time, which is typically 50–100 

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 G–M 

tube to a count rate of between 10 4 to 10 5 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. 



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. 



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. 



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. 



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 G–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 Geiger–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. 



Detection efficiency 

The efficiency of detection of a G–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/cm 2 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. 



Energy compensation 

If a G–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 G–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 2–3 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. 



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. 



Comparative response curves for GM tube with and without radiation energy 

compensation 



Thin-walled glass G–M tube showing a spiral wire cathode. The tape bands 

are for fixing compensating rings 









Thin-walled glass G–M tube with energy compensating rings fitted. The 

complete assembly fits into the aluminium housing. 

energy compensating rings- Absorb the 

low energy photon & enhanced high 

energy photon by production of secondary 

electron at ring internal 









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 Geiger– 

Müller 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. 


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

Plot of variation of ion pair current against applied voltage for a wire cylinder 

gaseous radiation detector. 





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. 



What is P10: 

Proportional counter with 90% Argon (inert gas) with 10% quenched gas 

(10% methane) 



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 Geiger–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. 



Plot of variation of ion pair current against applied voltage for a wire cylinder 

gaseous radiation detector. 



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 – creation of number 

ion pairs proportional to incident radiation energy. 

2. Avalanche region: in the immediate vicinity of the anode – 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. 



The generation of discrete Townsend avalanches in a proportional counter. 



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

region. 



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. 


https://www.orau.org/PTP/collection/proportional%20counters/Proportionalcounters.htm

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. 


https://www.orau.org/PTP/collection/proportional%20counters/bf3rclmk2mod201l.htm

Proportional Counter 



Proportional Counter 



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. 



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. 



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. 



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. 



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


Biological Half-life 


http://hyperphysics.phy-astr.gsu.edu/hbase/Nuclear/biohalf.html

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 from: 

1/T Effective = 1/T Physical + 1/T Biological 



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 

3 H 

4.5 x 10 3 

12 

12 

22 

P 

14.3 

1155 

14.1 

90 Sr 

1.1 x 10 4 

1.8 x 10 4 

6.8 x 10 3 

99m 

Tc 

0.25 

1 

0.20 



Tritium, 3 H, has a fairly long physical half life but clears from the body quickly, 

lessening the exposure. Phosphorous, 32 P, 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, 90 Sr, 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, 99m Tc, 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. 



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. 


https://en.wikipedia.org/wiki/Biological_half-life


More Reading 

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


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





Good Luck!

Understanding Neutron Radiography Post Exam Reading VIII-Part 1 of 2A

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