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Radiation Protection - Science topic

Radiation protection is very important in medcine, industry, for radiologists, and reactor research.
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I want to simulate plasma focus devices with all geometry and surface, with specific filling gas and predict x-ray spectrum.
i want to measure this parameter for estimate x-ray radiation shielding.
please help me for this issue by attach similar code.
i need practice to understand more about this problem.
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I think you can get the pulse height spectrum with Tally f 8
And extract other relevant information from the shape of the spectrum.
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Dear All,
I'd like to share my experience with our manuscript (theoretical concept) reviewing process. Currently, we are studying radiation shielding materials, particularly glass substances. The effect of different oxides and various glass types is under our scope. In literature, almost all researchers are trying to understand the same scope, as well. However, our recent paper submission was reviewed, and the reviewer concluded that the impact of heavy metal oxides has already been known, and no need to evaluate the manuscript for a possible publication. However, our paper did not only have radiation protection but also physical, mechanical, and optical properties.
What I am trying to ask you all is that this is a fair decision within radiation shielding studies? Is it not the correct way to investigate different glass types, as well as various oxide contributions in terms of theoretical radiation shielding parameters?
Thanks in advance for your valuable comments.
With warm regards.
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Also i consider the reviewers just consider the shielding performance, but as new materials, the comprehensive proeprties are needed
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I am asking about the effectiveness and accuracy of using the film badge dosemeters to record and measure the exposure for the medical staff nowadays as it is a very old way, but it cost nothing to make such a system for developing the films and read it by a densitometer
and the only guidelines that I have to implement such a system is the (IAEA) Safety series NO.8 Vienna 1962 but it is a very old one and can't get any other new versions or similar guidelines, so is there any such recent guidelines
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To date, there are no newer safety guidelines (IAEA) series No. 8 Vienna, 1962. But you can still use such dosimeters.
Dosimeters with film badges for recording and measuring exposure have been replaced in many laboratories by more modern thermoluminescent dosimeters.
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I am recently fabricating some composites for gamma radiation shielding. However, it is hard to achieve a uniform sample since the particles always lie in the bottom even with the dispersing agent. I am wondering is there a big difference in the radiation shielding ability between the uniform and ununiform samples? Is there any equation or theory I can find?
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The efficiency of the shielding of given material composition shouldn't critically depend on the degree of nonuniformity of the sample, nor on the its inhomogeneity. In contrast, it surely strongly depends on two main parameters : (i) the sample thickness and (ii) the energy of the emitted gamma rays.
For further information, you may get benefit by consulting chapters from the Encyclopedia of Materials: Science and Technology (Second Edition) 2001, by C.K.Gupta in Science Direct : , and chapters from the 3rd Edition, 2003, like, e.g., the chapter Effective Shielding of Gamma Radiation in Field Conditions by José Anisiode Oliveira e Silva, https://doi.org/10.1016/B978-0-444-89791-6.50140-2.
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Dear all,
I'd like to ask about the colorization or decolorization of a radiation shielding glass material after exposure to the different energy levels. As far as I see from some literature studies, such kinds of radiation shielding glass materials are affected by the radiation source's dose and exposure time. However, I did not totally understand the mechanism. Actually, is there any concept that predicts these color changes? Or even, color change possibilities against radiation sources?
Hope that I am clear about what I am trying to explain to you all.
Warm regards.
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Dear Recep Kurtulus, your thoughts are on the right track!
The color and intensity of the color is determined not only by the presence of specific ions, but also by their combinations, for example, impurities of variable valence, which play the role of electron traps (acceptors) in radiation processes: Ce4+, Fe3+ and Eu3+.
For example, transitions, which change the valence and causes changes in the color of the glass: V3+ (green) + е- → V2+ (pink).
In the case of Ce4+, Ce5+, Sb 5+ and Pb2+ ions, the change in valence does not cause a change in the color of the glass: Ce4+ (colorless) + е- → Ce3+ (colorless).
In your case with iron, you need to know exactly which ions, what valence, in what quantities and in what combinations are involved in the coloring process. Accurate calculation is possible, but not so easy.
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Dear All,
The search for alternative radiation shielding glass materials is very popular, nowadays. The literature has a variety of investigations in terms of experimental, theoretical, and simulation. At this point, what will be the next for radiation shielding glass? In a commercial manner, there are some products having PbO content for the intended photon energy level, however, is it possible to produce commercial novel lead-free glass compositions in the near future? What are your opinions?
Best regards.
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From a commercial point of view, a new solution is only viable if consistent. So, even with up-and-coming dense materials already available, if they demand a relatively short replacement time due to accelerated aging, the solution is not ideal. Pb doped glasses shielding rest for a lifetime; Dense materials will compete if the lifespan achieves similarities in protection and durability.
Best regards
WNM
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Dear All,
Currently, I am trying to understand the lead equivalency subject. That is, I have data for my glass series in terms of commonly investigated radiation shielding parameters (i.e. linear attenuation coefficient), and I want to convert this data to lead equivalent thickness. Therefore, is there anyone who can help with this conversion? I can give some data about half-value layer thickness for my glass series as below:
Code-1: 2.60 cm
Code-2: 2.45 cm
Code-3: 2.21 cm
Thanks in advance.
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LAC*HVL=Constant (at specified energy)
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What are the general principles of radiation protection that suggested by (ICRP)?
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Page 88-89 of ICRP 103, Three Principles of radiological protection:
Two principles are source-related and apply in all exposure situations (planned, emergency, and existing exposure situations.)
1.The principle of justification: Any decision that alters the radiation exposure situation should do more good than harm.
2. The principle of optimisation of protection: The likelihood of incurring exposures, the number of people exposed, and the magnitude of their individual doses should all be kept as low as reasonably achievable, taking into account economic and societal factors.
One principle is individual-related and applies in planned exposure situations.
3. The principle of application of dose limits: The total dose to any individual from regulated sources in planned exposure situations other than medical exposure of patients should not exceed the appropriate limits recommended by the Commission.
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Hi everyone,
I need to achieve some radiation protection calculation with MCNP6.2 but my partner is using Geant4. Does somebody know a simple way to swap from Geant4 to MCNP6.2?
Kind regards
Lucas
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Dear Lukas,
There is no way to convert those inputs since both have different style TOTALLY!. To make it more clear, one is like apple while the other is like orange. From the input structure to used language, all the essential steps are totally different. If you are going to use both of them, you will need to learn the basics of input files. However, as a personal recommendation, try to focus on one of them.
Truly
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A comparison between the protection against coronavirus and radiation protection according to the ALARA principle, " The smallest possible exposure time, as far away from other people as possible (or 'as socially distanced as possible'), and wearing a mask whenever possible".
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Hello All,
I agree that there is a danger in analogy, specifically, false equivalences. However, one can still see some similarities between radiation protection and SARS-COV-2 virus mitigation modalities. For example, droplets and aerosols exhibit short and longer range propagation behaviors, respectively, in a similar way to direct ionizing radiations (alpha and beta particles, short range due to Coulomb interactions with matter) and indirect ionizing radiations (gamma rays and neutrons, longer range due to no direct Coulomb interactions). In the case of radiative contamination, such as dispersed radioactive materials like plutonium contamination, the safety procedures are similar to the SARS-COV-2 virus procedures: full hazmat suits and washing down of these suits. And surprisingly, COVID-19 also has long term effects on the human body, which is also a hallmark of acute radiation exposure, even after the person is removed from the source of the exposure. Long-haulers, as some COVID-19 suffers are called, continue to exhibit unusual blood work and symptoms such as fatigue and headaches for weeks, months, and perhaps years after exposure to the SAR-COV-2 virus even if the titer of virus in their blood is too low to be detectable.
Of course, the real difference between radiation and the virus is that, hopefully, there is a vaccine against the virus.
Regards,
Tom Cuff
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ICRP used critical group concept when assess public dose in the past.
However, recently ICRP replaced critical group concept to representative person concept when assess public dose.
ICRP defined representative person is critical group's average person, but I think these two are very similar term.
What is difference of critical group and representative person in practical appliance?
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The ICRP provides concept of a “representative person” in Publication 101 (free access to its publications of 1928-2017 at http://www.icrp.org/page.asp?id=5). In accordance with the ICRP, the ‘representative person’ is an individual receiving a dose that is rep­resentative of the more highly exposed individuals in the population. This term is the equivalent of, and replaces, ‘average member of the critical group’ described in previous ICRP recommendations. However, the ICRP does not provide technical details on how to assess dose to the representative person in practice.
The IAEA, as a party responsible to more than 160 Member States, provided a guidance on how to use concept of the representative person in emergency preparedness and response to radiation emergencies (see https://www.iaea.org/publications/10362/actions-to-protect-the-public-in-an-emergency-due-to-severe-conditions-at-a-light-water-reactor ).
In accordance with the IAEA, the representative person is a theoretical construct defined to represent the highest doses reasonably expected to be received by any member of the public during an emergency. In most cases, no one would be expected to receive a dose approaching that calculated for the representative person for the relevant exposure scenario.
The representative person does not represent any specific person from a particular age group but is a theoretical construct defined by a combination of the limiting dose factors, breathing and ingestion rates, habits, and other characteristics (e.g., typical behavior) that are drawn from different age groups. Therefore, the doses are representative of the highest doses reasonably expected to be received by any member of the public during an emergency for the considered exposure scenario. Taking response actions based on the dose projected or received by the representative person gives reasonable assurance that the actions will also be appropriate to any member of the public.
Dosimetric parameters for the representative person are given in Table. For external exposure in the environment, there is little variability in dose with age, and so the dose to the representative person is based upon (a) the external exposure of an adult man to air submersion and (b) the external exposure of a child to ground deposition to account for the fact that their organs are closer to the ground. For the intake of radionuclides, the situation is more complicated due to the strong age-dependence of the dose factors and intake rates. For inhalation, a dosimetric model for reference adult men characterized by adult breathing rates and dose factors provides a reasonable upper bound of the dose to all age groups. For ingestion of food, milk, and water, a dosimetric model for reference 1-y-old infants (1-2 y), characterized by infant consumption rates and dose factors, provides a reasonable upper bound for all age groups. For inadvertent ingestion of soil (e.g., eating with dirty hands), the intake rates typical of a child playing outside arc combined with the dose factors for an infant. A specific dosimetric model for the reference adult pregnant woman, which is characterized by adult breathing and consumption rates and adult dose factors, was used for estimating the doses to the fetus due to an intake of radioactive material by a pregnant woman.
The Table is attached in pdf.
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can anyone also tell me about the history of PDE?
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There are no conceptual or theoretical differences between PDE and ADE (they are similarly toxicological limits based on classical risk assessment), but comparing European guidelines with American Risk_MaPP guidelines, PICs, the calculation methodologies at these different guidelines have some little different variable and approach for applying the uncertainty/adjustment factors.
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Let us assume that Betelgeuse is about to go supernova in the next years. To what extent will this affect the dose rate of ionising radiation on ground level on Earth (in orders of magnitude)? How would it affect the dose rate if the gamma-ray burst was directed in the direction of Earth?
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Hi, Oliver,
I assume you're thinking of "cosmic radiation" as in "cosmic rays", ie. accelerated protons (and other nuclei). In that case, there should be an increase in the general background flux, but it would be quite a long time in the future (like ~ 1000 years), due to how the charged particles travel through the galactic medium.
An increase in CR flux does increase the ground-level ionizing radiation flux, mostly in muons, but the change depends on the primary CR particle energy.
Davide is correct that we would expect no significant effect on life here due to a Betelgeuse supernova.
If Betelgeuse had its rotation axis pointed at us then in principle it could produce a gamma-ray burst, which would be a much more threatening event. In that case the main threat would be in the form of atmospheric chemistry changes leading to severe ozone depletion, along with a short-lived (seconds) increase in surface radiation during the burst. As far as I know Betelgeuse is not oriented to produce a GRB from our perspective.
Here are some relevant papers:
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Dear Sir/Madam, I invited as a guest editor from high quality journals to handle special issues. If anyone can prepare a review similar to my review papers, particularly about a natural product in cancer prevention with focus on the structure activity relationship and mechanism of action, please kindly let me know to send an official letter. At this stage you should just send the title, authors and affiliation and abstract. Please kindly let me know as soon as you can. The suggested deadline for sending review is about 3 month. Best wishes, Suggeted topic: Genotoxicity of different agents and possble protection. Reducing side effects of radiotherapy and chomotherapy. Next generation of cancer therapy; Natural products. Natural products as novel therapeutic compounds. Radiation protection.
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What is names of the journals
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Radiosurgery, the therapy of brain tumors, has long been made using a so called Gamma Knife with high activity sealed sources such as Cobalt-60. Nowadays the therapy can also be made with a linear accelerator such as a Cyber Knife. What are your experiences? Can you share advantages and disadvantages of each system. At the end do you think that the use of radioactive sources is still (medically)justified for radiosurgery given the alternative of a Linear accelerator?
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Linear
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Radiation Protection
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I guess the questions is whether you are referring to only Radon exposure, or whether you are also thinking about other radio-isotopes that can be found in high concentration is unique settings. In the United States there was recently exposure to members of the public while touring a museum like facility, because a bucket of "rocks" was actually a bucket with a high amount of Uranium.
We have enough (naturally abundant) radio-active Potassium in our bodies to allow determination of lean-body mass by whole body counting. We are not going to be able to change that exposure, but building in areas of high Radon exposure requires consideration of building design and use of space (and venting) in the underground areas.
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Hello Thai friends! I would like to know how Thailand regulate electronic device radiation both for ionizing and non-ionizing radiation. The following information will be much appreciated:
1. How does the government of Thailand regulate or control electronic product radiation (ionizing and non-ionizing)?
2. What offices/agencies are responsible for such regulations?
3. What are the basis for radiation regulation? (Laws, Acts, etc.)
4. What is the scope of regulation? (Manufacturing? Export/Import? Possession? Use?)
5. What kind/type of devices does the government of Thailand regulate?
4. Are inspection or monitoring activities done as part of regulation? How is this implemented?
Thank you very much!
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very good
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I am modeling radiation by Fluent DO Model. But the number of wavebands which could be defined in Fluent is limited to 10. How can I add more wavebands?
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In this way , you can have it
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It is important for radiation protection
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There is a nice book on: Introduction to radiological physics and radiation dosimetry, by Frank H. Attix.
In eq. 2.4 he describes the Kerma ( kinetic energy released in material) as the energy fluence times the mass-energy absorption coefficient of the absorbing material. For homogenous media being sufficiently large the Kerma is equal to the absorbed energy dose.
Your energy fluence is calculated via the photon flux ( e.g. in photons per area and per time) times the photon energy. The mass energy absorption coeffcient(s) can be picked up by NIST:
For non-monochromatic radiation you have to go for an integral.
I will send you the relevant page(s) of that book via RG messenger.
Good luck
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The ICRP Publication 101 "Assessing Dose of the Representative Person for the Purpose of Radiation Protection of the Public" defines the deterministic and probabilistic approaches to find the dose of the Representative Person from doses incurred by population members. Are results for members of the group of different ages included in forming the average or the 95th percentile of the dose distribution, or shall they be treated separately? I am not sure, although the example for the probabilistic approach in B.7.2 of ICRP 101 would suggest the first case. Simply put, at the end of the analysis, is there only one Representative Person with one Dose to the Representative Person, or are there Representative Persons for each age group, e.g. Representative Infant (1 y), Representative Child (10 y), Representative Adults (adult).
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You are on the right track but you need to understand the logic of the situation.
The need to involve various groups depends on the expected exposure situation. If you e.g. are dealing with staff > 20 years from a minor local onsite release, you limit your group accordingly. Observe that if you are concerned with releases to the general public you might still be OK with the average groups.
You can check it yourself by making your own parallel calculations. You might find that one infant + one adult etc. for all the groups, exposed at one situation may give you the same detriment/person as one individual receiving the same dose distributed over a lifetime.
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The effect of mobile phone radiation on human health is a subject of interest and study worldwide, as a result of the enormous increase in mobile phone usage throughout the world. Several studies have linked the radiations to cancer using animal models like rats. How can we prevent this, if true?
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may be at long run or up-normal using due to the thermal effects
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  1. attenuation
  2. radioactive sources
  3. thickness need
  4. dose 
  5. risk assessment
  6. build up factor
  7. effective atomic number
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there are many and you can use the equations or some standard curves for every shielding material for software GEANT , Gorbit can be used
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Hello every one i want to simulate the different modes of heat transfer through cement Kiln. Conductive heat transfer from clinker to inside wall of cement kiln .Convective heat transfer to outside air from cement kiln and radioactive heat transfer from the cement kiln surface. I want to enquirer which ansys module should be most relevant in modelling clinker geometry and and such modes of heat transfer.
I had tried ansys fluent for and steady state thermal module but not sure which one is more releavant
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Use the P1 model for the radiation with solar loading model turned off. For the combustion analysis turn on the species transport model and use the premixed combustion to simplify your chemical analysis (in this case you will need to create a pdf for the 'importchemkilm'. All these are available in versions 18.0 -19.2
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Collective dose is defined in the ICRP Draft TG 79 report as the dose integrated over an interval for a set amount of time and for all individuals affected. Collective dose is intended for operational planning. The draft clearly proscribes use of collective dose for risk assessment and assigning detriment to a population. Collective dose requires the linear no-threshold risk model with no distinction for radiation type, internal or external, and dose rate.
The purpose of collective dose is for optiimization of radiation protection. Optimization is for total dose to a group, while radiation protection is for individuals. Is collective dose and optimization useful in your radiation program or is simply another regulatory hoop to jump through. Please give examples.
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Well, I agree with those who deem collective dose at least of secondary importance. The only situation where I think it makes sense to calculate the potential collective dose as a consequence of the operation of a nuclear or radiological facility in the future is the site selection process. Even if it is a rather uncertain and inaccurate measure of harmfulness of a source for this particular case it can help select the better solution giving advantage to a site with lower collective dose estimated in advance. No other applications would be desirable, I think, none of those connected to the dose limits and constraints of individuals.
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For All depleted uranium slabs, the beta dose rate at contact is 240 mR/hour. Any mathematical derivation or reason if you know. Please share.
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The answer is very simple. You should calculate the surface beta dose rate of slabs with different thickness. Due to the self-absorbtion, the contribution of the "back side source layers" is decreasing when the thickness of the beta-radiating slab is increasing.
Note: the use of R (röntgen) is not conform with the topical ISO standards.
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Is it acting like magnetism?
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The full answer will be the addition of those of Joseph and Thor
- alpha, beta and gamma rays from radioactive nuclides are generally not energetic enough to produce nuclear reactions with a radioactive product.
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Dear All,
I have used a radiameter which displays results in micro-sievert per hour in order to measure leakage radiation dose around jaws and I have obtained a value of 40 micro-sievert per hour just after finishing patient radiotherapy treatment. My question is what's the threshold value in which there is a potential radiation leakage problem.
Thanks,
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Hi,
Concerning leakage radiation and a linac, I would recommend that you take a look at the IEC "linac standard", IEC 60601-2-1, Medical electrical equipment – Part 2-1: Particular requirements for the basic safety and essential performance of electron accelerators in the range 1 MeV to 50 MeV, Edition 3.1 2014-07. There is a separate subclause on "LEAKAGE RADIATION through BEAM LIMITING DEVICES", describing requirements on absorbed dose levels and corresponding tests.
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Does magnetic field can heat the human body?
According to the International Commission on Non-Ionizing Radiation protection (ICNIRP), there is a maximum allowable magnetic field exposure to our body. I don't understand why magnetic field can be harmful to our body. I understand that electric field can heat our body by ionic polarization and dipole rotation of an atom or molecules. Whereas, magnetic field can only heat magnetic materials.
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Most organs have almost the same relative permeability value as that of the free space (Ping-Ping Ding et al, IEEE Transactions on Magnetics, 2014; J. Malmivuo and R. Plonsey, Bioelectromagnetism, 1995). One factor influence the magnetic filed is the magnetic volume susceptibility of the tissue, but it is often very small although non-zero (Schenck JF, Progress in biophysics and molecular biology, 2005). Some non-zero susceptibility tissues have the thermal effect in alternative field. It also depends on the frequency of the field.
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Although the principles and safety regulations for radiation protection in general is the same as when applied for dentistry specific. However, the applications in dentistry is very unique to a certain extend. I would love to see that the dental world and the medical worlds collaborate to a greater extend in this regard. I am of the opinion that most oral health care workers needs to improve their knowledge, attitudes and practices in this regard.
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workers in dental X-ray equipment must have a basic knowledge of the inherent health risks associated with radiation and must have demonstrated familiarity with basic rules of radiation safety. Dental personnel must not expose any individual to the useful beam for training or demonstration purposes. In a rooms where X-ray equipment is used, post a sign (that may include the radiation symbol) stating: CAUTION X-RAYS. Dental X-ray equipment and imaging software should be operated only by individuals adequately instructed in safe operating procedures.
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Dear All,
I would like to know what are the measuring instruments used by physicist in nuclear medicine in order to perform some tasks as imaging quality assurance, equipments quality assurance, radiation protection, .....
Regards,
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nuclear medicine technology
certification board equipment list
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X-ray with energy of 40-100 kVp is radiated to shields with thickness of 1mm(shape of sample is sphere with radius of 10 Cm). Distance of source from detector is 100cm.  
Schematic set-up of  X-ray attenuation property measurement;
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Dear Gerhard,
Thanks a lot for your response,
Best regards!
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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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Why and How oxygen plays a role of a radiation sensitizer?
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The oxygen fixation hypothesis (OFH), developed in the late 1950s, is widely regarded as the most satisfactory explanation of why oxygen is a radiation sensitizer. Central to this hypothesis is the explicit belief that DNA lesions (originally called "target lesions") that are produced by x-rays with the chemical participation of oxygen pose a special threat to cell survival because these lesions cannot be chemically restored to an undamaged state. According to this hypothesis, oxygen sensitizes because these "nonrestorable" lesions ultimately increase the amount of stable DNA damage--and thus the extent of lethality--from a give dose.
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 I need this book or any  recent copy of  it, or any reference have the relation   of calculating  the  absorbed gamma dose  (D= Γ x A /d2) or any other dose.
the book: F. H. Attix, "Introduction to Radiological Physics and Radiation Dosimetry," New York, John Willy & Sons, (1986).
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For whole body monitoring thermo luminescence  TL, OSL Film RPL etc are used at chest level or so. Of  which material phosphor your dosimeter is made and what is frequency of use? 
What does it measure Hp(x)?
What type of radiations it can measure other than photons?
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In Romania  the Regulatory Body authorizes the systems of individual dosimetry and appoints the individual dosimetric services approved in compliance with the national regulations. The systems of dosimetry are used with the purpose of evaluating the individual penetrating dose equivalent, H(p)(10) and the individual superficial dose equivalent, H(p)(0,07).
The following types of dosimetric systems may be used with the purpose of the individual monitoring provided that they meet the requirements of these norms:
a) systems of dosimetry with film;
b) thermoluminiscent systems of dosimetry
c) electronic digital systems of dosimetry
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Area of human body is mass/age dependent. In Radiation Protection, it can (area ) in general be used to convert skin dose to whole body dose in case of non-uniform (shallow) exposure or  if some portion of skin receives shallow dose. What are mathematical models which can be used to estimate are area?
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Just to add that wT for skin is based on the nominal risk for non-melanoma skin cancer. Radiation-induced skin cancer is almost exclusively of the basal cell type. Melanoma skin cancer is a typical example where no clear evidence is available for association with radiation exposures (other examples include testicular cancer and chronic lymphocytic leukemia).
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This is a radiological protection question...thank you!
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As mentioned above, the equivalent dose limits are currently recommended only for the lens of the eye, the skin, the hands and feet. Historically, however, ICRP (or its predecessor IXRPC) had recommended specific “dose limits” for “blood forming organs”, “gonads”, “thyroid”, “bone”, “hands and forearms”, “feet and ankles”, and “head and neck”.
Precisely speaking, the term “dose limits” has been used for workers since the 1977 recommendations (in Publication 26), and for the public since the 1966 recommendations (in Publication 9), until which time the terms “tolerance dose” assuming thresholds for all radiation effects and “maximum permissible dose” assuming no thresholds for all radiation effects had been used. It was the 1977 recommendations that recommended dividing all radiation effects into tissue reactions (called nonstochastic effects at that time though) with a threshold, and stochastic effects with no thresholds, for which equivalent dose limits and effective dose limits are recommended.
For more details of such historical changes, please see the following paper (especially, Supplementary Tables 7 and 8).
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A lead shielding (µ=1.19) is used to protection against a source of caesium 157 with dose rate of 0.11 R/h. emitted gamma radiation has a 0.6 Mev energy. If lead shielding thickness is 0.02 m, what will the permissible exposure time (stay time)?
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It sounds like you have 0.11 R/hr, then you add 2 cm Pb shielding.  So you will calculate a new dose rate after the shielding.  Using 2 cm Pb, and Cs-137 gamma (0.66 MeV), I get a reduction of about 0.14: That is, the 0.11 rem/hr dose rate would be reduced to 0.14 of this, or about 0.016 mrem/hr.  In order to get a stay time, you need an allowable dose. If you assume 0.1 Rem (100 mrem or 1 mSv), then your stay time would be 0.1 rem / 0.016 rem/hr, or about 6 hours.
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I have access to a x-ray irradiator but not a gamma ray irradiator.
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Dear Julia,
without knowing your irradiator, a correct answer is hard to give. Maybe you can list some parameters of your X-ray irradiator, such as kVp (range of adjustable kVp), filtration and dose rate (range of possible dose rate).
As other researchers mentioned before, gammas and X-rays are in principle the same: they are photons with different origin (electron shell vs. nucleus). But, depending on energy and dose rate, the biological effect due to irradiation can vary. For example, gamma irradiators are mostly equipped with Co-60 or Cs-137, producing "high" energy photons (1173/1332 keV or 662 keV). So, substitution with "high" energy X-rays (150-300 kVp with thick filtration) should result in approx. the same biological effect using the same dose (keyword: RBE = relative biological effectiveness). However, substitution with "low" energy X-rays (<= 50 kVp) will not be appropriate, in my opinion, due to an increased RBE. Variation of dose rate can also result in different biological outcome because of time-dependent cell repair mechanisms.
Best regards
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I'm currently working with plants but I have a bigger interest in Astrobiology. I'm looking for someone who works in both areas or that is interested in this. I thought about a project to study radiation protection using medicinal plants. Can anyone help me? 
Thank you!
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Thank you Arvind, I will
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ICRP has continuously decreased dose limit from 150 mSv to 50mSv to 20 mSv per year in past. This was based on various studies mainly data of atomic bomb survivors. Since the risk information is almost complete from the study of atomic bomb survivors it is less likely that present risk factor of 5%/Sv is going to change. In view of this what could be the future changes in annual (average) dose limit of 20mSv . Will ICRP make it 10 or 30 mSv after 2035 or so?
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 Thank you Dr Peter Hill for raising this important point.
Comparisons of risks with other "safe" non-radiation occupations have been discussed in paragraphs 96–102 of ICRP Publication 26 and paragraphs C70-C77 in ICRP Publication 60. ICRP believes that, in assessing the implication of effective dose limits, the calculated rate at which fatal malignancies might be induced by occupational exposure to radiation should in any case not exceed the occupational fatality rate of industries recognized as having high standards of safety. ICRP then assumes that an annual occupational fatality probability of l0–3 might be taken as a reference risk for the effective dose limit, assuming that, in “safe” non-radiation occupations, the average annual fatality rate was about 100 per million workers and that subgroups with high risks might run a risk ten times the average (i.e., 10–3).
So, if risk in other "safe" occupations changes, a reference risk for dose limit may change.
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Are you measuring low energies from Electromagnetic waves emitted by human body and measuring high energy gamma rays emitted  from radionuclides in the human body. Did you identify the radionuclides? Does it vary with  age? Body weight?  Geographical location? Race, ethnicity, religion, diet, etc.?
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K-40, H-3, Na-22, C-14 are isotopes of essential elements and for all practical purposes, their concentration in the human body is determined unambigeously by a) the amount of K, H, Na and C in the body, and b) the isotopic ratios of these radioisotopes. Be-7, U, Th and their decay products may be influenced by dietary factors and might be enriched or depleted in the body due to different factors. Strong smokers might accumulate Ra decay products from tobacco, meat of deer is also enriched in Ra decay products with respect to other meat because plants that are growing slower in Northern regiins, accumulating more Ra. Meat from wild boars in Europe might still show high Cs levels from the Chernobyl accident (boar is enriching Cs-137 from mushrooms they eat), so regular eaters of boar meat will have also higher Cs-137  levels although it is excremented relatively fast, etc.
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I admire your research. I consider it very important. I have some general questions:
The DNA damage of lymphocytes conclude to cancer?
How do you measure the DNA damage?
The exposure to low-dose radiation from the uranium mining is more dangerous for children or the age does not matter? I know a city that is close to a lignite mine and many children have leukemia.
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Dose-response evidence is limited for an association between uranium exposure with subsequent leukemia. Please see paragraph 255 in the Annex D of the UNSCEAR 2016 Report, which (1) is dedicated to the biological effects of internal exposures due to uranium, (2) was published on 8 February 2017, and (3) is freely downloadable from
 
There are two more recent papers on leukemia risk in uranium miners.
[1] In German uranium miners, a positive but non-significant dose-response was observed for mortality from non-CLL (chronic lymphatic leukemia), and there was a significant increased mortality risk from chronic myeloid leukemia (non-CLL) but not from CLL ( https://www.ncbi.nlm.nih.gov/pubmed/27815431 ).
[2] In Canadian uranium miners, no association was observed for leukemia with increasing cumulative exposure to radon ( https://www.ncbi.nlm.nih.gov/pubmed/27651479 ).
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 Meaning glasses industry we can see ionizing radiation
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Radiation interacting with glasses will scintillate. And, there will be some materials you can make glasses from that will scintillate more strongly.
However, you will get one pulse of light for each gamma ray or beta particle interacting with the glasses. Given the small mass of glasses these interactions would be rare, except in situations where radiation dose rates would be so high you wouldn't want to be there anyway. I can't see any means by which these would be visible against ambient light - ie: wearing the glasses would only show the radiation if the room was dark, or you were wearing a hood to block out other light. Being able to see radiation but nothing else wouldn't seem to be useful. 
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i want use the tld with rando phanton in the chest and abdomen 
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A short tip to tld in randso. Drill small cylindrical holes, enough to put a small cylindrical pmma tube with a central hole for cylindrical tlds. You will get commercial hints by PTW in germany Freiburg.
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I am looking for an appropriate FDG PET SUVmax threshold to use. Can anybody provide me with some ideas?
Thank You. 
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SUV of 2.5 and for that matter any other value cannot be adopted as a threshold for distinguishing malignant lesions from benign abnormalities,  this is a terrible misconception and should be ignored completely, the size of the lesion and respiratory motion are very important factors in such measurements and the proponents of this threshold were totally misinformed about the physics of PET and their ignorance has resulted in mishandling many patients with cancer all over the world.    This has been a tragic addition to the literature and it is about time to abandon it for good,  we won't miss it for a second!!!
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My question concerns charcoal conditioned in Ukraine from the hardwood (oak and horn). In the "Specifications", it is described as having a Radionuclides (Radioactivity) - max 400 Bq / kg. I have not idea about the existence of any norms. I tried to find out on the IAEA site but found no data on the subject.
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I suspect, that the <400 Bq/kg is probably some regulatory limit. Such limits result form legislation, so in your case, Ukranian legislation, and, if it's imported, it has to be in accordance with EU regulations.
To derive such limits, the most precise way would be to calculate (for the radionuclides of concern) all possible paths (inhalation, ingestion, direct radiation); and a common way would be to demonstrate that  all theses path do typically sum up to a dose less than 0.001 mSv per year which is well below the natural background (de minimis principle).
However, just to set it into relation: charcoal (as well as any organic matter) has an activity of about 200 Bq/kg of natural C-14 resulting from nuclear reactions with cosmic radiation. So as  a first estimation, 400 Bq is just the double of the natural background, so it will be negligible (roughly spoken).    
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Due to the interaction of such radiation with matter, a small charge will be formed on the material surface. This charge must be safely leak out; therefore such shielding material must have good electrical properties.
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Materials become charged when electrons leave the surface of the material and when electrons from photo and Compton reactions enter the material. These external photons can be substantial for photons above 1 MeV. Hot cell windows irradiated with gamma only can shatter if no leakage method is provided.
Most counting laboratories do not have a problem with shield charging when non-conductive materials are included. Charge build up is easily leaked away as dose to the material is usually low. Nevertheless, surprises do occur in experimental configurations, particularly when there is low humidity. 
Most charging problems are seen in accelerators when targets build up substantial charges. See attached from an electron accelerator.
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1.       Can someone tell me, why there is no radiation emit from Helmholtz experiment? Is that because of low voltage? Can it produce the X-ray if this conduct this experiment with high voltage? Say 30 kV?
1.a. The electrons is interact with the glass (Pyrex glass),
2.       In X-Ray Tube, in order for electrons to escape into the partial vacuum, high voltage and heat are needed.
o   High Voltage: To accelerate the electrons
o   Heat (Filament): To emit the electrons by thermionic emission
o   High Current: To produce more electron hence the intensity will become higher (more electron means more count). That why in XRD application, if the intensity from the diffractogram is not high (within noise border), we can increase the mA to get the exact intensity and peak.
Correct me if I’m wrong.
3.       In FESEM, Electron Gun, the same concept was used, however, the Primary Electron Beam does not interact with the target material, for example, Titanium, Tungsten, Copper, etc. However, the Xray only can be produced if the beam is hitting the sample in the chamber. Is that correct?
4.       All in all, the Xray only can be produced, if the medium is high vacuum,  high voltage, heat (filament), and the electron must need to interact with matter?
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On your picture not FESEM (field emission SEM) displayed but thermo emission SEM. In FESEM electrons are extracted from filament without heat, by field. So, heat is not necessary.
Actually, SEM and X-ray tube have the same method of X-ray production. Voltages for SEM and tube are similar (for sem usually up to 30 kV, in some cases 50 kV). So they produce X-rays of exactly the same energy and should be shielded at all times. X-ray tube works quite often on Cu or Fe X-rays, which are 7-9 keV,  and they are very harmful for human beings.  Due to great variation of specimens in SEM it often produces X-rays of higher energy compared to typical X-ray tube X-rays. 
As for Helmholtz  coils, you can find elements that can emit X-rays of energy less than 200 eV (Be). So, if you put some Be behind coils you'll get its X-rays, but they can be stopped even by paper.  
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This question addresses specific issues associated with development of shielding optimization methodologies to accurately evaluate radiation protection strategies while accounting for the inherent multiple uncertainties associated with the problem. For most shield design efforts, effective dose for specific GCR or SPE radiation design environments has been calculated in order to determine the effectiveness of different shielding configurations to meet specfic vehicle radiation requirements.
NOTE: Question is related to shielding approaches of Spacecraft in Space .
e.g Water 
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For physical shielding hydrogen is the best material on a mass basis as it maximises the stopping power by maximising the number of target electrons per unit mass. It also minimises the production of secondaries such as neutrons. Water, fuel and polythene are the practical ways of employing hydrogen. On ISS approximately half the dose is from protons in the South Atlantic Anomaly and half from cosmic rays. They both produce significant secondary neutron fluxes. The SAA protons are very anisotropic and are shielded by structures . I suspect that the orientation of ISS during the SAA passes is the key issue rather than the attachment of the Shuttle which can afford additional shielding only over a small solid angle.
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According to my theory on cell response to radiation exposure, I got a weird simulation result of cell survival fraction curve. It is not like the traditional exponential-like curve, instead, it has a surprising lower region within a sort of exponential-like curve.
Is there anybody actually has ever observed cell survival fraction curve in actual radiation biology experiment like this one?
Some basic simulation information as below:
The dose listed in the figure was the total dose, but it is accumulated dose in 7 days.
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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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chernobyl nuclear radiation
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First of all, the pure (external) exposure of wood to nuclear radiation causes no problem as no radioactivity is generated in the wood by this radiation. However, wood from high contaminated areas will have a certain uptake of radionuclides, especially  Cs-137, which will still be present today.
There are estimations in IAEA Tecdoc1376 about different scenarios using wood from the most contaminated areas in the Chernobyl region (see link). Some of these scenarios can cause exposures above radiological permissible limits (table XVII). However, these scenarios  comprise mosty the use of wood ash for ferilisation and/or an extensive use of wood for building etc.
For all practical reasons, a single piece of wood from the Chernobyl region won't cause harmful doses, even at the highest radioactivity levels for wood given in table XVI. A piece of wood, let's say, one kg with an activity of 10000 Bq/kg  at a distance of 10 cm will roughly cause a dose of 0.1 mSv per year.  However, depending on its exact activity, you might be obliged to handle it within a controlled area. 
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There are several material which light in weight and transparent also they won't allow cosmic radiation to penetrate to it. I'm looking for that kind of material.
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In fact there are hardly any neutrons or gamma rays in primary cosmic rays incident on the top of the atmosphere. However they are produced within the atmosphere in a cascade of secondary particles that include everything known to man and probably many that are not. Primary cosmic rays are mainly energetic protons plus all heavier ions. For these, hydrogen is definitely the best shielding material. Use of fuel and water is proposed for interplanetary travel while polythene is used on the ISS.. If you are considering secondary particles near the earth's surface, these are mainly neutrons and mu-mesons. Hydrogen is again optimum buit in practice concrete and soil are used.
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There are many kinds of research going on the topic of radiation shielding for sustainable Human habitat in space. This is one of the important factors for the sustainable human life in space. So I have a doubt on choosing a suitable material for radiation shielding in outer space habitat which should protect human from both solar and cosmic radiation. 
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Detailed information on such materials are available at:
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Dear Moradi,
I am providing below the HVL calculation for all energies for typical absorber materials used for gamma ray attenuation
Refer to Handbook of Chemisty and Physics (Google search), latest Edition, 2016. Here you find table of mass attenuation coefficients of important elements for energies from 10 keV to 1 GeV. Taking lead (Z=82) as the attenuating element, we find:
µ/ρ for 1MeV = 7.1 x 10-2 cm2/gm;
µ/ρ for 1GeV = 11.5 x 10-2 cm2/gm;
Now HVL = 0.693/µ, where µ is the linear attenuation coefficient and µ/ρ is the mass attenuation coefficient.
Hence HVL in lead (ρ, density of lead = 11.34) for 1MeV is (0.693)/( 7.1 x 10-2 x 11.34) = 0.861 cm of Pb or 8.61 mm;
And  HVL in lead (ρ, density of lead = 11.34) for 1GeV is (0.693)/( 11.5 x 10-2 x 11.34) = 0.531 cm of Pb or 5.31mm.
The TVL (tenth value layer) values can be calculated using the relation, TVL = 3.32 HVL
As for your other question: how many HVL values are required, it depends on the radiation dose equivalent rate level at a particular point of the beam and the transmitted level desired after attenuation by the absorber (Remember, the radiation attenuation is exponential and no amount of absorber thickness will stop the beam completely). One HVL will reduce the dose rate level by one-half; one TVL will reduce the level by one-tenth.
We will take an illustrative example. A 1 MeV source (close to a Co-60 source, with an average energy of 1.25 MeV) is emitting gamma radiation whose dose equivalent rate is 1Sv/hr (roughly an air kerma rate of 100 rads/hr) at a specified distance from the source. A barrier of lead is required so that the dose equivalent rate on the other side of the barrier is safe for an occupational worker (20 mSv/yr or 10 µSv/hr, assuming the total working hours/year is 50 weeks/yr and 40 hrs/week). How many TVLs and HVLs are required?
1 TVL reduces the dose equivalent rate by a factor of 10
Therefore, 5 TVLs reduce the dose equivalent rate by a factor of 100000, i.e., the dose rate reduces to 10 µSv/hr
So the thickness of lead barrier required = 5 x 8.61 x 3.32 = 14.3 cm (about 17 HVLs)
The above computation assumes narrow beam geometry, for which the mass attenuation coefficients are given in the Handbook mentioned above.  In practice, there will be a lot of scatter from the absorber,etc. for a broad beam for which build-up corrections are required.
Note: I did not elaborate on the mathematics of the above computations because they are available in any book on radiation protection.
I hope the above will be of help. Please give your feedback.
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Consider a solid sphere of some material and a spherical shell of the same material with both having the same thickness in the sense that a straight-line path from outside to the center travels through the same thickness of material. These are intended to be used as radiation shields in a given external electron environment. The interest here is the ionizing radiation dose at the center of each. Because electrons do not follow straight-line paths when going through a shield, I am not surprised that the solid sphere and spherical shell will not produce identical doses at the center, but how much different should they be? And why? Is there some simple analytical argument (not Monte Carlo) that can roughly account for this difference?
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Thank you for the help but my question is not whether the spherical shell is the worst case. My question is whether there is a simple qualitative explanation as to why it is the worst case.
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Ionising radiation protection material, to incorporate natural fiber (plant-fiber).
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To block radiations, the key point is to use materials containing elements with high atomic number. that's why lead (Pb) is common in most radiation protection materials. however, it will all depends on the type and energy of the radiation you are trying to shield. type: is it particles (beta/alpha/neutron etc) or non-particles (xrays/gamma). non particle radiation can be easily stop with high atomic number elements. but with particle radiation, it is somewhat tricky ie. beta needs to be shield by both high and low atomic number elements. for energy: you need to improve the thickness of the materials. in the end, im not sure what kind of plant fiber you are using as i think most of the high atomic number elements are more abundant in ore minerals. but im not sure about this tho, since im not so familiar with plant fibers. hv you done elemental analysis to see what elements is in there? i think plants are mostly organic hence atomic numbers should be low.
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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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Assume that we would like to "measure" the background radiation level (e.g. dose rate), but we have a low radioactive source or a well-shielded medium source in the area/room. If we subtract the estimated/calculated dose rate given by such sources from the actual measured values, can we call the final result a "measured" value? or because we did some calculations (in this case, simple subtraction) on the values they cannot be considered as measured anymore?
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Of course!  Some of the radiation will originate in your detectors.  The detectors' responses need to be taken into account in detail.
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Will there be any composition change after irradiated with gamma rays ?
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Gamma interactions will alter the film chemically and physically. The amount of change will depend on the dose. Nevertheless, the film should have the same shielding properties as un irradiated PVA film as the number and distribution of the atoms will be the same.  You might have problems if the material is not structurally sound, Very large doses can cause plastics to crumble and flake.
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The methods of Partitioning and Transmutation (P & T) of spent nuclear fuel from Nuclear power plants are currently under research and development. Partitioning aims at Separation of the fuel in uranium, the actinide Plutonium (Pu), Neptunium (Np), Americium (Am) and Curium (Cm) and the fission and activation products, Transmutation on the conversion of Plutonium and of the minor actinides Neptunium, Americium and Curium in short-lived fission products. These methods provide the possibility that not only the total activity in the final repository for heat-generating waste decreases over time faster, but also their radiotoxicity.
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The application of P&T would, if fully implemented, result in a significant decrease in the transuranic inventory to be disposed of in geologic repositories. Currently, it is believed that the inventory and radiotoxicity can be reduced by a factor of 100 to 200 and that the time scale required for the radiotoxicity to reach reference levels (natural uranium) will be reduced from over 100 000 years to between 1000 and 5000 years. To achieve these results it is believed that it would be necessary for plutonium and neptunium to be multiple recycled and for americium (curium) to be incinerated in a single deep burn step.
Partitioning and Transmutation (P&T) is a complex technology which implies the availability of advanced reprocessing plants, facilities for fuel fabrication of TRUs and irradiation facilities beyond the present NPP facilities.
Partitioning is to a certain extent a broadening to other nuclides of the current reprocessing technique which has been operating at industrial level for several decades and for which the main facilities, at least in Europe and Japan, exist or can easily be extrapolated from present day nuclear plants. Partitioning is the technology which can be considered as a form of "super - reprocessing" by which, in addition to U, Pu and Iodine (I-129) also the Minor Actinides (MA) and the Long lived Fission products (LLFP: Tc-99, Zr-93, Cs-135, Pd-107 and Se-79) would be extracted from the liquid high level waste.
Transmutation requires fully new fuel fabrication plants and irradiation technologies which are to be developed and implemented on industrial scale. The existing NPP's could in principle be used for transmutation but many practical obstacles may arise: e.g. interference with daily operation of the plants, core safety considerations and lack of transmutation or incineration yields. New irradiation facilities: dedicated Fast Reactors (FR), Accelerator driven transmutation devices and even Fusion Reactors have been proposed for transmutation and "incineration" purposes.
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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 completed a construction of Radiation Shielding Concrete (basement and semi-basement) for Oncology treatment ( LINAC, Cyclotron and Cyberknife). I'd like to set guidelines and/or recommendations for future use of the facility. 
Construction materials:
Concrete - density is maintained 2.5 gm/cc, slump < 75 
CA - 2/3 in and 1/2 in stone chips
Admixture- Masterpolyheed
Cement - PPC
Ice was introduced to reduce hydration temperature
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I recommend you review the recommendations contained in the report by the National Council on Radiation  Protection and Measurements: Report No. 151 - Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities (2005).  Their website is: http://ncrponline.org/
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Some patients with head and neck cancers treated by radiotherapy with or without chemotherapy presents with an unacceptable morbidity of severe dry mouth, severe mucositis and, extreme toxicity to the neck skin, whereas, in some patient’s this does not happen with the same dosage and technique. This often leads to treatment interruption in many cases. So, it will be of interest to know and predict which patients are likely to develop these side effects of radiotherapy and thus, we can help improve the quality of life of patients by a predictive method.
What life style behaviors and biologic parameters that should be included to predict the risk of extreme radiation toxicity in patients receiving radiotherapy for head and neck cancers ?
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Some aspects influencing this scenario from our experience:
-          Older patients develop these complications (probably a large study group will enable you to underline an age threshold)
-          Body mass index is in direct distribution with the length and aggressive pathology, thus a patient with a normal BMI is less likely to develop these complications (maybe search for some data in research targeting nutrition problems in head and neck carcinoma)
-          Associated pathology such as diabetes predispose to these problems
However as a previous scientist answered it is very hard to standardize the data because ideally every patient should have a tailored radiation regimen so you need a very big lot of patients to draw conclusions on p<0.00001.
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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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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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We have small diffusion chambers from aluminum, inside this chamber a Ra-226 source is installed with a normal emanation coefficient. Do we need to isolate this box with a roof using a good isolation glue? 
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To seal my samples, I use high density polyethylene (HDPE) bottles with screw cap of the same material. You can see this in the attached presentation (page 6) to work: "NATURAL OCCURRING RADIONUCLIDES IN NOVEL SAND BEACHES FROM ESPÍRITO SANTO, STATE, BRAZIL" on my profile. In a similar to your experiment, I used a PVC tape on the threads and, after threaded, sealed with silicone. For more information, contact me at raquino@ipen.br
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Energy?
Power?
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Impact of the power?
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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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Why is it better than micro-nano-materials in low-energy photon is absorbed, but the size effect disappears with increasing energy?
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Hi Salman, I don't  understand your question very well. can you explain more?
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I am planning to conduct FMEA study for some of the existing equipment/device.This device has several models, most of them have almost same design features. I wish to find out the weak parts/sub component of these device type by conducting the study.
Since number of different models of this device are available in the market, it is not possible to conduct study on all the models.
Whether it would be appropriate to conduct FMEA study on specific model and conclude the results for the whole species of the devices?
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As long as the different models are similar in functions, it is more efficient to maintain a single FMEA to cover them. Within the single FMEA, it is possible to denote some failure modes as specific only to Model X.  Or you could add an extra column to input which are the applicable equipment for each line item.
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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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Is there any data available for potential exposure (#unexpected exposure) in the occupation of radiation facilities( non-nuclear facilities) like industrial radiography, food irradiation facility, radiotherapy etc. 
I want to compare my data with some of the study resulting in the probability of potential exposure in above facilities. OR  please give reference for the data for probability of accident in above facilities.
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