Accelerator Radiation Safety Newsletter
(all articles are to be considered personal/professional in nature and do not reflect the opinions of the institutions described unless otherwise stated)
An Official Publication of the
Health Physics Society's Accelerator Section
Quarter 2015 /
From The Secretary/Treasurer:
1990 was a great year. Although the Health Physics Society (HPS) Accelerator Section was still a year away from its 1991 founding, several of our section’s most illustrious members joined the Society in 1990. So it is that in 2015, section members Bob May, Scott Schwahn, and Marcia Torres will celebrate their 25th anniversary with the Society.
All of these folks have served on the Accelerator Section leadership team in various positions at various times. Without them and other long-time members, the section would not be the strong and vibrant organization it is today. If you see Bob, Scott, or Marcia at the annual meeting in Indianapolis, please congratulate and thank them for their continued support of the section.
But this brings up a question: how many section members are NOT HPS members? And why aren’t they? If this is you, please consider all the benefits of joining the Society (blatantly plagiarized from the HPS website):
What are you waiting for? Join the HPS today, and in 25 years you too will be part of Accelerator Section history! To join, see http://hps.org/aboutthesociety/howtojoin.html .
FROM THE CORRESPONDENTS
From Lorraine Day, CAMD Louisiana State University
I would like to personally thank the accelerator section for sponsoring my nomination for the Dade Moeller lectureship. This is quite an honor and I am blessed to have been considered. Though the HPS annual meeting is but a short distance away; I am still putting on some finishing touches for the plenary. I look forward to meeting with all my accelerator colleagues.
Every year; we honor select students with the Lutz Moritz and Wade Patterson awards. A group of Accelerator Section members gets together to review the ones they would like to hear more about. One of our nominated papers will not be able to make the short presentation as he has both military and family obligations before the HPS. This year; Jason Harris; Rich Brey; Reg Ronningen; Steve Frey and I acted as preliminary judges. Some of them will also be on hand for the final judging on the Sunday evening – just prior to the opening reception. Should other members wish to participate, we will be in the Capitol 1 room at the Westin from 4:30 until 6
After the selection has been completed, we will award the 2 plaques (along with some $$$) to the winners on Tuesday morning at the end of the Accelerator session. That evening; the winner will also be recognized at the awards banquet by Past-President Dr. Darrell Fisher. Please come out and support these deserving students and don’t forget to congratulate them! See you there.
CONTRIBUTION FROM FERMILAB
From J. Donald Cossairt, 630-840-3465
The purpose of this course is to instruct students in the radiation physics of accelerators, the regulatory framework in which the associated hazards must be controlled, and the elements of management needed to provide a climate for successful programmatic implementation. The intended audience for this course includes accelerator physicists, engineers, and radiation protection professionals who desire to improve their knowledge of accelerator radiation physics. If completed in accordance with the rules of the USPAS, this course carries 1.5 semester-hours of graduate credit. An application for continuing education credits will be made to the American Academy of Health Physics for this course. Such credits have been granted in the past.
Hazard Analysis and Decision Making
covers the theory and practical steps required to justify, design, and manage
personnel and machines to ensure safety. The course addresses both context
and content of complex issues associated with computer based safety systems.
This session will also discuss emerging risks such as cyber security as well
the latest design approaches.
From Pradyot Chowdbury at the
Canadian Light Source,
Production of Mo-99 during Commissioning Operation of a 40 kW 35 MeV Electron Linac: Approach to a Novel Pilot Scheme
Pradyot Chowdhury*, Mo Benmerrouche†, Mark de Jong, William Diamond, Hao Zhang, James dela Cruz and Michael James
Canadian Light Source Inc., 44 Inovation Boulevard, Saskatoon, SK S7N 2V3, CANADA
The production of 99Mo during commissioning operation of a high power electron linac with 40 kW, 35 MeV was carried out. By irradiating natural Mo at 10 kW, the measured saturation activity was found to be 80.75 GBq, that is 90% of the maximum estimated value. The electrons bombarded on a heavy metal converter generating Bremsstrahlung photons that undergo photonuclear reaction 100Mo(γ,n)99Mo with a Mo Target placed in the forward direction. Gamma spectroscopy detected 46Sc and 64Cu as long-lived activated products from Ti and Cu, respectively. Converters and targets are water cooled, and the radiation protection due to Bremsstrahlung and neutrons are achieved using iron, lead, polyethylene, concrete and earth as shielding materials. Monte Carlo simulations are performed with FLUKA to calculate the dose at the Converter, Target, Beam Dump and Shielding structures, as well as independent dose profiles for the electron, gamma and neutron.
In contrast to diagnostic radiology, nuclear medicine is used to image organ function and structure. It may be used to gather important medical information that may otherwise be unavailable, require surgery, or require more expensive diagnostic tests. One typically uses radioactive isotopes, generally called radiopharmaceuticals, which emit gamma rays that can be detected externally by gamma-cameras. The information is measured by multi-detector array of cameras placed in multiple locations and the data acquired with computer systems to convert the gamma ray generated signals into images producing information about the area of the body being examined (known as computed tomography) .
An acute shortage of 99Mo-99mTc since the first quarter of 2010 has not only impacted patient access to care, but has created an uncertainty about the future of nuclear medicine. An interruption at any point in the production, transport or delivery of 99Mo-99mTc generators would have substantial impacts on patient care. It was felt necessary to study alternative routes to produce 99Mo in order to ensure a constant reliable supply.
The Canadian Light Source (CLS) has undertaken a project funded by the Non-reactor based Isotope Supply contribution Program (NISP) to demonstrate the possibility of producing 99Mo radioisotopes using a linear electron accelerator. Licenses for this Class II Isotope Production Accelerator (Medical Isotope Production) by Canadian Light Source Incorporation have been obtained since 2011 from the Canadian Nuclear Safety Commission, Ottawa, ON . The design, construction and commissioning of the pilot 99Mo production facility at the CLS is based on a high-power linac producing 99Mo through a photo-nuclear reaction with the 100Mo. The pilot 99Mo facility at the CLS comprises a 35 MeV 40 kW electron linear accelerator, a water-cooled converter to produce a high flux of Bremsstrahlung radiation and a water-cooled target assembly with molybdenum targets, either natural molybdenum or enriched 100Mo for irradiation, and a remote handling system for transferring irradiated targets from the radiological controlled area into a shielded transport container.
The technique of linear accelerator based radioactive isotope production has not been exploited so far as most of the commercially available electron linear accelerators that are used for radiation therapy or industrial applications, work at energies below the threshold for photonuclear reactions. Recently, commercially available electron linac of energy and power suitable for the production of isotopes have become available that can produce large photon flux at a reasonable cost. One such electron linac that we have recently purchased is being used in our research and development program  for the production of 99Mo. The most frequently used medical isotope in Canada is 99mTc performing about 5500 scans each day. While 99mTc has a half-life of 6 hours, which is derived from their parents, 99Mo has a half-life of 66 hours. A solution containing 99Mo is milked on a 24 hour cycle to recover the 99mTc, when its activity reaches 94% of the saturation activity.
The special features of photonuclear production of isotope lies in relatively low reaction cross section and a great transport length of bremsstrahlung photons in a substance. These features restrict the radionuclide yields in both the gross and specific activity . At the same time, great ionization losses of heavy particles in the target quickly remove them from the resonance region. Therefore, in some cases the output of useful products in the photonuclear channel appears even higher than in the use of heavy particle beam .
In nuclear production of radionuclides for medicine with the use of heavy particles (n, p, ions), the region of nuclear reaction occurrence is limited mainly by the region of interaction between the primary particles and the target. However, in photonuclear production the delocalization of this region occurs due to the incorporation of an additional target, i.e., bremsstrahlung accelerator with the parameters typical of photonuclear production (> 20 MeV, ≥ 10 kW). Photo-neutrons may exert a substantial effect on the composition of the isotope product produced. Besides, the predominant yield of (γ, n) reactions limits the possibilities of obtaining a carrier-free isotope product. A separate problem in the process is the removal of heat from the converter and the target during their interaction with a concentrated high-power electron flux.
In view of these peculiarities, our results of isotope production at the electron accelerator would be utilized at the initial stage to optimize the production technology.
This report examines Mo-99 isotopes that can be produced by an electron accelerator and provides a quantitative estimate of yields; the concerns are discussed such as radiation shielding design, hazard analyses and their mitigations, calculated radiation shielding using IAEA technical document # 188, and FLUKA simulated Monte Carlo for dose and fluence calculations. The thematic of the process is presented pictorially in Figure 1, where the beam of electrons with a 35 MeV, 40 kW power hitting 4 water-cooled Tantalum converter plates. The colored yellow lines show high energy X-rays that are generated from the converter moving in the forward direction and irradiating the 18 water-cooled Mo-100 plates, where the desired isotope Mo-99 is produced. Finally, the gamma spectroscopy of the activated materials that are produced during the irradiation and the measured yield of Mo-99 are reported.
RESULTS AND DISCUSSION
Photonuclear Production of 99Mo
The parameters of the electron accelerator that was commercially manufactured for the MIP project has been used as the basis for further calculations in this paper for the production of 99Mo from the photonuclear reaction with 100Mo . This accelerator produces 35 MeV electrons at up to 40 kW or just over one mA average current (about 7 x 1015 e/s). Shvetson  has developed an effective photon yield equation (1) in the energy window of 8 to 20 MeV which overlaps with much of the photonuclear cross section (especially for heavier nuclei) given by:
Using the value of 0.25 photons per electron (0.22 at 30 MeV from equation-1 and adjusted by the amount calculated in  for 35 MeV compared to 30 MeV) in the energy window of 8 to 20 MeV times this quantity of electrons leads to a useful bremsstrahlung yield in the forward direction of about 1.6 x 1015 photons/s. A large fraction of this will be in a small cone with an area of less than one cm2 at the entrance of the isotope target and 1.5 to 2 cm2 at the exit. This is a very high photon flux and demonstrates why a potentially high isotope yield can be obtained with photonuclear reactions with typical cross sections in the range of 15 to 300 mb.
Figure 2 shows the photonuclear cross section  of the 100Mo(γ,n)99Mo reaction, with a threshold of 9 MeV and a maximum cross section of 150 mb at 14.5 MeV . These cross sections are generally lower than many of the charged particle cross sections such as (p,n) or (d,n) used to produce isotopes with a cyclotron. However, the cost of producing large quantities of photons is relatively low and the photons can penetrate windows with little power loss making it practical to separate the electron source from a converter target that converts the electrons into bremsstrahlung and further subdivide the isotope target from the converter target with another window. The figure also exhibits the typical Bremsstrahlung photon spectra with 20- and 35-MeV e-beams. The Bremsstrahlung is strongly focused into a forward cone  of a few degrees half-width, so some physical spacing between the converter and isotope target is practical. This photonuclear reaction is expected to produce Mo-99 yield and activity as given in the table-1 below:
Table-1: An estimation of the Mo-99 yield 
Radiation Shielding Concerns
The dose rate calculated using IAEA document “Technical Report Series – 188” , for Bremsstrahlung from a thick Tantalum converter in the forward direction is 4 x 105 Sv/h at one meter, and in the perpendicular direction 3 x 103 Sv/h at one meter. Assuming that the 35 MeV, 40 kW electron beam is stopped entirely in a thick heavy target, the amount of neutron yield would be about 5 x 1013 n/s. To keep the dose rate in the occupied areas as low as reasonably achievable, the number of tenth value layer of shielding required in the perpendicular direction are 8.3 for Bremsstrahlung and 6 for the neutrons, respectively. The shielding is achieved by using the following materials: iron, lead, polyethylene, concrete and earth.
The basic requirements for our facility are given by the physical size of the accelerator and target assembly, the shielding requirements and the space for personnel to do maintenance inside the facility. The radiation produced by an electron beam consists of a very intense photon field from the bremsstrahlung and a neutron field which is produced by photonuclear reactions. The neutron field will be comparable to that produced by a high-powered cyclotron while the photon field is much more intense. Figure 3 shows the photon field (rad/h at one m) per kW of electron beam power at zero and 90 degrees with respect to the beam direction for electron energies up to 100 MeV .
At 35 MeV and zero degrees with respect to the beam direction the dose rate is approximately: R0 = 1 x 106 rads/h/kW (1 x 104 Sv/h/kW) at one metre times 40 kW of electron-beam power to produce a photon field of 4 x 107 rads/h (4 x 105 Sv/h) at one metre. At 90ş the radiation dose rate is: R90 = 7500 rads/h/kW (75 Sv/h/kW) at one metre times 40 kW to produce a photon field of 3 x 105 rads/h (3 x 103 Sv/h) at one metre.
The total neutron yield has been estimated by using a conservative assumption that the 35 MeV, 40 kW electron beam is stopped entirely in a thick heavy target such as tantalum or lead. From Table XV in reference 9, the thick target neutron yield of a tantalum target is 1.2×1012 n/s/kW at ~ 35 MeV. This is multiplied by 40 kW to produce a neutron yield of: Yneutron ~ 5 x 1013 n/s
For a position outside the facility, a distance of 3 to 5 m from the target, a reduction of the bremsstrahlung field of between seven and ten decades are required, depending on the location and occupancy requirements.
For bremsstrahlung, the primary barrier transmission factor is determined using equation 2.1 in NCRP-151 :
Bpri = P(dpri)2 / WUT (2)
Where P is the shielding goal (expressed as dose equivalent) beyond the barrier, dpri = distance from the source to the dose point, W = photon dose at one m from the source, U = fraction of the workload that the beam is directed at the barrier in question, and T = occupancy factor at the dose point.
The number of 1/10th-values of shielding required:
n = - log (Bpri) (3)
Shielding barrier thicknesses have been evaluated based on ALARA with individual occupational doses not to exceed 1 mSv per year. Dose to individual members of the public is unlikely to exceed 50 µSv per year. Using these recommendations as the shielding goal (P) and a yearly operation of 2000 hours (U) leads to exposure rates of: Controlled areas 0.5 µSv/h, and Non-controlled areas 0.025 µSv/h. To keep the dose rate in the public occupied areas ALARA, the number of tenth value layer of shielding required in the perpendicular direction are 8.3 for Bremsstrahlung and 6 for the neutrons, respectively. The shielding is achieved by using the following materials: iron, lead, polyethylene, concrete and earth.
The cooling water at the converter and target as well as the room air is expected to be activated and produce ozone and hydrogen. The expected radioactive gases produced in air are 15O, 13N, and 41Ar, and in water are 15O, 11C, 7Be, and 3H (tritium). Adequate precautions are taken to mitigate these hazards. The tritium generated in the cooling water for the converter and target after 100 kW-hour of MIP Linac operation was found to be only marginal. Similarly, there was no Be-7 in the converter and target cooling water, nor any ozone production in the room air could be observed during this early phase of commissioning. However, we have experienced an elevated level of radiation from the converter and target holder material, where titanium has undergone nuclear reaction 48Ti(γ, pn)46Sc generating Sc-46, which emits two cascading gamma photons ~1 MeV detected by gamma spectroscopy, with a longer half-life of 83.8 days. The gamma spectra were recorded with a NaI scintillation spectrometer, GR-135. The Ti is being replaced by Cu at the target holder. The Cu also produced two short-lived 61Cu and 62Cu, and one long-lived 64Cu isotopes that decay by positron emission, which annihilates by emitting two photons at 511 keV. Figure 4 shows the gamma spectroscopy of Sc-46 and Cu-64 that were recorded after a few days of irradiation, when the majority of the short-lived isotopes decayed. Figure 5 shows the overall decay of the radiation level after the linac was turned off. The corresponding table-2 of isotopes included in the figure would suggest that a few short-lived and long-lived isotopes would be responsible for the two major decay half-lives that were derived from the decay curve as 6.42 minutes and 3 hours, respectively.
Estimating Activity by Direct Measurement
The measured Mo-99 activity and yield was estimated by using the relation A = Ḋ/Γ where
A = the activity (MBq)
Ḋ = the measured dose rate at 1 m (mSv/h), background subtracted; this measurement was made as the targets are placed in the shipping container.
Γ = the specific gamma ray dose constant = 3.052 x 10-5 (mSv/h)/MBq for 99Mo .
After irradiating natural molybdenum at 10 kW with the 35 MeV electron beam for 20 hours followed by a cooling period of 3.5 hours, the Mo-99 activity was measured at a distance of 1 m as 0.45 mSv/h. The measured dose rate corresponds to a Mo-99 activity as 14.75 GBq.
The end of bombardment activity (EOB) is expected to be
AEOB = As (1- exp (-λ t1)) (4)
where As is the saturation activity
λ is the decay constant for the calculated isotope, where λ = ln2/T1/2. T1/2 is the half-life of the isotope.
t1 is the irradiation period, i.e., 20 hours in present case.
During the cooling period, the decay would reduce the EOB activity exponentially, thus the measured activity would be:
Ameasured = AEOB (exp (-λ t2)) (5)
where Ameasured is the measured activity after the cooling period t2. Using the Mo-99 measured activity as 14.75 GBq, the As is computed to be 80.75 GBq, i.e. 8.1 GBq/kW. At the same rate of production, a 24-hour irradiation of natural Mo at 40 kW would have produced 72 GBq of Mo-99 activity. By comparing with the estimated 24 hours irradiation of natural molybdenum at a 40 kW power (Table-1) that would have generated 80 GBq, which produced about 90% of the theoretically estimated activity expected from the natural molybdenum. Thus, further tuning up of the beam alignment with the converter and target would improve the Mo-99 yield.
Monte Carlo Simulation with Fluka
FLUKA Simulation Code
FLUKA is a general-purpose tool for performing calculations describing particle transport and interactions with matter covering an extended range of applications from particle accelerator shielding to target design . The energy range covered by FLUKA is very wide and the code can transport photons and electrons over approximately 12 energy decades, from 1 PeV down to 1 keV . The electromagnetic part is fully coupled with the hadron sector, including the low-energy (i.e. <20 MeV) neutrons. The simulation of the electromagnetic cascade in FLUKA is very accurate, including the Landou–Pomeranchuk–Migdal effect and special treatment of the tip of the bremsstrahlung spectrum. Electron pairs and bremsstrahlung are sampled from the proper double differential energy-angular distributions, improving the common practice of using average angles . In FLUKA, the full set of Seltzer and Berger cross sections of accurate electron–nucleus and electron–electron bremsstrahlung has been tabulated in an extended form . The photonuclear interaction in FLUKA is enhanced from time to time [12, 14, and 15]. It can handle the photonuclear interactions over the entire energy range. Theoretically, the total bremsstrahlung photon yield is explained by the Bethe–Heitler equation . In this work, the bremsstrahlung photons are generated after the converter Ta materials bombarded with electrons at energy of 35 MeV. The geometry of the system used in FLUKA is shown in Figure 6. We assumed that an electron pencil beam travels in the +z direction and hits the converter target at normal incidence. The simulation has been performed with a statistics of 15 million electrons with beam energy of 35 MeV.
The geometry of the experimental facility contain a diamond window from which the electrons emerge and proceed to the converter block that contains four Tantalum converter plates surrounded by cooling water, where the bremsstrahlung photons are generated. The bremsstrahlung photons then hit the target block that contains 18 molybdenum target circular plates each 1 mm thick immersed in circulated cooling water and where the photonuclear reaction occurs. Finally, the non-attenuated photons pass through a vacuum region and are absorbed in the beam dump generating heat. All the above-mentioned components, as well as the bulk shielding material surrounding the inner core that includes iron and polyethylene shielding are depicted in the four quadrants of the Figure 6. While the second quadrant reveals the converter and target blocks, the third quadrant shows the beam dump. The color code is used to clarify the visualization: blue is used for water, brown for beam dump, dark brown for iron shielding and green for polyethylene. The outer surface of the beam dump is also water cooled. These four geometries are used for particle transport and calculating the resulting doses and energy deposition at various regions. Monte Carlo simulations are performed with FLUKA to calculate the dose at the converter, target, beam dump and shielding structures, as well as independent dose profiles for the electron, gamma and neutron in Figures 7-9. Also, we obtained the fluence map of the electron beam entrance at the diamond window and converter exit. The dose deposited in the beam dump is depicted in Figure 10. Finally, table-3 shows the fraction of total energy deposited in the various components of the materials that are penetrated by electrons and bremsstrahlung. The same table also includes results from our earlier Monte Carlo MCNP5 calculation for the converter, target, vacuum container and beam dump components only.
Hazards Associated with individual substances
The Medical Isotopes Production (MIP) facility has several potential hazards due to radioisotopes, ozone, or hydrogen gas produced in the cooling water or room air. We have demonstrated that these products would cause a negligible hazard taking into account the engineering and administrative controls that were developed and implemented to mitigate the hazard.
In the air, the main products of interest are 15O, 13N, 41A and ozone (O3). In the water they are 11C, 15O, 7Be, 3H (tritium) and hydrogen (H2). 11C, 15O, 13N, 7Be, and 41Ar are gamma emitters; tritium is a weak beta emitter that is considered hazardous only when it is ingested, whereas ozone is a highly reactive gas.
An additional hazard is the possibility of a sulfur hexafluoride (SF6) leak from the waveguides. Sulfur hexafluoride is much heavier than air and would create oxygen depletion in the room air causing asphyxiation.
Cooling Water Hazards Analysis
We calculated the activities of various elements for the Mo-99 production target water system with a 2.1 cm water path length as listed in Table-4 which shows the results of the calculated yield for all of the entries at 40 kW of electron beam power (in column 4). Column 5 shows the calculation of As x (1- exp (-λt=2000 h)) for each isotope - that will be representing the build-up in each year of operation for the various isotopes. Most of the active species except tritium and Be-7 will reach saturation in a few hours and decay as quickly once the beam is turned off. The volume of water in the closed system is about 42 US gallons or about 1.6 x 105 cm3. The dose rate at one meter from the center of the tank has been calculated to be 11.4 mSv/h. Both the O-15 and O-14 are positron emitters and should produce about the same dose rate as N-13. Tritium is a low energy beta emitter and therefore produces no external radiation exposure.
The same isotopes will be produced in the converter target system, but the saturation activities will be 50% smaller because of the shorter path length of the beam in water. The cooling water system that is used to cool the copper beam stop will have much smaller activities because of the greater distance and off-axis angle. Only 11C, 15O, 7Be, and tritium are considered because of the short half-lives and low production rates of the others. 15O is produced by the 16O(γ,n)15O reaction, the others are produced by spallation of 16O. In addition hydrogen will be produced by dissociation of the water.
The converter and target cooling systems will be confined totally within the MIP accelerator room.
We have initiated commissioning the Medical Isotope production system with an electron-linac of 35 MeV beam energy and is designed to operate with a maximum power of 40 kW. Issues of providing adequate radiation shielding and containment of the hazards due to activated products are presented. Converters and targets are water cooled and the radiation protection due to Bremsstrahlung and neutrons are achieved using iron, lead, polyethylene, concrete and earth as shielding materials. Monte Carlo simulations are performed with FLUKA code to generate the dose profiles of electron, gamma and neutron, as well as to obtain the dose at the Converter, Target, Beam Dump and Shielding structures.
The commissioning production of 99Mo was carried out by irradiating natural Mo at 10 kW, where the measured saturation activity was found to be 90% of the theoretically estimated value. The electrons that bombarded on a heavy metal converter generate Bremsstrahlung photons that undergo photonuclear reaction 100Mo (γ,n) 99Mo with a Mo Target placed in the forward direction. Gamma spectroscopy detected 46Sc and 64Cu as long-lived activated products from Ti and Cu, respectively. The radiation decay profile that was obtained after the linac turned off showed a fast decay component with half-life 6.42 minutes and a slow decay with a half-life of 3 hours.
Funding for the Medical Isotope Projects (MIP) from Natural Resources Canada for the Non-reactor based Isotope Supply contribution Program (NISP) and the Provincial Government of Saskatchewan are acknowledged. We acknowledge the help of Darin Street with the gamma spectrometer, GR-135.
*Corresponding Author: Email: Pradyot.email@example.com, Tel: 306-657-3850, Fax: 306-657-3535
†Current address: Brookhaven National Laboratory, Photon Sciences Division, Upton, NY, 11973, USA.
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Table-3: The distribution in percentage of total power deposited in various regions/materials, and that is compared with MCNP calculation.
Table-4: The produced nuclides in the cooling water, their half-life and expected dose with the
linac operation at 35 MeV and 40 kW .
Figure 1 – Pictorial Overview of Mo-99 Isotope Production 
Figure 2 – Photonuclear x-section of 100Mo and Bremsstrahlung spectra for 20- and 35-MeV
electron beam, reference 
Figure 3 – Bremsstrahlung generated radiation in the forward and sideways directions with
respect to the electron beam direction 
Figure 4: Gamma Spectroscopy of the Target Block and Converter Holder
Table-2 Short-lived Isotopes and their Decay half-life
Figure 5: Gamma Decay after End of Irradiation due to the tabled Short-Lived Isotopes
Figure 6: Four views of the geometry used for Fluka Monte Carlo simulation.
Figure 7: Electron Dose Profile with Stainless Steel and Polyethylene shielding
Figure 8: Gamma Dose Profile with Stainless Steel and Polyethylene shielding
Figure 9: Neutron Dose Profile with Stainless Steel and Polyethylene shielding
Figure 10: Copper Beam Dump with 2D Energy Deposition Profile