Science topic

Radiation Dosimetry - Science topic

The measurement and calculation of the absorbed dose in matter and tissue resulting from the exposure to indirect and direct ionizing radiation.
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Does anyone know whether it's possible to display the material number (or other characteristics) corresponding to a mesh point as a separate column when using the FMESH card?
Cheers
Carlo
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FMESH defines the mesh superimposed over the geometry, that means independent from the geometry definition, and so provides no access to the properties of the underlying geometry like the material number.
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In any OSL phosphor we require optical energy more that the thermal trap depth of that trap for optical stimulation. For example in case of Al2O3:C we require 2.6 eV photon to detrap the electron from the trap having 1.12 eV thermal trap depth. How are they related to each other?
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For a given trap, E(optical) is always > E(thermal), because of the Franck-Condon principle. As a result, transitions on a configurational coordinate diagram always take place vertically, meaning that the transition is much faster than the lattice relaxation time. Once ionized optically the defect’s lattice configuration relaxes to a new configurational coordinate via the emission of phonons. Thermal excitation, however, includes the phonon emission and lattice reconfiguration takes place simultaneously. Thus E(optical) = E(thermal) + E(phonons), with the latter term given by the Huang-Rhys factor.
If experimentally measured energies ( for example E(optical) using OSL, E(thermal) using TL) are either unphysically different or approximately the same, I would question whether the two methods are actually probing the same defect, and/or whether or not the E(optical) and E(thermal) values are correctly obtained from the data, before launching into detailed possible explanations.
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What is the logic or significance for using the inverse square factor for calibration of in vivo dosimeters (placed on surface) to the dose measured by ion chamber at dmax?
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Nazia Toor I do understand your point now, thanks for elucidating your thoughts.
While the addition of ISF in the dose calibration formulae may lack significance due to the cancellation during the calculation. However, I think the ISF further acts as a factor to correct for any (if at all) change during setup.
Also, I tried to find relevant articles that have evaluated the ISF effect in in-vivo dosimetry calibrations but to no avail. This may suggest again that both accuracy and precision are needed during calibration procedures.
I hope this helps!
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All organs are not equally damaged by same amount of radiation dosage. But, on which basis equivalency is measured? (i..e 1 gray in this organ equals 10 sievert) Is it arbitrarily qualitative or quantitative as well? Then what is the quantity? ( concentration of reactive oxygen species, DNA mutation frequency, Radiative cellular apoptosis..., percent Coagulation of biomolecules). But all humans are not equally affected by same amount of radiation energy applied on same organ. Then, does the equivalency chart vary from person-to-person, species-to-species, or year to year ?(i.e. refining of values with increasing precision) I f so, thaen how the equivalency are standardized?
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For derivation of dose limits, radiation weighting factors, tissue weighting factors etc, you may be interested in
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I know how to calculate the MU time but not sure how to get the cumulative dose. I have gone through the Radiation Physics book by Faiz. However no clear cut approach is shown for getting the cumulative dose? So my question is 1) Is there any approach by which cumulative dose can be calculated? or it is prescribed by the radiologist? 2) Do we need to optimize the dose distribution for telecobalt therapy?
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Francisco is correct in his comments. I will add a few as well.
Calculating MU/time and calculating dose are the inverse of one another. You need to be given MU to calculate dose, or you need prescribed dose to calculate MU. Typically one is tasked with calculating MU or time after the dose is prescribed by the Radiation Oncologist. In simple calculations the dose would be prescribed to a point in the body and the field apertures shaped to conform to the local anatomy, followed by normalizing to an isodose line that covers the desired area. All of this goes into the calculation of MU or time needed to get the prescribed dose. Modern techiniques involve the use of inverse planning utilizing arcs, dynamic MLC's, etc. It is always important to optimize the dose distribution, whether in linac based therapy or cobalt teletherapy. The ability to optimize the dose distribution is tied to the imaging available (2D, 3D, 4D, PET, MRI, etc.) the technology contained in the treatment planning system, and the skill of the planner.
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I need topics that can be researched on and suitable for PhD research work.
Thank you.
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BNCT
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Hi, everyone! I am studying the damage processes in crystalline ceramic samples under energetic ion irradiations. My samples were irradiated at room temperature. Now I want to consider the annealing effect induced by the temperature increase of samples during irradiation(Is this necessary?).  And the temperature change in samples during irradiation should be determined. To my knowledge, I should firstly determine how much energy of ions can be converted to the internal energy of target system, and then calculate the temperature change according to the equipartition theory (E=3*N*k*T) or Q=c*m*delta-T (c is the specific heat). Is that right? The question here is that how to determine the energy fraction of ions would be converted to the internal energy of system? Can anyone help to answer this question or give some other suggestions ? Many thanks to everyone!
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I work at the center that has a TLD-Reader model 4500. I know the that main sources of uncertainty in environmental dosimetry are: energy and angular dependence, temperature, humidity, linearity, ECC,RCF,Zero dose, ....
may you tell me the procedure of uncertainty calculation for these factors?
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There are many different possible sources of uncertainty for this approcah to measuring dose. Sources of variation include individual chip characteristics, and others sources you mention in your question. As a starting point, I would take an empirical approach and expose your chips to a known source to assess inter-TLD variation. I would repeat this process several times to assess intra-TLD variation. In terms of environmental monitoring, I would also consider using several TLDs together for an individual target and use the average of the readings after correcting for inter-TLD variation. In other words, the usual issues of taking environmental measurements apply and thus one must attempt to estimate the repeatability of a given set of measurements in order to quantify the uncertainty of a given measurement.
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I need some training on the use of Matlab, Monte Carlo, C++ for image processing, image reconstruction, analysis, radiation dosimetry and treatment planning.
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No problem! If you have more questions, you can continue on this thread or write to me personally.
Salutations,
Bruno
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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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It is well discussed but no answers
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The US National Council on Radiation Protection and Measurements (NCRP) has just started public consultation for the draft Commentary “Implications of Recent Epidemiologic Studies for the Linear-Nonthreshold Model and Radiation Protection” prepared by Scientific Committee 1-25 (SC 1-25) under Program Area Committee 1 (PAC 1). Please see “Documents in Review” at http://ncrponline.org/ to download the draft and post comments (due 16 October 2017). The information on SC 1-25 “Recent Epidemiologic Studies and Implications for the Linear-Nonthreshold Model” is available at http://ncrponline.org/program-areas/sc-1-25-recent-epidemiologic-studies-and-implications-for-the-linear-nonthreshold-model/ . The information on PAC 1 “Basic Criteria, Epidemiology, Radiobiology, and Risk” is available at http://ncrponline.org/program-areas/pac-1-basic-criteria-epidemiology-radiobiology-and-risk/ .
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What is meant by prodromal syndrome? At what total-body absorbed dose range this syndrome resulted? 
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5-hydroxytryptamine-3 receptor antagonists (5HT3 RAs), dexamethasone, metoclopramide, haloperidol, metoclopramide, dexamethasone and lorazepam have been used for prophylaxis of radiation-induced nausea and vomiting (RINV) following radiation therapy. Aprepitant (substance P neurokinin 1 receptor antagonist) is the newer antiemetic agent.
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Why and How does Fractionation introduces a "waste in dose", which is more pronounced for beams with a wide shoulder than for beams with a narrow shoulder in the survival curve??
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In addition to tumors refractory to conventional photons, carbon ion radiotherapy utilizes hypofractionation regimens for radioresponsive tumors, e.g., delivery of 40 or 50 Gy in 2 to 16 fractions for lung cancer, prostate cancer, and hepatocellular carcinoma.
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Which dosimetric term used when the source of radiation exposure is from radioactive material located within the body?
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If you are specifically asking internal exposure here, then the relevant radiation protection quantity is committed effective dose and committed equivalent dose. Committed dose quantities are necessary, because the intake of radionuclides leads to irradiation of the tissues over a period of time according to its physical half-life and biological half-time.
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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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I need to know if it is possible to measure any amino acid by any method after radiation exposure from serum of human blood. If yes, can I get the protocol? Are there any specific amino acids in particular that responds differentially to radiation exposure?
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You may be interested in recent radiation metabolomics studies of which several example papers are listed below.
 
[Human]
https://www.ncbi.nlm.nih.gov/pubmed/27815965   (PDF freely downloadable)
 
[Non-human primates]
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In the case of drugs, HED calculation has been well known. However, I think the absorbed dose of radiation is not applied for that, because it is just a wave of radiation and penetrates the whole body.
For example, I'd like to investigate the effect of low dose radiation below 100mSv (10cGy) on mice. Should I expect the same results of mice to human?
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Physically, the absorbed dose is the same for mice and humans. Biologically, however, effects of the physically same absorbed dose are not necessarily the same. Moreover, the radiosensitivity varies among strains of mice, and its choice depends on your endpoint (e.g., which type of cancer).
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The UK HPA recommends the use of the Pka to measure reference levels in CBCT examinations. We are measuring this in Brazil, but we have few results to compare.
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In your doctoral thesis you have used the phantom RS 250. I have great interest in making these measures for quality control. 
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Dear Researchers,
I need some real data of a patient treated with chemotherapy. This data includes the tumor reduction with time. For example, what is the tumor size (or the number of cancerous cells) just before the first dose, the second dose, the third dose,..,etc.? Any type of cancer is acceptable.
Any answer is appreciated.
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There are also free datasets in internet
Try this one
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Use of dicentric counting Chromosomal aberration test for biodosimety for ionizing radiation is very common technique. How much reliable method is this ? i m worried about this because of limited number of dicentrics formations in the exposed person.
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The Special Issue on the European RENEB project guest edited by Drs Ulrike Kulka and Andrzej Wojcik has just become available online as the January 2017 issue of International Journal of Radiation Biology. All papers in this special issue are freely downloadable.
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As we know, a high energy proton will continuously lost its energy along the travelling path in medium. For example, in proton therapy for cancer treatment, a proton beam irradiates a patient and releases all of its energy at the end of path in certain specific depth, so called Bragg peak. How do those existing protons which have no energy and stop at specific depth interact with surrounding molecules in tissue? Will it combine with a electron and become a hydrogen?
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If you are interested in the ultimate fate of the proton, think of it as an H+ ion (which is highly reactive) and how it will interact chemically with the tissue.
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The source consists of eight encapsulated Cobalt-60 pencils arranged in a cylinder. Each source pencil is of 1.4 cm outer diameter and 45.2 cm length. I want to compare the observed dose with theoretical calculated data using "Line source" formula not "point source". Is there any software available of dose calculation for "Line source"?
Please see the attached file for more information.
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Re: comments of Nelson Omi: Nowadays, cable-connected precision ion chamber-electrometer systems with miniature ion chambers(for high range) are available. The electrometer reading system can be therefore located in a radiation-free area. As per the table, the dose at 1m distance for the source configuration is about 600 rads/min which can be accurately estimated. The cables used are double-screened antimicrophonic type with minimum noise and radiation-induced conductivity. If needed, the cable can be routed so the exposed parts may be kept away from the radiation sources. Both dose and dose rates can be measured here. The errors due to the addition of doses contributed by raising and lowering of sources and room/table scatter can be taken into account by measuring the TOTAL dose to the sample under actual conditions. This method will thereby provide a secondary check on the Fricke dosimetry which involves some assumptions for spectrophotometer dosimetry calculation. Besides, ion chamber dosimetric systems calibrated against primary standards in NIST/BIPM/NPL laboratories are available from several companies. Hope this helps. Thanks.
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Reading this paper about the metabolic fate of chondroitin sulfate, where 24% of "Administered radioactivity" ends up in feces after 24h of administration
Single dose in rats of 16 mg/kg and 90 Fci/kg, activity of 12.5 mci/mg, 
3H-chondroitin sulfate (3H-CS) was prepared by reduction with sodium 3H-borohydride.
I've never taken any radioactivity/physics courses in my degree, so Im lost on how to determine the actual concentration that they found in the feces in a molarity or w/v form. Could anyone help me understand the calculations I have to do? 
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Are interested in the radioactivity or the amount of chondroitin?
The radioactivity traces the chondroitin. The chondroitin was administered at 16 mg/kg. The radioactivity traces the fraction of the dose in various compartments. You will need mass of the feces and mass of the total administered (body mass x 16 mg/kg). If 25% went to the feces then concentration is 0.25 of total administered divided by fecal mass.
3H  has 28.8 Ci/mmol. 10-15Ci = fCi. mmole = 90*10-15/28.8
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From kV and mAs values and without the use of any detectors, how can you find the dose delivered to patients? When using an X-ray, for diagnosis purposes, if you know only the two values of mAs and kV, can you determine the dose or maybe even estimate the dose delivered to the patient? Without the use of any modern detectors. 
I have portable 30mA X ray machine with the following specifications:
Output     30mA at 52 KV
                20mA at 68 KV
                15mA at 85 KV
Timer        0.06mA to 6.0
                 23 Steps
Tube         1.5 mm sq. Focal Spot X-Ray tube
Input          230V, 15 Amps
L. V. Compensation     210 to 250V
Beam Limination Cone with Centering Device
Weight        15 Kgs.
Dimension (mm)   250x175x250
Can somebody guide me to calculate the dose in gray (Gy)?
Thanking you in anticipation.
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Mr Sandeep Mittal
I am working out below two independent methods to arrive at the output of x-rays for a given combination of kV and mA. Please note that both these are only approximate but they do indicate the different approaches to the answers. Let us take the distance from the x-ray tube target to the point at which the output is required as 100 cm.
Method 1:
Let us assume kV = 60 and mA = 10; then power dissipated on the x-ray target is:
60 x 10 watts or 600 joules/sec = 600 x 107 ergs/sec ……….(1)
The fraction of electron energy converted to bremmstralung (f) is given approximately by:
f = Ee x Z x 10-6 (Z is the atomic number of target = 74 for tungsten; Ee is the electron energy in keV)(from medical physics literature).
Since the average energy of the unfiltered x-ray photon energy is one-third of the electron energy from Kramer’s equation, the photon energy available as bremmstralung therefore is:
Hence f = (60/3) x 74 x 10-6 = 0.148% …………………………..(2)
Therefore the x-ray energy output at the target = (1) x (2) above = 0.888  x 107 ergs/sec …… (3)
Assuming the x-ray distribution from the target is isotropic (valid for a thick target), the energy output at 100 cm is: [(3) above] / 4π (1002) = 0.0705 x 103  ergs/sec/cm2 ………………(4)
Assuming the total (inherent + added) filtration in the tube is 3 mm:
For 60kV, 3mm filtration offers 2HVLs, since the first HVL is 1mm and the homogeneity coefficient is 0.5 (obtained from literature); Hence the attenuation offered by 3 mm Al = ¼ (0.25)
This leads to an output after filtration = [(4) above]/4  = 0.0176 x 103 ergs/sec/cm2 ……….(5)
To get the output air in R/sec at 1 meter, we have to multiply (5) above by the mass energy absorption coefficient of air for 60 kV with 3 mm filtration. For an unfiltered x-ray beam, Kramer’s Law gives an average energy of one-third kV; for a filtered(hardened) beam, let us assume the average energy as 30 keV (this is a valid approximation). The mass energy absorption coefficient µen /ρ for 30 keV x-rays = 0.1501 cm2 /gm [From Hubbel’s 1982 revised 2012 data]…………….(6)
Therefore the output at 1m = [(5) x (6)]/ 87.7 [1 R = 87.7 ergs/gm of air] = 0.0301 R/sec  = 1.81 R/min…..(7)
This can be converted to air kerma in rads/sec by multiplying by 0.877
It may be noted the above derivation involves a number of approximations. Moreover, the values of  (µen /ρ) vary widely over the low keV region.
 Method II
This  method uses the RadPro calculator available free on-line. Here you have to enter the values of kV (60), mA (10), filtration(3 mm Al) and distance(100 cm) from the target to the point of measurement. This calculation then yields a value of 1.8719 x 108 µR/hr = 187.19 R/hr = 187.19/3600 R/sec = 0.052 R/sec = 3.12 R/min ………………………….(8)
It may be noted that these values (7) and (8) above are obtained by altogether two different approaches and both involve approximations. Both these denote typical outputs of a fluoroscopic unit.
Please consult any radiation physics text or website for references cited above.
Hope this helps.Good Luck.
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In calculation of the absorbed dose in radionuclide scintigraphy, the absorbed dose in tissue T from radionuclide  in a single source organ S is given by:
D(T <-- S)= As x S (T <-- S)
where As is the cumulated activity. 
I want to calculate the absorbed dose of the kidney in renal scintigraphy using 99mTc-DMSA. I have measured the activity of the rat's kidney using dose calibrator. for calculation of the absorbed dose in a period of time, I need the S-value for rat's kidney and 99mTc-DMSA.
How can I find or calculate it?
Or:
Is there another way to calculate the absorbed dose in rat's kidney from measured activity?
Thanks
Kaveh
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What is the best method to assess the absorbed dose by the target organs during an X-ray computed tomography?
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Dear Slimane,
During an X-ray computed tomography exposure, the most common parameter
used to estimate and minimize patient dose is the CT dose index or CTDIw (Computerised Tomography Dose  Index). 
CTDI is measured with a pencil ionization chamber ( 10cm) and  2 cylindrical phantoms (  the measurements are only an approximation of the patient dose). 
These phantoms refer, respectively, the head examinations (with a diameter of 16 cm) and body examinations  (with a diameter of 32 cm ).
After measuring CTDIw = CTDIw = (1/3)CTDI100,center + (2/3)CTDI100,peripheral
>>>>>>> CTDIvol = CTDIw/Pitch
One approach (actually an approximation):
E= DLP * k
where :
* E = Effective Dose in mSv
* DLP is Dose Length Product given by: CTDIvol* length of scan [in mGy*cm] and 
* k= .0023 for head exams , k =0.015 for abdomen.
we can have easily average absorbed dose to tissue if we have: 
  • wT= tissue weighting factor (next page)
  • wR= radiation weighting coefficient
* CTDI underestimates dose from contiguous scans (helical) by not capturing scatter tails.
* CTDI overestimates dose from axial scan with no table motion because scatter tails included.
Best regards.
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Hello all,
I have made an excel program for differentiating the global radiation values into direct and diffuse radiation for vacuum tube solar collectors. In order to validate my calculations and equations, I need values of global radiation and corresponding values of direct radiation, both measured at the same time, location and conditions.
If anyone is willing to share those, it will be of great help. Also, please suggest how do I cite the data while using it in my project.
Thank You.
Best regards,
Aparna
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Please contact me at agu.laguarda@gmail.com
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In the literature it is said that the grainless nature of radiochromic films is responsible for their high resolution. What exactly does this mean?
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Yes, it is very helpful. Thank you very much.
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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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I want to do dosimetry with dosxyznrc code and for this purpose I need to make egsphantom with ctcreate code from thorax phantom dicom images. My phantom is from thorax region and includes three material; the lungs, the heart and the spinal cord.
thanks in advance to your attention or maybe your answer.
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You are very welcome Milad. Did you work?
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We have measured the radiation doses at different positions of a Cobalt-60 gamma irradiator room using Fricke and Ceric-cerous dosimeters. Now, we want to compare and validate the obtained dose with established formula (to calculate radiation dose) and dose calculated by Monte Carlo simulation. I need help to calculate the dose using Monte Carlo simulation. It will be a great pleasure for me, if there is anyone who has expertise in Monte Carlo simulation and interested to help me.
Thank you very much.
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Dear Firoz
There are a number of MC codes such as MCNP, GEANT,FLUKA etc. that can be used to do any parametric studies including high dose dosimetry. There is an article that covers MC simulation, high dose dosimetry and benchmarking it is the following:
  Sohrabpour, M., Hassanzadeh, M., Shahriari, M., Sharifzadeh, M., “Gamma irradiation dose mapping simulation using the MCNP code and benchmarking with dosimetry”, 2002, J. Appl. Radn. Isot.Vol. 57, No. 4, PP. 537-542
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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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I tried searching it in google indeed, but just found some old and new articles on modeling of radiation environment that were of no use to my problem. What I'm exactly looking for is a set of information like what we have for gravitational field (Spherical harmonic coefficients) that describes the magnetic field strength and charge densities in proportion to solar activity.
Let's put it this way: I want a model to use it to calculate the amount of radiation dosage absorbed by spacecraft in a trajectory around Jupiter (or Earth) for a certain amount of time.
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I believe ICRP Publication 123 has the correct information. For simulations, I recommend using SPENVIS.  https://www.spenvis.oma.be
Jupiter specific information:
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The typical properties of them is the 0.5 mT of splittings from the central line at both sides, but what are the other evidence; MW power saturation features, decay, kinetic features,... etc.
Thanks in advance
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The spin-flip transitions are less probable than the main transitions, so, usually they are more resistant to MW power saturation than the main lines.
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Many radiotherapy modalities use small fields or beam overlapping during treatment. Dose measurement under such circumstances can be accomplished by systems planning or direct measures (ion chambers, radiochromic films, etc,.)
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After sending the above answer, I happened to come across an ad in the current Journal of Medical Physics of a 1mm spatial resolution water-equivalent Extradin W1 scintillator-based instrument for small field dosimetry,  marketed by Standard Imaging company. The instrument has built-in automatic correction for Cherenkov Effect in the light guide.
I hope this instrument may be ideal for your work. 
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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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One idea to improve a TLD response sensitivity and other its characteristics is to add co-dopants to the phosphors. Is there any rule for co-dopant selection in a specific phosphor which already has a dopant? Is there any way to predict the effects of adding a co-dopant to a TLD phosphor?
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In the case of lanthanide doping, look for the papers by Pieter Dorenbos, e.g.
Dorenbos, P. and Bos, A.J.J., 2008. Lanthanide level location and related thermoluminescence phenomena. Radiat. Meas. 43, 139-145.
Other co-dopants can also increase the TL intensity by introducing new defects or acting as charge compensators, increasing the incorporation of the dopants. In the materials we synthesized, Li co-doping often increases the TL intensity by many orders of magnitude:
Oliveira, L.C., Milliken, E.D. and Yukihara, E.G., 2013. Development and characterization of MgO:Nd,Li synthesized by Solution Combustion Synthesis for 2D Optically Stimulated Luminescence dosimetry. J. Lumin. 133, 211-216.
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According to the definition, CTDIvol is obtained by dividing CTDIw by pitch factor. 
The use of tube current modulation results in different CTDIw values for every rotation. Even though CTDIw is not constant during the whole scanning, the CT scanner displays a CTDIvol value at the end of the examination. How is this value calculated? I'm especially interested in how Siemens Somatom machines calculate it (in case different CT producers have different methods).
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I have been using the unit: umol m-2 s-1 to talk about photosynthetically active radiation, but is it wrong in the International System of Units?
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Yes, but the journals ask for the use of SI!
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Does anyone know that : How to measure the decay time of scintillator material? I would like to measure the decay time and afterglow of scintillators which are used for X-ray detection (particularly with synchrotron source).
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You can use a transient recorder to record and evaluate single pulses, see link. There are cards for PCs, and depending on what exactly you need, these are not very expensive anymore.
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I was just wondering if anybody has any experience with using Ar + Cl mixtures as a fill gas in GM counters? What pleateau slope, length and starting voltage can one expect as a function of increasing Cl concentration in such mixtures at say about 100 mbar? As far as I know, even low concentrations of Cl, say between 0.1%  - 1%  are sufficient to obtain the required quenching. I would appreciate any feedback on this. Thanks in advance !
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I think in the absence of some literature that applies to your setup you will simply have to plot the slope and establish the plateau voltage yourself. Chlorine is the most common halogen quench gas and it will allow you to operate at a lower GM voltage than using an organic quench gas. Depending on the dimensions and pressure (density) of your detection volume, your voltage may range anywhere from +-400 to +-1400 V.  I suggest modifying the proportion of quench gas (Cl) based on your dead time calculations only and not in effort to change your operating voltage.
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Hi every body
In our radiotherapy center, there is a Varian Linac 2100C/D with energy  6MV. However the Flattening Filter Free (FFF) mode is not available. I need PDD, profiles  at several depth measured for fields 2x2, 4x4, 6x6, 8x8, 10x10, 20x20, 30x30 and 40x40 and also Sc and Scp for Varian 2100C/D 6MV energy without Flattening Filter (FFF) for comparing to simulation data obtained by Beannrc code.
Best regards
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Dear Sergio Faermann 
Thanks for your response. you are right two accelerators are not same and of course we can not remove FF, I like to comprise some simulation code , however for validation of each Linac I need PDD and profile measuring data. So for this data, we are able to compare two code which are validated by same measuring.
by the way, thanks for your attention.
beast regards
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OR, does the use of these radionuclide tracers, themselves, add a significant risk of causing cancer in patients?
The dose of ionizing-radiation from the tracer used in one PET scan, for example, typically exposes the patient to about 25% of the maximum allowable annual radiation exposure permitted for nuclear workers (which is a VERY high limit = to over 200 standard/modern medical chest xrays, meaning a patient is getting exposed to the equivalent of about 50x chest xrays ALL AT ONCE for each PET test).
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Dear Bob,
Actually, some of them appear to be not quite safe. For example, there are risks of a Gallbladder Radionuclide Scan.
The Risks of a Gallbladder Radionuclide Scan
There is a risk of exposure to radiation with this test. The gallbladder radionuclide scan uses small amounts of radioactive tracers. However, this test has been used for over 50 years and there are no known long-term side effects from such low doses of radiation. The benefits of the tests outweigh the risk of radiation exposure (Radiology, 2012).
There is, however, a rare chance of an allergic reaction, which is typically mild.
On the other hand, others were proved to be safe. For example, in the bone scan:
The amount of the radionuclide injected into your vein for the procedure is small enough that there is no need for precautions against radioactive exposure. The injection of the tracer may cause some slight discomfort. Allergic reactions to the tracer are rare, but may occur.
Hoping this will be helpful,
Rafik
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for example
if there is a patient planned to treated with 25 fractions with 200 cGy, for 5 days per week (5 weeks), let he absent for 3 days for the first week and absent 2 days for the second and finally absent for 10 days in the last one,  how to calculate the actually received dose and the remaining
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Dear Dr. Gawad,
Although I have heard of the Ellis NSD-concept, I have more experience with the LQ-model (by Fowler and Barendsen). The literature on missed radiation fractions is extensive; for practical purposes I recommend the papers by Nuran Bese et al. (2007) and by Roger Dale et al (2002) [1,2]. 
If you do not have access to these two papers, do not hesitate to send me an e-mail, so I can forward them as pdf-attachments.
There are several (more or less freeware) LQ-model software packages, that also have a a module for compensation of missed fractions.
See amongst others: http://www.eyephysics.com/tdf/
Sincerely yours,
Lukas Stalpers, radiation oncologist
AMC - University of Amsterdam, The Netherlands
1. Bese NS, Hendry J, Jeremic B. Effects of prolongation of overall treatment time due to unplanned interruptions during radiotherapy of different tumor sites and practical methods for compensation. Int J Radiat Oncol Biol Phys. 2007;68(3):654-61
2. Dale RG, Hendry JH, Jones B, Robertson AG, Deehan C, Sinclair JA. Practical methods for compensating for missed treatment days in radiotherapy, with particular reference to head and neck schedules. Clin Oncol (R Coll Radiol). 2002; 14: 382–393
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I think direct measurement with anthropomorphic phantom is the best way, but this method needs expensive phantoms and is time consuming according to calibrate and read TLDs. Is an alternative method to asses organ doses with less difficulties?
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You might want to look at the method developed by Walter Huda and me.  Using the free-in-air isocenter kerma, we measured the ratio to organ doses in multiple anthropomorphic phantoms.  Very easy approach to organ dose:
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Hello,
has anyone used 4+4Gy TBI for NSG mice irradiation and engraftment of:
1.Human leukemia cell lines
2.Human CD34 stem cells?
Any less than that? 4Gy only?
Thank you
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For a human T-cell xenograft experiment in the NSG model, we used 2Gy TBI. Despite a 24 hour delay of T-cell transfer post TBI to ameliorate inflammation and potential GvHD, we observed profound splenomegaly in addition to clear behavioural indications of GvHD at experimental termination 2 weeks post T-cell transfer. Had we continued along this experimental vein, we would have considered reducing the TBI dosage to 1 Gy for subsequent models. T-cell dose used was 4x10^6 transgenic T-cells/animal.
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In radiation dosimetry for acceptance test of computed tomography;   papers speaking about ' CTDI ' more than 'MSAD' ;
seem that MSAD is theoretical concept but maybe accurate than other index to estimation of dose in QC.
How be measured this index in practice?
Thanks in advance
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In clinical practice (scanning patients), only CTDIvol and DLP computed by the scanner are displayed on the console and recorded in the DICOM RDSR, so most non-physicist folks have never heard of MSAD. For acceptance testing (which is what you asked about), AAPM Report 39 (1993) has a lot of discussion of MSAD, but I don't think that IAEA or ImPACT or ACR–AAPM Technical Standard even mention it,
I quote from the McCollough et al article mentioned in an earlier comment: "In the early days of CT, direct measurement of the MSAD was a labor-intensive process ... the introduction of CTDI by Shope et al provided a much more practical method with which to estimate the MSAD".
See also the discussion in AAPM Reports 96 and 111.
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I want to use Fricke dosimetry (liquid and gel) system for alpha radiation, but the issue is G-value of the system. I want to know how to increase the G-value?
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 1. Chemical Dosimetry during Alpha Irradiation: A Specific System for UV-Vis in Situ Measurement /Cedric Costa, Johan Vandenborre, Francis Crumière, Guillaume Blain, Rachid Essehli, Massoud Fattahi
2. 
Radiolysis of 0.4 M sulfuric acid solutions with fission fragments from dissolved californium-252. Estimated yields of radical and molecular products that escape reactions in fission fragment tracks / Ned E. Bibler J. Phys. Chem., 1975, 79 (19), pp 1991–1995
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in radiotherapy procedures fractions of dose delivering to the tumor from different planes, i would like to know is it possible to measure the accurate absorbed dose using TLD material. If possible what are the exact location of placing the TLD material on the patients body, in case of different therapy procedures. 
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Under the precondition of correct calibration you can measure doses at each place on the patient. Your problem will be to estimate or measure inner doses in the tumor volume if your tlds stay at the surface (skin) of the patient. The main reason ist the entering radiation field which can be unpredictably be contaminated with secondary radiation (scattered photons, secondary electrons and so on). And measurement in the inner part of the patient is only possible if you enter the cavities eg vagina, mouth, nose, rectum. In all other cases tld measurements will stay very unsecure estimates.
Another situation exists if you use human phantoms like rando phantoms with a lot of tld holes. This technique is a well defined experimental situation and technique I often used to check up the value of planning results and to minimize patients exposure.
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I have undergone chest x-rays for 8 times in only 9 months. I am told each time u go to an x-ray your life span is short by 3 years. So, I imagine I have lost 27 years of life ahead and predisposed to cancer attack risk.  But again I am told that if I take  antioxidants like resveratrol those free radicals as a result of x-ray exposure can be removed. When I ask doctors they don't give me a clear answer; I am now asking the experts who know what's on ground . What can be an advise to me looking at the above concern?
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As documented by other colleagues above, the dose received for each chest X.rays is very low and thus the risk of cancer is so low that no study can demonstrate it. The information given to you about the reduction of 3 years of your life for each X ray is non sense.
Your question is also to know if antioxydants could remove the free radicals created by the interaction of X rays with your tissue. The reponse is clearly NO just because the life of these radicals is less than a fraction of second. This is the reason why radiation protection relies on prevention and is based on the justification and the optimisation of medical exposures. Thus the main question in your case is  to know why you had 8 chest X rays in 9 months.  Were all the examinations useful ?  Have they changed the disgnosis or the treatment ? If not they were useless and should not have been performed? Ask you doctors !
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I'm using GM tube LND7121 
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A short answer to differing count rates. Either you see dead time effects possibly caused by an electronic change, or your detectors are not really identical, possible cause for different count rates can be entrance window thickness. Or third reason could be  a low power supply in the low dose rate detector. My central question is, don´t you use calibration sources to check the operability of your system?
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As, i am new in this feild, I want to know how can i use fricke dosimetry to determine the x ray from synchrotron radiation
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Fricke dosimetery may not be ideal for determining the X ray measurement from a synchrotron.  Synchrotron radiation is broad band and ranges form the visible through the X-ray spectrum.  Measurement of the change of Ferrous to ferric ion in a solution may not be the best way a the reaction is likely to be very energy dependent, nevertheless it could be done, but you would need to calibrate it against a standard by measuring the ionisation in air, produced by the beam and then calibrating against the F++ to F+++ reaction rate. 
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Im looking to study bile acid transport via apical sodium codependent bile acid transporter (ASBT) in CaCo-2 cells. I have read several papers with brief descriptions of radiolabled bile acids in a transwell assay.
A) I was wondering if any on has done this via fluorescence instead of radioactivity?
B) I have never done radioactivity. Different paper use different labeled bile acids (i.e. 14C, 3H, ect.)....is there a specific reason?
C) Does anyone have a detail protocol for this assay? 
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For radioactivity, you will have to deal with a number of regulations including training, regular testing and proper disposal. The radioisotope C14 is a beta emitter and beta emissions are easily quenched by the air and even lab gloves. Tritium on the other hand is gamma emitter and gamma particles are higher energy and not as easy to quench. In terms of experimental development, you won't need to deal with the regulatory hassles and safety concerns when you do fluorescence. The sensitivity of of fluorescence may not be as good which is one disadvantage. 
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I am looking for how to apply the dose constraints for OARs in the case of a RC3D hypofraction treatment.
Example; RC3D case of lung cancer treatment is a vector of 66 Gy, for the DVHS v20 is applied to the healthy lung <30% and the V30, <20%
My question is when the doctor decided to do a treatment with a dose of 30 Gy in 10 seance. Is what you apply according to the same dose constraints given that doses, per seance, are not the same?
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merci pour vos réponse  
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I am working on electromagnetic field effects on pregnant rats, so I want to now how can I measure SAR for rat's bodies.
Thanks for your help.
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Thank you so much Maciej M Kmiec and Somayeh Asadi for your help. I  will wait you Dr. Somayeh Asadi. 
best regards
Ali
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Dear Richard,
thats exactly the point. We should differentiate between screening techniques like mammography screening, TB etc and the indicated examination in a special case. In Germany screening must be approved. And the concerned circle is the population. Effective dose could be right even with all limitation of risk models.  Abusus of X-ray examination because of laziness of the examiners must be avoided and restricted.
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We would like to investigate the effects of diagnostic x ray irradiation on lipid peroxidation in different animals tissues. Therefore we would like to know if somebody had already determined a dose of radiation (Gy) exerting meseurable changes in lipid peroxidation.
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Dear dr Kamal Hadad,
thank you very much for the information provided.
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The International Commission on Radiation Protection (ICRP) proposed a set of operational quantities defined to allow for calibration of ionizing radiation protection instruments for measurements to show compliance with the system of protection quantities. These measurable quantities are the ambient dose equivalent, the directional dose equivalent, and the personal dose equivalent.
An earlier question "What is the difference between Sievert and Gray? A practical question concerning the SI units for ionizing radiation?" addressed the confusion of Sievert and Gray and its use in radiation protection programs. This question is a continuation and addresses the practical aspects of calibrating and interpreting instruments used for radiation protection.
The ICRP asserts it has proposed measurable quantities, but have defined them by calculation. The calculation is ideal and impractical for measurement as a parallel expanded beam of a single energy is not possible to produce. The point of dose is at a depth in a sphere or slab, a location not accessible to an instrument. Actual calibration must be performed free-in-air with a non-uniform beam and with physical constraints that may not be negligible. Calibration is to an instrument that is energy dependent and does not have the backscatter characteristics of a sphere, slab, or human body.
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Dear Joseph,
I first read our old discussion again, in order not to repeat to much.
I think you had a look at my chapter about doses and dose units. The central point there is, you can calibrate a dose meter using exaggerating geometries like enlarged and adjusted fields in special phantoms. That means that your calibration factors contain  attenuation and scattering from the phantom and maximal enlarged fields. If you now add the radiation quality factors Q you indeed can factor in the kind of radiation (light or dense ionizing).
I have strange problems to use the unit Sievert (Sv) for "physical" operational doses used for measurements and body doses like organ dose and effective dose which are used to estimate risks.
For the information of other participants in this question I add again my textbook chapter, where you can find the calibration geometries, phantoms and dose definitions etc.
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use ImageJ software to get the ADC value, but, when I get the ADC map, the resolution is low, i can not draw the ROI around tumor, and before the software told the image is not 32 bit..I used 1T Bruker ICON Siemens, b value is 0, 800! And also, I wonder ImageJ can get ADC value from only two b values or multiple b values? Do anyone know about software can use multiple b value for ADC value? (DTIstudio?). Please refer the attached file and help me sold the problem! Thanks!
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You can compute it yourself. DWI and ADC map are usually low resolution image when compared with T2W images. As reference : http://www.sciencedirect.com/science/article/pii/S001048251500058X
in fact the ADC can be expressed as:
ADC = - ( ln( Sb1 / sb0 ) / b1 )
with b0 = 0 and b1 = 800, Sb1 and Sb0 are the intensities of each image at b0 and b1.
If you have more than two b, you can perform semi-logarithmic regression.
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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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if you know an article addressing this question with quantitative results will be great
treatment site , GYN, skin ,etc  
thank you in advance
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Many articles are available on this issue.
Two articles are attached for your perusal.
1. DeWard et al., A dosimetric uncertainty analysis for photon-emitting brachytherapy sources: Report of AAPM Task Group No. 138 and GEC-ESTRO (published 14 Jan 2011)
This report addresses uncertainties pertaining to brachytherapy single-source dosimetry preceding clinical use. The International Organization for Standardization (ISO) Guide to the Expression of Uncertainty in Measurement (GUM) and the National Institute of Standards and Technology (NIST) Technical Note 1297 are taken as reference standards for uncertainty formalism. Uncertainties in using detectors to measure or utilizing Monte Carlo methods to estimate brachytherapy dose distributions are provided with discussion of the components intrinsic to the overall dosimetric assessment. Uncertainties provided are based on published observations and cited when available.
2. Antony Palmer et al., Physics-aspects of dose accuracy in high dose rate (HDR) brachytherapy: source dosimetry, treatment planning, equipment performance and in vivo verification techniques.
Journal of Contemporary Brachytherapy (2012/volume 4/number 2) 81-91
This study provides a review of recent publications on the physics-aspects of dosimetric accuracy in high dose rate (HDR) brachytherapy. The discussion of accuracy is primarily concerned with uncertainties, but methods to improve dose conformation to the prescribed intended dose distribution are also noted. The main aim of the paper is to review current practical techniques and methods employed for HDR brachytherapy dosimetry.
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I want estimate effective dose from cardiac CT with PMMA phantom and ionization chamber and a dose calculator software. is any other way to estimate E dose? 
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To physically measure the dose (rather than modelling), you can use a RANDO or ATOM phantom and use either TLDs or MOSFETs, this is measuring the organ doses directly, from which you can calculate E. This is more accurate than any modelling method, however it is significantly more time consuming, so it would depend on the reason that you are calculating E.
As well as the method of using the E/DLP factor, there is also the IMPACT Excel spreadsheet, which uses the SR250 montecarlo dataset to calculate E from the selected parameters and the CTDI values, the SR250 MC dataset is £50, but the spreadsheet is free.
If you are interested, I have attached an article from one of my PhD students on the variety of ways to calculate and measure E.
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Hi,
    I have created a DICOM file in MATLAB and hope to import into Pinnacle TPS. I have double check this DICOM file can be opened in ImangJ or DICOM viewer, but when this file is imported into TPS, this file can be recognised as DICOM but warning message is shown: the data dose not exsit./ the path and/or file predix is invalid. 
    dose anybody meet the similar thing before? I paste the head file below. please could you help me to check? thanks.
                          Filename: 'IM00010-CT'
                       FileModDate: '08-Feb-2015 15:30:42'
                          FileSize: 525882
                            Format: 'DICOM'
                     FormatVersion: 3
                             Width: 512
                            Height: 512
                          BitDepth: 16
                         ColorType: 'grayscale'
    FileMetaInformationGroupLength: 210
        FileMetaInformationVersion: [2x1 uint8]
           MediaStorageSOPClassUID: '1.2.840.10008.5.1.4.1.1.2'
        MediaStorageSOPInstanceUID: '1.3.6.1.4.1.9590.100.1.2.87336183112909916105497931640152611773'
                 TransferSyntaxUID: '1.2.840.10008.1.2.1'
            ImplementationClassUID: '1.3.6.1.4.1.9590.100.1.3.100.7.1'
         ImplementationVersionName: 'MATLAB IPT 7.1'
              SpecificCharacterSet: 'ISO_IR 100'
                         ImageType: 'ORIGINAL\PRIMARY\AXIAL\HELIX'
              InstanceCreationDate: '20141120'
              InstanceCreationTime: '164243'
                       SOPClassUID: '1.2.840.10008.5.1.4.1.1.2'
                    SOPInstanceUID: '1.3.6.1.4.1.9590.100.1.2.87336183112909916105497931640152611773'
                         StudyDate: '20141120'
                   AcquisitionDate: '20141120'
                       ContentDate: '20141120'
               AcquisitionDateTime: '20141120164237+1030'
                         StudyTime: '163950'
                   AcquisitionTime: '164237'
                       ContentTime: '164241.465000'
                   AccessionNumber: ''
                          Modality: 'CT'
                      Manufacturer: 'Philips'
                   InstitutionName: 'RAH ONCOLOGY'
                InstitutionAddress: 'ADELAIDE'
            ReferringPhysicianName: [1x1 struct]
                       StationName: 'HOST-7520'
                  StudyDescription: 'PHYSICS'
                 SeriesDescription: ''
       InstitutionalDepartmentName: 'Radiology'
                      OperatorName: [1x1 struct]
             ManufacturerModelName: 'Brilliance Big Bore'
                       PatientName: [1x1 struct]
                         PatientID: 'a201114'
                  PatientBirthDate: ''
                        PatientSex: 'O'
                        PatientAge: ''
                       ScanOptions: 'HELIX'
                    SliceThickness: 2
                               KVP: 140
            DataCollectionDiameter: 600
                   SoftwareVersion: '3.5.5'
                      ProtocolName: 'QA Laser Physics/PHYSICS'
            ReconstructionDiameter: 500
                GantryDetectorTilt: 0
                       TableHeight: 159
                 RotationDirection: 'CW'
                      ExposureTime: 800
                   XrayTubeCurrent: 250
                          Exposure: 200
                        FilterType: 'B'
                 ConvolutionKernel: 'B'
                   PatientPosition: 'HFS'
                  StudyInstanceUID: '1.2.840.113704.1.111.5696.1416455815.16'
                 SeriesInstanceUID: '1.2.840.113704.1.111.2888.1416463943.6'
                           StudyID: '18591'
                      SeriesNumber: 2
                 AcquisitionNumber: []
                    InstanceNumber: 1
              ImagePositionPatient: [3x1 double]
           ImageOrientationPatient: [6x1 double]
               FrameOfReferenceUID: '1.2.840.113704.1.111.2888.1416463822.3'
                        Laterality: ''
        PositionReferenceIndicator: ''
                     SliceLocation: -66
                     ImageComments: ''
                   SamplesPerPixel: 1
         PhotometricInterpretation: 'MONOCHROME2'
                              Rows: 512
                           Columns: 512
                      PixelSpacing: [2x1 double]
                     BitsAllocated: 16
                        BitsStored: 16
                           HighBit: 15
               PixelRepresentation: 1
           SmallestImagePixelValue: 24
            LargestImagePixelValue: 2718
                      WindowCenter: [2x1 double]
                       WindowWidth: [2x1 double]
                  RescaleIntercept: -1024
                      RescaleSlope: 1
          PerformedProcedureStepID: '1859188'
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Each system has pre-defined required meta data. So, assuming you have a valid image matrix, it might be the header-meta. One issue I have seen (different vendor not phillips) is the data type mismatch. I see yours as uint8, and commonly are maybe uint16. Mess with header meta data a little.
Matlab defaults are not enough to "fool" the vendor products... to accept the image to open.
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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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If you are a clinical medical physicist, what are the major challenges in your department in order to perform invivo dosimetry for each individual patient when he/she is receiving radiotherapy or undergoing for a diagnostic radiological procedure?
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Hi Shakardokht,
I don't have any experience with dosimetry in diagnostic radiology but I have been doing radiotherapy in vivo dosimetry (TLDs and GaF film) for a while now.  Narayan's work is really intra cavitory dosimetry, which, these days is not performed regularly as modern planning systems can predict/estimate the doses to OARs quite accurately. I have found eye lens dosimetry to be the most challenging one. In vivo dosimetry in this case uses surrogates on the skin to estimate dose to lens at depth. The other issue the uncertainty in dose estimation using TLDs due to the angular dependence of TLDs. Not sure if a phantom exists where one could place TLDs in place of lens and then performs dosimetry by putting TLDs on the skin and comparing the doses from on the skin TLDs to the TLDs placed inside the phantom where the lens is located. The other anatomical areas where in vivo can be challenging are the nose and ear, especially when MV electron are used to treat skin tumours in these areas. The issues to consider here are the electron backscatter, uneven surface and the lack of tissue. There probably are other areas/issues such as dosimeter preparation readout etc. which can also be challenging.
Cheers
Vinod 
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I am working with dose issues in CT. In my former institute we used TLD measurements inside the Rando phantom for accurate dose estimations. Besides the fact that these measurements are very time consuming, the phantom + equipment cost a lot of money. Do you have suggestion about any alternatives? May be simulations?
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We started a PhD on this subject. We will employ MCNP and try also EGSnrcMP. That gives our aspect on the theme.
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I have a Geiger counter that not only displays counts but also the dose rate in Gray/h. The dose rate is calibrated to gamma radiation from a Cs-131 source.
If I'd have a U-238 source instead, can I estimate the dose rate from the rate the Geiger counter is displaying?
If I'm correct U-238 gammas have about 50 and 110 keV, while Cs-131 is 662 keV. Does that mean that as a rule of thumb I could divide the displayed rate by 6 to get an estimate of the actual dose rate?
I think it should be possible, if the detection rate is the same for both gammas and pure U-238 is present. Is that correct?
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Ahhh gosh...... I should always read questions before answering..............Starting again its not Cs-131 but Cs-137 yes I know you mentioned ~662keV and U-238 50&110keV but one reads over that! I am afraid that a GM tube can "see a difference" between these energies unless it has energy compensation http://en.wikipedia.org/wiki/File:GM_tube_compensation.gif and this is because of the interactions that occur in matter. The photoelectric effect depends on the energy and the material in the GM tube and that is the prominent effect up to about 50-60keV then the Compton effect starts. I would urge caution in using an uncompensated tube "calibrated" for Cs-137 to measure lower energy gammas, even with compensation you can see that as the energy gets lower response diminishes. The example from wikipedia even suggests that there is a difference between 50 and 110 keV!
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In a first look, both are fluorescence, giving immediate light emission on irradiation.  Mechanism wise both are similar. If both are same then why different terminology? Or is there any technical difference?   
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I´ll try a quick answer. Radioluminescence is caused by irradiation of certain materials with radioactive particles and photons. This radioluminescence can be prompt, called fluorescence, or delayed, called phosphorescence. Distinctions are the processes in the solid. Primarily no statement to delay.
Scintillation is the prompt emission of "visible" light of irradiated substances, the scintillators. So if you want, scintillations are a subgroup of radioluminescence.
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Is there any literature evidence on symmetric planning and its advantages over asymmetric planning in ldr brachy. Thanks
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Most prostates are bi-laterally symmetric in their external contour.  Asymmetric planning may arise from two causes:
     1.  If the trans-rectal ultrasound (TRUS) volume study has the prostate shifted left or right from midline, you should center the prostate for the treatment plan. The patient is not sedated for the TRUS volume study, but for the implant in the operating room (OR) with the patient anesthetized, the brachytherapist can more accurately align the prostate to the grid.
     2.  If the prostate is asymmetric, your needle placement approach may affect outcomes.  The two dominant philosophies either consider the target as the prostate or as the prostate plus explicit margins.  The latter is recommended by ASTRO, ABS, and GEC-ESTRO guidelines, but many centers still consider the prostate alone as the target.  Our margins are 4, 5, and 6 mm for low, intermediate, and high-risk patients, respectively, but they are also influenced by mapping biopsy results.  If the prostate is centered and margins added, the resulting planning target volume (PTV) can be made bilaterally symmetric by subsuming prostate invaginations and nodules within the PTV.
For reasons of speed, predictability, and simplicity, all of my plans, about 2,800 patients, have needle placements and offsets that are bi-laterally symmetric.  If the plan calls for a needle 2 cm left of midline and 1 cm offset from the base, the brachytherapist will expect a matching needle on the right.  By limiting the parameters that change in the OR you limit the possibility of mistakes.  Even though the needle positions are symmetric, the contents of the needles may differ in terms of number of sources and their spacing.  Not one of my plans has ever been symmetric in terms of seed placement.
I have seen no credible clinical comparisons of symmetric versus asymmetric needle placement, and I expect no difference in outcomes.  There are also no formal comparisons between outcomes of centers treating the prostate plus margins versus those treating prostate alone.  In the latter case, the 100% isodose is often pushed outside the prostate, but the extent is rarely documented.  For low-risk patients, differences in biochemical progression-free survival are within the uncertainty expected in inter-institutional comparisons.  However, for high-risk men, the differences are substantial.  All the brachytherapists, whose outcomes for high-risk men are comparable to ours, treat with large, explicit margins.  (Refer to GS Merrick et al., "Time to failure after definitive therapy for prostate cancer: implications for importance of aggressive local treatment," J Contemp Brachytherapy, 5: 215-221 (2013).)
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I want to built a miniphantom for SC measuring using EBT2 film for small field 6MV photon beam,  2 plexiglass cubes (density 1.19g/cm3) dimensions 3x3 with height of 5 cm for top and 10cm for bottom as a holder.
1- Which materials do you recomment for top for electron contamination in small field film dosimetry? Brass top or plexiglass
2- Is it necessary to calculate the equivalent thickness for top (plexi or brass)? 
3- What dimensional did you consider for your phantom? lateral dimensions and top thickness?
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It dependent to your protocol.  For estro is 5 cm lateral and 10 cm depth of phantom. Brass used for 10 mv and higher. In iaea protocol use of build up phantom. You can use of fc65 instead of ebt2.
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Treatment time (for teleterapy unit) is similar with monitor unit (for LINAC).
Manual calculation for MU or time depends on dose delivery technique, there are SSD tech. and isocentric tech.
To calculate MU in LINAC,
If we calculate in SSD tech., we used data of PDD.
if we calculate in isocentric tech, we used data of TMR.
How does to calculate treatment time Cobalt-60 teletherapy unit for in SAD/isocentric techique?
we use TAR or TMR ?
if we have data of TAR and PSF Cobalt-60 from BJR Supplement 25, to calculate time in isocentric tech, we use TAR data? or must convert TAR to TMR ?
Thanks a lot for your discussion.
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To calculate treatment time Cobalt-60 teletherapy unit for in SAD/isocentric technique
we can use
1) TAR if dosimetry system is calibrated interms of dose to air (NDair) which is then be converted to NDwater (IAEA TRS 277)
2) TMR if dosimetry system is calibrated interms of dose water (NDwater) using IAEA TRS 398 
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AAPM Report Task Group 71 described recommendation for manual calculation of Monitor Unit calculation for photon and electron beams. In this report, d0 =10 cm (depth of normalization) used as reference depth. It not use dmax (depth of maximum dose) as reference depth.
How does if we do manual calculation of treatment time cobalt?
Is AAPM Report TG 71 applicable for calculation of treatment time cobalt-60 ?
can we use this recommendation?
if yes, how many reference depth that we must choose?
dmax
d0=10cm
or d0=5cm
thanks a lot for your discussion.
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For proper measurement of the beams, some reference depth other than depth of maximum dose is recommended. Although for Co-60, 4MV-9MV 5cm is sufficient and for energy more than 9MV, 10cm is recommended by IAEA and other protocols but we can use 10cm or more and the at Dmax for any beam, the calculated dose will be same.
dose at Dmax of Co-60= Measured dose at any depth x 100 / PDD at that depth
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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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I am beginner to FLUKA, need help to make geometry and run beam. 
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One way would be to use the DICOM import facility and to use a CT scan of the RANDO phantom. You could then use the CT numbers to define the material. As Piotr says, Flair is nice and can handle the DICOM import for you. The dose scoring can be done using standed scoring cards in FLUKA. One potential problem is the large number of voxels necessary to represent the phantom, but it should work.
I think the RANDO phantom would be a huge job to implement with combinatorial geometry. If you want to go that route, then maybe a simpler phantom would be necessary. 
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