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Cadmium Zinc Telluride (CZT)
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Gerhard Martens Thank you so much!
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Hi,
My question is how to determine the efficiency of a "NaI(TI) Detector" its size is 3 × 3.
Radiation detection is done at different distances for various amount of energy. The results calculated from this method is similar with previous work.
But can't find the next step to calculate its efficacy.
Please help me
Regards,
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For an experimental efficiency calculation you need to rely on calibrated reference samples of which the activity is known and certified.
With such reference samples you perform a measurement with your detector and compare the measured activity (taking into account the geometry of your system) with the certified activity of the samples (compensated by the half life of the isotope from the date the samples were certified).
In this way you get the efficiency for a given energy line, but because the efficiency is energy dependent, you should repeat the measurement for several energy lines until you calibrate the whole range of your detector
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I am aware CR-39 (Columbia Resin 39) is mostly used in optical lenses, but I want to use them as ion track detectors. Hence, I need to adquire film-like or sheets of this plastic detector. I have no idea where to buy them! Any information is highly appretiated.
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I am now running a small production line manufacturing various types of CR-39 or PADCs - mostly to custom spec or buy in larger bulk from main manufactures. Worked for TASL for many years and still involved with the track etch technology.
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The website and manual are too general and do not explain well. I have the radiation data from sensors. Have to make a quality check. How can I use the BSRN toolbox to create station to archive file? The report and website are too naive' to explain.
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did you get a solution Sir ?
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Pl see the attached image. I want to calculate the detection limit using the following formula LD = 2.71 +3.29 (BACKGROUND COUNTS)^0.5
In 200 keV region there is a visible peak (assume >LC). In 300 keV region there is no visible peak (assume <LC).
What value for 'BACKGROUND COUNTS' should I use in both cases? In 200 keV energy, should I use gross counts as 'BACKGROUND COUNTS' i.e. a+b and what about in the case of 300 keV case? Pl reply.
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The attached image is of a Cs-137 spectrum with NaI.
Is this an example for any spectrum?
Are you seeking an estimate of MDA for reporting purposes, to state that nothing was found in the region above the detection limit?
A detection limit should be for a given radionuclide for a given analysis matrix. The actual background is determined from the matrix and its sources of noise in the peak region.
A rough background for a non-specific photo peak would be the total counts in the region of interest.
The equation you are using for the MDA is inadequate at few total counts.
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n+ contact of p type Ge is ~700µm thick compared to p+ contact of n type Ge only ~0.3µm. The process of adding the contact is different. n+ contact Li is added by diffusion process but p+ contact B is added by ion implantation technique. In diffusion we can get only thick contact but in ion implantation we can achieve thin contact, Fine. With a thin (~0.3µm) Li ion implantation, low energy efficiency of p type HPGe detector can be on par with n type HPGe detector (as per the attached fig).
Why Li is also not added as contact by ion implantation method? Is it a mandatory requirement of having thick (~700µm) n+ contact or only due to technology used for doping, we are getting thick (~700µm) contact?
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Hi, is it possible to test HPGe detector at room temperature, what is the resistance or diode test results should be. Thanks
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Hello
there is a p-type of HpGe detectors, this kind is characterized by the litium dead layer wich existe in the outer side of the detector cristal and it is increased with the passage of time ( dead layer= 0,7 mm if the detector is new ).
On the other hand, the n-ype of these detectors is characterized by a very thin dead layer ( in order of 10E-4 mm ) in the outer side and gross daed layer in the detector cavity.
Flowing our monte simulation of the n-type HpGE using MCNP gives a clair contrast with the expiremental results. where :
MCNP effeciency /Experemental effeciency <1 .
The insertion of the a dead layer ( dead layer depth= 0,08 mm) in the simulation improve the results especially in the <100 keV energy range.
My quation is, do what i did is correct ? and this value of the dead layer is reasonable after 20 years of functioning ? especially that in the leterature, the study of the dead layer of the n-type detector is rare and it is limited for the dead layer existing in the inere part (the cavity).
Regards
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Please share me the best answer might you get...
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In CT scan variable intensity X Ray is passed through the object and an 3D image is recorded (Computed Tomography). Brain is covered by skull, made of high Z of Calcium. Inside skull brain is made up of low Z of C, H, O. How X ray is passing through skull and carries necessary information for image of brain and coming out from the opposite side, where again skull is present? Confused.
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Gerhard Martens detailed really well how the attenuation works along a line, resulting in a single projection (you may consider this a classical X-ray image), a bit more on how the image is formed from this:
The basic idea behind the CT is that the set of projections (X-ray images from various angles around the head) combined together yields the Radon-transform of the µ(x,y) image (the map of Z-density).
This 'set of projections' (called a sinogram -not the examination of the sinuses, in this case the name comes from the fact that the projections of a single 2D point yield a shape of a sine (or cosine) function) essentially means, that a series of line integrals of µ(x,y) are recorded by rotating the X-ray source and the detector around the object (head), at least for each rotation angle from the [0;pi] range (the density of sampling may be varied, i.e. steps of 1°, or higher may be used obviously with penalities on image quality).
From the resulting sinogram, the CT-image (µ(x,y)) is obtained by inversion. There are several approaches yielding the inverse Radon-transformation, the simplest is the filtered back-projection (FBP), modern CT-scanners generally use some iterative approach (e.g. a maximum likelihood -ML-EM- approach), some approaches use Legendre-polinomials for the inversion, etc...
The resulting CT-image is then scaled to reflect the Z-density in a conventional way, showing Hounsfield-units (HU) (defined by HUair=-1000, HUwater=0).
All in all, remember, that with high enough energy, the X-ray is not completely absorbed in the skull, only attenuated and by measuing the attenuation from all directions around the head, the Z-density can be reconstructed with the appropriate mathematical framework.
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I did ROS assay using H2-DCFDA but could not find the function of detecting 2 wavelengths ( ex: 485, em: 520) at the same time as single reading in microplate reader so I got readings separately at the above mentioned wavelengths. I am not familiar well with this wavelength thing so I don't know how to analyze my result as I did ROS assay first time and don't even have idea that how it looks as combined reading. Can anyone please help me how to analyze the data? Any formula to calculate ROS from these wavelength values?
I have 2 separate readings (480nm and 520nm) for ROS assay.
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Hi Saurabh,
Your question is fine! No such thing as a bad question!
If you don't have the option to read the fluorescence, you may be able to get some meaningful data out of an absorbance measurement. Typically, your best bet is to measure the absorbance in the *excitation region*, since you can be sure the dye/assay molecule absorbs light in this region. Briefly, fluorescence relies on the principle of an electron absorbing energy from the incident photon and entering an excited state. The emission we measure is a result of the relaxation of this excited state. If you can't detect the emitted photons (fluorescence), your best bet is to measure the absorbance, which should have a peak around the excitation wavelength. Then, assuming minimal interference, you should be able to prepare a standard curve of concentration vs. absorbance and be able to quantify some data thanks to the Beer-Lambert law.
Now I recommend doing some googling or other lit research on your specific assay--it's possible, even likely, someone has optimized an absorbance based protocol, perhaps using an alternate wavelength or adding a 2nd reading to correct for some background or other effect.
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Ruby (Al2O3 doped with Cr3+ ions), when exposed to ionizing radiation (for e.g. X-rays), emits luminescence. What is the mechanism that causes this 'Radio-luminescence? Does Cr3+ accept electrons or give up?
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@Raju: It may happen that we use such conversion for understanding but possibly no such experimental evidences are available in many cases. Even if conversion is there, it may be well below detection limit and negligible. See sizes of different states, they may find difficult to adjust in lattice etc.In CaSO4:Dy we assume Reduction as per Nambi Model, but experimental evidence is not there. There is no change in colour evan after huge irradiation for CaSO4:Dy, if reduction holds, colour should change. You may contact Dr. B. C. Bhatt, who is also on ResearchGate and has done pioneering work in these fields. I am also sharing the question with him.
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1. Can Compton Scattering like scattering happens for K shell electron like it happens for valence shell electron for X and Gamma Ray?
2. If yes, probability is more for K or Valence Shell electron? And more importantly why its high (either K or valence shell)?
Thank you.
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The energy of the gamma-quantum needed for Compton scattering is equal to the electron mass, which is approximately 0.5 MeV. The deepest K levels for very heavy atoms like Uranium is of order of 0.1 MeV, for lighter atoms it is smaller. Therefore, while considering the scattering the atomic potential is just a small correction. All electrons may be treated as free.
If there is a small correction, it is for the decrease of the cross section as the discrete electronic structure of the atom imposes only limitations.
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Compton scattering do occur for low Z elements.
Cross-section for PE is prop to Z^5, for CS it is z^1.CS happens for z^1, then it should happen for for z^5 also.
Binding energy of K shell e has smaller value for low Z elements.
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In the photoelectric effect, the photon transfers all the energy to the atomic electron. The atomic electron must be tightly bound so that the photon can transfer all the energy to the atomic electron (Head-to-head collision); otherwise Compton scattering occurs.  Hence the photoelectric effect occurs with tightly bound electrons especially K-shell. The Binding energy of the K-shell electrons, which are most tightly bound, decreases for low Z elements. So the photoelectric effect is reduced for low Z media.
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We want to buy an underwater detector for radiation protection purpose. Differentiating between natural (like radon and thoron) and artificial (like iodine and cesium) radio-nuclides is of more important in this project. We just want to buy one detector to do some test and get familiar with. After verification we plan to deploy a network in the sea. So currently I need no accessory or central server, I just need a stand alone device which I can communicate with a laptop.
 
I checked out AT6104DM(atomex), sara water(envinet) and Katerina. I contacted with the companies. Now for financial purpose I need to know price of them.
If any body has bought one of these devices or similar, kindly give me a clue about the price of them.
Also a training course by experts of company would be useful, Can you guess how does it cost?
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Qiang li: can you give us more details?
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How to estimate risk of stochastic effects mainly cancer for a given population exposed to low doses (5-10) mGy per year prevalent in High Background Radiation Areas without using the concept of collective dose as well as ICRP's accepted risk factor of ~5%/Sv?
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Different lifestyles in different populations would indeed be important.
ERR in each study is obtained after consideration of various lifestyle factors and other confounders, but the background incidence of disease varies among countries and also within the country.
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Dear experts, I would like to ask that I am perform calculate shooting a ion beam to a thin target (several cm) and I use a sphere + cosine card (F1). but where should I locate the center of the sphere? in the middle of the target or on the surface behind of target.
Further more, Some one guide me how to plot graphic and geometry?
Any advice.
Thank you very much.
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Hello;
I'am not really expert, but I'am a user of MCNP code and I dealt few years ago with F Tally card and I think that for symmetry reasons your sphere center should correspond with a middle of the target, in aim to obtain normalized number of particles / surfaces, elsewhere you should make some geometrical calculations to plot a accurate particle spectra.
for plot graphic and geometry, there are two ways:
- direct ay ith Vised editor: an graphical interface of MCNP5 (under windows and I think there an available release under other OS like linux...)
- if you are working with line command terminal, you have to use line command as:
> mcnp ip inputfile.name 
many options are available on MCNP user guide, it just depends which release are you using (MCNP5 or MCNPX)
Probably there will be more complete answers than mine, hope I could help a litte;
good luck;
Saladin
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Can anyone provide references to radiation measurement data collected on patients who have undergone various NM procedures?
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Search the NRC website (nrc.gov) or Google for the following articles;
NUREG CR6345 Radiation Dose Estimates for Radiopharmaceuticals
and
Regulatory Guide 8.39 - Release of Patients Administered Radioactive Materials
I think these will aid in what you are looking for.
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In the commercial available charge sensitive preamplifier  for semiconductor radiation detector.Generally the HV(High Voltage) bias given to the semiconductor detector  is  through the charge sensitive preamplifier. Why it is done like this?
Why HV bias is not given separately to the semiconductor detector?
I will be very much thankful for the reply.
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Dear Manoj,
If I understand your question correctly, you are referring to the SHV input into a commercial preamplifier commonly designated "bias" with BNC output commonly designated "energy". 
If you look carefully at the preamplifier schematic, the DC bias is supplied to the detector through one or more load resistors. The detector input signal is usually capacitively coupled to the preamplier, where the coupling capacitor is located between the DC bias input and the preamplifier. Hence, the bias voltage is actually not being applied through the preamplifier circuit. The coupling capacitor serves as a filter to block DC current while passing AC signals from the detector to the preamplifier.
You can fashion a preamplifier of your own and apply the bias separately, but  a commercial preamplifier offers a convenient method of both biasing the detector while coupling the detector to the preamplifier.
Sincerely,
Douglas
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Dear All,
I have tried several plastic tubes for closed loop radon measurements but I found that the radon can penetrate trough so many of them. It causes significant radon loss if the sampling is continuous.
Do you have any experience in case of different plastic tubes? Which one is the best?
Many thanks
Zoli
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I've had a similar issue, and after some testing I found that pore-free, gasproof FEP-tubes work really well. I've experienced no measurable radon loss during continuous testing of several hours. These tubes are also quite flexible and not too expensive.
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In the radiation detector the pulse shaping circuit consists of CR-RC circuit and amplifier.The function of pulse shaping amplifier is to improve Signal to Noise ratio.It also act as a filter to remove the noise since it consists of CR-RC circuit which represent differentiator and Integreator . The preamplifier input given to the pulse shaping circuit generates the  Gaussian Pulse as the output response. I have seen that the tailing edge of the Gaussian pulse consists undershoot. There is a need for baseline restore through Pole-Zero circuit.Why this undershoot occur in the output of pulse shaping amplifier?How does Pole-Zero adjustment removes this undershoot?
I will be very much thankful for the reply.
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Dear Manoj,
Welcome,
It is so that the differentiator is effective while the integrator is not since the signal appeared as a differentiated pulse, please see the given link below.
The c-r differentiator has the response Fd(S)=  r 1sc1/ 1+sc1r1= sc1r 1for 1>>sc1r1> That is your differentiator has wcut off > wmin of the pulse. With wcutoff= 1/c1r1, While your integrator Fi(S)= 1/ 1 + sc2r2,
To be effective sc2r2 must be much greater than 1 and consequently, Fi(S)= 1/ sc2r2 which means that that  wmax>> w cutoff. where wctoff= 1/c2r2 This means that the filters must be adjusted to pass the pulse with slight shaping from wmin to wmax.
Multiplying the two filters to get the overall transfer characteristics one obtains:
Fi Fd approximately =  sc1r 1/ sc2r2 = c1r1/c2r2 which is the condition of compensation. since the pulse in this case will preserve its shape. while removing out of band noise and interference.
This may be a scenario for the understanding and solving the problem.
Best wishes
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Generally there are three types of noise is present in the semiconductor radiation detector.These are Thermal Noise,Shot noise and Flicker Noise.The thermal noise is removed by operating the detector under low temperature.The Flicker noise can be removed by using High pass circuit.But How to eliminate the Shot noise in the Semiconductor radiation detector?The shot noise in the semiconductor radiation detector is due to the discrete flow of carrier in the device. I will be very much thankful for the help if any one suggest some method to eliminate the shot noise?
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I don't believe you can eliminate shot noise, but if you can increase quantum efficiency it will improve the signal to noise ratio - other things being equal.
In the limit where you generate one carrier for each incident photon, you will be limited by photon arrival statistics - generally described by a similar Poisson distribution to shot noise.
If you generate multiple carriers from each incident photon, there is additional noise to that defined by the photon arrival rate.
If there is a bias current which is independent of incident radiation, this will also contribute shot noise.  Cooling the detector should reduce this.
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I use a NaI detector which is powered by a digibase (integrated bias supply, preamplifier and MCA) and I find that its linearity is not quite good .( 5-6%).
I'm measuring radioactivity in granites and If I use 609  and 2614 Kev for linear calibration E=a*Ch +b , I get a difference 80KeV in the 1461KeV (Im interested in linear and not second order calibration)
If anyone uses such a system I would appreciate if he could send me information about the linearity of his system
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Dear Mira,
The non-linearity that you are observing is most likely an intrinsic property of NaI:Tl. Measurements have shown that the light yield is higher per unit energy deposited as the total energy decreases. These measurements are commonly normalized near 450 keV, and show a difference in relative light yield as high as 20% as the absorbed energy approaches 20 keV. I suggest that you read the following:
W. Megesha et al.,  IEEE Trans. Nucl. Sci., 45 (1998) p. 456.
L. Swiderski, et al., IEEE Trans. Nucl. Sci., 56 (2009) p. 934.
S.A. Payne et al., IEEE Trans. Nucl. Sci., 58 (2011) p. 3392.
Some explanations are offered for this nonlinear behavior here:
J.E. Jaffe et al., Nucl. Instrum. Meth., A570 (2007) p. 72.
This nonlinear behavior is also part of the reason that energy resolution is typically limited to 6% FWHM or greater (662 keV) for common NaI:Tl detectors.
Because nonlinear behavior is expected from NaI:Tl, I suggest that you compare your data with that of the aforementioned publications to determine if that is the problem. If so, then you might be able to simply match your data to your known measured energies and develop an empirical correction with curve fitting software, as our colleague Dashty suggested.
I hope that this helps.
Sincerely,
Douglas
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In the reliability analysis of repairable and redundant safety systems someone needs to consider the effect of maintenance program. We are developing a Markov model for the ECCS system of a typical PWR reactor and for transition rates calculation we need the typical values of test interval and test duration for the ECCS system of a PWR nuclear reactor.
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In the USA, Regulatory Guides (RGs) 1.79 and 1.79.1 relate to the preoperational and startup testing of emergency core cooling systems (ECCS) for Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). These regulatory guides identify the ECCS functions that are to be tested as those necessary to ensure the specified design functions of the ECCS are met during conditions of normal operation, anticipated operating occurrences and postulated accident conditions.
In Japan, METI (Ministry of Economy, Trade and Industry) periodic inspections are conducted every 13 months in accordance with the EUIL (Electric Utility Industry Law) for light water reactors. These inspections include ISI (in-service inspection), system performance tests, containment leakage rate tests, and overhauls of mechanical equipment. In addition, many voluntary maintenance activities, most of which are periodic overhauls for mechanical equipment, are planned and conducted by electric utilities during the plant refueling outages.
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  • To determine the neck thickness I need to calculate the forward range of electrons which might be produced by the Ion beam interactions.Since the range of Ion beam itself is in Microns,Is it safe to assume that if I keep my neck thickness around 2-3mm the electrons emitted in the forward direction won't leave the cup surface?.
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OK, I got your situation.
The electron range get shorter when the electron energy get lower.
You can check the homepage below.
Then the stopping range of 300-eV electron is much shorter than 10-keV electron (0.5 micro-m in copper).
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Is there any mathematical formula there for the ideality factor of the diodes?
I will be very much thankful for the above. Please send the mathematical expression for it.
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Dear Manoj,
I would like to somewhat elaborate the answer of the the colleagues Andres and Dreher. At first, there is analytical formula fro the ideality factor. You can find them by following the link:http://journals.tubitak.gov.tr/physics/issues/fiz-07-31-1/fiz-31-1-2-0609-3.pdf.
However, the ideality factor is best estimated from the dark i-v characteristics at the intermediate current range where the both Rsh and Rs are negligible. In this current range the the diode equation follows the equation:
I= I0 exp V/nVt, with the symbols have their usual meaning.
In this region one selects two points I1,V1 and I2,V2. and substite these value in the equation one gets;
V2-V1 = n ln I2/I1,
n =( V2-V1)/  ln I2/I1,
This may be the best approach to estimate a most probable value for the ideality factor n.
Best wishes
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I was measuring radon with AlphaE (product of Saphymo)
On the screen, the "NOISE" response appeared.
I tried to start another measurement but the "NOISE" was still on it during several days...
I could not find sufficient info in the instrument manual...
Many thanks for your help
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I fear Paula and coworkers mismatched the instrument's type, AlphaE is battery powered an with at least under that profile it seems to be very, robust.
As to "NOISE", I never saw that message, having used the instrument in several environments with high acustic noise, high humidity and so on.
A first question is: are you sure your instrument is measurring only radon or could be present other dispersed isotopes in air?
Another source of noise in the detector could be mechanical vibration, which could be or even not, induced by acustic noise since the instrument is very light.
The issue is enteresting, if I have news I'll write.
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Physics experts in gamma and beta radiation detection.
Medical Physics.
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You could measure the distribution of beta+ emitters (e.g. I-123 or Sr-83) with positron emmission tomography (PET) and re-calculate the absorbed dose for I-131 and Sr-90. This would even work in an actual patient, if you tracer your therapeutic substance with the PET nuclides. Or am I missing a point? 
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I performed the radiation dose calculations and radiological consequences of a hypothetical accident by HotSpot, How to calculate values of error bars shown in figures? How to calculate combined relative error?
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HotSpot is an excellent tool for preparing response to radiological accidents. In particular, its ability to predict instrument response is very useful for planning purposes. HotSpot is woefully inadequate for dose calculations as are all other air dispersion programs. Actual conditions for any accident are manifestly impossible to predict. Placing error bars or propagated relative error is not possible. Nevertheless, you can estimate reasonable upper and lower limits. Use your judgement, but do not pretend that you have some statistically justified error calculations.
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Hello people!!!
I am currently trying to understand the theory of cosmic radiations. I also want to know the magnitude of cosmic radiations in Germany for the last 6 years. Can any one suggest me a good source? Further, can you say me what is the unit for measurement of cosmic radiations?
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A large component of the radiation dose to the human population arises from cosmic radiation entering the Earths atmosphere (Grasty and La Marre, 2004), The Intensity of the radiation due to cosmic rays depends mainly on the thickness of the atmosphere above the measurements location, and therefore mainly on altitude above sea level. Changes in barometric pressure and temperature and associated differences in atmospheric attenuation also cause small fluctuations of short term nature.
The annual effective dose rate due to the outdoor cosmic effect was calculated from a digital terrain model averaging topographic height within a cell size of 0.829 Km x 0.890 for the studied region using the following
Cosmic (mSv/a)= 37*exp (0.38*Altitude (km))* 0.877)/100
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What are the formulas for calculating the activity and activity concentrations for Cs- 137?
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Dear Nelson. The content of the attachment could help you.
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The characteristics graph of Geiger Muller Counter always keeps going up and does not drop down . It may remain constant over an interval but does not drop down on the graph scale. Why it does not come down after going up?
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It is not clear what characteristic of GM counter Mr Shahid refers. If it is the graph between the charge collected by the anode wire versus the applied voltage for an ionizing event, it is a steadily rising curve. In the GM region the gas amplification is very large: 10 to 1010for a single ionizing particle of any kind (alpha, beta or gamma). That is why the GM counter, unlike an ion chamber or proportional counter, cannot be used to identify the ionizing particle. What really happens here is the primary avalanche initiates further avalanches due to ionization and excitation of the atoms of the gas, producing UV photons which are also ionizing. The +ve ions being heavy are initially localised at one point of the anode wire.  Ultimately due to secondary electron production due to photoionization by UV, the whole anode will be covered by a sheath of +ve ions which results in the reduction of electric field, thus terminating the discharge. The +ve ions then drift toward the cathode releasing electrons from the cathode wall,etc and the discharge will become self-perpetuating. To arrest this recycling, quenching agents are used. Dr Pekko has elaborated on these aspects well.
The other characteristic plotted for the GM counter is the relation between count rate and applied voltage for a given radioactive source placed below the counter, Here, there is a slow rise, followed by a plateau where the count rate remains fairly constant over a range of voltages. Further increase of voltage results in the production of multiple counts. 
I can discuss quenching agents and other concepts like dead time, recovery time, etc but these are available from standard texts (see e.g., Nuclear Radiation Detection by Wlliam J Price or Radiation Detection and Measurement by Glenn F.Knoll).
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Dear colleagues. One contaminated herself with final [18F]FDG product. Hand-Foot monitor with plastic scintillator coated with ZnS shows 7kcps alpha contamination. I have doubts that there was alpha particles. Detector is covered with thick plastic wrap foil and 18F should be pure. It was obtained in 18O(p,n)18F reaction after purification on several cartriges. Is that alpha detection a false positive result? Is it possible that monitors pulse discriminator or high voltage on PMT is badly configured ?
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In general, using pulse shape discrimination (PSD) , a pile up of 2 beta events can mimick an alpha event. Especially if you have high count rates. There have been attempts to tackle the problem eg with neuronal networks (see link), but as far as I know the common PSD uses pulse width as discrimination parameter, so no matter how you configure your PSD, high beta count rates will generate false alpha signals.
For practical purpose, you can try to verify the effect measuring a dilution series of an F-18 compound. If you reduce activity to the half, pile ups should be reduced to about 1/4.
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Dear engineer Why win-q disconnected to quantulus 1220?.what can i do to solve this problem.I turn on and off system and counter but win-q display disconnected massage in application.I reinstalled app. But win-q is disconnect .at first its error was conveyor clearing but after turn off & on it shows disconnected. we check the plates and remove trays and check them.they dont have any problem. please answer my question. I need to answer .best regard((when counter time was over the elevator took vial at tray and after that we hearing the pneumatic pomp tone(Eu-152,STD capsule) more and more.after that we turn on & off several times.but after that application showed disconnected))
best regard.
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برای کاهش دماتا این اندازه ای که مد نظرتون باید با استفاده از دو لوله مسی که در پشت قسمت خنک کننده تعبیه شده جریان آب برقرار کنید چون همنطور ک می دوندی قسمت خنک کننده فقط از جریان هما استفاده می کنه و رسیدن به دمای 5 با برقراری فقط جریان هوا فکر نمی کنم ممکن باشه.
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Metals like copper or silver are never known to emit UV at room temperature. It is now possible by gamma irradiation of metal.
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Dr. Sankaran Ananthanarayanan,  1. YOU ARE POOR IN BASIC RADIATION PHYSICS. 2. YOUR ANSWER LACKS CLARITY AND VERY VAGUE.  I am very sure about my experimental discovery of UV dominant optical emission following Cu or Ag X-rays from within the same excited metal atoms of Variable Energy X-ray Source, AMC 2084, U.K. (Braz. J. Phy., 40, no 1, 38-46,2010)
http://www.scielo.br/scielo.php?script=sci_arttext&pid=S0103-97332010000100007).  I have already  explained with unprecedented detail how the optical emission takes place in radioisotopes and XRF sources in the same paper.   
FIRST AND BEST REVIEW OF THE RESEARCH PAPER
Margaret West,*a Andrew T. Ellis,b Philip J. Potts,d Christina Streli,c Christine Vanhoof,e Dariusz Wegrzynekf and Peter Wobrauschekc, Atomic spectrometry update-X-ray fluorescence spectrometry, J. Anal. At. Spectrom., 2011, 26, 1919.
DOI: 10.1039/c1ja90038b http://pubs.rsc.org | doi:10.1039/C1JA90038B
Refer under the title: 2.3 Spectrum analysis, matrix correction and calibration procedures
Words of citation: The phenomenon of optical emission predominantly in the UV, which accompanies the emission of X-rays, gamma rays, and beta radiation from radioisotope sources and X-ray tubes was investigated by Rao. It was the first work in which the emission of UV radiation was confirmed experimentally and a possible explanation for the mechanism of the UV emission was given by the author.  https://www.researchgate.net/publication/273124068_24_FIRST_AND_BEST_REVIEW_OF_THE_RESEARCH_PAPER_publis
The reviewers already cited my plausible explanation for UV emission following X-rays. PLEASE NOTE IN MY QUESTION, I DID NOT ASK FOR ALTERNATE EXPLANATION, YET YOU TRIED TO PROVIDE ONE. 
I never came across any reference on what you said, 'photo and Compton electrons exciting valence electron to emit UV'.  And you have not cited any reference, when you said, "(photo and Compton) electrons which, in turn, in favourable cases, can excite the valence electrons to emit UV or visible light.In a particular metal, these are decided by transitions of energy levels allowed  by quantum mechanical rules.  
When I mentioned my work with Cu or Ag X-rays from AMC2084,UK, you provided your own explanation  on gamma rays. Your comment is totally irrelevant and very vague. 
You said, "The line emission spectrum of various elements are available from reference data published by National Standard Laboratories".  What you quoted is optical spectrum of metals at high temperatures.   What I have reported is UV dominant  optical emission from Cu, Ag, and Mo XRF sources and 57-Co  notably at room temperature. Please note radioisotopes and XRF sources emit a new class of atomic spectra of  solids (radioisotopes and  XRF sources) unprecedented at room temperature. If I understand correctly, you are trying to downplay the work done and confuse the readers. In this process, your ignorance of the subject  is exposed.
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Durridge Inc. RAd7 monitor measures radon concentration from alpha energy spectrum.
Removing inlet filter should in principle allow to measure at least time variations in equlibrium factor, since a fraction of radon daughters is expected to be trapped in the drying column.
Anyone who tried or found literature about this? Thank you in advance for any suggestion
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The RAD7 measures Rn-222. It measures Rn-222 indirectly by measuring Po-218 after sufficient equilibrium with Po-218 as Yasser Ebaid stated. It measures other progeny as well as Po-218 and can even be used to measure Rn-220. 
The RAD7 could be used to measure the disequilibrium of Rn-222 progeny if there were no filtering and no inlet restrictions. Such a configuration would require calibration for the specific conditions (moisture, dust loading, collection time, ...). Such a configuration is not recommended as the detector and chamber will degrade and and require cleaning and recalibration, often.
The RAD7 measures Rn-222 by measuring Rn-222 progeny, but it is not designed for quantification of Rn-222 progeny.
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I am doing experiments where I am feeding Arabidopsis plants with radioactive phosphate (33P) to measure rate of Pi uptake. We will measure the uptake by scintillation counter which will give the data in counts per minute. How to convert counts per minute data into nanomoles phosphate per mg of plant tissues? Please let me know. Thanks.
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The company has given you the the molar concentration of P-33.
Usually they provide information as P-33 in X molar H3PO4. You need the molar concentration of H3PO4. The stable H3PO4 is phosphate concentration you need. When you calculate the uptake it will be the stable phosphate. Typically the supplier describes the product as
Phosphorus-33 Radionuclide, 25mCi (925MBq), Specific Activity: 8500-9120Ci (314-337TBq)/mMole, Orthophosphoric Acid in Water, Concentration:>500mCi/mL
You want the specific activity as X Ci/mMole.
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Radiation detection, gamma spectroscopy
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Just let you know that Ac-227 calibrated solution is available.
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I am conducting a radiation dose exposure and risk assessment  on native peoples in the USA and several other countries. I am interested to learn if anyone has been reporting low dose results and in particular in relation to indigenous peoples.
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From what I can tell, you would be interested in studies in India (Kerala) [1] and China (Yangjiang) [2] on (what are more or less indigenous) populations with large background radiation (>70mSv/year) that have been done, with no effects seen on cancer or mortality, but they did detect some chromosomal aberration. Also, 1 million people undergo radiation therapy each year, without a statistical increase in secondary (caused by the radiation as opposed to the one they were trying to cure) cancers. And it has been shown that low dose radiation increase protective functions [3]. However, you might want to look for studies on the Navajo Uranium miners.[4] Lots of lung caners, but this was really due to inadequate precautions taken for airborne alpha radiation, which is well known to cause lung cancer, and is actually likely the main reason smokers develop lung cancer from the Polonium in tobacco [5], which of course also disproportionately effects native populations, who are known to smoke more.
I am convinced that the Linear No-Threshold low-dose has been pretty well debunked for cancer and early mortality (see review below)[6], and in fact several studies have shown benefits (hormetic) to low-dose exposure (10 mGy), with no ill effects below 100 mGy. Hiroshima bomb survivors also saw no effects below 150 mSv and other studies have shown no increase in cancer or mortality below 200 mSv. Chernobyl data showed an increase in thyroid cancer, but not below 200 mSv as well, however, children are much more susceptible, and you can get an increase in leukemia from iodine doses that would not affect adults. Airplane crews get a substantial increase in dose and see no ill effects for their dose levels (greater than 50 mSv/year).
Good luck, hope this helps!
[1] Population study in the high natural background radiation area in Kerala, India. Nair MK, Nambi KS, Amma NS, Gangadharan P, Jayalekshmi P, Jayadevan S, Cherian V, Reghuram KN, Radiat Res. 1999 Dec; 152(6 Suppl):S145-8
[2] Cancer mortality in the high background radiation areas of Yangjiang, China during the period between 1979 and 1995. Tao Z, Zha Y, Akiba S, Sun Q, Zou J, Li J, Liu Y, Kato H, Sugahara T, Wei L J Radiat Res. 2000 Oct; 41 Suppl():31-41
Effect of high-level natural radiation on chromosomes of residents in southern China. Hayata I, Wang C, Zhang W, Chen D, Minamihisamatsu M, Morishima H, Wei L, Sugahara T Cytogenet Genome Res. 2004; 104(1-4):237-9
[3] (linked below) Doss, Mohan. “Low Dose Radiation Adaptive Protection to Control Neurodegenerative Diseases.” Dose-Response 12.2 (2014): 277–287. PMC. Web. 2 Mar. 2016. http://www.ncbi.nlm.nih.gov/pmc/articles/PMC4036399/
[4] Brugge, Doug, and Rob Goble. “The History of Uranium Mining and the Navajo People.” American Journal of Public Health 92.9 (2002): 1410–1419. Print.http://www.ncbi.nlm.nih.gov/pmc/articles/PMC3222290/
[5] Cigarette Smoke Radioactivity and Lung Cancer Risk
Hrayr S. Karagueuzian, Celia White, James Sayre,and Amos Norman, Nicotine Tob Res (2012) 14 (1): 79-90. doi: 10.1093/ntr/ntr145  http://ntr.oxfordjournals.org/content/14/1/79.abstract
[6] (linked below) Tubiana, Maurice et al. “The Linear No-Threshold Relationship Is Inconsistent with Radiation Biologic and Experimental Data.” Radiology 251.1 (2009): 13–22. PMC. Web. 2 Mar. 2016. http://www.ncbi.nlm.nih.gov/pmc/articles/PMC2663584/
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I am looking for collaboration on natural radioactivity measurement and environmental impact
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about the measurement of activity concentrations of primordial (238U, 232Th, 226Ra and 40K) and anthropogenic (137Cs) radionuclide's and gamma dose rate in environmental samples using HPGe and NaI (Tl) dedector.
Natural radioactivity in soil is mainly due to 238U, 40K, 232Th and 226Ra, which causes external and internal radiological hazards due to emission of gamma rays and inhalation of radon ant its daughters (UNSCEAR, 1988). Measurement of external gamma dose due to terrestrial sources is necessary not only due to its contributions to the collective dose but also due to variations of the individual dose related to the pathway. These doses strongly depends on the concentrations of 238U, 232Th, their progenies and 40K, presents in rocks and soil, which in turns depends upon the geology of the regions
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How will it make a difference if gamma spectroscopy is  carried out using powder rather than crystal. I think resolution will be poor but why I am not able to make it out.
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The answers by Hanno, Erik, Raymond, and Phillip are correct. The multiple surfaces of a powder cause multiple refections of the emitted light at the multiple surfaces. Note that a black powder can appear to be white, that is why a scratch on a black surface is particularly visible. Many single crystal detectors use a powder as the reflection medium. 
Bicron made shock resistant NaI detectors of various shapes by forming the detectors using granules of NaI. The gaps between the granules were filled with a plastic having the same index of refraction as NaI. The resolution and efficiency was very close to single-crystal NaI.
A scintillator powder must have a filler with matching index of refraction if used as more than a thin layer. 
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Particularly I am interested in threshold values for 226Ra, 232Th and 40K activities in nitrogen, phosphorus and potassium fertilizers.
Any input is highly appreciated.
Thank you
Rüdiger
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Dears
peace upon you
I am interested in this topic and I found some regulation and recommendation regarding this matter which I can summarized as follow:
§Canada
-- The Canadian Soil Quality Guidelines for Uranium set a guideline value for uranium in soil for agricultural areas of  23ppm  (mg/kg) (Canadian Council of Ministers of the Environment, 2007), (285.2 Bq/kg).
-- the recommended Canadian Soil Quality Guidelines for the protection of environmental and human health are 23 mg/kg for agricultural land use, 23 mg/kg for residential/parkland land use, 33 mg/kg for commercial land use, and 300 mg/kg for industrial land use.
§USA
-- According to the US Environmental Protection Agency (US EPA), the use of phosphogypsum for agricultural purposes is permitted as long as the Ra-226 concentration is less than 10 pCi/g (370 Bq/kg) (US EPA).
-  EPA Standards
     For Airborne Emissions of Radionuclides (40CFR61) specifies a limit of  20 pCi/m^2s (0.7 Bq/m^2s) from phosphogypsum stacks, and a limit on annual emissions of Po-210 of 2 Ci (0.07 TBq) from elemental
phosphorous plants.
- Uranium concentrations in phosphate ores found in the U.S. range from 20 - 300 parts per million (ppm) (or 7 - 100 picocuries per gram (pCi/g)). while Thorium occurs at essentially background levels, between 1 - 5 ppm (or about 0.1 - 0.6 pCi/g)(EPA).
 §European Union
 NORM processing and disposal falls under controls if radioactivity levels exceed 1kBq/kg.
 §UNSCEAR
 - The maximum value of Ra(eq) recommended internationally for building materials is 370 Bq/ kg (UNSCEAR, 1982)
§CHINA
-Ra-226 content in phosphate fertilizer and its compound fertilizer shall not be higher than 500 Bq kg-1 (GB 8921-2011, 2011)
 
§SOUTH AUSTRALIA
-In South Australia the regulatory limit is set by the Environmental Protection Agency (EPA) and is equivalent to 200 ppm uranium. When material is over this 200 ppm limit (2480 Bq/kg), regulatory controls require management strategies such as blending material to below this level to ensure worker and community safety at all times.
 
According to the regulations for the Safe Transport of Radioactive Material, Safety Requirements No.TS-R-1, International Atomic Energy Agency (IAEA), Vienna, 1996 Edition (Revised in 2000)
and  Regulations for the Safe Transport of Radioactive Material, Safety Requirements No.TS-R-1,
International Atomic Energy Agency (IAEA), Vienna, 2005  Edition.
 •Material containing more than 10 Bq/g of Th-232 will be a subject to international transport regulations. If it is known that Ra-226 is in equilibrium with its parent U-238, the same 10 Bq/g activity concentration limit appears to be applicable. If, however, U-238 has been removed (or not present – as in oil and gas sludge), the limit for Ra-226 will be 100 Bq/g (assuming that an exemption from para 107(e) of the regulations is applicable to a particular material).
 •ICRP(17) suggests that for the control of public exposure an appropriate value for the dose constraint is 0.3 mSv in a year. In keeping with this suggestion the Canadian NORM guidelines have adopted 0.3 mSv/a as its first investigation level. Tables 5.1 and 5.2 list the amounts of radioactive materials that if released to the environment without further controls will not cause doses in excess of 0.3 mSv/y.
Alzahrany
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Why do electrons produce more scintillation light than heavy charged particles?
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Dear Prof. Ghal-Eh,
we can easily get around the range argument by assuming that the electrons or HCPs  are completely stopped in the scintillator. Then let us assume that the incident particles carry identical amounts of energy into the scintillator. Since they are eventually completely stopped they deposit identical amounts of energy into the scintillator. Fluorescence requires that electrons in the scintillating material (in your example PPO) get excited. For each excitation you need at least the excitation energy. On the average there is a certain amount of energy loss of the incident particle necessary to produce an excitation and a resulting photon. If this were all that can happen you just divide the kinetic energy of the incident particle by that average energy loss and you get the average number of photons per incident particle: identical numbers  whether it is an electron or a heavy charged particle.
Now, as I wrote,only a small fraction of the kinetic energy lost by particles in a scintillator is converted to photon production. Most of the energy lost is dissipated without radiation emission. So, electron excitation is not all that can happen. There are also collisions with kinetic energy transfer to atoms/molecules which produces heat.
The dissipation of energy by non-radiative processes  is much more likely for HCP than for electrons. HCP spend most of their energy to heat the material (in your example all three components of the liquid scintillator) and little is left for electronic excitation. Electrons can transfer energy into electronic exciation much more efficiently.
The physical picture behind this is: by the incident particle kinetic energy has to be transferred to a bound electron in the scintillating material to finally produce a photon. An incident electron can ultimately transfer all its energy to another electron in one collision (colliding particles of equal mass in a head-on collision). A heavy particle  can only transfer a fraction of its kinetic energy to an electron just because of kinematic reasons. In a head-on collision of two particles with masses m1 and m2 the (maximum) energy transfer is 4m1 m2 Ekin/ (m1+m2)^2. For incident electrons we have m1=m2=me and the full kinetic energy Ekin can be transferred. If we assume an incident alpha particle the maximum energy that can be transferred to an electron in one collision is approximately Ekin/2000. Hence much of the energy loss of HCP goes into collisions with similarly heavy constituents in the scintillator and these are the atoms that get kinetic energy which means the scintillator is heated. All these arguments are particularly valid at low kinetic energies where nuclear stopping of HCP prevails.
The same arguemnts are basically responsible for the size of atomic excitation cross sections. At kinetic energies of approximately two to  three times the threshold energy electrons have their maximum excitation cross section. At these (low) energies the excitation cross sections  of HCP can almost be negelcted compared to those of the electrons (smaller by many orders of magnitude).
I hope this explanation is sufficient to answer your original question.
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Nuclear track detector CR-39 is made of is a sample of a solid material, so if we irradiate this detector by He-Ne laser so what is the effect of that on track detector properties?
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Laser can affect the physico-chemical properties of CR-39, which one can measure by spectrophotometric techniques such as UV-Vis (where you can find band-gap from Tauc's plot), FTIR (to see the behaviour of functional groups). 
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I saw some pre-amplifier specifications and all of them say noise in terms of energy ( for example 2.8 keV for Si). If I convert this energy to charge (4eV per e-h pair in silicon) and convert the charge to average current taking an integration time of 10 us, it turns out to be around 11 pA. 
Does that mean the minimum signal current that the preamplifier detects and then integrates is 11 pA?
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Yes, nA currents are expected in pre-amplifiers.
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This number is very low but is not reported in literature. It depends on the system, obviously, but an order of magnitude in at least one case could be useful.
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It will depend on the dye concentration. But sorry,  don't know of any measurements either. Have you found an answer? If you are going to do some measurements yourself we might be interested in a collaboration!
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PPC cement, coarse aggregate ( mixture of 3/4 th inch & 1/2 inch downgraded stone chips), fine aggregate (coarse sand), and admixture (MasterPolyheed) is used for this construction. 
Fresh density is kept above 2.4 gm per cc. The hardened density requirement is above 2.35 gm per cc. 
Ice is being used to decrease the mixing water temperature. (To avoid hairline crack in the future)
Key concern is to avoid radiation leakage.
The rooms are being constructed in the basement with wall thickness of 5 to 8 feet and slab thickness of 4 to 8.5 feet. Rooms will be used for Oncology treatment and/or Tomotherapy. 
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In addition to Dr Newhauser's comments, here are some other considerations:
Use high density coarse aggregates, for eg, granite in place of limestone
High workability of concrete and sufficiently spaced reinforcement
Monitoring and controlling heat of hydration during casting and subsequently. Cement replacement materials like fly ash and silica fume are good choices as they not only decrease the rate of hydration, but have filler properties resulting in fewer pores in the matrix.
Sufficient shrinkage reinforcement
Substantial and sustained curing
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Delete please
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Another paper on this topic:
M. D. Tarasov et al, 'Efficiency of Radioluminescence of Water under the Action of Accelerated Electrons', Instruments and Experimental Techniques, 2007
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Hi,
We are looking at electrons, protons and ions generated from our LPP x-ray source which passes through an electrostatic analyzer before reaching the detector. Can anyone tell us, which is best to use, the electron multiplier tube or the micro channel plates?
Many thanks
Rad 
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EMT is relatively less expensive compared with MCP. if using MCP go for CSI deposition on the front surface as it helps for the X-ray photons. Fast counting is possible in both.
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I measured some water samples from Kabul-Afghanistan by means of ICP-MS to look for the concentrations of radionuclide, I measured considerable amount of Uranium-238 (Still under the WHO permissible limit), but as a very active element of Uranium series progeny, I couldn't see any counts for Ra-226,  I couldn't find reasonably good interpretation of the absence of Ra-226, any body have any good explanation for this absence ? I am a physicist not a chemist, so please enlighten me about the issue.
Thank you very much in advance
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Dear Mohammad,
do not mix the concentration (µg/L) with the activity (Bq/L). Of course, Uranium decays to Radium (mainly U-238 --> Ra-226 via several steps of alpha decay, but the absolute mass is quite different, assuming radiiactive equilibrium. Let us assume you have a concentration of 3 µg/L for Uranium typical for seawater (> 99 % is U-238) you have roughly an activity of 0.037 Bq U-238, which would finally result in a Ra-226 activity of about the same, assiuming no geochemical loss. This would be equivalent to a mass of  1.02 10E-12 g or 1.02 pg. In other words, the relation between the masses of these two isotopes would be a factor of 2.793 E-6. Consequently, you need to lower your detection limit by about a factor of 6 orders of magnitude.
In addition, you might get some geochemical reactions from the intermediate elements like Th-234 and Th-230, which could result in even lower concentration of the element radium.
I hope this example is easy to understand and I hope I understood you correctly.
Good luck!
Hartmut
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ًWhat is the effect of gamma source activity on HPGe detector efficincy
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High dead-time can cause broadening of the peaks. The amount of broadening is electronics dependent.
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Geiger counters work in environments with different kinds of radiation types.  How dose it measures radiation dose in sivert unit?
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Sievert is a unit of risk, not dose. A Sievert estimates risk of cancer as effective dose. Instruments are calibrated to Sieverts based on one of several definitions. These calibrations have little resemblance to actual exposure situations. Nevertheless, rules and practices are such that the measurements with an appropriate instrument will meet regulatory guidelines. Regulatory guidelines are set sufficiently low that there should be no concern about the actual risk, including instrument flaws. The suggested reference by Pedro Almendral more or less embraces this condition, but does not answer the question in the reference.
Compensated GM detectors do a better job than uncompensated GM detectors. Most detectors, in particular GM detectors, measure correctly only for the calibration conditions. Detectors can be constructed to and with effort used to measure actual dose conditions. In general the effort is huge and not worth the effort.
An uncompensated GM is one of the worst choices for knowing the approximate dose rate for a location. Some scintillation detectors are worse. It all depends on the energy mix and the direction of radiation.
I show the definitions of dose and the response of detectors to calibration conditions and simple environmental exposure.
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Measurements of Uranium radon and radium concentrations by using SSNTDs. 
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Thank you so much Dr.  Mushtaq Ahmed for your attachments.
Best Regards
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Is there any reference data available for reliability  and availability of GM based radiation survey meters which are used for routine survey purpose in industrial or medical institutions. 
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Thanks Lung-Kwang Pan for your views, but the facilities which really use only GM for practical applications, it may not be appropriate to use data of other monitors/detectors.
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Any informations or suggestions are welcome. 
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Dear Dr. Naima
If you want to detect neutrons (energy, number etc.) in some volume, you should add multifunctional detector with primitive scorers and appropriate filters in the module 'Geometry' and place it in logical volume of your body. Then derive it through Run and RunAcction. Also you can do it using macros. The best examples are  examples/extended/runAndEvent and /examples/extended/hadronic.
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Knoll's book says "An estimate can be made of the amount of inherent fluctuation by assuming that the formation of each charge carrier is a Poisson processes".
What is the basis of Poisson process?
Does this assumption holds true in every detector, if not why does it fail?
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the basis of the radiation detection by a detector is to deposit the charged particle's energy and convert the deposited energy to a large number of charge carriers.
for example, when the energy of a compton recoiled electron is 100keV and deposited in HpGe, it will produce about 30k e-h pairs. The number of 30k is determined by the the many collisions of the 100keV electron with the atoms of Ge in HpGe. If the number of collisons is very large and the probability of producing an e-h pair in each collision is very low, the number (here: 30k ) of total e-h produced by the 100keV would obey poisson statistics.
In fact, the production of the 30k e-h paris is not independant, i.e., some e-h paris are the "son " or "grandson" of others, then the 30k is not exactly the poisson distribution. Fano distribution give the right result.  Usually, for gaseous and semiconductor detectotor, fano factor is less than 1 and for  scintillator detection , fano factor is about 1.
So, the "poisson assumption" does not hold for gaseous and semiconductor detectors, and approximately hold for scintillator detector..
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Some researchers insist on the presence of a NIST standards in each Neutron Activation Analysis, saying: “ Certified Reference Material GBW 0xxxx (for example) cannot be used as standard. It can be only used as Reference Material for the estimation of precision and accuracy of results”.
Why? Is it really inevitable to use a NIST standards, if so what is the use in the IAEA standards - for example-?
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NAA is for elemental analysis, and calibration is therefore best carried out using weighed amounts of pure elements from any reliable supplier of analytical reagents.
For QC/QA it is recognized practice to demonstrate validity by analyzing a CRM with approximately the same composition as the samples you analyze. All these matters are clarified if you read ISO 17025 or any similar local standard used for accredtitation.
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High energy astrophysicsts  and Nuclear Physicists
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Rather than answering your question in frequency, it would be more realistic to answer in energy, simply the frequency is so high that gamma ray behave more like a particle (photon) than a wave. The CGRO and other satellite experiment can detect gamma ray only up to several 100s GeV or < 1TeV = 1.E12 eV. Higher than that, the flux is so low that satellite instruments loss detection power or their discrimination power to separate gamma from much higher flux of cosmic rays. The Ground gamma ray telescope can detect gamma photons interaction with atmosphere via indirect measurement. The highest energy of gamma ray  of those experiments can reach approximately 1.E14 eV, in terms of frequency ~ 2.4E28 Hertz.
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Using the method of Laplace Inverse Transform, proposed by B.R. Archer et al. (1982-1988), we can get an x-ray spectrum from the Transmission Data where we need to use the values of mass attenuation coefficients. Not only this method but other methods also involve the use of the mass attenuation coefficients for the beam attenuating materials and these values are very sensitive to get an accurate x-ray spectrum back. If we use the mass attenuation values from NIST (that includes scattering, photoelectric absorption and pair production), can we also use the same values in the backward calculation? By backward calculation I mean the calculation of the Transmission Data using the X-ray spectrum in the following way:
T1(x1)=F1(E1)*exp(-mu1(E1)*x1)+F2(E2)*exp(-mu2(E2)*x1)+...+Fn(En)*exp(-mun(En)*x1)
T2(x2)=F1(E1)*exp(-mu1(E1)*x2)+F2(E2)*exp(-mu2(E2)*x2)+...+Fn(En)*exp(-mun(En)*x2)
....
Tm(xm)=F1(E1)*exp(-mu1(E1)*xm)+F2(E2)*exp(-mu2(E2)*xm)+...+Fn(En)*exp(-mun(En)*xm)
Or any other factors related to the geometry of the x-ray machine (that contribute the secondary x-rays passing through the attenuators and/or detectors) also need to be considered?
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You can try using FFAST tabulated data (http://www.nist.gov/pml/data/ffast/) of x-ray mass attenuation coefficients, which is the most accurate tabulation (I found) compared with experiment. Experimental geometry is an issue, where x-rays are attenuated by background matters (in between the monitor and the detector), detector gas and detector windows, that can be modelled to remove their effects; the efficiency of a detector can be estimated using attenuation data measured using that particular detector. Of course, theoretical data of x-ray mass attenuation coefficients do not provide information about the oscillatory part (XAFS) around an absorption edge; it just provides the background comparable for regions away from the structural parts! 
 
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92.3 & 92.8 keV gamma peak shall appear as the single peak as the FWHM is 2-3 keV in HPGe detector.
If X ray of 93.350 keV contributes...what will be the % of contribution (% w.r.t gamma)?
Assuming no x ray cut off shield is used.
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I agree with Dr Kuzminov.
It is of significance that you mesure the efficiency of your HPGe D correctly and specifically at the energy range wanted. Try also MC as well if you do not have the experiemntal conditions and accuracy needed. 
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If the results of activity size distribution (AMD and GSD) depend on the ventilation rate, sampling location, and room size. why we make comparison in some papers? what is the value of measurment reating in some cases?
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The activity size distribution Z is important regarding the distribution of radon progeny in respect to the three modes that Porstendoefer first introduced: Nucleation mode, Accumulation, Coarse mode. Depending on Z, the,  what was traditionally called free and attached progeny, is altered. It is very difficult to measure Z because it necessitates advanced techniques, for example aerosol spectroscopy (~100 kEuro purchase) . Otherwise, other methods should be employed, such as discrimination through grids of various size values. Differentiation in Z, imposes alteration in distribution of all progeny nuclei. Monte-Carlo is a good method in that.
This constitutes a personal aspect. I hope that it assisted.
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as a phosphor for X-ray imaging applications 
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When incorporated into Gd2O2S (if this is possible based on the similarity between the ionic radii of the dopant and Gd3+), the ions you mention will be in the Cs+ and In3+ valence state. Then, these ions have a closed valence shell configuration. E.g. Cs+ has the Xe electronic configuration. Consequently, the first excited state - obtained by promoting an electron to an unoccupied shell, is too high in energy to give near-UV or visible emission. Similarly, In3+ has a closed 4d10 configuration.
When synthesized, these phosphors might give some luminescence, especially when defects are expected to be formed. E.g. 'forcing' Cs+ onto a Gd3+ lattice site will lead to charge compensating defects which can induce some luminescence, although often with low efficiency. 
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Leakage as I know is the ratio of activity lost through glue or rubber and so on>>
= radon activity pass through material with out Bg/total Activity with out Bg?
is this true?
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Leakage is through loss of air from the sealed test system. Evacuate the system and observe the in-leakage rate. If you have no substantial change over the time of the planned experiment you will have negligible radon leakage. Pressurize the system and observe pressure changes. If there are no vacuum or pressure losses, there will be no leakage losses at normal temperature and pressure.
Diffusion through the seals can be tested by placing the radon source in the chamber and observing the final air concentration of the system. A measurable difference from the expected value indicates the loss. With this method you calibrate the chamber. The chamber calibration is the basis for comparing all measurements. There is no need to account for loss from the chamber if it is calibrated and sealing is reproducible.
A note on calculation of the diffusion constant. Several authors have introduced the term back diffusion. The diffusion equation describes a stochastic process. Atoms go in all directions in and out of the material. The diffusion constant accounts for 'back diffusion.' Inclusion of the term back diffusion is incorrect.
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what is the important application of radon diffusion in irradiation polymer materials?
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Thank you,  I agree with Prof. Virk,  the radon problem is help human beings to determine the health risks due to radon doses. 
We use polymers to study the effects of alpha radiation or energy in the polymer material to help us in calculating that we want to calculate in our studies.
the same effects of radiation on the polymer may give us an evaluation to what will happen in the living cells such as lungs or stomach .
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Henry constant
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For measuring correctly you need a good calibration before you conduct the measurements.  You must be careful about the leakage of gas from the can or container in which you measure the concentration. 
It is also depends on the procedure of sample collection. 
There is a ratio between indoor radon and Radon releasing from water depending on the radium content in the water.
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radon primary standard source is very important in the field of radon measurments to calibtate devises and instruments
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First: What is a primary standard: See "International vocabulary of
metrology — Basic and general concepts and associated terms (VIM)" at BIPM website.
5.4 (6.4)
primary measurement standard
primary standard measurement standard established using a primary reference measurement procedure, or created as an artifact, chosen by convention
In case of Rn-222 you need a radon gas standard done by an absolute method. This would be for example 
J.L. Picolo, Nucl. Instr. and Meth. A 369 (1996) 452
R. Dersch, Appl. Radiat. Isot. 60 (2004) 387
See also for further information: Dersch, R., 1998.  Appl. Radiat. Isotopes, 49, pp.1171-1174.
PTB and METAS have implemented the absolute method.
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For long term fading study (2 years post irradiation) of a TL material we have seen about 50% decrease in TL signal for the samples irradiated with the doses up to 2 Gy but a 30% increase in TL signal for doses above 3 Gy – 25 Gy when compared with the results of 24 hours post irradiation. which mechanism can explain it?
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Here is a half-baked, preliminary thought for how "negative fading" can possibly take place. In several materials, there is an effect of "non-monotonic dose dependence". This means that up to a certain dose, the TL increases with the dose and at higher doses, the TL intensity decreases with the dose. A possible theoretical explanation has been given using numerical simulation by:
It is possible that negative fading can be considered as an effect opposite to dose dependence at high doses so that the thermal loss of trapped carriers may cause an increase in the emitted TL. Of course, this is merely a questionable hand-waving argument. Following appropriate simulations (or theoretical considerations) it may or may not turn out to be valid.
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Can anyone explain me how does the detector parameters (like capacitance, leakage current etc.) affect the choice of feedback capacitance in the design of a charge sensitive pre-amplifier for radiation detection?
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experimental or literature informations. thanks 
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thank you  very much for your kind answer but we are pretty expert in measuring radon-222 activity concentrations in different environmental matrices. what we are looking for is some informations about if a measurement of the radon partition coefficient has been ever made. thank you anyway
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Suppose the energy required to produce a single electron-hole pair in a semiconductor is given by 'ε' Then the number of electron-hole pairs produced by a particle depositing energy E in the material is E/ε. Let us say this quantity 'N'. Now what is the total amount of charge inside the material: is it (a) N*e or is it (b) 2*N*e ? (e= charge on an electron).
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Actually it is 2*N*e. Of course the signal you see depends on which charge you collect. If you only collect electrons you will see N*e, even though 2*N*e was produced.
An interesting example of this is the following.
While characterizing Silicon Photomultipliers (SiPM) I measured the photon detection efficiency (PDE). The canonical method, used by everyone till that moment, was to illuminate the photosensor with a monochromatic light source, inside an integrating sphere where a calibrated photodiode is installed. By measuring the photocurrent (after subtracting the dark noise current) and normalizing to the calibrated diode one gets the PDE. This is wrong, for two reasons.
1) With this procedure one includes also the cross-talk in the SiPM.
2) The overall photocurrent is due to electrons AND holes.
I employed a new method, i.e. by counting the single photons with the SiPM and, on top of a better control of the cross-talk, I got the expected result for the PDE (1/2 of what measured with the photocurrent).
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I have an ultraradiacTM-plus personal radiation meter (a type of GM tube), give a pre-settable alarms for both instantaneous rate and cumulative dose. I want to know if there is any research using this meter, or its practical field uses, if anybody uses it .
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For information and practice of the handling of the UltraRadiac/Intensimeter 28, please try our Android app: https://play.google.com/store/apps/details?id=com.RadiaDroid
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I need it to be relevant for the measurement of gamma radiation exposure.
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According to the IAEA, ionizing radiation has a direct action on the complex vital molecules (for example the DNA) within the cell by breaking the bonds between the atoms. Ionization in non-vital molecules (for example, water molecules) produces very active chemicals (free radicals) which attack vital molecular systems.The damage may change the coded information in the cell nucleus, disrupt the cell’s chemistry and function or physically rupture membranes some of which contain the digestive enzymes. Natural mechanisms are capable of identifying and repairing limited damage to improve the cell’s chance of survival.
However, incorrect or incomplete repairs are also a possibility: these may affect the cell’s longer term viability or performance.
At very high whole body doses in excess of 15 Gy, swelling (oedema) of the brain and generalized shock affecting the cardiovascular system leads to coma and death.
The range of doses associated with death from acute exposure of the blood system, the gastrointestinal tract and the central nervous system is based upon sparse human data, supplemented by knowledge of the dose-response relationship derived from animal experiments. No individual would be expected to die after receiving an acute whole body dose at or below 1 Gy unless the person was seriously ill before irradiation. In an exposed population of 100 people, about 5 individuals would probably die after receiving about 2 Gy and about 50 would die within 60 days of receiving a homogeneous whole body dose of 3.5 Gy. This is called the Lethal Dose,
LD50/60. When the correct treatment is provided by a specialized hospital, the survival rate improves and the LD50/60 increases to between 4 and 5 Gy. Most individuals would be expected to die after receiving an acute exposure to a whole body dose of between 6 and 10 Gy unless they receive treatment to prevent infection and bleeding. Above about 10 Gy death is most likely, even after attempts to stimulate the bone marrow or administration of a bone marrow transfusion from a compatible donor.
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Generally X-ray detectors have only energy information.
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Dear prashanth,
Any detector as such is not designed for giving out only timing or only energy type of signals. The fundamental signal that any detector gives out as a response to radiation interaction in its active volume is a momentary pulse of current. Based on the way in which you process this momentary current pulse the system as a whole either acts as a timing based ( used only for counting purpose) or as a energy signal (used for extracting energy spectrum). My suggestion would be to continue to use whatever detector you have. To get the timing information you can just change the front end electronics from charge sensitive (current integrator) pre-amplifier to a current sensitive(current to voltage converter) pre-amplifier.
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Potassium, uranium, thorium, and radium.
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If you want to conduct measurements at the well site you may need to consider regulations relating to potentially explosive atmospheres, and I don't know of a gamma spectrometry system that would be certified for use in such environments. If there are hydrocarbon residues in pipes etc this may be a consideration even away from the well head.
Not a problem if you take samples to a lab.
210Pb has a low intensity 46.5keV gamma ray, and would require HPGe spectrometry of thin samples, and a detector with good response at low energy. You can conduct alpha spectrometry for the 210Po daughter, but this requires considerable chemical extraction and questions about yield estimation (but, that may be bias because I specialise in gamma spec).
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Like other solid state scintillation detector viz NaI(Tl), CsI(Tl) etc detector why ZnS(Ag) detector can't be used for Gamma counting also? Why ZnS(Ag) is used specifically or can be used specifically for alpha counting only? Can we used NaI(Tl) detector with suitable modifications in the design as alpha counting?
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ZnS is a polycristalline phosphor/scintillator. Largest available grain size are some 40-50 um. Hence, one is directed to screen use only. NaI can be made in large bulk material. ZnS has among the highest intrinsic efficiencies available ( approx 20%) but has a low transparency to its own light. High extrinsic efficiency for gamma for polycrystalline (powder) phosphors can be obtained for instance by Gd2O2S:X (X=Tb, Eu,Sr).
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So far, X-ray source such as XRF source or X-ray tube used in hospital is known for X-ray emission alone. The question is whether X-rays are followed by UV from within the same excited atoms of X-ray source?
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Dear Scott Williams, Your view “UV radiation is still part of bremsstrahlung spectra” is untrue. There was no literature on experimental finding or theoretical prediction on UV emission from X-ray tube or XRF source, prior to my research work with XRF sources published from 1997 onwards. High energetic electron is needed for generating bremsstrahlung. As charge around proton is also very high, electron loses energy in keV while passing through Coulomb space resulting into bremsstrahlung with energy in keV. Significant charge around proton in Coulomb space does not allow electron to lose energy just at eV level and generate UV. Therefore, it is not a matter of UV absorption in target material.
By virtue of fine X-ray energy, XRF sources emit UV dominant atomic spectra that I have verified with narrow and optical filters. Though the previously unknown phenomenon of X-rays causing UV (Fig.6 of Braz.Jour.Phy, March 2010) is the same even with X-ray tube, I could not extend my study further to record UV spectrum from X-ray tube before my retirement in 1997. I hope our discussion will arouse interest in readers to study the exact nature of spectrum.
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Solid salt such as rubidium sulfate is known to emit atomic emission of light only when subjected to high temperatures. Recent study has unfolded that gamma irradiation of the salt can cause atomic emission of light notably at room temperature. What is the technique involved?
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On irradiation of Rb, Ba or Tb salt with gamma rays from 241Am, correspondingly Rb, Ba or Tb X-rays are emitted, as does happen in the case of Variable Energy X-ray Source, AMC2084,U.K.. The latest findings reveal that from within the same excited atoms two more generations of X-rays are produced. X-rays first generate Bharat radiation with energy higher than that of UV at eV level that in turn generates a new class of atomic spectrum of the salt regardless of temperature. The nature of atomic spectrum (percentage of UV, visible (VIS) and near infrared (NIR) radiation intensities in gross light intensity) thus produced depends upon the X-ray energy. In clear words the nature of spectrum does not depend upon the element whether it is Rb, Ba or Tb.
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So far radioisotopes are known for ionizing radiation emissions such as alpha, beta, gamma, and characteristic X-rays. The question is whether gamma, beta, and characteristic X-ray emissions in radioisotopes cause nonionizing radiation, light ?
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The recent publication in 2010 reported experimental discovery of UV dominant optical emission from radioisotopes (radiochemicals such as 137-Cs, and metal sources like 57-Co). Beta, gamma, and Characteristic X-ray emissions first cause Bharat Radiation (predicted), which in turn causes UV dominant optical emission by valence excitation from within the same excited atoms of radioisotopes by a previously unknown atomic phenomenon, now known as Padmanabha Rao effect.
In 2013, discovery of Bharat Radiation in 12.87 to 31 nm in solar spectrum was reported.
Reference:
M.A. Padmanabha Rao,
UV dominant optical emission newly detected from radioisotopes and XRF sources,
Braz. J. Phy., 40, no 1, 38-46,2010.
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