Health Phys.Abstracts，Volume 118,Number 6

DETECTIONANDANALYSISOFUNCHARGEDPARTICLESUSINGCONSUMER-GRADECCDS

John A.Cummings1,James W.Deaton2,Charles T.Hess2,Samuel T.Hess2

(1.Department of Physics and Astronomy,University of Maine,120 Bennett Hall,Orono,ME 04469-5709)

Abstract:In the current climate of increased global terrorism,the threat of a radiological incident is becoming more realistic than ever,and as such,the necessity of early-warning detection is paramount to national security.To assist with this need,we have investigated the detection of uncharged particle emissions from radiological sources using charged-coupled devices (CCDs),which are contained within a variety of products,including consumer cellphones and traffic cameras.Because the CCD is intrinsically sensitive to charge accumulation as a result of linear energy transfer by the incident particles,each event can be counted and quantified using video-image processing and an estimated energy band assessed by the properties of the pixels.In an effort to make this process applicable to the widest possible range of CCDs available,this experiment was conducted using low-quality CCDs contained within consumer-grade,budget web cameras.Within a239Pu-Be neutron howitzer,particles were detected using several camera models:Gigaware X76,Z76 and Logitech C170,C270.Particle detection events were counted by post-processing with Matlab,and an efficiency for each CCD was determined relative to both a theoretical flux model and a calibrated3He tube detector.The relative detection efficiencies for the cameras tested fell within the range 14%-18% and showed a linear correlation between incident energy and pixel response.

Keywords:

Health Phys.118(6):583-592;2020

ASSESSMENTOFPATIENTRADIATIONDOSEDURINGCONVENTIONALDIAGNOSTICX-RAYEXAMINATIONSINTHREEPUBLICHOSPITALSINNORTHERNJORDANUSINGTLDS

Zaid Q.Ababneh1,2,Anas M.Ababneh1,Majd S.Bani Hani1,Riad S.Ababneh1,Khaled M.Aljarrah2,3

(1.Physics Department,Faculty of Science,Yarmouk University,Irbid,Jordan;2.College of Applied Medical Sciences,King Saud bin Abdulaziz University for Health Sciences,Al-Ahsa,Saudi Arabia;3.Physics Department,Jordan University of Science and Technology,Irbid,Jordan)

Abstract:This study aims to measure entrance surface doses during routine chest and abdomen X-ray examinations of adult and child patients.Radiation dose measurements were performed using thermoluminescent dosimeters TLD-100s in three major public hospitals in northern Jordan on a total of 100 patients.Wide variations in entrance surface doses were observed within and between hospitals,which might be attributed to significant variations of the selected exposure parameters.For adult patients,the results have shown that the majority of entrance surface dose values from both chest and abdomen examinations were within recommended values of diagnostic reference levels.For child patients,the mean entrance surface dose from chest examinations in three age groups were 0.131 mGy (0-1 y),0.136 mGy (1-5 y),and 0.191 mGy (5-10 y).These values were considered relatively high compared to the European reference levels and published results in the literature.However,for abdomen examinations,entrance surface dose values were relatively lower than European reference levels.Patient effective doses were estimated using a PCXMC 2.0 Monte Carlo program.The results for both adults and children were found to be relatively lower than the values reported by international publications.Due to the wide variations of entrance surface dose and the higher radiation doses delivered to child patients,this study recommends implementing a quality assurance program in such hospitals to achieve optimization between good image quality and minimum dose according to the as low as reasonably achievable principle.Moreover,the results of this work will provide a useful base for establishing local diagnostic reference levels for chest and abdomen examinations in Jordan.

Keywords:dose assessment;dose equivalent,effective;dosimetry,thermoluminescent;radiation,medical

Health Phys.118(6):593-599;2020

NEUTRONANDPHOTONDOSERATESINAD-TNEUTRONGENERATORFACILITY:MCNPSIMULATIONSANDEXPERIMENTS

Xu Xu1,Chang Yi1,Tang Wanyue1,Sun Yuanming1,Lu Jingbin1,Liu Yumin1,Zhao Long1,Liu Jiaxi1,Li Xiaoyi1

(1.College of Physics,Jilin University,Changchun,China)

Abstract:The deuterium-tritium neutron generator is a common neutron source for fast neutron activation analysis.The 14.1 MeV neutrons emitted from a deuterium-tritium neutron generator are difficult to shield due to their strong penetrability and the induced secondary gamma rays in the shield.A rough calculation based on attenuation factors shows that when 14.1 MeV neutrons with a yield of 1 × 108ns-1penetrate the designed shielding layers,which consist of a 0.5-m-thick concrete layer and a 0.5-m-thick water layer,the neutron ambient dose equivalent rate is 2.48 μSv h-1.A geometric model of a neutron shielding room is constructed based on the calculation.Monte Carlo simulations indicate that the highest neutron ambient dose equivalent rate outside the neutron shielding room is 0.73 μSv h-1,and the neutron ambient dose equivalent rate at the detector position in the shielding room is 2.12 μSv h-1.The experimental results show that the highest neutron ambient dose equivalent rate outside the neutron shielding room is 1.43 μSv h-1,and the neutron ambient dose equivalent rate at the detector position inside the shielding corridor is 2.74 μSv h-1.Comparative investigations show that the experimental results are basically consistent with the results of the Monte Carlo simulations,except for some positions with large proportions of fast neutrons where it is too difficult for the neutron dose equivalent meter to provide reliable values.Moreover,the radiation dose rate outside the designed shielding room is lower than the occupational exposure dose limit,which is in line with the design expectations.Finally,the gamma spectrum at the position of the gamma detectors is measured by a high-purity germanium detector.The analyzed results show that many secondary gamma rays are generated by the interaction of neutrons with the shield materials and detector probe crystals,and some gamma rays are produced from natural background radionuclides such as40K,208Tl,212Bi,214Bi,212Pb,214Pb,and228Ac.

Keywords:dose equivalent,effective;gamma radiation;neutron activation;radiation protection

Health Phys.118(6):600-608;2020

REDUCTIONOFOCCUPATIONALEXPOSUREUSINGANOVELTUNGSTEN-CONTAININGRUBBERSHIELDININTERVENTIONALRADIOLOGY

Kenta Kijima1,Anchali Krisanachinda2,Mikoto Tamura1,Hajime Monzen1,Yasumasa Nishimura3

(1.Department of Medical Physics,Graduate School of Medical Sciences,Kindai University,Osaka,Japan;2.Department of Radiology,Faculty of Medicine,Chulalongkorn University,Bangkok,Thailand;3.Department of Radiation Oncology,Faculty of Medicine,Kindai University,Osaka,Japan)

Abstract:This study investigates whether a novel tungsten-containing rubber shield could be used as substitute shielding material in interventional radiology to reduce the occupational exposure of operators to scattered radiation from a patient.The tungsten-containing rubber is a lead-free radiation-shielding material that contains as much as 90% tungsten powder by weight.Air kerma rates of scattered radiation from solid-plate phantoms,simulating a patient,were measured with a semiconductor dosimeter at the height of the operator’s eye (1,600 mm from the floor),chest (1,300 mm),waist (1,000 mm),and knee (600 mm) with and without tungsten-containing rubber shielding (1-5 mm thickness).The tungsten-containing rubber and a commercial shielding material (RADPAD) were affixed onto the phantom on the operator’s side,and reductions in air kerma rates were compared.Reduction rates for tungsten-containing rubber shielding with thicknesses of 1,2,3,4,and 5 mm at each height level were as follows:70.37 ± 0.40%,72.17 ± 0.29%,72.95 ± 0.31%,72.58 ± 0.35%,and 73.63 ± 0.63% at eye level;76.36 ± 0.19%,77.13 ± 0.10%,77.36 ± 0.14%,77.62 ± 0.25%,and 77.66 ± 0.14% at chest level;67.78 ± 0.31%,68.12 ± 0.19%,68.88 ± 0.28%,68.97 ± 0.14%,and 68.85 ± 0.45% at waist level;and 0.14 ± 0.94%,0.72 ± 0.56%,1.08 ± 0.74%,1.77 ± 0.80%,and 1.79 ± 1.82% at knee level,respectively.Reduction rates with RADPAD were 61.80 ± 0.67%,60.33 ± 0.61%,64.70 ± 0.25%,and 0.14 ± 0.66% at eye,chest,waist,and knee levels,respectively.The shielding ability of the 1 mm tungsten-containing rubber was superior to that of RADPAD.The tungsten-containing rubber could be employed to minimize an operator’s radiation exposure instead of the commercial shielding material in interventional radiology.

Keywords:exposure,occupational;exposure,personnel;diagnostic radiology;radiation protection

Health Phys.118(6):609-614;2020

ESTIMATIONOFEXTERNALDOSERATESTOHOTELWORKERSFROMBEDLINENSCONTAMINATEDBY131IPATIENTS

C.Foreman1,S.Dewji1

(1.Department of Nuclear Engineering,Center for Nuclear Security Science and Policy Initiatives,Texas A&M University,College Station,TX 77843)

Abstract:Iodine-131 is commonly used in medical diagnosis and therapy for patients with hyperthyroidism or differentiated thyroid cancer.Following treatment,patients may recuperate in a hotel room to avoid exposing family members.The main purpose of this study was to estimate external effective dose rate coefficients to a hotel worker who handles potentially contaminated bed linens due to secretions from131I patients as sweat or urine.The external dose rate estimates were derived using Monte Carlo radiation transport code and the phantom with movable arms and legs to model a housekeeper standing in an upright position holding a pile of bed linens.Simulations further integrated the body burden of time-dependent biokinetic metabolism of131I in the patient’s body,differentiating between biokinetic excretion models of hyperthyroid vs.cancer patients.Organ absorbed dose rate and effective dose rate coefficients were calculated for three scenarios of bed linen contamination and estimated out to 5 d postadministration and compared to past131I patient contamination measurements.

Keywords:131I;contamination,external;nuclear medicine;radiation therapy

Health Phys.118(6):615-622;2020

EFFECTOFTEFLONTRANSMITTANCEONSENSITIVITYOFTHERMOLUMINESCENCEDOSIMETERCARDS

V.B.Podobedov1,A.Romanyukha2,C.C.Miller1,A.Hoy2

(1.National Institute of Standards and Technology,Gaithersburg,MD;2.Naval Dosimetry Center,Bethesda,MD)

Abstract:Thermoluminescence dosimeter cards purchased by the US Navy in recent years have different radiation sensitivities,e.g.,they exhibit a different amount of light per dose unit.Presented tests indicate that the optical transparency of the Teflon encapsulation is partially responsible for the significant variation of the DT-702/PD radiation sensitivity.It was confirmed also that the Teflon transparency is in fact a primary cause of the radiation sensitivity increase in the most recently produced dosimetric cards.This conclusion is based on the correlation found between the calibrated radiation sensitivity of the dosimeter card element and the optical transparency of its Teflon encapsulation.The transparency measurements were performed at the wavelength of 400 nm within a 10 nm spectral interval effectively covering the spectral range of the thermoluminescence.It is anticipated that the experimentally determined correlation will help to approve the acceptance of new thermoluminescence dosimeter cards in the Naval Dosimetry Center inventory as well as improve the production process.

Keywords:calibration;dosimetry,personnel;dosimetry,external;dosimetry,thermoluminescence

Health Phys.118(6):623-628;2020

DIFFERENCEDISTRIBUTIONSAPPLICABLETOCERTAINHEALTHPHYSICSMEASUREMENTS

Alan L.Justus1

(1.Los Alamos National Laboratory,Los Alamos,NM)

Abstract:This paper discusses calculational methods for the determination of the difference distributions associated with certain health physics measurements.These measurements include the check-source response counts relative to an initial reference count,the Albatross (i.e.,HPI model 2080B) neutron tube counts relative to the gamma tube counts,and those that involve the use of the automatic background subtraction feature of portable health physics instrumentation.Examples are provided that illustrate the methods for a few specific measurements.For the comparison of a daily source count to its previously determined reference value,minimum counts for various scenarios were presented in order to reliably meet required tolerance limits of ±10% and ±20%.In either case,it was found beneficial that the initial reference readings be established using a counting interval of longer length than the daily interval.For the comparison of Albatross neutron counts to the gamma counts,it was seen that the relative error in the difference distribution was still related to that of the parent distribution.It was seen,therefore,that an effective way of reducing the gamma influence on the Albatross was to increase the counting interval used,hence yielding a significantly larger mean count per interval.For the automatic background subtraction feature,it was noted that net count values near 0 counts would almost always have the negative values of the difference distribution truncated to 0 counts by commercially available off-the-shelf instrumentation,whereas significant net count values would be displayed correctly but with a larger associated variance than the gross count itself.This paper therefore also provides a technical basis for the necessary source strength of a check source in order to meet daily limits,the gamma field limitations of the HPI 2080B Albatross,as well as the consequences of automatic background subtraction.

Keywords:analysis,statistical;gamma radiation;instrumentation;radiation protection

Health Phys.118(6):629-646;2020

EVALUATIONOFSKYSHINEFROMANACCELERATORFACILITY:DEPENDENCEONDISTANCEANDANGLE

Taiee Ted Liang1,James C.Liu1,Sayed H.Rokni1

(1.SLAC National Accelerator Laboratory,Menlo Park,CA)

Abstract:Neutron skyshine from Linac Coherent Light Source II 4 GeV electron beam operation at SLAC National Accelerator Laboratory can contribute to prompt radiation exposure to the public at distances far beyond the accelerator tunnel housing.One of the shielding design requirements at SLAC is that the annual dose to a member of the public is no more than 0.05 mSv y-1.This study uses Monte Carlo code FLUKA to simulate the generation of neutrons from 4 GeV electron beam losses on a thick copper target inside a generalized geometry of the Linac Coherent Light Source II Beam Transport Hall accelerator tunnel section.The effective dose from neutron skyshine was characterized as a function of both distance from the tunnel wall (up to 1 km away) and angle relative to the beam direction (between 0° and 180°).This new methodology for evaluating neutron skyshine dose is applicable to high-energy GeV-range electron accelerator facilities both at SLAC and elsewhere.

Keywords:accelerators;dose,equivalent;Monte Carlo;neutrons

Health Phys.118(6):647-655;2020

DISCOVERYOFRADIOCESIUM-BEARINGPARTICLESINMASKSWORNBYMEMBERSOFTHEPUBLICINFUKUSHIMAINSPRING2013

Shogo Higaki1,Yuichi Kurihara2,Yoshio Takahashi2

(1.Isotope Science Center,University of Tokyo,Tokyo,Japan;2.School of Science,University of Tokyo,Tokyo,Japan)

Abstract:To investigate the publics’ internal exposure by inhalation of radiocesium from the Fukushima Daiichi nuclear disaster in 2011,we examined the activity of radiocesium and radiocesium-bearing particles adhering to nonwoven fabric masks worn daily by members of the public in spring 2013 and 2014.We found a maximum cumulative137Cs activity of 4.58 ± 0.15 Bq in 4 wk of spring 2013,which is 20.8% of the activity measured for the same subject in spring 2012 using the same method.This decrease was faster than the physical decay of radiocesium.Radiocesium was detected in 21 of 722 masks in 2013;three of these included type A radiocesium-bearing particles.The activity ratio of the radiocesium-bearing particles on the mask to the total radiocesium was at most approximately 20%.The two radiocesium sources,radiocesium-bearing particles and fugitive dust,are both insoluble particles.The largest internal dose from inhalation was 7.6 μSv in spring 2013,which is negligible compared to the dose limit recommended for members of the public by the International Commission on Radiological Protection.

Keywords:137Cs;134Cs;dose,internal;inhalation

Health Phys.118(6):656-663;2020

ASSESSMENTMODELOFRADIATIONDOSESFROMEXTERNALEXPOSURETOTHEPUBLICAFTERTHEFUKUSHIMADAIICHINUCLEARPOWERPLANTACCIDENT

Shogo Takahara1,Masashi Iijima1,Masatoshi Watanabe1

(1.Japan Atomic Energy Agency,Naka-gun,Tokai-mura,Japan)

Abstract:Radiation exposure is one of most important factors to manage following a nuclear emergency.Actual measurement is the best way to obtain information concerning the dose received by the people in terms of accuracy and reliability.However,in practice,it is difficult to collect measurements from all people affected by nuclear accidents over the whole period of exposure from past to future.Therefore,probabilistic assessment using a model is needed.An assessment model of radiation doses from external exposures was developed based on the actual measurement of individual doses and ambient dose equivalent rates inside and outside houses in Fukushima City.A survey of behavioral patterns was also performed for the same purpose.In addition to our measurement and survey,we took into account the latest insights from the experiences of the Fukushima Daiichi nuclear power plant accident.Comparisons between the assessed and measured results revealed that the time-dependence of doses and the distribution of doses obtained using the developed models agree well with the results of actual measurements.Thus,our probabilistic approach was validated.Based on both our assessment and on our actual measurements,no participants were observed to receive doses in excess of 1 mSv y-1as of 8 y after the Fukushima Daiichi nuclear power plant accident in Fukushima City.

Keywords:accidents,nuclear;dose,external;modeling,dose assessment;Fukushima Daiichi

Health Phys.118(6):664-677;2020

DETERMININGANDCOMPARINGNEUTRONSPECTRAATTAIWANPHOTONSOURCEBEFOREANDAFTERTHEINSTALLATIONOFLOCALINJECTIONSHIELDING

Yu-Chi Lin1,2,Ang-Yu Chen1,Rong-Jiun Sheu2,3

(1.National Synchrotron Radiation Research Center,101 Hsin-Ann Road,Hsinchu Science Park,Hsinchu,Taiwan;2.Institute of Nuclear Engineering and Science,National Tsing Hua University,101,Sec.2,Kuang-Fu Road,Hsinchu,Taiwan;3.Department of Engineering and System Science,National Tsing Hua University,101,Sec.2,Kuang-Fu Road,Hsinchu,Taiwan)

Abstract:Accurate neutron spectrum measurements at light source facilities are difficult to perform because of relatively low and time-varying neutron intensities.A homemade Bonner cylinder spectrometer was used to determine the energy spectra of neutrons outside the lateral shielding wall of the Taiwan Photon Source before and after the installation of local injection shielding.The spectrometer,similar to the design of conventional Bonner spheres,features (1) highly sensitive neutron detection and (2) a wide-range response to neutrons with energies up to the GeV range.Neutron measurements were conducted by intentionally parking the injected 3-GeV electrons at the septum of the storage ring.On the basis of high-fidelity FLUKA simulations,neutron spectra at the measurement location under the experimental conditions were obtained and adopted as an initial guess for spectrum unfolding.The neutron spectra determined before and after the local shielding were comprehensively compared in terms of their intensities and characteristics.The local shielding resulted in overall reductions of approximately 44% and 38% in total neutron flux and dose rate,respectively.Both before and after the local shielding,high-energy neutrons (>10 MeV) were the dominant component of the radiation field,which contributed approximately 30% to 35% of the total neutron flux and 55% to 59% of the total neutron dose rate.

Keywords:operational topics;accelerators;Monte Carlo;neutron detection

Health Phys.118(6):693-701;2020

RadiationSafetyConsiderationsintheTreatmentofCanineSkeletalConditionsUsing153Sm,90Y,and117mSn

Richard E.Wendt III1,Kimberly A.Selting2,Jimmy C.Lattimer3,Janine Wong4,Jaime Simón5,Nigel R.Stevenson6,Stanley D.Stearns7

(1.Department of Imaging Physics,The University of Texas MD Anderson Cancer Center,Houston,Texas 77030;2.College of Veterinary Medicine,The University of Illinois at Urbana-Champaign,Urbana,Illinois 61802;formerly in the Department of Veterinary Medicine and Surgery,College of Veterinary Medicine,The University of Missouri,Columbia,Missouri 65211;3.Department of Veterinary Medicine and Surgery,College of Veterinary Medicine,The University of Missouri,Columbia,Missouri 65211;4.The University of Texas Health Science Center School of Public Health,Houston,Texas 77030;formerly in the Department of Imaging Physics,The University of Texas MD Anderson Cancer Center,Houston,Texas 77030;5.IsoTherapeutics Group,LLC,Angleton,Texas 77515;6.Exubrion Therapeutics,Inc.,Buford,Georgia 30518;formerly with Serene,LLC,The Woodlands,Texas 77381;7.The Gabriel Institute,Houston,Texas 77055)

Abstract:The treatment of pets,service animals,and pre-clinical research subjects with radionuclides raises concern for the safety of the people who interact with the animals after their treatment.Three treatments of skeletal conditions in dogs are considered in this study:153Sm-1,4,7,10-tetraazacylcododecanetetramethylenephosphonic acid,which is a bone-seeking radiopharmaceutical;unencapsulated90Y permanent interstitial implants,which are sometimes called “l(fā)iquid brachytherapy”;and117mSn radiosynoviorthesis,which is also called radiosynovectomy.External exposure rate readings of the153Sm and117mSn treatments,and Monte Carlo simulations of117mSn at a distance of 1 m and of all three in direct contact with tissue were analyzed for doses.Dogs that have received any of these treatments using typically administered activities may be released from radiation safety isolation immediately after treatment from the standpoint of external exposure.People should avoid prolonged close proximity,such as sleeping with a treated dog,for three weeks following an90Y interstitial implant or for a month following117mSn radiosynoviorthesis.No such avoidance is necessary after treatment with153Sm-1,4,7,10-tetraazacylcododecanetetramethylenephosphonic acid.

Keywords:dogs;medical radiation;nuclear medicine;radiopharmaceuticals

Health Phys.118(6):702-710;2020