Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

International Nuclear Information System (INIS)

Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

2012-01-01

Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

Shield calculations, optimization vs. paradigm

International Nuclear Information System (INIS)

Cornejo D, N.; Hernandez S, A.; Martinez G, A.

2006-01-01

Many shieldings have been designed under the criteria of 'Maximum dose rates of project'. It has created the paradigm of those 'low dose rates', for the one which not few specialists would consider unacceptable levels of dose rate superior to the units of μSv.h -1 , independently of the exposure times. At the present time numerous shieldings are being designed considering dose restrictions in real times of exposure. After these new shieldings, the dose rates could be notably superior to those after traditional shieldings, without it implies inadequate designs or constructive errors. In the work significant differences in levels of dose rates and thickness of shieldings estimated by both methods for some typical facilities. It was concluded that the use of real times of exposure is more adequate for the optimization of the Radiological Protection, although this method demands bigger care in its application. (Author)

Shielding synchrotron light sources: Advantages of circular shield walls tunnels

Energy Technology Data Exchange (ETDEWEB)

Kramer, S.L. [Design and Accelerator Operations Consulting, 568 Wintergreen Ct Ridge, NY 11961 (United States); Ghosh, V.J.; Breitfeller, M. [NSLS-II, Brookhaven National Laboratory, Upton, NY 11973 (United States)

2016-08-11

Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produce significantly higher neutron component dose to the experimental floor than lower energy injection and ramped operations. High energy neutrons produced in the forward direction from thin target beam losses are a major component of the dose rate outside the shield walls of the tunnel. The convention has been to provide thicker 90° ratchet walls to reduce this dose to the beam line users. We present an alternate circular shield wall design, which naturally and cost effectively increases the path length for this forward radiation in the shield wall and thereby substantially decreasing the dose rate for these beam losses. This shield wall design will greatly reduce the dose rate to the users working near the front end optical components but will challenge the beam line designers to effectively utilize the longer length of beam line penetration in the shield wall. Additional advantages of the circular shield wall tunnel are that it's simpler to construct, allows greater access to the insertion devices and the upstream in tunnel beam line components, as well as reducing the volume of concrete and therefore the cost of the shield wall.

Gamma dose from activation of internal shields in IRIS reactor.

Science.gov (United States)

Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

Gamma dose from activation of internal shields in IRIS reactor

International Nuclear Information System (INIS)

Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

Shielding calculations. Optimization vs. Paradigms

International Nuclear Information System (INIS)

Cornejo Diaz, Nestor; Hernandez Saiz, Alejandro; Martinez Gonzalez, Alina

2005-01-01

Many radiation shielding barriers in Cuba have been designed according to the criterion of Maxi-mum Projected Dose Rates. This fact has created the paradigm of low dose rates. Because of this, dose rate levels greater than units of Sv.h-1 would be considered unacceptable by many specialists, regardless of the real exposure times. Nowadays many shielding barriers are being designed using dose constraints in real exposure times. Behind the new barriers, dose rates could be notably greater than those behind the traditional ones, and it does not imply inadequate designs or constructive errors. In this work were obtained significant differences in dose rate levels and shield-ing thicknesses calculated by both methods for some typical installations. The work concludes that real exposure time approach is more adequate in order to optimise Radiation Protection, although this method should be carefully applied

Gamma dose rate calculations for conceptual design of a shield system for spent fuel interim dry storage in CNA 1

International Nuclear Information System (INIS)

Blanco, A; Gomez S

2012-01-01

After completing the rearrangement of the Spent Fuel Elements (SFE) into a compact arrangement in the two storage water pools, Atucha Nuclear Reactor 1 (ANR 1) will leave free position for the wet storage of the SFE discharged until December 2014. Even, in two possible scenarios, such as extending operation from 2015 or the cessation of operation after that date, it will be necessary to empty the interim storage water pools transferring the SFE to a temporary dry storage system. Because the law 25.018 'Management of Radioactive Wastes' implies for the first scenario - operation beyond 2015 - that Nucleoelectrica Argentina S.A. will still be in charge of the dry storage system and for the second - the cessation of operation after 2015 - the National Commission of Atomic Energy (CNEA) will be in charge by the National Management Program of Radioactive Wastes, the interim dry storage system of SNF is an issue of common interest which justifies go forward together. For that purpose and in accordance with the criticality and shielding calculations relevant to the project, in this paper we present the dose rate calculations for shielding conceptual design of a system for dry interim storage of the SFE of ANR 1. The specifications includes that the designed system must be suitable without modification for the SFE of the ANR 2. The results for the calculation of the photon dose rate, in touch and at one meter far, for the Transport Module and the Container of the SFE, are presented, which are required and controlled by the National Regulatory Authority (NRA) and the International Atomic Energy Agency (IAEA). It was used the SAS4 module of SCALE5.1 system and MCNP5. As a design tool for the photon shielding in order to meet current standards for allowable dose rates, a radial and axial parametric analysis were developed based on the thickness of lead of the Transport Module. The results were compared and verified between the two computing systems. Before this

Dose rate measuring device and dose rate measuring method using the same

International Nuclear Information System (INIS)

Urata, Megumu; Matsushita, Takashi; Hanazawa, Sadao; Konno, Takahiro; Chiba, Yoshinori; Yumitate, Tadahiro

1998-01-01

The device of the present invention comprises a scintillation fiber scope having a shape elongated in the direction of the height of a pressure vessel and emitting light by incident of radiation to detect radiation, a radioactivity measuring device for measuring a dose rate based on the detection of the fiber scope and a reel means for dispensing and taking up the fiber scope, and it constituted such that the dose rate of the pressure vessel and that of a shroud are determined independently. Then, when the taken out shroud is contained in an container, excessive shielding is not necessary, in addition, this device can reliably be inserted to or withdrawn from complicated places between the pressure vessel and the shroud, and further, the dose rate of the pressure vessel and that of the shroud can be measured approximately accurately even when the thickness of them is different greatly. (N.H.)

Dose rate measuring device and dose rate measuring method using the same

Energy Technology Data Exchange (ETDEWEB)

Urata, Megumu; Matsushita, Takashi; Hanazawa, Sadao; Konno, Takahiro; Chiba, Yoshinori; Yumitate, Tadahiro

1998-11-13

The device of the present invention comprises a scintillation fiber scope having a shape elongated in the direction of the height of a pressure vessel and emitting light by incident of radiation to detect radiation, a radioactivity measuring device for measuring a dose rate based on the detection of the fiber scope and a reel means for dispensing and taking up the fiber scope, and it constituted such that the dose rate of the pressure vessel and that of a shroud are determined independently. Then, when the taken out shroud is contained in an container, excessive shielding is not necessary, in addition, this device can reliably be inserted to or withdrawn from complicated places between the pressure vessel and the shroud, and further, the dose rate of the pressure vessel and that of the shroud can be measured approximately accurately even when the thickness of them is different greatly. (N.H.)

Shielding augmentation of roll-on shield from NAPS to Kaiga-2

International Nuclear Information System (INIS)

Pradhan, A.S.; Kumar, A.N.

2000-01-01

Extensive radiation field surveys were conducted in NAPS and KAPS reactor buildings as a part of commissioning checks on radiation shielding. During such surveys, dose rate higher than the expected values were noticed in fuelling machine service areas. A movable shield, separating high field area fuelling machine vault and low field area fuelling machine service area, known as roll-on shield was identified as one of the causes of high field in fuelling machine service area along with weaker end-shield. This paper discusses systematic approach adopted in bringing down the dose rates in fuelling machine service area by augmentation of roll-on shield. (author)

Developing A Directional High-Dose Rate (d-HDR) Brachytherapy Source

Science.gov (United States)

Heredia, Athena Yvonne

Conventional sources used in brachytherapy provide nearly isotropic or radially symmetric dose distributions. Optimizations of dose distributions have been limited to varied dwell times at specified locations within a given treatment volume, or manipulations in source position for seed implantation techniques. In years past, intensity modulated brachytherapy (IMBT) has been used to reduce the amount of radiation to surrounding sensitive structures in select intracavitary cases by adding space or partial shields. Previous work done by Lin et al., at the University of Wisconsin-Madison, has shown potential improvements in conformality for brachytherapy treatments using a directionally shielded low dose rate (LDR) source for treatments in breast and prostate. Directional brachytherapy sources irradiate approximately half of the radial angles around the source, and adequately shield a quarter of the radial angles on the opposite side, with sharp gradient zones between the treated half and shielded quarter. With internally shielded sources, the radiation can be preferentially emitted in such a way as to reduce toxicities in surrounding critical organs. The objective of this work is to present findings obtained in the development of a new directional high dose rate (d-HDR) source. To this goal, 103Pd (Z = 46) is reintroduced as a potential radionuclide for use in HDR brachytherapy. 103Pd has a low average photon energy (21 keV) and relatively short half -life (17 days), which is why it has historically been used in low dose rate applications and implantation techniques. Pd-103 has a carrier-free specific activity of 75000 Ci/g. Using cyclotron produced 103Pd, near carrier-free specific activities can be achieved, providing suitability for high dose rate applications. The evolution of the d-HDR source using Monte Carlo simulations is presented, along with dosimetric parameters used to fully characterize the source. In addition, a discussion on how to obtain elemental

Evaluation of radiation shielding rate of lead aprons in nuclear medicine

Energy Technology Data Exchange (ETDEWEB)

Han, Sang Hyun; Han, Beom Heui; Lee, Sang Ho [Dept. of Radiological Science, Seonam University, Asan (Korea, Republic of); Hong, Dong Heui [Dept. of Radiological Science, Far East University, Eumseong (Korea, Republic of); Kim, Gi Jin [Dept. of Nuclear Medicine, Konyang University Hospital, Daejeon (Korea, Republic of)

2017-03-15

Considering that the X-ray apron used in the department of radiology is also used in the department of nuclear medicine, the study aimed to analyze the shielding rate of the apron according to types of radioisotopes, thus γ ray energy, to investigate the protective effects. The radioisotopes used in the experiment were the top 5 nuclides in usage statistics {sup 99m}Tc, {sup 18}F, {sup 131}I, {sup 123}I, and {sup 201}Tl, and the aprons were lead equivalent 0.35 mmPb aprons currently under use in the department of nuclear medicine. As a result of experiments, average shielding rates of aprons were {sup 99m}Tc 31.59%, {sup 201}Tl 68.42%, and {sup 123}I 76.63%. When using an apron, the shielding rate of {sup 13}'1I actually resulted in average dose rate increase of 33.72%, and {sup 18}F showed an average shielding rate of –0.315%, showing there was almost no shielding effect. As a result, the radioisotopes with higher shielding rate of apron was in the descending order of {sup 123}I, {sup 201}Tl, {sup 99m}Tc, {sup 18}F, {sup 131}I. Currently, aprons used in the nuclear medicine laboratory are general X-ray aprons, and it is thought that it is not appropriate for nuclear medicine environment that utilizes γ rays. Therefore, development of nuclear medicine exclusive aprons suitable for the characteristics of radioisotopes is required in consideration of effective radiation protection and work efficiency of radiation workers.

Equivalent-spherical-shield neutron dose calculations

International Nuclear Information System (INIS)

Russell, G.J.; Robinson, H.

1988-01-01

Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab

Evaluation of radiation shielding rate of lead aprons in nuclear medicine

International Nuclear Information System (INIS)

Han, Sang Hyun; Han, Beom Heui; Lee, Sang Ho; Hong, Dong Heui; Kim, Gi Jin

2017-01-01

Considering that the X-ray apron used in the department of radiology is also used in the department of nuclear medicine, the study aimed to analyze the shielding rate of the apron according to types of radioisotopes, thus γ ray energy, to investigate the protective effects. The radioisotopes used in the experiment were the top 5 nuclides in usage statistics "9"9"mTc, "1"8F, "1"3"1I, "1"2"3I, and "2"0"1Tl, and the aprons were lead equivalent 0.35 mmPb aprons currently under use in the department of nuclear medicine. As a result of experiments, average shielding rates of aprons were "9"9"mTc 31.59%, "2"0"1Tl 68.42%, and "1"2"3I 76.63%. When using an apron, the shielding rate of "1"3'1I actually resulted in average dose rate increase of 33.72%, and "1"8F showed an average shielding rate of –0.315%, showing there was almost no shielding effect. As a result, the radioisotopes with higher shielding rate of apron was in the descending order of "1"2"3I, "2"0"1Tl, "9"9"mTc, "1"8F, "1"3"1I. Currently, aprons used in the nuclear medicine laboratory are general X-ray aprons, and it is thought that it is not appropriate for nuclear medicine environment that utilizes γ rays. Therefore, development of nuclear medicine exclusive aprons suitable for the characteristics of radioisotopes is required in consideration of effective radiation protection and work efficiency of radiation workers

A new design of a lead-acrylic shield for staff dose reduction in radial and femoral access coronary catheterization

Energy Technology Data Exchange (ETDEWEB)

Eder, H. [Deptartment of Radiation Protection (Germany); Seidenbusch, M.C.; Treitl, M. [Muenchen Univ. Clinical Center (Germany). Inst. for Clinical Radiology; Gilligan, P. [Mater Private Hospital, Dublin (Ireland). Medical Physics

2015-10-15

Today's standard radiation protection during coronary angiography and percutaneous coronary interventions is the combined use of lead acrylic shields and table-mounted lower body protection. Ambient dose measurements, however, have shown that these protection devices need improvement. Using an anthropomorphic physical phantom, various scenarios were investigated with respect to personnel exposure: (a) enlarging the shield (b) adding a flexible protective curtain to the bottom side of the shield, and (c) application of radioprotective patient drapes. For visualization of the dose reduction effect, Monte Carlo simulations were performed. The flexible curtain in contact with the patient's body reduces the ambient dose rate at the operator's position by up to (87.5 % ± 7.1) compared to the situation with the bare shield. The use of both the flexible curtain and the patient drape reduces the ambient dose rate by up to (90.8 % ± 7). Similar results were achieved for the assisting personnel when they were positioned next to the operator. In addition, the enlarged shield provides better protection of the head region of tall operators. Adding a flexible protective curtain to the bottom side of the shield can protect operators from high doses, especially for body parts which are not protected by lead aprons, e.g. head, and eye lenses. This may be important with respect to lower dose limits for eye lenses in future. The protective effect in real-life working conditions is still being evaluated in an ongoing clinical study.

Peripheral dose measurement for CyberKnife radiosurgery with upgraded linac shielding

International Nuclear Information System (INIS)

Chuang, Cynthia F.; Larson, David A.; Zytkovicz, Andrea; Smith, Vernon; Petti, Paula L.

2008-01-01

The authors investigated the peripheral dose reduction for CyberKnife radiosurgery treatments after the installation of a linac shielding upgrade. As in a previous investigation, the authors considered two treatment plans, one for a hypothetical target in the brain and another for a target in the thorax, delivered to an anthropomorphic phantom. The results of the prior investigation showed that the CyberKnife delivered significantly higher peripheral doses than comparable model C Gamma Knife or IMRT treatments. Current measurements, after the linac shielding upgrade, demonstrate that the additional shielding decreased the peripheral dose, expressed as a percentage of the delivered monitor units (MU), by a maximum of 59%. The dose reduction was greatest for cranial-caudal distances from the field edge less than 30 cm, and at these distances, the CyberKnife peripheral dose, expressed as a percentage of the delivered MU, is now comparable to that measured for the other treatment modalities in our previous investigation. For distances between 30 and 70 cm from the field edge, the additional shielding reduced the peripheral dose by between 20% and 55%. At these distances, the CyberKnife peripheral dose remains higher than doses measured in our previous study for the model C Gamma Knife and IMRT

Concrete spent fuel storage casks dose rates

International Nuclear Information System (INIS)

Bace, M.; Jecmenica, R.; Trontl, K.

1998-01-01

Our intention was to model a series of concrete storage casks based on TranStor system storage cask VSC-24, and calculate the dose rates at the surface of the casks as a function of extended burnup and a prolonged cooling time. All of the modeled casks have been filled with the original multi-assembly sealed basket. The thickness of the concrete shield has been varied. A series of dose rate calculations for different burnup and cooling time values have been performed. The results of the calculations show rather conservative original design of the VSC-24 system, considering only the dose rate values, and appropriate design considering heat rejection.(author)

Testicular shield for para-aortic radiotherapy and estimation of gonad doses.

Science.gov (United States)

Ravichandran, R; Binukumar, J P; Kannadhasan, S; Shariff, M H; Ghamrawy, Kamal El

2008-10-01

For radiotherapy of para-aortic and abdominal regions in male patients, gonads are to be protected to receive less than 2% of the prescribed dose. A testicular shield was fabricated for abdominal radiotherapy with 15 MV X-rays ((Clinac 2300 CD, Varian AG) with low melting point alloy (Cerroband). The dimensions of the testicular shield were 6.5 cm diameter and 3.5 cm depth with 1.5 cm wall thickness. During treatment, this shield was held in position by a rectangular sponge and Styrofoam support. Phantom measurement was carried out with a humanoid phantom and a 0.6 cc ion chamber. The mean energy of the scattered photon was calculated for single scattering at selected distances from the beam edge and with different field dimensions. One patient received radiotherapy with an inverted Y field and gonad doses were estimated using calibrated thermo-luminescent detector (TLD) chips. Measured doses with the ion chamber were 7.1 and 3.5% of the mid-plane doses without a shield at 3 and 7.5 cm off-field respectively. These values decreased to 4.6 and 1.7% with the bottom shield alone, and to 1.7 and 0.8% with both bottom and top shields covering the ion chamber. The measured doses at the gonads during the patient's treatment were 0.5-0.92% for the AP field (0.74 +/- 0.17%, n = 5) and 0.5-1.2% for the PA field (0.88 +/- 0.24%, n = 5). The dose received by the testis for the full course of treatment was 32 cGy (0.8%) for a total mid-plane dose of 40 Gy. The first-scatter energy estimated at the gonads is around 1.14 MeV for a primary beam of 15 MV for a long axis dimension of 37 cm of primary field. During the patient's treatment, the estimated absorbed doses at the gonads were comparable with reported values in similar treatments. The testicular shield reported in this study is of light weight and could be used conveniently in treatments of abdominal fields.

Testicular shield for para-aortic radiotherapy and estimation of gonad doses

Directory of Open Access Journals (Sweden)

Ravichandran R

2008-01-01

Full Text Available For radiotherapy of para-aortic and abdominal regions in male patients, gonads are to be protected to receive less than 2% of the prescribed dose. A testicular shield was fabricated for abdominal radiotherapy with 15 MV X-rays ((Clinac 2300 CD, Varian AG with low melting point alloy (Cerroband. The dimensions of the testicular shield were 6.5 cm diameter and 3.5 cm depth with 1.5 cm wall thickness. During treatment, this shield was held in position by a rectangular sponge and Styrofoam support. Phantom measurement was carried out with a humanoid phantom and a 0.6 cc ion chamber. The mean energy of the scattered photon was calculated for single scattering at selected distances from the beam edge and with different field dimensions. One patient received radiotherapy with an inverted Y field and gonad doses were estimated using calibrated thermo-luminescent detector (TLD chips. Measured doses with the ion chamber were 7.1 and 3.5% of the mid-plane doses without a shield at 3 and 7.5 cm off-field respectively. These values decreased to 4.6 and 1.7% with the bottom shield alone, and to 1.7 and 0.8% with both bottom and top shields covering the ion chamber. The measured doses at the gonads during the patient′s treatment were 0.5-0.92% for the AP field (0.74 ± 0.17%, n = 5 and 0.5-1.2% for the PA field (0.88 ± 0.24%, n = 5. The dose received by the testis for the full course of treatment was 32 cGy (0.8% for a total mid-plane dose of 40 Gy. The first-scatter energy estimated at the gonads is around 1.14 MeV for a primary beam of 15 MV for a long axis dimension of 37 cm of primary field. During the patient′s treatment, the estimated absorbed doses at the gonads were comparable with reported values in similar treatments. The testicular shield reported in this study is of light weight and could be used conveniently in treatments of abdominal fields.

Study on the evaluation method of radiation dose rate around spent fuel shipping casks

International Nuclear Information System (INIS)

Yamakoshi, Hisao

1986-01-01

This study aims at developing a simple calculation method which can evaluate radiation dose rate around casks with high accuracy in a short time. The method is based on a concept of the radiation shielding characteristics of cask walls. The concept was introduced to replace for ordinary radiation shielding calculation which requires a long calculation time and a large memory capacity of a computer in the matrix calculation. For the purpose of verifying the accuracy and reliability of the new method, it was applied to the analysis of the dose rate distribution around actual casks, which had been measured. The results of the analysis revealed that the newly proposed method was excellent for the forecast of radiation dose rate distribution around casks in view of the accuracy and calculation time. The short calculation time and high accuracy by the proposed method were attained by dividing the whole procedure of ordinary fine radiation shielding calculation into the calculation of radiation dose rate on a cask surface by the matrix expression of the characteristic function and the calculation of dose rate distribution using the simple analytical expression of dose rate distribution around casks. The effect of the heterogeneous array of spent fuel in different burnup state on dose rate distribution around casks was evaluated by this method. (Kako, I.)

Characterization of a lead breast shielding for dose reduction in computed tomography

Energy Technology Data Exchange (ETDEWEB)

Correia, Paula Duarte; Brochi, Marco Aurelio Corte; Azevedo-Marques, Paulo Mazzoncini de, E-mail: pauladuarte@usp.br [Universidade de Sao Paulo (FM/RSP), Ribeirao Preto, SP (Brazil). Faculdade de Medicina; Granzotti, Cristiano Roberto Fabri; Santos, Yago da Silva [Universidade de Sao Paulo (FFCLRP/RSP), Ribeirao Preto, SP (Brazil). Faculdade de Filosofia, Ciencias e Letras

2014-07-15

Objective: several studies have been published regarding the use of bismuth shielding to protect the breast in computed tomography (CT) scans and, up to the writing of this article, only one publication about barium shielding was found. The present study was aimed at characterizing, for the first time, a lead breast shielding. Materials and methods: the percentage dose reduction and the influence of the shielding on quantitative imaging parameters were evaluated. Dose measurements were made on a CT equipment with the aid of specific phantoms and radiation detectors. A processing software assisted in the qualitative analysis evaluating variations in average CT number and noise on images. Results: the authors observed a reduction in entrance dose by 30% and in CTDIvol by 17%. In all measurements, in agreement with studies in the literature, the utilization of cotton fiber as spacer object reduced significantly the presence of artifacts on the images. All the measurements demonstrated increase in the average CT number and noise on the images with the presence of the shielding. Conclusion: as expected, the data observed with the use of lead shielding were of the same order as those found in the literature about bismuth shielding. (author)

GARDEC, Estimation of dose-rates reduction by garden decontamination

International Nuclear Information System (INIS)

Togawa, Orihiko

2006-01-01

1 - Description of program or function: GARDEC estimates the reduction of dose rates by garden decontamination. It provides the effect of different decontamination Methods, the depth of soil to be considered, dose-rate before and after decontamination and the reduction factor. 2 - Methods: This code takes into account three Methods of decontamination : (i)digging a garden in a special way, (ii) a removal of the upper layer of soil, and (iii) covering with a shielding layer of soil. The dose-rate conversion factor is defined as the external dose-rate, in the air, at a given height above the ground from a unit concentration of a specific radionuclide in each soil layer

Daily dose and shielding optimization in work performance at 'Ukrytie' object

International Nuclear Information System (INIS)

Batij, V.G.; Derengovskij, V.V.; Egorov, V.V.; Kuz'menko, V.A.; Rud'ko, V.M.; Sizov, A.A.; Stoyanov, A.I.

2000-01-01

The procedure of daily dose and shielding optimization in work performance at 'Ukryttia' object is offered. The recommendations allowing reducing collective effective doze according to the optimization principle are submitted. The technique of shielding optimization is given at stabilization works realization. The optimum shielding calculation example for the strengthening support is given

The use of the long modular diagnostics shield module to mitigate shutdown dose rates in the ITER diagnostics equatorial ports

Science.gov (United States)

Juárez, R.; Guirao, J.; Kolsek, A.; Lopez, A.; Pedroche, G.; Bertalot, L.; Udintsev, V. S.; Walsh, M. J.; Sauvan, P.; Sanz, J.

2018-05-01

The ITER equatorial port plugs are submitted to a drained weight limit of 45 T. This limitation can conflict with their radiation shielding demands, although some weight margin is being discussed. The port interspaces are subject to a shutdown dose rate limit of 100 µSv h‑1 after 106 s of cooling time. To meet it, the port plugs must show a neutron flux attenuation comparable to their neighborhood, despite considering penetrations to host systems. Most of this task relies on the drawer shield module (DSM). In this work, two DSM concepts are analyzed with this perspective: the box-based DSM and the modular DSM. Regardless the penetrations, the box-based DSM leads to unsatisfactory port plugs to meet both weight and SDDR requirements. On the contrary, the modular DSM shows a performance which allows for the adoption of such DSM concept, or equivalent, a port may comply with both requirements at the same time, provided the penetrations are well designed.

The dose penumbra of a custom-made shield used in hemibody skin electron irradiation.

Science.gov (United States)

Rivers, Charlotte I; AlDahlawi, Ismail; Wang, Iris Z; Singh, Anurag K; Podgorsak, Matthew B

2016-11-08

We report our technique for hemibody skin electron irradiation with a custom-made plywood shield. The technique is similar to our clinical total skin electron irradiation (TSEI), performed with a six-pair dual field (Stanford technique) at an extended source-to-skin distance (SSD) of 377 cm, with the addition of a plywood shield placed at 50 cm from the patient. The shield is made of three layers of stan-dard 5/8'' thick plywood (total thickness of 4.75 cm) that are clamped securely on an adjustable-height stand. Gafchromic EBT3 films were used in assessing the shield's transmission factor and the extent of the dose penumbra region for two different shield-phantom gaps. The shield transmission factor was found to be about 10%. The width of the penumbra (80%-to-20% dose falloff) was measured to be 12 cm for a 50 cm shield-phantom gap, and reduced slightly to 10 cm for a 35 cm shield-phantom gap. In vivo dosimetry of a real case confirmed the expected shielded area dose. © 2016 The Authors.

Effectiveness of Bismuth Shield to Reduce Eye Lens Radiation Dose Using the Photoluminescence Dosimetry in Computed Tomography

International Nuclear Information System (INIS)

Jung, Mi Young; Kweon, Dae Cheol; Kwon, Soo Il

2009-01-01

The purpose of our study was to determine the eye radiation dose when performing routine multi-detector computed tomography (MDCT). We also evaluated dose reduction and the effect on image quality of using a bismuth eye shield when performing head MDCT. Examinations were performed with a 64MDCT scanner. To compare the shielded/unshielded lens dose, the examination was performed with and without bismuth shielding in anthropomorphic phantom. To determine the average lens radiation dose, we imaged an anthropomorphic phantom into which calibrated photoluminescence glass dosimeter (PLD) were placed to measure the dose to lens. The phantom was imaged using the same protocol. Radiation doses to the lens with and without the lens shielding were measured and compared using the Student t test. In the qualitative evaluation of the MDCT scans, all were considered to be of diagnostic quality. We did not see any differences in quality between the shielded and unshielded brain. The mean radiation doses to the eye with the shield and to those without the shield were 21.54 versus 10.46 mGy, respectively. The lens shield enabled a 51.3% decrease in radiation dose to the lens. Bismuth in-plane shielding for routine eye and head MDCT decreased radiation dose to the lens without qualitative changes in image quality. The other radiosensitive superficial organs specifically must be protected with shielding.

Distribution and characteristics of gamma and cosmic ray dose rate in living environment

International Nuclear Information System (INIS)

Nagaoka, Toshi; Moriuchi, Shigeru

1991-01-01

A series of environmental radiation surveys was carried out from the viewpoint of characterizing the natural radiation dose rate distribution in the living environment, including natural and artificial ones. Through the analysis of the data obtained at numbers of places, several aspects of the radiation field in living environments were clarified. That is the gamma ray dose rate varies due to the following three dominant causes: 1) the radionuclide concentration of surrounding materials acting as gamma ray sources, 2) the spatial distribution of surrounding materials, and 3) the geometrical and shielding conditions between the natural gamma ray sources and the measured point; whereas, the cosmic ray dose rate varies due to the thickness of upper shielding materials. It was also suggested that the gamma ray dose rate generally shows an upward tendency, and the cosmic ray dose rate a downward one in artificial environment. This kind of knowledge is expected to serve as fundamental information for accurate and realistic evaluation of the collective dose in the living environment. (author)

Parameters calculation of shielding experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-02-01

The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt

SMART, Radiation Dose Rates on Cask Surface

International Nuclear Information System (INIS)

Yamakoshi, Hisao

1989-01-01

1 - Description of program or function: SMART calculates radiation dose rate at the center of each cask surface by using characteristic functions for radiation shielding ability and for radiation current back-scattered from cask wall and cask cavity of each cask, once cask-type is specified. 2 - Method of solution: Matrix Calculation

Electron, electron-bremsstrahlung and proton depth-dose data for space-shielding applications

Science.gov (United States)

Seltzer, S. M.

1979-01-01

A data set has been developed, consisting of depth-dose distributions for omni-directional electron and proton fluxes incident on aluminum shields. The principal new feature of this work is the accurate treatment, based on detailed Monte Carlo calculations, of the electron-produced bremsstrahlung component. Results covering the energy region of interest in space-shielding calculations have been obtained for the absorbed dose (a) as a function of depth in a semi-infinite medium, (b) at the edge of slab shields, and (c) at the center of a solid sphere. The dose to a thin tissue-equivalent detector was obtained as well as that in aluminum. Various results and comparisons with other work are given.

ITER Generic Diagnostic Upper Port Plug Nuclear Heating and Personnel Dose Rate Assessment

International Nuclear Information System (INIS)

Feder, Russell E.; Youssef, Mahmoud Z.

2009-01-01

Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of a large aperture diagnostic were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture. The neutronics discrete-ordinates code ATTILA(reg s ign) and SEVERIAN(reg s ign) (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER 'Brand Model' MCNP benchmark model. A biased quadrature set equivalent to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and Large Aperture cases. The Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 (micro)Sv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 (micro)Sv/hr but fell below the limit to 90 (micro)Sv/hr 2-weeks later. The Large Aperture style shielding model exhibited higher and more persistent dose rates. After 1

Shield calculations, optimization vs. paradigm; Calculos de blindajes, optimizacion vs. paradigma

Energy Technology Data Exchange (ETDEWEB)

Cornejo D, N.; Hernandez S, A.; Martinez G, A. [Centro de Proteccion e Higiene de las Radiaciones, Calle 20 No. 4113 e/41 y 47 Playa C.P. 11300 LaHabana (Cuba)]. e-mail: nestor@cphr.edu.cu

2006-07-01

Many shieldings have been designed under the criteria of 'Maximum dose rates of project'. It has created the paradigm of those 'low dose rates', for the one which not few specialists would consider unacceptable levels of dose rate superior to the units of {mu}Sv.h{sup -1}, independently of the exposure times. At the present time numerous shieldings are being designed considering dose restrictions in real times of exposure. After these new shieldings, the dose rates could be notably superior to those after traditional shieldings, without it implies inadequate designs or constructive errors. In the work significant differences in levels of dose rates and thickness of shieldings estimated by both methods for some typical facilities. It was concluded that the use of real times of exposure is more adequate for the optimization of the Radiological Protection, although this method demands bigger care in its application. (Author)

Radiation dose reduction at a price: the effectiveness of a thyroid shield during head CT scanning

International Nuclear Information System (INIS)

Fu Qiang; Lu Tao; Zhang Ling

2008-01-01

Objective: To assess radiation dose to the thyroid in patients undergoing head CT scanning and to evaluate dose reduction to the thyroid by load shielding. Methods: A post-morterm was scanned by different model and study was undertaken to evaluate the dose reduction by thyroid lead shields and assess their practicality in a clinical setting. (a)No thyroid shields and (b) thyroid shield. One thermoluminescent dosimeters (TLDs)were placed over the thyroid gland center, A thyroid lead shield (Pb eq 0.5mm)was placed around the neck of post-morterm. Scan parameter, CTDIw and DLP were recorded. Results: (a) 0.207mSv; (b) 0.085mSv. A mean effective radiation dose reduction of 58% was seen in the shielded versus the unshielded. Conclusion: Thyroid exposure to scattered radiation from head CT scanning only once is associated with a low but not negligible risk of cancer, but accumulatived doses to the thyroid are serious, highlighting the need for increased awareness of patient radiation protection. Thyroid lead shielding yields significant radiation protection, which should be used routinely during head CT scan. (authors)

Coronary calcium scoring with MDCT: The radiation dose to the breast and the effectiveness of bismuth breast shield

International Nuclear Information System (INIS)

Yilmaz, Mehmet Halit; Yasar, Dogan; Albayram, Sait; Adaletli, Ibrahim; Ozer, Harun; Ozbayrak, Mustafa; Mihmanli, Ismail; Akman, Canan

2007-01-01

Objective: The purpose of our study was to determine the breast radiation dose during coronary calcium scoring with multidetector computerized tomography (MDCT). We also evaluated the degree of dose reduction by using a bismuth breast shield when performing coronary calcium scoring with MDCT. Materials and methods: The dose reduction achievable by shielding the adult (35 years or older) female breasts was studied in 25 women who underwent coronary calcium scoring with MDCT. All examinations were performed with a 16-MDCT scanner. To compare the shielded versus unshielded breast dose, the examinations were performed with (right breast) and without (left breast) breast shielding in all patients. With this technique the superficial breast doses were calculated. To determine the average glandular breast radiation dose, we imaged an anthropomorphic dosimetric phantom into which calibrated dosimeters were placed to measure the dose to the breast. The phantom was imaged using the same protocol. Radiation doses to the breasts with and without the breast shielding were measured and compared using the Student's t-test. Results: The mean radiation doses with and without the breast shield were 5.71 ± 1.1 mGy versus 9.08 ± 1.5 mGy, respectively. The breast shield provided a 37.12% decrease in radiation dose to the breast with shielding. The difference between the dose received by the breasts with and without bismuth shielding was significant, with a p-value of less than 0.001. Conclusion: The high radiation during MDCT greatly exceeds the recommended doses and should not be underestimated. Bismuth in plane shielding for coronary calcium scoring with MDCT decreased the radiation dose to the breast. We recommend routine use of breast shields in female patients undergoing calcium scoring with MDCT

Benchmarking residual dose rates in a NuMI-like environment

Energy Technology Data Exchange (ETDEWEB)

Igor L. Rakhno et al.

2001-11-02

Activation of various structural and shielding materials is an important issue for many applications. A model developed recently to calculate residual activity of arbitrary composite materials for arbitrary irradiation and cooling times is presented in the paper. Measurements have been performed at the Fermi National Accelerator Laboratory using a 120 GeV proton beam to study induced radioactivation of materials used for beam line components and shielding. The calculated residual dose rates for the samples studied behind the target and outside of the thick shielding are presented and compared with the measured ones. Effects of energy spectra, sample material and dimensions, their distance from the shielding, and gaps between the shielding modules and walls as well as between the modules themselves were studied in detail.

Serial measurement of radiation leakage dose rates in safekeeping at the Gammaknife room

International Nuclear Information System (INIS)

Baba, Sadaaki; Nozaki, Kenichi; Toyoda, Tatsuya; Wakamatsu, Osamu; Machida, Toru

2006-01-01

We report the serial measurement of leakage dose rates in safekeeping at the Gammaknife room during the past 4 years and 9 months by scintillation survey meter. The leakage dose rates at the radiation boundaries were the same as the natural background levels. Leakage dose rates at each shield calculation point from two 90 Sr calibration sources contained in the storehouse were negligible compared with those from 60 Co sources of the Gammaknife. 60 Co sources of the Gammaknife are arranged in 201 pieces at 10 degree interval on the circumference and in five lines within an arc of 35 degrees. Its shield container is made of iron at least 43 cm thick. We got leakage dose rates less than 40% of the calculated values. We think it is caused by the difference of each actual distance and shield thickness because 60 Co sources are usually considered as a point source in the shield calculation. There are shutters opening up and down when patients go in and out to the direction of the couch. The leakage values to this direction were about twice as much as the calculated value. So, we knew the thickness of those shutters was thinner than 43 cm. The half life time of 60 Co source calculated from the serial measurements of leakage dose rates was 4.93 years on average. It is 94% of the physical half life value of 5.27 years. We judged it was acceptable considering the difficulty of measuring low dose rate level with the radiation survey meter. Very strong correlation was observed between the decrease of 60 Co dose rate acquired from one minute measurement at the center of 18 cm diameter polysterene phantom gotten from December 2000 to August 2005 and that of computation based on the physical half life time. Likewise there was strong and more correlation with leakage dose rate in the Gammaknife room. From this, we deduce the leakage dose rate decreases according to the theory of the disintegration of radioactivity with passage of time. Revised radiation related laws took effect

Field measurement and interpretation of beta doses and dose rates

International Nuclear Information System (INIS)

Selby, J.M.; Swinth, K.L.; Hooker, C.D.; Kenoyer, J.L.

1983-01-01

A wide variety of portable survey instruments employing GM, ionization chamber and scintillation detectors exist for the measurement of gamma exposure rates. Often these same survey instruments are used for monitoring beta fields. This is done by making measurements with and without a removable shield which is intended to shield out the non-penetrating component (beta) of the radiation field. The difference does not correspond to an absorbed dose rate for the beta field due to a variety of factors. Among these factors are the dependence on beta energy, source-detector geometries, mixed fields and variable ambient conditions. Attempting to use such measurements directly can lead to errors as high as a factor of 100. In many instances correction factors have been derived, that if properly applied, can reduce these errors substantially. However, this requires some knowledge of the beta spectra, calibration techniques and source geometry. This paper discusses some aspects of the proper use of instruments for beta measurements including the application of appropriate correction factors. Ionization type instruments are commonly used to measure beta dose rates. Through design and calibration these instruments will give an accurate reading only for uniform irradiation of the detection volume. Often in the field it is not feasible to meet these conditions. Large area uniform distributions of activity are not generally encountered and it is not possible to use large source-to-detector distances due to beta particle absorption in air. An example of correction factors required for various point sources is presented when a cutie pie ionization chamber is employed. The instrument reading is multiplied by the appropriate correction factor to obtain the dose rate at the window. When a different detector is used or for other geometries, a different set of correction factors must be used

The effect of patient shield position on gonad dose during lumbar spine radiography

International Nuclear Information System (INIS)

Clancy, Conor L.; O'Reilly, Geraldine; Brennan, Patrick C.; McEntee, Mark F.

2010-01-01

Background: In an effort to standardise radiological practices in the Republic of Ireland, current legislation states that 'written protocols for every type of standard radiological practice shall be established'. In order to fulfil this requirement the Irish Medical Council recommends the protocols issued by the Commission of European Communitees (CEC) for adoption in the country. Whilst this document does provide good guidance with regard to various radiographic factors, patient shielding instructions are notably ambiguous. The aim of this study was to remove some of this ambiguity by defining the optimal method of positioning patient shielding in antero-posterior (AP) and lateral lumbar spine radiographic examinations. These projections were chosen on the basis of their area of coverage being in close to and in some cases including the reproductive organs. They also represent the highest source of collective population dose of any conventional radiographic examination carried out in the UK. Method: A dosimetry study was devised to establish organ dose to the male testes and female ovaries using various clinically advocated methodologies for positioning patient shielding these included: no apron; tube-side apron; receptor-side apron and a wrap-around apron. The study was carried out using a direct digital radiography unit, an anthropomorphic phantom, various lead aprons and lithium thermoluminescent dosimeters (TLD). Results: For the AP projection, a statistically significant testes dose reduction of 42% (p ≤ 0.01) was observed when a tube-sided apron was used. No testes dose reductions were noted for the lateral projection. Ovary dose savings were not observed for any of the shielding methods investigated. Conclusion: This study found that the testes dose in AP examinations was reduced by 42% when patient shielding was positioned inferior to the imaged field and on the tube-side of the patient. This result validates the shielding methods used at the majority

Peripheral dose to the testes: the design and clinical use of a practical and effective gonadal shield

International Nuclear Information System (INIS)

Fraass, B.A.; Kinsella, T.J.; Harrington, F.S.; Galtstein, E.

1985-01-01

A simple and practical gonadal shield has been developed for use near megavoltage radiation fields. The lead shield encloses only the testes, allowing its use with nearly any radiation field that does not include the testes. The dose to the testes with and without the shield has been measured extensively both in phantoms and on patients. The gonadal shield allows a 3 to 10-fold reduction in dose to the testes depending primarily on the distance from the field edge to the gonads. When the shield is used, the gonadal dose is always less than 1% of the patient's prescription dose. Based on our patient studies of testicular injury following conventionally-fractionated irradiation, a dose of less than 50 cGy should preserve normal testicular function

Influence of lead apron shielding on absorbed doses from cone-beam computed tomography

International Nuclear Information System (INIS)

Rottke, Dennis; Andersson, Jonas; Ejima, Ken-Ichiro; Sawada, Kunihiko; Schulze, Dirk

2017-01-01

The aim of the present work was to investigate absorbed and to calculate effective doses (EDs) in cone-beam computed tomography (CBCT). The study was conducted using examination protocols with and without lead apron shielding. A full-body male RANDO"R phantom was loaded with 110 GR200A thermoluminescence dosemeter chips at 55 different sites and set up in two different CBCT systems (CS 9500"R, ProMax"R 3D). Two different protocols were performed: the phantom was set up (1) with and (2) without a lead apron. No statistically significant differences in organ and absorbed doses from regions outside the primary beam could be found when comparing results from exposures with and without lead apron shielding. Consequently, calculating the ED showed no significant differences between the examination protocols with and without lead apron shielding. For the ProMax"R 3D with shielding, the ED was 149 μSv, and for the examination protocol without shielding 148 μSv (SD = 0.31 μSv). For the CS 9500"R, the ED was 88 and 86 μSv (SD = 0.95 μSv), respectively, with and without lead apron shielding. The results revealed no statistically significant differences in the absorbed doses between examination with and without lead apron shielding, especially in organs outside the primary beam. (authors)

Efficacy of a radiation absorbing shield in reducing dose to the interventionalist during peripheral endovascular procedures: a single centre pilot study.

Science.gov (United States)

Power, S; Mirza, M; Thakorlal, A; Ganai, B; Gavagan, L D; Given, M F; Lee, M J

2015-06-01

This prospective pilot study was undertaken to evaluate the feasibility and effectiveness of using a radiation absorbing shield to reduce operator dose from scatter during lower limb endovascular procedures. A commercially available bismuth shield system (RADPAD) was used. Sixty consecutive patients undergoing lower limb angioplasty were included. Thirty procedures were performed without the RADPAD (control group) and thirty with the RADPAD (study group). Two separate methods were used to measure dose to a single operator. Thermoluminescent dosimeter (TLD) badges were used to measure hand, eye, and unshielded body dose. A direct dosimeter with digital readout was also used to measure eye and unshielded body dose. To allow for variation between control and study groups, dose per unit time was calculated. TLD results demonstrated a significant reduction in median body dose per unit time for the study group compared with controls (p = 0.001), corresponding to a mean dose reduction rate of 65 %. Median eye and hand dose per unit time were also reduced in the study group compared with control group, however, this was not statistically significant (p = 0.081 for eye, p = 0.628 for hand). Direct dosimeter readings also showed statistically significant reduction in median unshielded body dose rate for the study group compared with controls (p = 0.037). Eye dose rate was reduced for the study group but this was not statistically significant (p = 0.142). Initial results are encouraging. Use of the shield resulted in a statistically significant reduction in unshielded dose to the operator's body. Measured dose to the eye and hand of operator were also reduced but did not reach statistical significance in this pilot study.

Design of emergency shield

International Nuclear Information System (INIS)

Soliman, S.E.

1993-01-01

Manufacturing of an emergency movable shield in the hot laboratories center is urgently needed for the safety of personnel in case of accidents or spilling of radioactive materials. In this report, a full design for an emergency shield is presented and the corresponding dose rates behind the shield for different activities (from 1 mCi to 5 Ci) was calculated by using micro shield computer code. 4 figs., 1 tab

Calculation of shielding and radiation doses for PET/CT nuclear medicine facility

International Nuclear Information System (INIS)

Mollah, A.S.; Muraduzzaman, S.M.

2011-01-01

Positron emission tomography (PET) is a new modality that is gaining use in nuclear medicine. The use of PET and computed tomography (CT) has grown dramatically. Because of the high energy of the annihilation radiation (511 keV), shielding requirements are an important consideration in the design of a PET or PET/CT imaging facility. The goal of nuclear medicine and PET facility shielding design is to keep doses to workers and the public as low as reasonably achievable (ALARA). Design involves: 1. Calculation of doses to occupants of the facility and adjacent regions based on projected layouts, protocols and workflows, and 2. Reduction of doses to ALARA through adjustment of the aforementioned parameters. The radiological evaluation of a PET/CT facility consists of the assessment of the annual effective dose both to workers occupationally exposed, and to members of the public. This assessment takes into account the radionuclides involved, the facility features, the working procedures, the expected number of patients per year, and so on. The objective of the study was to evaluate shielding requirements for a PET/CT to be installed in the department of nuclear medicine of Bangladesh Atomic Energy Commission (BAEC). Minimizing shielding would result in a possible reduction of structural as well as financial burden. Formulas and attenuation coefficients following the basic AAPM guidelines were used to calculate un-attenuated radiation through shielding materials. Doses to all points on the floor plan are calculated based primarily on the AAPM guidelines and include consideration of broad beam attenuation and radionuclide energy and decay. The analysis presented is useful for both, facility designers and regulators. (author)

American National Standard: neutron and gamma-ray flux-to-dose rate factors

International Nuclear Information System (INIS)

Anon.

1977-01-01

This Standard presents data recommended for computing biological dose rates due to neutron and gamma-ray radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are given; the energy range for the gamma-ray conversion factors is 0.01 to 15 MeV. Specifically, this Standard is intended for use by shield designers to calculate wholebody dose rates to radiation workers and the general public. Establishing dose-rate limits is outside the scope of this Standard. Use of this Standard in cases where the dose equivalents are far in excess of occupational exposure guidelines is not recommended

Radiation dose reduction to the male gonads during MDCT: the effectiveness of a lead shield.

Science.gov (United States)

Hohl, Christian; Mahnken, Andreas H; Klotz, Ernst; Das, Marco; Stargardt, Achim; Mühlenbruch, Georg; Schmidt, Thorsten; Günther, Rolf W; Wildberger, Joachim E

2005-01-01

Our study was designed to quantify the effect of a standard gonad shield on the testicular radiation exposure due to scatter during routine abdominopelvic MDCT. Routine abdominopelvic MDCT was performed in 34 patients with gonadal lead shielding and 32 patients without this shielding; the testes were not exposed to the direct beam during the examination. We estimated the testicular dose administered with thermoluminescent dosimetry, taking into account each patient's body weight and body mass index (BMI). With a 1-mm lead shield, the mean testicular dose was reduced from 2.40 to 0.32 mSv, a reduction of 87%. The difference was found to be statistically significant (p Shielding the male gonads reduces the testicular radiation dose during abdominopelvic MDCT significantly and can be recommended for routine use.

SU-C-BRB-06: Utilizing 3D Scanner and Printer for Dummy Eye-Shield: Artifact-Free CT Images of Tungsten Eye-Shield for Accurate Dose Calculation

International Nuclear Information System (INIS)

Park, J; Lee, J; Kim, H; Kim, I; Ye, S

2015-01-01

Purpose: To evaluate the effect of a tungsten eye-shield on the dose distribution of a patient. Methods: A 3D scanner was used to extract the dimension and shape of a tungsten eye-shield in the STL format. Scanned data was transferred into a 3D printer. A dummy eye shield was then produced using bio-resin (3D systems, VisiJet M3 Proplast). For a patient with mucinous carcinoma, the planning CT was obtained with the dummy eye-shield placed on the patient’s right eye. Field shaping of 6 MeV was performed using a patient-specific cerrobend block on the 15 x 15 cm 2 applicator. The gantry angle was 330° to cover the planning target volume near by the lens. EGS4/BEAMnrc was commissioned from our measurement data from a Varian 21EX. For the CT-based dose calculation using EGS4/DOSXYZnrc, the CT images were converted to a phantom file through the ctcreate program. The phantom file had the same resolution as the planning CT images. By assigning the CT numbers of the dummy eye-shield region to 17000, the real dose distributions below the tungsten eye-shield were calculated in EGS4/DOSXYZnrc. In the TPS, the CT number of the dummy eye-shield region was assigned to the maximum allowable CT number (3000). Results: As compared to the maximum dose, the MC dose on the right lens or below the eye shield area was less than 2%, while the corresponding RTP calculated dose was an unrealistic value of approximately 50%. Conclusion: Utilizing a 3D scanner and a 3D printer, a dummy eye-shield for electron treatment can be easily produced. The artifact-free CT images were successfully incorporated into the CT-based Monte Carlo simulations. The developed method was useful in predicting the realistic dose distributions around the lens blocked with the tungsten shield

SU-C-BRB-06: Utilizing 3D Scanner and Printer for Dummy Eye-Shield: Artifact-Free CT Images of Tungsten Eye-Shield for Accurate Dose Calculation

Energy Technology Data Exchange (ETDEWEB)

Park, J; Lee, J [Program in Biomedical Radiation Sciences, Department of Transdisciplinary Studies, Graduate School of Convergence Science and Technology, Seoul National University, Seoul (Korea, Republic of); Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Kim, H [Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Kim, I [Department of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Ye, S [Program in Biomedical Radiation Sciences, Department of Transdisciplinary Studies, Graduate School of Convergence Science and Technology, Seoul National University, Seoul (Korea, Republic of); Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Department of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Advanced Institutes of Convergence Technology, Seoul National University, Suwon (Korea, Republic of)

2015-06-15

Purpose: To evaluate the effect of a tungsten eye-shield on the dose distribution of a patient. Methods: A 3D scanner was used to extract the dimension and shape of a tungsten eye-shield in the STL format. Scanned data was transferred into a 3D printer. A dummy eye shield was then produced using bio-resin (3D systems, VisiJet M3 Proplast). For a patient with mucinous carcinoma, the planning CT was obtained with the dummy eye-shield placed on the patient’s right eye. Field shaping of 6 MeV was performed using a patient-specific cerrobend block on the 15 x 15 cm{sup 2} applicator. The gantry angle was 330° to cover the planning target volume near by the lens. EGS4/BEAMnrc was commissioned from our measurement data from a Varian 21EX. For the CT-based dose calculation using EGS4/DOSXYZnrc, the CT images were converted to a phantom file through the ctcreate program. The phantom file had the same resolution as the planning CT images. By assigning the CT numbers of the dummy eye-shield region to 17000, the real dose distributions below the tungsten eye-shield were calculated in EGS4/DOSXYZnrc. In the TPS, the CT number of the dummy eye-shield region was assigned to the maximum allowable CT number (3000). Results: As compared to the maximum dose, the MC dose on the right lens or below the eye shield area was less than 2%, while the corresponding RTP calculated dose was an unrealistic value of approximately 50%. Conclusion: Utilizing a 3D scanner and a 3D printer, a dummy eye-shield for electron treatment can be easily produced. The artifact-free CT images were successfully incorporated into the CT-based Monte Carlo simulations. The developed method was useful in predicting the realistic dose distributions around the lens blocked with the tungsten shield.

A three-dimensional computed tomography-assisted Monte Carlo evaluation of ovoid shielding on the dose to the bladder and rectum in intracavitary radiotherapy for cervical cancer

International Nuclear Information System (INIS)

Gifford, Kent A.; Horton, John L.; Pelloski, Christopher E.; Jhingran, Anuja; Court, Laurence E.; Mourtada, Firas; Eifel, Patricia J.

2005-01-01

Purpose: To determine the effects of Fletcher Suit Delclos ovoid shielding on dose to the bladder and rectum during intracavitary radiotherapy for cervical cancer. Methods and Materials: The Monte Carlo method was used to calculate the dose in 12 patients receiving low-dose-rate intracavitary radiotherapy with both shielded and unshielded ovoids. Cumulative dose-difference surface histograms were computed for the bladder and rectum. Doses to the 2-cm 3 and 5-cm 3 volumes of highest dose were computed for the bladder and rectum with and without shielding. Results: Shielding affected dose to the 2-cm 3 and 5-cm 3 volumes of highest dose for the rectum (10.1% and 11.1% differences, respectively). Shielding did not have a major impact on the dose to the 2-cm 3 and 5-cm 3 volumes of highest dose for the bladder. The average dose reduction to 5% of the surface area of the bladder was 53 cGy. Reductions as large as 150 cGy were observed to 5% of the surface area of the bladder. The average dose reduction to 5% of the surface area of the rectum was 195 cGy. Reductions as large as 405 cGy were observed to 5% of the surface area of the rectum. Conclusions: Our data suggest that the ovoid shields can greatly reduce the radiation dose delivered to the rectum. We did not find the same degree of effect on the dose to the bladder. To calculate the dose accurately, however, the ovoid shields must be included in the dose model

Peripheral dose to the testes: the design and clinical use of a practical and effective gonadal shield.

Science.gov (United States)

Fraass, B A; Kinsella, T J; Harrington, F S; Glatstein, E

1985-03-01

A simple and practical gonadal shield has been developed for use near megavoltage radiation fields. The lead shield encloses only the testes, allowing its use with nearly any radiation field that does not include the testes. The dose to the testes with and without the shield has been measured extensively both in phantoms and on patients. The gonadal shield allows a 3 to 10-fold reduction in dose to the testes depending primarily on the distance from the field edge to the gonads. When the shield is used, the gonadal dose is always less than 1% of the patient's prescription dose. Based on our patient studies of testicular injury following conventionally-fractionated irradiation, a dose of less than 50 cGy (1% of a typical 5000 cGy treatment regimen) should preserve normal testicular function.

Shielding structure analysis for LSDS facility

Energy Technology Data Exchange (ETDEWEB)

Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization.

Shielding structure analysis for LSDS facility

International Nuclear Information System (INIS)

Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong

2014-01-01

The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

McKissock, B.I.; Bloomfield, H.S.

1990-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

Mckissock, B.I.; Bloomfield, H.S.

1989-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

Evaluation of induced activity, decay heat and dose rate distribution after shutdown in ITER

Energy Technology Data Exchange (ETDEWEB)

Maki, Koichi [Hitachi Ltd., Ibaraki (Japan). Hitachi Research Lab.; Satoh, Satoshi; Hayashi, Katsumi; Yamada, Koubun; Takatsu, Hideyuki; Iida, Hiromasa

1997-03-01

Induced activity, decay heat and dose rate distributions after shutdown were estimated for 1MWa/m{sup 2} operation in ITER. The activity in the inboard blanket one day after shutdown is 1.5x10{sup 11}Bq/cm{sup 3}, and the average decay heating rate 0.01w/cm{sup 3}. The dose rate outside the 120cm thick concrete biological shield is two order higher than the design criterion of 5{mu}Sv/h. This indicates that the biological shield thickness should be enhanced by 50cm in concrete, that is, total thickness 170cm for workers to enter the reactor room and to perform maintenance. (author)

Neutron dose rate in the upper part of a PWR containment. Comparison between measurements and TRIPOLI-2 calculations

International Nuclear Information System (INIS)

Vergnaud, T.; Bourdet, L.; Gonnord, J.; Nimal, J.C.; Champion, G.

1984-01-01

Conception of a reactor building requires large openings in the primary concrete shield for a postulated loss-of-coolant accident. Through these openings neutrons escape and produce dose rates in several parts of the reactor building. Some calculations using ANISN, DOT and essentially TRIPOLI-2 codes allow to compute the neutron dose rates at several places such as reactor containment operating floor and containment annulus. Some complementary shields are provided and the instrumentations are placed in area where the dose rate is lower. Comparisons are presented between measurements and calculations

Electron dose distributions caused by the contact-type metallic eye shield: Studies using Monte Carlo and pencil beam algorithms

Energy Technology Data Exchange (ETDEWEB)

Kang, Sei-Kwon; Yoon, Jai-Woong; Hwang, Taejin; Park, Soah; Cheong, Kwang-Ho; Jin Han, Tae; Kim, Haeyoung; Lee, Me-Yeon; Ju Kim, Kyoung, E-mail: kjkim@hallym.or.kr; Bae, Hoonsik

2015-10-01

A metallic contact eye shield has sometimes been used for eyelid treatment, but dose distribution has never been reported for a patient case. This study aimed to show the shield-incorporated CT-based dose distribution using the Pinnacle system and Monte Carlo (MC) calculation for 3 patient cases. For the artifact-free CT scan, an acrylic shield machined as the same size as that of the tungsten shield was used. For the MC calculation, BEAMnrc and DOSXYZnrc were used for the 6-MeV electron beam of the Varian 21EX, in which information for the tungsten, stainless steel, and aluminum material for the eye shield was used. The same plan was generated on the Pinnacle system and both were compared. The use of the acrylic shield produced clear CT images, enabling delineation of the regions of interest, and yielded CT-based dose calculation for the metallic shield. Both the MC and the Pinnacle systems showed a similar dose distribution downstream of the eye shield, reflecting the blocking effect of the metallic eye shield. The major difference between the MC and the Pinnacle results was the target eyelid dose upstream of the shield such that the Pinnacle system underestimated the dose by 19 to 28% and 11 to 18% for the maximum and the mean doses, respectively. The pattern of dose difference between the MC and the Pinnacle systems was similar to that in the previous phantom study. In conclusion, the metallic eye shield was successfully incorporated into the CT-based planning, and the accurate dose calculation requires MC simulation.

Shielding assessment for the proposed HRIBF upgrade to the National ISOL Facility

International Nuclear Information System (INIS)

Slater, C.O.; Olsen, D.K.; Johnson, J.O.; Lillie, R.A.; Gabriel, T.A.

1997-04-01

An upgrade of the existing ORNL Holifield Radioactive Ion Beam Facility (HRIBF) to the National Radioactive Ion Beam Isotope Separator On Line (RIB ISOL) Facility is being proposed. Part of the upgrade involves increasing the source proton energy and current, resulting in more intense, higher energy radiation. Shielding requirements for the proposed upgrade to the HRIBF have been assessed with respect to weight, space, and dose-rate constraints. Shielding assessments were made for operating, shutdown, and accident conditions. The results indicate reasonable shielding solutions for the target room except for the marginal dose rate on the roof. Shielding requirements in the target room were greatly reduced by decisions to move the target to a more interior room and to direct the proton beam downward into the target. A slightly more difficult shielding problem arises for proton beam extraction losses from the cyclotron. Here, the assumed isotropic beam losses (hence, neutron emissions) mean higher roof dose rates than those over the target room unless substantial localized shielding is placed over the cyclotron. Shutdown dose rates were found to present no problems. While dose rates through the sides of the facility during accident conditions will probably satisfy the accident dose-rate constraints, dose rates above the roof will be well above the constraints unless a solution is devised to shield the locations where beam losses are likely to occur. Ground activation analysis was postponed for this study

Influence of lead apron shielding on absorbed doses from cone-beam computed tomography.

Science.gov (United States)

Rottke, Dennis; Andersson, Jonas; Ejima, Ken-Ichiro; Sawada, Kunihiko; Schulze, Dirk

2017-06-01

The aim of the present work was to investigate absorbed and to calculate effective doses (EDs) in cone-beam computed tomography (CBCT). The study was conducted using examination protocols with and without lead apron shielding. A full-body male RANDO® phantom was loaded with 110 GR200A thermoluminescence dosemeter chips at 55 different sites and set up in two different CBCT systems (CS 9500®, ProMax® 3D). Two different protocols were performed: the phantom was set up (1) with and (2) without a lead apron. No statistically significant differences in organ and absorbed doses from regions outside the primary beam could be found when comparing results from exposures with and without lead apron shielding. Consequently, calculating the ED showed no significant differences between the examination protocols with and without lead apron shielding. For the ProMax® 3D with shielding, the ED was 149 µSv, and for the examination protocol without shielding 148 µSv (SD = 0.31 µSv). For the CS 9500®, the ED was 88 and 86 µSv (SD = 0.95 µSv), respectively, with and without lead apron shielding. The results revealed no statistically significant differences in the absorbed doses between examination with and without lead apron shielding, especially in organs outside the primary beam. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

International Nuclear Information System (INIS)

Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

2003-01-01

One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

Efficacy of a Radiation Absorbing Shield in Reducing Dose to the Interventionalist During Peripheral Endovascular Procedures: A Single Centre Pilot Study

International Nuclear Information System (INIS)

Power, S.; Mirza, M.; Thakorlal, A.; Ganai, B.; Gavagan, L. D.; Given, M. F.; Lee, M. J.

2015-01-01

PurposeThis prospective pilot study was undertaken to evaluate the feasibility and effectiveness of using a radiation absorbing shield to reduce operator dose from scatter during lower limb endovascular procedures.Materials and MethodsA commercially available bismuth shield system (RADPAD) was used. Sixty consecutive patients undergoing lower limb angioplasty were included. Thirty procedures were performed without the RADPAD (control group) and thirty with the RADPAD (study group). Two separate methods were used to measure dose to a single operator. Thermoluminescent dosimeter (TLD) badges were used to measure hand, eye, and unshielded body dose. A direct dosimeter with digital readout was also used to measure eye and unshielded body dose. To allow for variation between control and study groups, dose per unit time was calculated.ResultsTLD results demonstrated a significant reduction in median body dose per unit time for the study group compared with controls (p = 0.001), corresponding to a mean dose reduction rate of 65 %. Median eye and hand dose per unit time were also reduced in the study group compared with control group, however, this was not statistically significant (p = 0.081 for eye, p = 0.628 for hand). Direct dosimeter readings also showed statistically significant reduction in median unshielded body dose rate for the study group compared with controls (p = 0.037). Eye dose rate was reduced for the study group but this was not statistically significant (p = 0.142).ConclusionInitial results are encouraging. Use of the shield resulted in a statistically significant reduction in unshielded dose to the operator’s body. Measured dose to the eye and hand of operator were also reduced but did not reach statistical significance in this pilot study

Efficacy of a Radiation Absorbing Shield in Reducing Dose to the Interventionalist During Peripheral Endovascular Procedures: A Single Centre Pilot Study

Energy Technology Data Exchange (ETDEWEB)

Power, S.; Mirza, M.; Thakorlal, A.; Ganai, B.; Gavagan, L. D.; Given, M. F.; Lee, M. J., E-mail: mlee@rcsi.ie [Beaumont Hospital, Imaging and Interventional Radiology Department (Ireland)

2015-06-15

PurposeThis prospective pilot study was undertaken to evaluate the feasibility and effectiveness of using a radiation absorbing shield to reduce operator dose from scatter during lower limb endovascular procedures.Materials and MethodsA commercially available bismuth shield system (RADPAD) was used. Sixty consecutive patients undergoing lower limb angioplasty were included. Thirty procedures were performed without the RADPAD (control group) and thirty with the RADPAD (study group). Two separate methods were used to measure dose to a single operator. Thermoluminescent dosimeter (TLD) badges were used to measure hand, eye, and unshielded body dose. A direct dosimeter with digital readout was also used to measure eye and unshielded body dose. To allow for variation between control and study groups, dose per unit time was calculated.ResultsTLD results demonstrated a significant reduction in median body dose per unit time for the study group compared with controls (p = 0.001), corresponding to a mean dose reduction rate of 65 %. Median eye and hand dose per unit time were also reduced in the study group compared with control group, however, this was not statistically significant (p = 0.081 for eye, p = 0.628 for hand). Direct dosimeter readings also showed statistically significant reduction in median unshielded body dose rate for the study group compared with controls (p = 0.037). Eye dose rate was reduced for the study group but this was not statistically significant (p = 0.142).ConclusionInitial results are encouraging. Use of the shield resulted in a statistically significant reduction in unshielded dose to the operator’s body. Measured dose to the eye and hand of operator were also reduced but did not reach statistical significance in this pilot study.

Dose reduction during CT scanning in an anthropomorphic phantom by the use of a male gonad shield.

Science.gov (United States)

Price, R; Halson, P; Sampson, M

1999-05-01

Shielding the radiosensitive gonads during X-ray exposure has been advocated for plain film radiography for many years. In the UK, gonad shields are not widely employed in routine CT scanning, possibly owing to a perceived difficulty in protecting the gonads from a multidirectional X-ray source. The increasing numbers of CT scanners in the UK, with the large doses they deliver to patients, make potential dose reduction methods an important issue. This study measures the dose reduction achievable by shielding the male gonads with a lead wrap-around protection device. The reductions in dose when shielded both from direct radiation and from indirect radiation scattered from local tissues were studied. The use of the device resulted in a statistically significant reduction in the absorbed testicular dose from both direct and scattered radiation, with no increase in the dose measured in surrounding tissues. In three clinically relevant experimental protocols where the testes were not irradiated directly, the testicular absorbed dose from indirect scatter was reduced by 77-93% of the corresponding non-shielded figure. In these three experiments, image quality was unaltered by the use of the shield. A larger dose reduction was obtained when the shield was used to protect the testes from direct irradiation. However, this was achieved at the expense of considerable image degradation from streak artefact that would effectively prevent the clinical use of the device in this setting.

Dose rate calculations for a reconnaissance vehicle

International Nuclear Information System (INIS)

Grindrod, L.; Mackey, J.; Salmon, M.; Smith, C.; Wall, S.

2005-01-01

A Chemical Nuclear Reconnaissance System (CNRS) has been developed by the British Ministry of Defence to make chemical and radiation measurements on contaminated terrain using appropriate sensors and recording equipment installed in a land rover. A research programme is under way to develop and validate a predictive capability to calculate the build-up of contamination on the vehicle, radiation detector performance and dose rates to the occupants of the vehicle. This paper describes the geometric model of the vehicle and the methodology used for calculations of detector response. Calculated dose rates obtained using the MCBEND Monte Carlo radiation transport computer code in adjoint mode are presented. These address the transient response of the detectors as the vehicle passes through a contaminated area. Calculated dose rates were found to agree with the measured data to be within the experimental uncertainties, thus giving confidence in the shielding model of the vehicle and its application to other scenarios. (authors)

Calculation of the external dose rate in the spent fuel pool for the case to use compact racks

International Nuclear Information System (INIS)

Passos, E.M. dos; Alves, A.S.M.

1988-01-01

The possible introduction of compact racks in the spent fuel pool of the Angra 1 Nuclear Power Plant largely inreases its storage capacity, but originates an increase of the gamma radiation sources. The precise evaluation of the effects of the adoption of this option on the external gamma dose rates and also on the thickness of the concrete shielding requires the utilization of sofisticated computer codes (QAD, ANISN), which allow the calculation of the gamma dose rates through thick shielding walls. This paper describes the utilized methodology for the calculation of the modified pool shieldings, showing the obtained results for the Angra 1 NPP case. The gamma dose rate was calculated with the point Kernel model, first analytically, and later through utilization of the tridimensional multigroup QAD computer code. (author) [pt

Electron dose distributions caused by the contact-type metallic eye shield: Studies using Monte Carlo and pencil beam algorithms.

Science.gov (United States)

Kang, Sei-Kwon; Yoon, Jai-Woong; Hwang, Taejin; Park, Soah; Cheong, Kwang-Ho; Han, Tae Jin; Kim, Haeyoung; Lee, Me-Yeon; Kim, Kyoung Ju; Bae, Hoonsik

2015-01-01

A metallic contact eye shield has sometimes been used for eyelid treatment, but dose distribution has never been reported for a patient case. This study aimed to show the shield-incorporated CT-based dose distribution using the Pinnacle system and Monte Carlo (MC) calculation for 3 patient cases. For the artifact-free CT scan, an acrylic shield machined as the same size as that of the tungsten shield was used. For the MC calculation, BEAMnrc and DOSXYZnrc were used for the 6-MeV electron beam of the Varian 21EX, in which information for the tungsten, stainless steel, and aluminum material for the eye shield was used. The same plan was generated on the Pinnacle system and both were compared. The use of the acrylic shield produced clear CT images, enabling delineation of the regions of interest, and yielded CT-based dose calculation for the metallic shield. Both the MC and the Pinnacle systems showed a similar dose distribution downstream of the eye shield, reflecting the blocking effect of the metallic eye shield. The major difference between the MC and the Pinnacle results was the target eyelid dose upstream of the shield such that the Pinnacle system underestimated the dose by 19 to 28% and 11 to 18% for the maximum and the mean doses, respectively. The pattern of dose difference between the MC and the Pinnacle systems was similar to that in the previous phantom study. In conclusion, the metallic eye shield was successfully incorporated into the CT-based planning, and the accurate dose calculation requires MC simulation. Copyright © 2015 American Association of Medical Dosimetrists. Published by Elsevier Inc. All rights reserved.

The effect of patient shield position on gonad dose during lumbar spine radiography

Energy Technology Data Exchange (ETDEWEB)

Clancy, Conor L.; O' Reilly, Geraldine [Department of Medical Physics and Bioengineering, Garden Hill House, St. James Hospital, James Street, Dublin (Ireland); Brennan, Patrick C. [Discipline of Medical Radiation Sciences, C42 - Cumberland Campus, The University of Sydney, Lidcombe NSW 2141 (Australia); McEntee, Mark F. [School of Medicine and Medical Science, Health Sciences Centre, University College Dublin, Belfield, Dublin 4 (Ireland)], E-mail: mark.mcentee@ucd.ie

2010-05-15

Background: In an effort to standardise radiological practices in the Republic of Ireland, current legislation states that 'written protocols for every type of standard radiological practice shall be established'. In order to fulfil this requirement the Irish Medical Council recommends the protocols issued by the Commission of European Communitees (CEC) for adoption in the country. Whilst this document does provide good guidance with regard to various radiographic factors, patient shielding instructions are notably ambiguous. The aim of this study was to remove some of this ambiguity by defining the optimal method of positioning patient shielding in antero-posterior (AP) and lateral lumbar spine radiographic examinations. These projections were chosen on the basis of their area of coverage being in close to and in some cases including the reproductive organs. They also represent the highest source of collective population dose of any conventional radiographic examination carried out in the UK. Method: A dosimetry study was devised to establish organ dose to the male testes and female ovaries using various clinically advocated methodologies for positioning patient shielding these included: no apron; tube-side apron; receptor-side apron and a wrap-around apron. The study was carried out using a direct digital radiography unit, an anthropomorphic phantom, various lead aprons and lithium thermoluminescent dosimeters (TLD). Results: For the AP projection, a statistically significant testes dose reduction of 42% (p {<=} 0.01) was observed when a tube-sided apron was used. No testes dose reductions were noted for the lateral projection. Ovary dose savings were not observed for any of the shielding methods investigated. Conclusion: This study found that the testes dose in AP examinations was reduced by 42% when patient shielding was positioned inferior to the imaged field and on the tube-side of the patient. This result validates the shielding methods used at the

Fetal shielding combined with state of the art CT dose reduction strategies during maternal chest CT

Energy Technology Data Exchange (ETDEWEB)

Chatterson, Leslie C., E-mail: lch088@mail.usask.ca [Department of Diagnostic Imaging, University of Saskatchewan (Canada); Leswick, David A.; Fladeland, Derek A. [Department of Diagnostic Imaging, University of Saskatchewan (Canada); Hunt, Megan M.; Webster, Stephen [Saskatchewan Ministry of Labour Relations and Workplace Safety (Canada); Lim, Hyun [Department of Community Health and Epidemiology, College of Medicine, University of Saskatchewan (Canada)

2014-07-15

Purpose: Custom bismuth-antimony shields were previously shown to reduce fetal dose by 53% on an 8DR (detector row) CT scanner without dynamic adaptive section collimation (DASC), automatic tube current modulation (ATCM) or adaptive statistical iterative reconstruction (ASiR). The purpose of this study is to compare the effective maternal and average fetal organ dose reduction both with and without bismuth-antimony shields on a 64DR CT scanner using DASC, ATCM and ASiR during maternal CTPA. Materials and methods: A phantom with gravid prosthesis and a bismuth-antimony shield were used. Thermoluminescent dosimeters (TLDs) measured fetal radiation dose. The average fetal organ dose and effective maternal dose were determined using 100 kVp, scanning from the lung apices to the diaphragm utilizing DASC, ATCM and ASiR on a 64DR CT scanner with and without shielding in the first and third trimester. Isolated assessment of DASC was done via comparing a new 8DR scan without DASC to a similar scan on the 64DR with DASC. Results: Average third trimester unshielded fetal dose was reduced from 0.22 mGy ± 0.02 on the 8DR to 0.13 mGy ± 0.03 with the conservative 64DR protocol that included 30% ASiR, DASC and ATCM (42% reduction, P < 0.01). Use of a shield further reduced average third trimester fetal dose to 0.04 mGy ± 0.01 (69% reduction, P < 0.01). The average fetal organ dose reduction attributable to DASC alone was modest (6% reduction from 0.17 mGy ± 0.02 to 0.16 mGy ± 0.02, P = 0.014). First trimester fetal organ dose on the 8DR protocol was 0.07 mGy ± 0.03. This was reduced to 0.05 mGy ± 0.03 on the 64DR protocol without shielding (30% reduction, P = 0.009). Shields further reduced this dose to below accurately detectable levels. Effective maternal dose was reduced from 4.0 mSv on the 8DR to 2.5 mSv on the 64DR scanner using the conservative protocol (38% dose reduction). Conclusion: ASiR, ATCM and DASC combined significantly reduce effective maternal and fetal

Fetal shielding combined with state of the art CT dose reduction strategies during maternal chest CT

International Nuclear Information System (INIS)

Chatterson, Leslie C.; Leswick, David A.; Fladeland, Derek A.; Hunt, Megan M.; Webster, Stephen; Lim, Hyun

2014-01-01

Purpose: Custom bismuth-antimony shields were previously shown to reduce fetal dose by 53% on an 8DR (detector row) CT scanner without dynamic adaptive section collimation (DASC), automatic tube current modulation (ATCM) or adaptive statistical iterative reconstruction (ASiR). The purpose of this study is to compare the effective maternal and average fetal organ dose reduction both with and without bismuth-antimony shields on a 64DR CT scanner using DASC, ATCM and ASiR during maternal CTPA. Materials and methods: A phantom with gravid prosthesis and a bismuth-antimony shield were used. Thermoluminescent dosimeters (TLDs) measured fetal radiation dose. The average fetal organ dose and effective maternal dose were determined using 100 kVp, scanning from the lung apices to the diaphragm utilizing DASC, ATCM and ASiR on a 64DR CT scanner with and without shielding in the first and third trimester. Isolated assessment of DASC was done via comparing a new 8DR scan without DASC to a similar scan on the 64DR with DASC. Results: Average third trimester unshielded fetal dose was reduced from 0.22 mGy ± 0.02 on the 8DR to 0.13 mGy ± 0.03 with the conservative 64DR protocol that included 30% ASiR, DASC and ATCM (42% reduction, P < 0.01). Use of a shield further reduced average third trimester fetal dose to 0.04 mGy ± 0.01 (69% reduction, P < 0.01). The average fetal organ dose reduction attributable to DASC alone was modest (6% reduction from 0.17 mGy ± 0.02 to 0.16 mGy ± 0.02, P = 0.014). First trimester fetal organ dose on the 8DR protocol was 0.07 mGy ± 0.03. This was reduced to 0.05 mGy ± 0.03 on the 64DR protocol without shielding (30% reduction, P = 0.009). Shields further reduced this dose to below accurately detectable levels. Effective maternal dose was reduced from 4.0 mSv on the 8DR to 2.5 mSv on the 64DR scanner using the conservative protocol (38% dose reduction). Conclusion: ASiR, ATCM and DASC combined significantly reduce effective maternal and fetal

Effect of compositional variation in plutonium on process shielding design

Energy Technology Data Exchange (ETDEWEB)

Brown, T.H.

1997-11-01

Radiation dose rate from plutonium with high {sup 239}Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations.

Effect of compositional variation in plutonium on process shielding design

International Nuclear Information System (INIS)

Brown, T.H.

1997-11-01

Radiation dose rate from plutonium with high 239 Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations

Dose rate in the reactor room and environment during maintenance in fusion reactors

International Nuclear Information System (INIS)

Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

1995-01-01

According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

Testicular shield for para-aortic radiotherapy and estimation of gonad doses

OpenAIRE

Ravichandran, R.; Binukumar, J. P.; Kannadhasan, S.; Shariff, M. H.; Ghamrawy, Kamal El

2008-01-01

For radiotherapy of para-aortic and abdominal regions in male patients, gonads are to be protected to receive less than 2% of the prescribed dose. A testicular shield was fabricated for abdominal radiotherapy with 15 MV X-rays ((Clinac 2300 CD, Varian AG) with low melting point alloy (Cerroband). The dimensions of the testicular shield were 6.5 cm diameter and 3.5 cm depth with 1.5 cm wall thickness. During treatment, this shield was held in position by a rectangular sponge and Styrofo...

Modelling the Influence of Shielding on Physical and Biological Organ Doses

CERN Document Server

Ballarini, Francesca; Ferrari, Alfredo; Ottolenghi, Andrea; Pelliccioni, Maurizio; Scannicchio, Domenico

2002-01-01

Distributions of "physical" and "biological" dose in different organs were calculated by coupling the FLUKA MC transport code with a geometrical human phantom inserted into a shielding box of variable shape, thickness and material. While the expression "physical dose" refers to the amount of deposited energy per unit mass (in Gy), "biological dose" was modelled with "Complex Lesions" (CL), clustered DNA strand breaks calculated in a previous work based on "event-by-event" track-structure simulations. The yields of complex lesions per cell and per unit dose were calculated for different radiation types and energies, and integrated into a version of FLUKA modified for this purpose, allowing us to estimate the effects of mixed fields. As an initial test simulation, the phantom was inserted into an aluminium parallelepiped and was isotropically irradiated with 500 MeV protons. Dose distributions were calculated for different values of the shielding thickness. The results were found to be organ-dependent. In most ...

Analysis of shield for the nuclear ship MUTSU

International Nuclear Information System (INIS)

Fuse, Takayoshi; Takeuchi, Kiyoshi; Yamaji, Akio

1975-01-01

On the nuclear ship MUTSU, a higher-than-expected level of radiation was found, with output raised to 1.4 per cent. To investigate the radiation leakage, the analysis of the shielding problem utilized a four-step sequence of PALLAS-2DCY cylindrical r-z calculations with fixed sources distributions in the core. The neutron dose contours show the importance of streaming in the gap between the reactor vessel and the primary shield. Dominant consideration of thermal insulation exclude shielding from this area resulting in an imbalance in the shielding effectiveness. The neutron dose rate at the upper part of the reactor vessel is increased by neutrons incident on the head from cavity scattering. The calculation indicates that the neutron dose rate at the top of the primary shield is 5 rem/hr at 100 per cent output. (auth.)

Organ dose assessment of nuclear medicine practitioners using L-block shielding device for handing diagnostic radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Kang, Se Sik; Kim, Jung Hoon [Dep. of Radiological Science, College of Health Science, Catholic University of Pusan, Busan (Korea, Republic of); Cho, Yong In [Dept. of Diagnostic Radiology, Dongnam Institute of Radiological and Medical Science, Busan (Korea, Republic of)

2017-03-15

In the case of nuclear medicine practitioners in medical institutions, a wide range of exposure dose to individual workers can be found, depending on the type of source, the amount of radioactivity, and the use of shielding devices in handling radioactive isotopes. In this regard, this study evaluated the organ dose on practitioners as well as the dose reduction effect of the L-block shielding device in handling the diagnostic radiation source through the simulation based on the Monte Carlo method. As a result, the distribution of organ dose was found to be higher as the position of the radiation source was closer to the handling position of a practitioner, and the effective dose distribution was different according to the ICRP tissue weight. Furthermore, the dose reduction effect according to the L-block thickness tended to decrease, which showed the exponential distribution, as the shielding thickness increased. The dose reduction effect according to each radiation source showed a low shielding effect in proportion to the emitted gamma ray energy level.

Shielding considerations for neutral-beam injection systems

International Nuclear Information System (INIS)

de Seynes, X.

1983-03-01

Results of a study on the geometry of an FED-A Neutral Beam Injector beamline duct shield are presented. Also included is a calculation of dose rates, as a function of time, from an activated NBI. The shielding investigations consisted of varying the parameters of the geometry and transporting particles through it using the MCNP Monte-Carlo code. The dose rates were calculated by the ACDOS3 code using realistic MCNP results. A final-to-incident flux ratio of 6.5 x 10 -7 can be achieved through the use of a 65.5 cm reentry duct. This is for a realistic source and pure water shielding material. The activated NBI produced a dose rate of 15.9 mrem/hr two and a half days after shutdown of the reactor

Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

International Nuclear Information System (INIS)

Ueki, Kohtaro; Kawakami, Kazuo; Shimizu, Daisuke

2003-01-01

The Monte Carlo coupling technique with the coordinate transformation is used to evaluate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the 'SSW-SSR/CRT calculation system'. In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60 Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house

Shielding effect of building to natural radiation and its influence to population dose evaluation

International Nuclear Information System (INIS)

Nakamura, Takashi; Itoh, Kazuo; Yoshimura, Toshiaki.

1980-01-01

This work investigated the shielding effect of the building which is indispensable for the accurate evaluation of the population dose of external exposure from natural radiation. At first, the attenuation coefficients of various building materials were measured and found to agree with the calculated values within 10% errors. The shielding factors of these materials were calculated from the calculated attenuation coefficients and buildup factors. The shielding factors of the wall, window, roof and floor were calculated separately by settling the model houses and combining the shielding factors of the building materials used, and then the shielding factor of the whole building was obtained by use of the opening fraction of the wall and the fractions of the wall, roof and floor areas to the total floor area. The influence of the shielding effect of the building is well represented by the occupancy factor which is the ratio of the group doses including that shielding effect to those excluding it. The occupancy factor lies between 0.9 and 1.0 for four specified districts, Tokyo, Osaka, Ibaraki and Nagano. (author)

Comparison between measured and design dose rate equivalents on board of Nuclear Ship Mutsu

International Nuclear Information System (INIS)

Yamaji, Akio; Sakamoto, Yukio

1993-01-01

The power-up test of the Nuclear Ship Mutsu was restarted in March 1990 and completed successfully in February 1991. The experimental voyages were carried out for about one year and all experiments were completed in February 1992. A comparison between the measured and design dose rate equivalents on board is described with showing a modified method in the shielding design. The measured values were obtained extensively in the cavity between the primary and secondary shields, in the double bottom, outside the secondary shield, and on the surface of the main coolant loop. The shielding design calculations were made with the most conservative geometries and material compositions within the allowed tolerance. In addition, a conservative model was adopted in case of performing the approximation due to the geometrical restriction of calculation code. The computational accuracies were evaluated based on various experimental analyses. The evaluated value was used as the design value. The shield structures were determined with a judgement that the real value does not exceed the design value. The adequacy of the judgement was confirmed by measurements on board. The measured dose rate equivalents in all positions on board satisfied the design criteria. (author)

MicroShield 6.20 computations to evaluate dose rate in unprotected materials

International Nuclear Information System (INIS)

Slaveikova, M.; Stanev, I.

2013-01-01

An analysis of compliance with the requirement of Art. 36 of the Regulation on the conditions and procedure of transport of radioactive material is made.This analysis is carried out in connection with the construction of sites for temporary storage of radioactive materials and radioactive wastes from decommissioning activities of the Kozloduy NPP units 1-4. The aim is to assess the dose in unprotected materials. An analysis of the conformity with the requirements of the Bulgarian legislation to assess the dose rate of material in the absence of physical or other barriers. Many calculations are carried out to assess the dose rate around a piece of metal from the dismantling of the primary circuit, which is conservatively assumed that contamination is greatest

A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

International Nuclear Information System (INIS)

Lee, Yoon Hee

2006-02-01

Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon Hee

2006-02-15

Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

Neutron shielding and activation of the MASTU device and surrounds

Energy Technology Data Exchange (ETDEWEB)

Taylor, David, E-mail: david.taylor@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lilley, Steven; Turner, Andrew [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Davis, Andrew [Now at College of Engineering, University of Wisconsin, Madison, WI 53706 (United States)

2014-10-15

Highlights: We model neutron shielding for the planned MASTU device; nadequacies in the existing shielding design are remedied; Levels of public exposure are considered; We model activated gamma emission for the device under a worst case scenario. Abstract: A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day. The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.

Shielding evaluation of neutron generator hall by Monte Carlo simulations

Energy Technology Data Exchange (ETDEWEB)

Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)

2017-04-01

A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)

Shielding study of a fusion machine. Elaboration of a global shielding calculation scheme for the Tokamak tore Supra

International Nuclear Information System (INIS)

Diop, C.M'B.

1984-01-01

This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr

Advanced local dose rate calculations with the Monte Carlo code MCNP for plutonium nitrate storage containers

International Nuclear Information System (INIS)

Quade, U.

1994-01-01

Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de

Shield device for controlling the dose of x-rays applied in an x-ray machine

International Nuclear Information System (INIS)

Charrier, P.

1983-01-01

This invention provides an improved shield for use with an x-ray machine. The shield can control the dose of x-rays applied by the machine in different areas without affecting the power of the x-rays. This is achieved with a shield especially designed and positioned to intercept with x-rays for longer or shorter periods in different areas during the taking of the picture, but not for the whole period of time necessary for taking this picture. Each area of the subject being x-rayed is exposed to full power x-rays. However, owing to the shield, the areas that require smaller dose receive these full power x-rays for a shorter portion of the time required to take the picture while the other areas that require larger dose of x-rays, receive the full power x-rays for a longer portion of the full period of time required to take the picture. To ensure this differential exposure, the shield is placed through the path of the x-rays and rotated about an axis which is generally transverse to the direction of travel of the x-rays to cut out some of said x-rays for different portions of the period of time necessary for taking the picture. The shield is preferably shaped to intercept x-rays for a longer period in some areas than in others depending on the required doses. A plurality of differently shaped shields can be provided to suit different picture taking situations

Simple shielding reduces dose to the contralateral breast during prone breast cancer radiotherapy

Energy Technology Data Exchange (ETDEWEB)

Goyal, Uma, E-mail: uma.goyal@gmail.com; Locke, Angela; Smith-Raymond, Lexie; Georgiev, Georgi N.

2016-07-01

Our goal was to design a prone breast shield for the contralateral breast and study its efficacy in decreasing scatter radiation to the contralateral breast in a prone breast phantom setup receiving radiation therapy designed for breast cancer. We constructed a prone breast phantom setup consisting of (1) A thermoplastic mask with a left-sided depression created by a water balloon for a breast shape; (2) 2 plastic bags to hold water in the thermoplastic mask depression; (3) 2000 mL of water to fill the thermoplastic mask depression to create a water-based false breast; (4) 1-cm thick bolus placed in the contralateral breast holder; (5) 2 lead (Pb) sheets, each 0.1-cm thick for blocking scatter radiation in the contralateral bolus-based false breast; (6) a prone breast board to hold the thermoplastic mask, water, bolus, and lead; (7) 9 cm solid water on top of the breast board to simulate body; (8) a diode was used to verify dose for each treatment field of the treated water-based breast; (9) metal–oxide–semiconductor-field effect transistor (MOSFET) dosimeters to measure dose to the contralateral bolus-based breast. The phantom prone breast setup was CT simulated and treatment was designed with 95% isodose line covering the treated breast. The maximum dose was 107.1%. Megavoltage (MV) port images ensured accurate setup. Measurements were done using diodes on the treated water-based breast and MOSFET dosimeters at the medial and lateral sides of the contralateral bolus-based breast without and with the Pb shield. Five treatments were done for each of the 3 data sets and recorded individually for statistical purposes. All treatments were completed with 6 MV photons at 200 cGy per treatment. The dose contributions from each of the 3 data sets including 15 treatments total without and with the prone lead shield to the medial and lateral portions of contralateral bolus-based breast were averaged individually. Unshielded dose means were 37.11 and 2.94 cGy, and

Shielding tests for a new ship for the transport of spent nuclear fuels

International Nuclear Information System (INIS)

Ito, D.; Kitano, T.; Akiyama, H.; Ueki, K.; Sanui, T.

1998-01-01

a new ship for the transport of spent nuclear fuels which uses serpentine concrete as its major shielding material has been constructed. The shielding calculations are based on DOT3.5 code (CCC-276) and the DLC23). Experiments with Cf-252 and Co-60 sources were carried out to confirm the validity of this method of calculating the shielding effectiveness of serpentine concrete. In these experiments, neutron and gamma-ray dose equivalent rates were measured in various arrangements to simulate the shielding structures of the ship, the calculations for each arrangement were performed by this shielding calculation method. For both neutron and gamma-rays, the calculation results agreed with the experiments very well, confirming that this calculation method used in the ship's shielding design is valid. Two kinds of on-board gamma-ray shielding tests were performed to confirm the ship's actual shielding effectiveness. In one kind of test, gamma-ray dose equivalent rates were measured for each shielding wall using Co-60 sources. In the other kind of test, gamma-ray dose equivalent rates in the ship's accommodation area were measured when a strong Co-60 source was placed in a loaded shipping cask's position. In both gamma-ray shielding tests all measured dose equivalent rates were less than the calculated values, confirming that the ship's actual shielding is sufficient to meet safety requirements. (authors)

Quantitative assessment of selective in-plane shielding of tissues in computed tomography through evaluation of absorbed dose and image quality

International Nuclear Information System (INIS)

Geleijns, J.; Veldkamp, W.J.H.; Salvado Artells, M.; Lopez Tortosa, M.; Calzado Cantera, A.

2006-01-01

This study aimed at assessment of efficacy of selective in-plane shielding in adults by quantitative evaluation of the achieved dose reduction and image quality. Commercially available accessories for in-plane shielding of the eye lens, thyroid and breast, and an anthropomorphic phantom were used for the evaluation of absorbed dose and image quality. Organ dose and total energy imparted were assessed by means of a Monte Carlo technique taking into account tube voltage, tube current, and scanner type. Image quality was quantified as noise in soft tissue. Application of the lens shield reduced dose to the lens by 27% and to the brain by 1%. The thyroid shield reduced thyroid dose by 26%; the breast shield reduced dose to the breasts by 30% and to the lungs by 15%. Total energy imparted (unshielded/shielded) was 88/86 mJ for computed tomography (CT) brain, 64/60 mJ for CT cervical spine, and 289/260 mJ for CT chest scanning. An increase in image noise could be observed in the ranges were bismuth shielding was applied. The observed reduction of organ dose and total energy imparted could be achieved more efficiently by a reduction of tube current. The application of in-plane selective shielding is therefore discouraged. (orig.)

Application of the personnel photographic monitoring method to determine equivalent radiation dose beyond proton accelerator shielding

International Nuclear Information System (INIS)

Gel'fand, E.K.; Komochkov, M.M.; Man'ko, B.V.; Salatskaya, M.I.; Sychev, B.S.

1980-01-01

Calculations of regularities to form radiation dose beyond proton accelerator shielding are carried out. Numerical data on photographic monitoring dosemeter in radiation fields investigated are obtained. It was shown how to determine the total equivalent dose of radiation fields beyond proton accelerator shielding by means of the photographic monitoring method by introduction into the procedure of considering nuclear emulsions of division of particle tracks into the black and grey ones. A comparison of experimental and calculational data has shown the applicability of the used calculation method for modelling dose radiation characteristics beyond proton accelerator shielding [ru

Influence of bismuth shielding use in the dose CT and the imaging quality in standard brain protocol

International Nuclear Information System (INIS)

Sanchez Carnoma, G.; Urena Llinares, A.; Santos Rubio, A.; Haro Madero, G.; Herrador Cordoba, M.

2006-01-01

The purpose of this work is to evaluate the reduction of lens doses, CT dose index (CTDI) and image quality when bismuth shielding is used. Dose indexes nCDTIs,p, w were measured for a head phantom (d=16 cm) in three situations: first, without shielding, second, shielded with bismuth over the phantom surface and, last, with 2 cm between both. Four regions of interest were selected, two including the theoretical eyes position and the others 4-6 cm below them. Noise and grey level in Housfield units were also measured in all geometries. A bismuth shielding in contact or 2 cm above surface provides a dose reduction of 36% y 29% respectively. In the first case an importation of image quality appears, increasing a 55% the Housfield units in the superficial regions of interest and a growth of noise 6 times greater. Bismuth protection provides an important reduction in lens dose., in the case of dire ct incidence radiation beam. (Author)

Radiation dose reduction at a price: the effectiveness of a male gonadal shield during helical CT scans

International Nuclear Information System (INIS)

Dauer, Lawrence T; Casciotta, Kevin A; Erdi, Yusuf E; Rothenberg, Lawrence N

2007-01-01

It is estimated that 60 million computed tomography (CT) scans were performed during 2006, with approximately 11% of those performed on children age 0–15 years. Various types of gonadal shielding have been evaluated for reducing exposure to the gonads. The purpose of this study was to quantify the radiation dose reduction to the gonads and its effect on image quality when a wrap-around male pediatric gonad shield was used during CT scanning. This information is obtained to assist the attending radiologist in the decision to utilize such male gonadal shields in pediatric imaging practice. The dose reduction to the gonads was measured for both direct radiation and for indirect scattered radiation from the abdomen. A 6 cm 3 ion chamber (Model 10X5-6, Radcal Corporation, Monrovia, CA) was placed on a Humanoid real bone pelvic phantom at a position of the male gonads. When exposure measurements with shielding were made, a 1 mm lead wrap-around gonadal shield was placed around the ion chamber sensitive volume. The use of the shields reduced scatter dose to the gonads by a factor of about 2 with no appreciable loss of image quality. The shields reduced the direct beam dose by a factor of about 35 at the expense of extremely poor CT image quality due to severe streak artifacts. Images in the direct exposure case are not useful due to these severe artifacts and the difficulties in positioning these shields on patients in the scatter exposure case may not be warranted by the small absolute reduction in scatter dose unless it is expected that the patient will be subjected to numerous future CT scans

Radiation dose reduction at a price: the effectiveness of a male gonadal shield during helical CT scans.

Science.gov (United States)

Dauer, Lawrence T; Casciotta, Kevin A; Erdi, Yusuf E; Rothenberg, Lawrence N

2007-03-16

It is estimated that 60 million computed tomography (CT) scans were performed during 2006, with approximately 11% of those performed on children age 0-15 years. Various types of gonadal shielding have been evaluated for reducing exposure to the gonads. The purpose of this study was to quantify the radiation dose reduction to the gonads and its effect on image quality when a wrap-around male pediatric gonad shield was used during CT scanning. This information is obtained to assist the attending radiologist in the decision to utilize such male gonadal shields in pediatric imaging practice. The dose reduction to the gonads was measured for both direct radiation and for indirect scattered radiation from the abdomen. A 6 cm3 ion chamber (Model 10X5-6, Radcal Corporation, Monrovia, CA) was placed on a Humanoid real bone pelvic phantom at a position of the male gonads. When exposure measurements with shielding were made, a 1 mm lead wrap-around gonadal shield was placed around the ion chamber sensitive volume. The use of the shields reduced scatter dose to the gonads by a factor of about 2 with no appreciable loss of image quality. The shields reduced the direct beam dose by a factor of about 35 at the expense of extremely poor CT image quality due to severe streak artifacts. Images in the direct exposure case are not useful due to these severe artifacts and the difficulties in positioning these shields on patients in the scatter exposure case may not be warranted by the small absolute reduction in scatter dose unless it is expected that the patient will be subjected to numerous future CT scans.

Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

International Nuclear Information System (INIS)

Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

2015-01-01

As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

Shielding for Critical Organs and Radiation Exposure Dose Distribution in Patients with High Energy Radiotherapy

International Nuclear Information System (INIS)

Chu, Sung Sil; Suh, Chang Ok; Kim, Gwi Eon

2002-01-01

High energy photon beams from medical linear accelerators produce large scattered radiation by various components of the treatment head, collimator and walls or objects in the treatment room including the patient. These scattered radiation do not provide therapeutic dose and are considered a hazard from the radiation safety perspective. Scattered dose of therapeutic high energy radiation beams are contributed significant unwanted dose to the patient. ICRP take the position that a dose of 500mGy may cause abortion at any stage of pregnancy and that radiation detriment to the fetus includes risk of mental retardation with a possible threshold in the dose response relationship around 100 mGy for the gestational period. The ICRP principle of As Low As Reasonably Achievable (ALARA) was recommended for protection of occupation upon the linear no-threshold dose response hypothesis for cancer induction. We suggest this ALARA principle be applied to the fetus and testicle in therapeutic treatment. Radiation dose outside a photon treatment filed is mostly due to scattered photons . This scattered dose is a function of the distance from the beam edge, treatment geometry, primary photon energy, and depth in the patient. The need for effective shielding of the fetus and testicle is reinforced when young patients are treated with external beam radiation therapy and then shielding designed to reduce the scattered photon dose to normal organs have to considered. Irradiation was performed in phantom using high energy photon beams produced by a Varian 2100C/D medical linear accelerator (Varian Oncology Systems, Polo Alto, CA) located at the Yonsei Cancer Center. The composite phantom used was comprised of a commercially available anthropomorphic Rando phantom (Phantom Laboratory Inc., Salem, YN) and a rectangular solid polystyrene phantom of dimensions 30cm x 30cm x 20cm. The anthropomorphic Rando phantom represents an average man made from tissue equivalent materials that is

Shielding for Critical Organs and Radiation Exposure Dose Distribution in Patients with High Energy Radiotherapy

Energy Technology Data Exchange (ETDEWEB)

Chu, Sung Sil; Suh, Chang Ok; Kim, Gwi Eon [Yonsei Univ., Seoul (Korea, Republic of)

2002-03-15

High energy photon beams from medical linear accelerators produce large scattered radiation by various components of the treatment head, collimator and walls or objects in the treatment room including the patient. These scattered radiation do not provide therapeutic dose and are considered a hazard from the radiation safety perspective. Scattered dose of therapeutic high energy radiation beams are contributed significant unwanted dose to the patient. ICRP take the position that a dose of 500mGy may cause abortion at any stage of pregnancy and that radiation detriment to the fetus includes risk of mental retardation with a possible threshold in the dose response relationship around 100 mGy for the gestational period. The ICRP principle of As Low As Reasonably Achievable (ALARA) was recommended for protection of occupation upon the linear no-threshold dose response hypothesis for cancer induction. We suggest this ALARA principle be applied to the fetus and testicle in therapeutic treatment. Radiation dose outside a photon treatment filed is mostly due to scattered photons . This scattered dose is a function of the distance from the beam edge, treatment geometry, primary photon energy, and depth in the patient. The need for effective shielding of the fetus and testicle is reinforced when young patients are treated with external beam radiation therapy and then shielding designed to reduce the scattered photon dose to normal organs have to considered. Irradiation was performed in phantom using high energy photon beams produced by a Varian 2100C/D medical linear accelerator (Varian Oncology Systems, Polo Alto, CA) located at the Yonsei Cancer Center. The composite phantom used was comprised of a commercially available anthropomorphic Rando phantom (Phantom Laboratory Inc., Salem, YN) and a rectangular solid polystyrene phantom of dimensions 30cm x 30cm x 20cm. The anthropomorphic Rando phantom represents an average man made from tissue equivalent materials that is

Calculation of parameters for an iron shield experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-01-01

In this text is carreid out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gama-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The tranpsort calculations were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reaction and doses rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented reasonable concordance with the experimental measurements. Finally, is presented a proposal for setting up of an experimental arrangement, using the IEA-R1 reactor, with the purpose of lay down a shielding benchmark. (Author) [pt

Beta particle dose rates to micro-organisms in soil

International Nuclear Information System (INIS)

Kabir, M.; Spiers, F.W.; Iinuma, Takeshi.

1977-01-01

Studies were made to estimate the beta-particle dose rates to micro-organisms of various sizes in soil. The small insects and organisms living in soil are constantly exposed to beta-radiation arising from naturally occuring radionuclides in soil as in this case no overlying tissue shields them. The technique of measuring beta-particle dose rate consisted of using of a thin plastic scintillator to measure the pulse height distribution as the beta particle traverses the scintillator. The integrated response was determined by the number and size of the photomultiplier pulses. From the data of soil analyses it was estimated that typically about 29% of the beta particles emitted per gm. of soil were contributed by the U/Ra series, 21% by the Th series and about 50% by potassium. By combining the individual spectra of these three radionuclides in the proportion found in a typical soil, a resultant spectrum was computed representing the energy distribution of the beta particles. The dose rate received by micro-organisms of different shape and size in soil was derived from the equilibrium dose rates combined with a 'Geometrical Factor' of the organisms. For small organisms, the dose rates did not vary between the spherical and cylindrical types, but in the case of larger organisms, the dose rates were found to be greater for the spherical types of the same diameter. (auth.)

EURISOL-DS multi-MW target unit: Neutronics performance and shielding assessment, dose rate and material activation calculations for the MAFF configuration

CERN Document Server

Romanets, Y; Kadi, Y; Luis, R; Goncalves, I F; Tecchio, L; Kharoua, C; Vaz, P; Ene, D; David, J C; Rocca, R; Negoita, F

2010-01-01

One of the objectives of the EURISOL (EURopean Isotope Separation On-Line Radioactive Ion Beam) Design Study consisted of providing a safe and reliable facility layout and design for the following operational parameters and characteristics: (a) a 4 MW proton beam of 1 GeV energy impinging on a mercury target (the converter); (b) high neutron fluxes (similar to 3 x 10(16) neutrons/s) generated by spallation reactions of the protons impinging in the converter and (c) fission rate on fissile U-235 targets in excess of 10(15) fissions/s. In this work, the state-of-the-art Monte Carlo codes MCNPX (Pelowitz, 2005) and FLUKA (Vlachoudis, 2009; Ferrari et al., 2008) were used to characterize the neutronics performance and to perform the shielding assessment (Herrera-Martinez and Kadi, 2006; Cornell, 2003) of the EURISOLTarget Unit and to provide estimations of dose rate and activation of different components, in view of the radiation safety assessment of the facility. Dosimetry and activation calculations were perfor...

Spatial variation of natural terrestrial gamma-ray dose rates in Brunei

International Nuclear Information System (INIS)

Hu, S.J.; Lai, K.K.; Manato, S.; Kodaira, K.

1998-01-01

A carbon survey of natural terrestrial gamma-rat dose rates along the main roads of the western part of Brunei Darussalam was carried out using two portable type 1.5 φ x 4 NaI(TI) and 1 φ x 2 NaI(TI) scintillation counters. A series of semicontinuous count rates measurements were performed inside a moving vehicle. This yielded equal-distance data which were analysed statistically to obtain the spatial variation of the natural terrestrial gamma-ray dose rates. The equal-distance data of dose rates were obtained by correcting for shielding effect of the car. The thickness of the pavement and the contribution from the pavement material were estimated from a correlation curve between the dose rates measured on pavements and on the nearby soils. A spectral analysis of the equal-distance data enabled us to clarify the structure of the spatial variation in dose rates. The data could be reasonably smoothened by removing the random noise components in a higher wave number region. (author). 6 refs., 7 figs., 1 tab

Spatial variation of natural terrestrial γ-ray dose rates in Brunei

International Nuclear Information System (INIS)

Hu, S.J.; Lai, K.K.

1998-01-01

A carborne survey of natural terrestrial y-ray dose rates along the main roads of the western part of Brunei Darussalam was carried out using two portable type 1.5'φx4' NaI(T1) and 1'φx2' NaI(T1) scintillation counters. A series of semicontinuous count rates measurements were performed inside a moving vehicle. This yielded equal-distance data which were analysed statistically to obtain the spatial variation of the natural terrestrial γ-ray dose rates. The equal-distance data of dose rates were obtained by correcting for shielding effect of the car. The thickness of the pavement and the contribution from the pavement material were estimated from a correlation curve between the dose rates measured on pavements and on the nearby soils. A spectral analysis of the equal-distance data enabled us to clarify the structure of the spatial variation in dose rates. The data could be reasonably smoothened by removing the random noise components in a higher wave number region

Usefulness assessment of secondary shield for the lens exposure dose reduction during radiation treatment of peripheral orbit

International Nuclear Information System (INIS)

Kwak, Yong Kuk; Hong, Sun Gi; Ha, Min Yong; Park, Jang Pil; Yoo, Sook Hyun; Cho, Woong

2015-01-01

This study presents the usefulness assessment of secondary shield for the lens exposure dose reduction during radiation treatment of peripheral orbit. We accomplished IMRT treatment plan similar with a real one through the computed treatment planning system after CT simulation using human phantom. For the secondary shield, we used Pb plate (thickness 3mm, diameter 25mm) and 3 mm tungsten eye-shield block. And we compared lens dose using OSLD between on TPS and on simulation. Also, we irradiated 200 MU(6 MV, SPD(Source to Phantom Distance)=100 cm, F·S 5×5 cm)on a 5 cm acrylic phantom using the secondary shielding material of same condition, 3 mm Pb and tungsten eye-shield block. And we carried out the same experiment using 8 cm Pb block to limit effect of leakage and transmitted radiation out of irradiation field. We attached OSLD with a 1cm away from the field at the side of phantom and applied a 3mm bolus equivalent to the thickness of eyelid. Using human phantom, the Lens dose on IMRT treatment plan is 315.9 cGy and the real measurement value is 216.7 cGy. And after secondary shield using 3mm Pb plate and tungsten eye-shield block, each lens dose is 234.3, 224.1 cGy. The result of a experiment using acrylic phantom, each value is 5.24, 5.42 and 5.39 cGy in case of no block, 3mm Pb plate and tungsten eye-shield block. Applying O.S.B out of the field, each value is 1.79, 2.00 and 2.02 cGy in case of no block, 3 mm Pb plate and tungsten eye-shield block. When secondary shielding material is used to protect critical organ while irradiating photon, high atomic number material (like metal) that is near by critical organ can be cause of dose increase according to treatment region and beam direction because head leakage and collimator and MLC transmitted radiation are exist even if it's out of the field. The attempt of secondary shield for the decrease of exposure dose was meaningful, but untested attempt can have a reverse effect. So, a preliminary inspection

Dosimetric evaluation of photon dose calculation under jaw and MLC shielding

International Nuclear Information System (INIS)

Fogliata, A.; Clivio, A.; Vanetti, E.; Nicolini, G.; Belosi, M. F.; Cozzi, L.

2013-01-01

Purpose: The accuracy of photon dose calculation algorithms in out-of-field regions is often neglected, despite its importance for organs at risk and peripheral dose evaluation. The present work has assessed this for the anisotropic analytical algorithm (AAA) and the Acuros-XB algorithms implemented in the Eclipse treatment planning system. Specifically, the regions shielded by the jaw, or the MLC, or both MLC and jaw for flattened and unflattened beams have been studied.Methods: The accuracy in out-of-field dose under different conditions was studied for two different algorithms. Measured depth doses out of the field, for different field sizes and various distances from the beam edge were compared with the corresponding AAA and Acuros-XB calculations in water. Four volumetric modulated arc therapy plans (in the RapidArc form) were optimized in a water equivalent phantom, PTW Octavius, to obtain a region always shielded by the MLC (or MLC and jaw) during the delivery. Doses to different points located in the shielded region and in a target-like structure were measured with an ion chamber, and results were compared with the AAA and Acuros-XB calculations. Photon beams of 6 and 10 MV, flattened and unflattened were used for the tests.Results: Good agreement between calculated and measured depth doses was found using both algorithms for all points measured at depth greater than 3 cm. The mean dose differences (±1SD) were −8%± 16%, −3%± 15%, −16%± 18%, and −9%± 16% for measurements vs AAA calculations and −10%± 14%, −5%± 12%, −19%± 17%, and −13%± 14% for Acuros-XB, for 6X, 6 flattening-filter free (FFF), 10X, and 10FFF beams, respectively. The same figures for dose differences relative to the open beam central axis dose were: −0.1%± 0.3%, 0.0%± 0.4%, −0.3%± 0.3%, and −0.1%± 0.3% for AAA and −0.2%± 0.4%, −0.1%± 0.4%, −0.5%± 0.5%, and −0.3%± 0.4% for Acuros-XB. Buildup dose was overestimated with AAA, while Acuros-XB gave

Radiation dose reduction at a price: the effectiveness of a male gonadal shield during helical CT scans

Directory of Open Access Journals (Sweden)

Erdi Yusuf E

2007-03-01

Full Text Available Abstract Background It is estimated that 60 million computed tomography (CT scans were performed during 2006, with approximately 11% of those performed on children age 0–15 years. Various types of gonadal shielding have been evaluated for reducing exposure to the gonads. The purpose of this study was to quantify the radiation dose reduction to the gonads and its effect on image quality when a wrap-around male pediatric gonad shield was used during CT scanning. This information is obtained to assist the attending radiologist in the decision to utilize such male gonadal shields in pediatric imaging practice. Methods The dose reduction to the gonads was measured for both direct radiation and for indirect scattered radiation from the abdomen. A 6 cm3 ion chamber (Model 10X5-6, Radcal Corporation, Monrovia, CA was placed on a Humanoid real bone pelvic phantom at a position of the male gonads. When exposure measurements with shielding were made, a 1 mm lead wrap-around gonadal shield was placed around the ion chamber sensitive volume. Results The use of the shields reduced scatter dose to the gonads by a factor of about 2 with no appreciable loss of image quality. The shields reduced the direct beam dose by a factor of about 35 at the expense of extremely poor CT image quality due to severe streak artifacts. Conclusion Images in the direct exposure case are not useful due to these severe artifacts and the difficulties in positioning these shields on patients in the scatter exposure case may not be warranted by the small absolute reduction in scatter dose unless it is expected that the patient will be subjected to numerous future CT scans.

Calculation of neutron and gamma-ray flux-to-dose-rate conversion factors

International Nuclear Information System (INIS)

Kwon, S.G.; Lee, S.Y.; Yook, C.C.

1981-01-01

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute (ANSI) N666. These data are used to calculate the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoenergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions. (author)

External dosimetry sources and shielding

International Nuclear Information System (INIS)

Calisto, Washington

1994-01-01

A definition of external dosimetry r external sources dosimetry,physical and mathematical treatment of the interaction of gamma radiation with a minimal area in that direction. Concept of attenuation coefficient, cumulated effect by polyenergetic sources, exposition rate, units, cumulated dose,shielding, foton shielding, depth calculation, materials used for shielding.Beta shielding, consideration of range and maximum β energy , low stopping radiation by use of low Z shielding. Tables for β energy of β emitters, I (tau) factor, energy-range curves for β emitters in aqueous media, gamma attenuation factors for U, W and Pb. Y factor for bone tissue,muscle and air, build-up factors

Comparison of predicted versus measured dose rates for low-level radioactive waste cask shipments

International Nuclear Information System (INIS)

Macher, Martin S.

1992-01-01

Shippers of low-level radioactive waste must select casks which will provide sufficient shielding to keep dose rates below the federal limit of 10 mr/hr at 2 meters from the vehicle. Chem-Nuclear Systems, Inc. uses a cask selection methodology which is based on shielding analysis code predictions with an additional factor of safety applied to compensate for inhomogeneities in the waste, uncertainties in waste characterization, and inaccuracy in the calculational methods. This proven cask selection methodology is explained and suggested factors of safety are presented based on comparisons of predicted and measured dose rates. A safety factor of 2 is shown to be generally appropriate for relatively homogeneous waste and a safety factor of between 3 and 4 is shown to be generally appropriate for relatively inhomogeneous wastes. (author)

Dose field research of analysis room for in-hospital neutron irradiator

International Nuclear Information System (INIS)

Zhang Zizhu; Song Mingzhe; Li Wei; Chen Jun; Yang Yong; Li Yiguo

2012-01-01

Neutron equivalent dose rate and y ray dose rate inside the analysis room of the in-hospital neutron irradiator (IHNI) and outdoor were measured. The results show that γ ray dose rate inside the analysis room exceeds calculation value many times and γ/ ray dose rate outdoor is higher than supervision region dose limit of 7.5 μSv/h. According to the measurement results and the Monte Carlo simulation, the following shielding plan was adopted. Lead shielding with thickness of 16 cm was installed on the wall, which faces the neutron beam, to shield γ ray, and lithium polyethylene plate with thickness of l cm was installed on all the wall (not including ceiling and floor) to shield scattering neutron. After shielding transformation, the highest γ ray dose rate point inside the analysis room decreased 277 times, the neutron equivalent dose rate decreased 5.8 times, and the outdoor γ/ray dose rate decreased nearly 90 times. (authors)

Shielding calculation for bremsstrahlung from β-emitters

International Nuclear Information System (INIS)

Ichimiya, Tsutomu

1990-01-01

Accompanying the revision of radiation injury prevention law, the shielding calculation method for photon corresponding to the dose equivalent was shown. However, regarding the electron from β decay nuclide and bremsstrahlung caused by shielding material, the shielding calculation method corresponding to the 1 cm dose equivalent has not been reported, hence, in this report, the spectrum of β-ray is calculated and the 1 cm dose equivalent transmission rate of the bremsstrahlung was calculated for three kinds of shielding materials (iron, lead, concrete). As the result of consideration, it is sufficient to think about the bremsstrahlung due to negative electron emission accompanying β-decay. In β-decay, electrons which constitute the continuous spectrum with maximum energy are emitted. The shape of the spectrum differs with nuclides. The maximum energy of β-ray of generally used nuclides is mostly below 3MeV and, besides, the electron ray itself is easily shielded, while the strength of bremsstrahlung depends on the atomic number of shielding materials and its generating mechanism is complicated. In this report, the actual shielding calculation method for bremsstrahlung is shown with regard to the most frequently used β-decay nuclides. (M.T.)

Comparison of Radiation Dose Rates with the Flux to Dose Conversion Factors Recommended in ICRP-74 and ICRP-116

International Nuclear Information System (INIS)

Jeong, Hae Sun; Kil, A Reum; Lee, Jo Eun; Jeong, Hyo Joon; Kim, Eun Han; Han, Moon Hee; Hwang, Won Tae

2016-01-01

The evaluation of radiation shielding has been performed for the design and maintenance of various facilities using radioactive sources such as nuclear fuel, accelerator, and radionuclide. The conversion of flux to dose mainly used in nuclear and radiation fields has been generally made with the dose coefficients presented in ICRP Publication 74 (ICRP- 74), which are produced based on ICRP Publication 60. On the other hand, ICRP Publication 116 (ICRP-116), which adopts the protection system of ICRP Publication 103, has recently been published and provides the dose conversion coefficients calculated with a variety of Monte Carlo codes. The coefficients have more than an update of those in ICRP-74, including new particle types and a greatly expanded energy range. In this study, a shielding evaluation of a specific container for neutron and gamma sources was performed with the MCNP6 code. The dose rates from neutron and gamma-ray sources were calculated using the MCNP6 codes, and these results were based on the flux to dose conversion factors recommended in ICRP-74 and ICRP-116. As a result, the dose rates evaluated with ICRP-74 were generally shown higher than those with ICRP-116. For neutrons, the difference is mainly occurred by the decrease of radiation weighting factors in a part of energy ranges in the ICRP-116 recommendations. For gamma-rays, the ICRP-74 recommendation applied with the kerma approximation leads to overestimated results than the other assessment

Benchmark experiments of dose distributions in phantom placed behind iron and concrete shields at the TIARA facility

International Nuclear Information System (INIS)

Nakane, Yoshihiro; Sakamoto, Yukio; Tsuda, Shuichi

2004-01-01

To verify the calculation methods used for the evaluations of neutron dose at the radiation shielding design of the high-intensity proton accelerator facility (J-PARC), dose distributions in a plastic phantom of 30x30x30 cm 3 slab placed behind iron and concrete test shields were measured by using a tissue equivalent proportional counter for 65-MeV quasi-monoenergetic neutrons generated from the 7 Li(p,n) reactions with 68-MeV protons at the TIARA facility. Dose distributions in the phantom were calculated by using the MCNPX and the NMTC/JAM-MCNP codes with the flux-to-dose conversion coefficients prepared for the shielding design of the facility. The comparison results show the calculated results were in good agreement with the measured ones within 20%. (author)

Scatter radiation breast exposure during head CT: impact of scanning conditions and anthropometric parameters on shielded and unshielded breast dose

Energy Technology Data Exchange (ETDEWEB)

Klasic, B. [Hospital for pulmonary diseases, Zagreb (Croatia); Knezevic, Z.; Vekic, B. [Rudjer Boskovic Institute, Zagreb (Croatia); Brnic, Z.; Novacic, K. [Merkur Univ. Hospital, Zagreb (Croatia)

2006-07-01

Constantly increasing clinical requests for CT scanning of the head on our facility continue to raise concern regarding radiation exposure of patients, especially radiosensitive tissues positioned close to the scanning plane. The aim of our prospective study was to estimate scatter radiation doses to the breast from routine head CT scans, both with and without use of lead shielding, and to establish influence of various technical and anthropometric factors on doses using statistical data analysis. In 85 patient referred to head CT for objective medical reasons, one breast was covered with lead apron during CT scanning. Radiation doses were measured at skin of both breasts and over the apron simultaneously, by the use of thermo luminescent dosimeters. The doses showed a mean reduction by 37% due to lead shielding. After we statistically analyzed our data, we observed significant correlation between under-the-shield dose and values of technical parameters. We used multiple linear regression model to describe the relationships of doses to unshielded and shielded breast respectively, with anthropometric and technical factors. Our study proved lead shielding of the breast to be effective, easy to use and leading to a significant reduction in scatter dose. (author)

Scatter radiation breast exposure during head CT: impact of scanning conditions and anthropometric parameters on shielded and unshielded breast dose

International Nuclear Information System (INIS)

Klasic, B.; Knezevic, Z.; Vekic, B.; Brnic, Z.; Novacic, K.

2006-01-01

Constantly increasing clinical requests for CT scanning of the head on our facility continue to raise concern regarding radiation exposure of patients, especially radiosensitive tissues positioned close to the scanning plane. The aim of our prospective study was to estimate scatter radiation doses to the breast from routine head CT scans, both with and without use of lead shielding, and to establish influence of various technical and anthropometric factors on doses using statistical data analysis. In 85 patient referred to head CT for objective medical reasons, one breast was covered with lead apron during CT scanning. Radiation doses were measured at skin of both breasts and over the apron simultaneously, by the use of thermo luminescent dosimeters. The doses showed a mean reduction by 37% due to lead shielding. After we statistically analyzed our data, we observed significant correlation between under-the-shield dose and values of technical parameters. We used multiple linear regression model to describe the relationships of doses to unshielded and shielded breast respectively, with anthropometric and technical factors. Our study proved lead shielding of the breast to be effective, easy to use and leading to a significant reduction in scatter dose. (author)

Scan Quality and Entrance Skin Dose in Thoracic CT: A Comparison between Bismuth Breast Shield and Posteriorly Centered Partial CT Scans

International Nuclear Information System (INIS)

Tappouni, Rafel; Mathers, Bradley

2012-01-01

Objectives. To compare the effectiveness of the bismuth breast shield and partial CT scan in reducing entrance skin dose and to evaluate the effect of the breast shield on image quality (IQ). Methods. Nanodots were placed on an adult anthropomorphic phantom. Standard chest CT, CT with shield, and partial CT were performed. Nanodot readings and effective doses were recorded. 50 patients with chest CTs obtained both with and without breast shields were reviewed. IQ was evaluated by two radiologists and by measuring Hounsfield units (HUs) and standard deviation (SD) of HU in anterior subcutaneous region. Results. Breast shield and the partial CT scans reduced radiation to the anterior chest by 38% and 16%, respectively. Partial CT increased dose to the posterior chest by 37% and effective dose by 8%. Change in IQ in shield CT was observed in the anterior chest wall. Significant change in IQ was observed in 5/50 cases. The shield caused an increase of 20 HU (P = 0.021) and a 1.86 reduction in SD of HU (P = 0.027) in the anterior compared to posterior subcutaneous regions. Summary. Bismuth breast shield is more effective than the partial CT in reducing entrance skin dose while maintaining image quality

Preliminary study for development of low dose radiation shielding material using liquid silicon and metallic compound

Energy Technology Data Exchange (ETDEWEB)

Jang, Seo Goo; Lee, Sung Soo [Dept. of Medical Science, Graduate School of Soonchunhyang University, Asan (Korea, Republic of); Han, Su Chul [Div. of Medical Radiation Equipment, Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Kang, Sung Jin [SoonChunHyang University Hospital, Seoul (Korea, Republic of); Lim, Sung Wook [Graduate school of SeJong University, Seoul (Korea, Republic of)

2017-09-15

This study measured and compared the protective clothing using Pb used for shielding in a diagnostic X-ray energy range, and the shielding rates of X-ray fusion shielding materials using Si and TiO{sub 2}. For the experiment, a pad type shielding with a thickness of 1 mm was prepared by mixing Si-TiO{sub 2}, and the X-ray shielding rate was compared with 0.5 mmPb plate of The shielding rate of shielding of 0.5 mmPb plate 95.92%, 85.26 % based on the case of no shielding under each 60kVp, 100kVp tube voltage condition. When the shielding of Si-TiO{sub 2} pad was applied, the shielding rate equal to or greater than 0.5 mmPb plate was obtained at a thickness of 11 mm or more, and the shielding rate of 100% or more was confirmed at a thickness of 13 nn in 60kVp condition. When the shielding of Si-TiO{sub 2} pad was applied, the shielding rate equal to or greater than 0.5 mmPb plate was obtained at a thickness of 17 mm or more, and a shielding rate of 0.5 mmPb plate was observed at a thickness of 23 mm in 100kVp condition. Through the results of this study, We could confirm the possibility of manufacturing radiation protective materials that does not contain lead hazard using various metallic compound and liquid Si. This study shows that possibility of liquid Si and other metallic compound can harmonize easily. Beside, It is flexible and strong to physical stress than Pb obtained radiation protective clothes. But additional studies are needed to increase the shielding rate and reduce the weight.

[Trial manufacture of a plunger shield for a disposable plastic syringe].

Science.gov (United States)

Murakami, Shigeki; Emoto, Takashi; Mori, Hiroshige; Fujita, Katsuhisa; Kubo, Naoki

2008-08-20

A syringe-type radiopharmaceutical being supplied by a manufacturer has a syringe shield and a plunger shield, whereas an in-hospital labeling radiopharmaceutical is administered by a disposable plastic syringe without the plunger shield. In cooperation with Nihon Medi-Physics Co. Ltd., we have produced a new experimental plunger shield for the disposable plastic syringe. In order to evaluate this shielding effect, we compared the leaked radiation doses of our plunger shield with those of the syringe-type radiopharmaceutical (Medi shield type). Our plunger shield has a lead plate of 21 mm in diameter and 3 mm thick. This shield is equipped with the plunger-end of a disposal plastic syringe. We sealed 99mTc solution into a plastic syringe (Terumo Co.) of 5 ml with our plunger shield and Medi shield type of 2 ml. We measured leaked radiation doses around syringes using fluorescent glass dosimeters (Dose Ace). The number of measure points was 18. The measured doses were converted to 70 microm dose equivalent at 740 MBq of radioactivity. The results of our plunger shield and the Medi shield type were as follows: 4-13 microSv/h and 3-14 microSv/h at shielding areas, 3-545 microSv/h and 6-97 microSv/h at non-shielding areas, 42-116 microSv/h and 88-165 microSv/h in the vicinity of the syringe shield, and 1071 microSv/h and 1243 microSv/h at the front of the needle. For dose rates of shielding areas around the syringe, the shielding effects were approximately the same as those of the Medi shield type. In conclusion, our plunger shield may be useful for reducing finger exposure during the injection of an in-hospital labeled radiopharmaceutical.

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

2016-11-15

Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

Decontamination evaluation based on radioactivity measurement instead of air dose rate

International Nuclear Information System (INIS)

Shozugawa, Katsumi

2013-01-01

Air dose rate at 1 m above the ground comes from gamma radiations emitted from vast area ranging over several ten meters of the contaminated field from the counter. After showing the actual example of the difference between air dose rate data and Cs 137 distribution map made by using a shielded NaI-scintillation counter within and around a contaminated sinkhole (a ditch or trench) near Fukushima Daiichi Nuclear Power Plants, the author proposes to make a decontamination program according to the radioactivity distribution measurement instead of air dose rate measurement. Furthermore, he explains some problems arising from a point and plane radiation source, and also difficulties accompanied by movement of Cs 137 atoms in the soils according to the absorption characteristics of the existing minerals but these are also important to consider for performing an effective decontamination. (S. Ohno)

Shielding of manned space stations against Van Allen Belt protons: a preliminary scoping study

International Nuclear Information System (INIS)

Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Corbin, J.M.

1986-09-01

Calculated results are presented to aid in the design of the shielding required to protect astronauts in a space station that is orbiting through the Van Allen proton belt. The geometry considered - a spherical shell shield with a spherical tissue phantom at its center - is only a very approximate representation of an actual space station, but this simple geometry makes it possible to consider a wide range of possible shield materials. Both homogeneous and laminated shields are considered. Also, an approximation procedure - the equivalent thickness approximation - that allows dose rates to be estimated for any shield material or materials from the dose rates for an aluminum shield is presented and discussed

Status of the ITER tokamak nuclear shielding and radiological protection design

Energy Technology Data Exchange (ETDEWEB)

Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Chaffard, P.Y.; Izquierdo, J. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Juarez, R. [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Pampin, R.; Portone, A. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain)

2016-11-01

Highlights: • Comprehensive review of design status of the ITER tokamak regarding nuclear shielding. • Investigation of shield design options and streaming mitigation measures. • Review of state-of-the-art shutdown dose rate analyses for selected port systems. - Abstract: Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

Monte Carlo validation and optimisation of detector packaging for spectroscopic dosimetry for in vivo urethral dosimetry during low dose rate brachytherapy

International Nuclear Information System (INIS)

Nourbehesht, L.K.; Cutajar, D.L.; Guatelli, S.; Rosenfeld, A.B.

2015-01-01

The urethral mini-dosimeter, developed by the Centre for Medical Radiation Physics, University of Wollongong, uses spectroscopic dosimetry to provide real time point dose measurements along the urethra during low dose rate prostate brachytherapy. Spectroscopic dosimetry uses the measured spectrum of the treatment isotope to estimate the dose rate at the point of measurement, however, the silicon mini-detectors employed in the urethral mini-dosimeter require water proof encapsulation which must be capable of providing electromagnetic shielding without greatly increasing the size of the probe. The introduction of non-tissue equivalent materials within the encapsulation can change the spectrum of radiation incident on the detector, which may influence the application of spectroscopic dosimetry within the urethral dosimeter. The Monte Carlo code Geant4 was adopted to study the effect of encapsulation on the operation of the urethral mini-dosimeter, as well as to determine whether an appropriate thickness of aluminium shielding was possible for electromagnetic screening. The depth dose response and angular dependence of the urethral mini-dosimeter with three thicknesses of aluminium shielding (20, 50, 100 µm) was compared with the urethral mini-dosimeter without aluminium shielding. The aluminium shielding had the effect of increasing the depth dose response (up to 3 % within 30 mm and up to 5 % within 50 mm), slightly reduced the azimuth angular dependence and slightly increased the polar angular dependence. The 100 µm thick shielding provided the least azimuth angular dependence (±2 %) and provided a polar angular dependence of ±1.4 % within the angles of −45° to 45°.

Field measurement and interpretation of beta doses and dose rates

International Nuclear Information System (INIS)

Selby, J.M.; Swinth, K.L.; Hooker, C.D.; Kenoyer, J.L.

1983-01-01

A large number of portable survey instruments employing G.M., ionization chamber, and scintillation detectors used for gamma measurements are also used for monitoring in beta fields by using removable shields to separate the beta and gamma components of the radiation field. The difference does not correspond to an absorbed dose rate for the beta field due to a variety of factors. Among these factors are the dependence on beta energy, source-detector geometries, mixed fields and variable ambient conditions. Attempting to use such measurements directly can lead to errors as high as a factor of 100. Appropriate calibrations and correction factors can be used to reduce the errors in beta measurements to a tolerable level

Calculation of the neutrons shielding in cyclotron accelerator

International Nuclear Information System (INIS)

Ribeiro, Martha S.; Sanches, Matias P.; Rodrigues, Demerval L.

2000-01-01

The objective of radioprotection in cyclotron facilities is to reduce the dose levels in the workplaces to classify them like supervised areas. In this way, the radiation dose rates in areas occupied by workers during cyclotron operations should not exceed 7,5 μSv/h. In controlled areas these levels are not observed and some rigorous controls must be exerted by administrative procedures or protection mechanisms. The Cyclotron Laboratory at IPEN-CNEN/SP has a cyclotron model Cyclone 30, 30 MeV, used for research and it is also used for radioisotopes production for medical diagnosis and therapeutical applications. Among them, 123 I, 67 Ga and 18 F can be pointed. When accelerator is operating, failures in perforations and paths that conduce to room accelerator can be occur and thus, the dose levels are higher than that established by law. For this reason, a review for shielding structure was necessary in order to optimize radiation dose. The purpose of this work was to determine the shielding thickness and adequate material to diminish the dose rates in workplaces to a value below 7,5 μSv/h. It was used a method to employ the equivalent dose value in the facility areas for neutrons fluency rate for the principal reactions in target irradiation processes. The purposed shielding for the vault doors ensures dose levels lower than established limits to supervised areas. (author)

Parameters calculation of a shielding experiment and evaluation of calculation methodology

International Nuclear Information System (INIS)

Gavazza, S.; Otto, A.C.; Gomes, I.C.; Maiorino, J.R.

1986-01-01

In this text is carried out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gamma-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The transport calculation were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reactions and dose rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented coherence with the experimental measurements. (Author) [pt

Shielding calculations and collective dose estimations with the point-kernel-code VISIPLAN registered for the example of the project ZENT

International Nuclear Information System (INIS)

Boehlke, S.; Niegoth, H.

2012-01-01

In the nuclear power plant Leibstadt (KKL) during the next year large components will be dismantled and stored for final disposal within the interim storage facility ZENT at the NPP site. Before construction of ZENT appropriate estimations of the local dose rate inside and outside the building and the collective dose for the normal operation have to be performed. The shielding calculations are based on the properties of the stored components and radiation sources and on the concepts for working place requirements. The installation of control and monitoring areas will depend on these calculations. For the determination of the shielding potential of concrete walls and steel doors with the defined boundary conditions point-kernel codes like MICROSHIELd registered are used. Complex problems cannot be modeled with this code. Therefore the point-kernel code VISIPLAN registered was developed for the determination of the local dose distribution functions in 3D models. The possibility of motion sequence inputs allows an optimization of collective dose estimations for the operational phases of a nuclear facility.

Temperature and neutron dose rate measurements at a spent fuel shipping cask

International Nuclear Information System (INIS)

Krause, F.

1982-01-01

Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

Evaluation of bismuth shielding effectiveness in reducing breast absorbed dose during thoracic CT scan

Energy Technology Data Exchange (ETDEWEB)

Alonso, T. C.; Mourao, A. P.; Santana, P. C.; Silva, T. A. [Federal University of Minas Gerais, Program of Nuclear Science and Techniques, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil)

2015-10-15

Computed Tomography (CT) is an essential method for tracking neoplasia and efficiently diagnosing a wide variety of thoracic diseases. CT is generally considered the most accurate choice for lung examination. Due to the growing use of CT, breast and other superficial and radiosensitive organs are unnecessarily irradiated during radiological procedures, thus requiring the development of strategies appropriate to optimize and, if possible, to reduce the radiation dose. The use of bismuth shielding to reduce radiation dose absorbed by breast during thoracic CT examinations has been the subject of many studies recently published by Brazilian and foreign authors of various fields. The purpose of this paper is both to accurately determine the glandular dose when breast is exposed to radiation and to assess the reduction in absorbed dose during thoracic CT examinations, using a set of Thermoluminescent Dosimeters, an anthropomorphic phantom and bismuth shielding. (Author)

Evaluation of bismuth shielding effectiveness in reducing breast absorbed dose during thoracic CT scan

International Nuclear Information System (INIS)

Alonso, T. C.; Mourao, A. P.; Santana, P. C.; Silva, T. A.

2015-10-01

Computed Tomography (CT) is an essential method for tracking neoplasia and efficiently diagnosing a wide variety of thoracic diseases. CT is generally considered the most accurate choice for lung examination. Due to the growing use of CT, breast and other superficial and radiosensitive organs are unnecessarily irradiated during radiological procedures, thus requiring the development of strategies appropriate to optimize and, if possible, to reduce the radiation dose. The use of bismuth shielding to reduce radiation dose absorbed by breast during thoracic CT examinations has been the subject of many studies recently published by Brazilian and foreign authors of various fields. The purpose of this paper is both to accurately determine the glandular dose when breast is exposed to radiation and to assess the reduction in absorbed dose during thoracic CT examinations, using a set of Thermoluminescent Dosimeters, an anthropomorphic phantom and bismuth shielding. (Author)

Shielding technology for high energy radiation production facility

International Nuclear Information System (INIS)

Lee, Byung Chul; Kim, Heon Il

2004-06-01

In order to develop shielding technology for high energy radiation production facility, references and data for high energy neutron shielding are searched and collected, and calculations to obtain the characteristics of neutron shield materials are performed. For the evaluation of characteristics of neutron shield material, it is chosen not only general shield materials such as concrete, polyethylene, etc., but also KAERI developed neutron shields of High Density PolyEthylene (HDPE) mixed with boron compound (B 2 O 3 , H 2 BO 3 , Borax). Neutron attenuation coefficients for these materials are obtained for later use in shielding design. The effect of source shape and source angular distribution on the shielding characteristics for several shield materials is examined. This effect can contribute to create shielding concept in case of no detail source information. It is also evaluated the effect of the arrangement of shield materials using current shield materials. With these results, conceptual shielding design for PET cyclotron is performed. The shielding composite using HDPE and concrete is selected to meet the target dose rate outside the composite, and the dose evaluation is performed by configuring the facility room conceptually. From the result, the proper shield configuration for this PET cyclotron is proposed

Radiation shield analysis for a manned Mars rover

International Nuclear Information System (INIS)

Morley, N.J.; ElGenk, M.S.

1991-01-01

Radiation shielding for unmanned space missions has been extensively studied; however, designs of man-rated shields are minimal. Engle et al.'s analysis of a man-rated, multilayered shield composed of two and three cycles (a cycle consists of a tungsten and a lithium hydride layer) is the basis for the work reported in this paper. The authors present the results of a recent study of shield designs for a manned Mars rover powered by a 500-kW(thermal) nuclear reactor. A train-type rover vehicle was developed, which consists of four cars and is powered by an SP-100-type nuclear reactor heat source. The maximum permissible dose rate (MPD) from all sources is given by the National Council on Radiation Protection and Measurements as 500 mSv/yr (50 rem/yr) A 3-yr Mars mission (2-yr round trip and 1-yr stay) will deliver a 1-Sv natural radiation dose without a solar particle event, 450 mSv/yr in flight, and an additional 100 mSv on the planet surface. An anomalously large solar particle event could increase the natural radiation dose for unshielded astronauts on the Martian surface to 200 mSv. This limits the MPD to crew members from the nuclear reactor to 300 mSv

Dose reduction using bismuth shielding during paediatric CT examinations in Slovakia

International Nuclear Information System (INIS)

Gbelcova, L.; Nikodemova, D.; Horvathova, M.

2011-01-01

Considering the massive increase of computer tomography (CT) examinations in Slovakia during the last 10 y, it can be expected that a higher radiation load may be observed in the Slovak population. Since child population is more sensitive to radiation than adult population, a monitoring has started to see how high the radiation dose is for paediatric patients during CT examinations in chosen departments in Slovakia. The CT examination of the head is one of the most frequently done examinations in Slovakian departments and that is why measurements were done to clarify how usage of bismuth shields for eyes and thyroid can affect the eye and thyroid doses. For simulation, 215 thermoluminescent dosimeters were exposed on anthropomorphic phantom of a child with and without usage of bismuth shields. The result was that only two of the three chosen departments confirmed a reduction. On the other hand, one of the departments confirmed that the reduction can be up to 56-65 %, which is significant. (authors)

An innovative method for on-power radiometry of end-shields of nuclear power plants

International Nuclear Information System (INIS)

Kumar, Gaurav; Gupta, Pankaj; Nawal, Shriram; Gautam, Mahesh; Kakkar, Aman Deep; Yadav, Umed

2012-01-01

Every lndian PHWR reactor calandria is sandwiched within a pair of shield on either side. These shields are perpendicular to the coaxial axis of calandria and are called end-shields. These provide shielding from leakage radiation from reactor core in escaping out to Fuelling Machine vault, thereby significantly reducing the dose rates in the vaults. This has got a direct impact on radiation field in accessible areas. By maintaining low dose rates in accessible areas, the individual and collective doses of radiation workers can be effectively controlled well within the stipulated limits. Thus, it is of utmost importance to ensure adequacy of shielding provided by end-shields. In this context, a limited radiometry exercise is executed after filling of end-shields with steel balls and prior to their installation at designated place. This exercise provides limited inputs along the periphery of end-shield due to limited strength of radiation source, its handling provisions and dose constraints to the individual. In order to ascertain an in-depth analysis of shielding adequacy on-power, different methodologies have been adopted and have certain limitation in precisely pinpointing the affected area/location besides limitation on number of locations that can be monitored at a single stretch. To overcome these important anomalies, a computer based setup has been indigenously designed. The setup essentially comprises of a radiation monitor with wide energy, measuring, temperature and humidity range; a custom designed 25 m long compatible cable with suitable connectors; a laptop with additional cooling arrangement; a configurable interfacing software; thermal shielding for the detector and tying/fixing provisions. The radiation monitor after being properly shielded for thermal impacts is installed on the head of Fuelling Machine. It is connected through long cable to a laptop kept at Fuelling Machine service area with due cooling provisions (as temperature in the area will

Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

International Nuclear Information System (INIS)

Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

2013-01-01

An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18 O(p, n) 18 F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18 F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18 F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H 2 18 O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection. - Highlights: • Neutron doses were verified using TLD approach. • Neutron doses were increased at cyclotron centers. • Revised L-shaped shield suppresses effectively the neutrons. • Neutron dose can be attenuated to 1.13×10 6 %

Shielding NSLS-II light source: Importance of geometry for calculating radiation levels from beam losses [Shielding Synchrotron Light Sources: Importance of geometry for calculating radiation levels from beam losses

International Nuclear Information System (INIS)

Kramer, S. L.; Ghosh, V. J.; Breitfeller, M.; Wahl, W.

2016-01-01

Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produces significantly higher neutron component dose to the experimental floor than a lower energy beam injection and ramped operations. Minimizing this dose will require adequate knowledge of where the miss-steered beam can occur and sufficient EM shielding close to the loss point, in order to attenuate the energy of the particles in the EM shower below the neutron production threshold (<10 MeV), which will spread the incident energy on the bulk shield walls and thereby the dose penetrating the shield walls. Designing supplemental shielding near the loss point using the analytic shielding model is shown to be inadequate because of its lack of geometry specification for the EM shower process. To predict the dose rates outside the tunnel requires detailed description of the geometry and materials that the beam losses will encounter inside the tunnel. Modern radiation shielding Monte-Carlo codes, like FLUKA, can handle this geometric description of the radiation transport process in sufficient detail, allowing accurate predictions of the dose rates expected and the ability to show weaknesses in the design before a high radiation incident occurs. The effort required to adequately define the accelerator geometry for these codes has been greatly reduced with the implementation of the graphical interface of FLAIR to FLUKA. This made the effective shielding process for NSLS-II quite accurate and reliable. Lastly, the principles used to provide

Influence of lead apron shielding on absorbed doses from panoramic radiography.

Science.gov (United States)

Rottke, D; Grossekettler, L; Sawada, K; Poxleitner, P; Schulze, D

2013-01-01

This study investigated the absorbed doses in a full anthropomorphic body phantom from two different panoramic radiography devices, performing protocols with and without applying a lead apron. A RANDO(®) full body phantom (Alderson Research Laboratories Inc., Stamford, CT) was equipped with 110 thermoluminescent dosemeters at 55 different sites and set up in two different panoramic radiography devices [SCANORA(®) three-dimensional (3D) (SOREDEX, Tuusula, Finland) and ProMax(®) 3D (Planmeca, Helsinki, Finland)] and exposed. Two different protocols were performed in the two devices. The first protocol was performed without any lead shielding, whereas the phantom was equipped with a standard adult lead apron for the second protocol. A two-tailed paired samples t-test for the SCANORA 3D revealed that there is no difference between the protocol using lead apron shielding (m = 87.99, s = 102.98) and the protocol without shielding (m = 87.34, s = 107.49), t(54) = -0.313, p > 0.05. The same test for the ProMax 3D showed that there is also no difference between the protocol using shielding (m = 106.48, s = 117.38) and the protocol without shielding (m = 107.75, s = 114,36), t(54) = 0.938, p > 0.05. In conclusion, the results of this study showed no statistically significant differences between a panoramic radiography with or without the use of lead apron shielding.

A gonadal shield with appropriate wall thicknesses for Co-60-teletherapy

International Nuclear Information System (INIS)

Kahlhoefer, J.

1982-01-01

A gonadal shield for men has been designed especially for use in Co-60-teletherapy of ventral fields in supine position. It has been made by simple means in the clinical workshop. The thickness of the top, sides, and botton are equal to about 3 half value layers for scattered radiation incident in the respective directions. Radiation dose rate to the testicles during irradiation of a large abdominal field ('inverted Y') was measured in a phantom and was found to be 7.1% of tumour dose rate without shield and 0.9% with gonadal shield. (orig.) [de

The evaluation of radiation dose to embryo/fetus and the design of shielding in the treatment of brain tumors

Energy Technology Data Exchange (ETDEWEB)

Cho, Woong; Huh, Soon Nyung; Chie, Eui Kyu; Ha, Sung Whan; Park, Yang Gyun; Park, Jong Min [Seoul National Univ., Seoul (Korea, Republic of); Park, Suk Won [Chungang Univ., Seoul (Korea, Republic of)

2006-12-15

Purpose : To estimate the dose to the embryo/fetus of a pregnant patient with brain tumors, and to design an shielding device to keep the embryo/fetus dose under acceptable levels. Materials and Methods : A shielding wall with the dimension of 1.55 m height, 0.9 m width, and 30 mm thickness is fabricated with 4 trolleys under the wall. It is placed between a patient and the treatment head of a linear accelerator to attenuate the leakage radiation effectively from the treatment head, and is placed 1 cm below the lower margin of the treatment field in order to minimize the dose to a patient from the treatment head. An anti-patient scattering neck supporters with 2 cm thick Cerrobend metal is designed to minimize the scattered radiation from the treatment fields and it is divided into 2 section. They are installed around the patient neck by attach from right and left sides. A shielding bridge for anti-room scattered radiation is utilized to place 2 sheets of 3 mm lead plates above the abdomen to setup three detectors under the lead sheets. Humanoid phantom is irradiated with the same treatment parameters, and with and without shielding devices using TLD, and ionization chambers with and without a build-up cap. Results : The dose to the embryo/fetus without shielding was 3.20, 3.21, 1.44, 0.90 cGy at off-field distances of 30, 40, 50, and 60 cm. With shielding, the dose to embryo/fetus was reduced to 0.88, 0.60, 0.35, 0.25, cGy, and the ratio of the shielding effect varied from 70% to 80%. TLD results were 1.8, 1.2, 0.8, 1.2, and 0.8 cGy. The dose measured by the survey meter was 10.9 mR/h at the patient's surface of abdomen. The dose to the embryo/fetus was estimated to be about 1 cGy during the entire treatment.

Study of shielding options for lower ports for mitigation of neutron environment and shutdown dose inside the ITER cryostat

International Nuclear Information System (INIS)

Pampin, Raul; Suarez, Alejandro; Arnould, Anne; Casal, Natalia; Juarez, Rafael; Martin, Alex; Moro, Fabio; Mota, Fernando; Polunovskiy, Eduard; Sabourin, Flavien

2016-01-01

Highlights: • Mitigation of the radiation environment inside the cryostat needed to reduce ITER coil heating and occupational exposure. • Cryopump and diagnostics lower ports are significant contributors, shielding options for both are explored. • Shielding performance studied in terms of neutron transmission and nuclear heating to coils for a range of options. • Benefits/constraints discussed together with other engineering parameters. - Abstract: Mitigation of the neutron environment inside the cryostat, and of the subsequent decay gamma dose field from activated materials, is necessary in order to reduce heating of coils and occupational exposure, thereby facilitating smooth operation and maintenance of ITER. Several lines of action are currently being explored to mitigate crucial contributions, such as the leakage through the lower ports. Results are presented here for the two types of lower ports in ITER: cryopump ports and remote-handling ports. Different shielding configurations and material options are investigated and compared in terms of neutron attenuation, coil heating and shutdown dose rate reduction, whilst also considering other engineering constraints such as weight or pumping power. Results enable informed decision-making of best compromise solutions for subsequent design and integration.

Study of shielding options for lower ports for mitigation of neutron environment and shutdown dose inside the ITER cryostat

Energy Technology Data Exchange (ETDEWEB)

Pampin, Raul, E-mail: raul.pampin@f4e.europa.eu [Fusion For Energy, Josep Pla 2, Barcelona 08019 (Spain); Suarez, Alejandro; Arnould, Anne; Casal, Natalia [ITER Organization, Route de Vinon sur Verdon, 13067 Saint Paul lez Durance Cedex (France); Juarez, Rafael [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Martin, Alex [ITER Organization, Route de Vinon sur Verdon, 13067 Saint Paul lez Durance Cedex (France); Moro, Fabio [ENEA, Via Enrico Fermi, Frascati, Rome (Italy); Mota, Fernando [CIEMAT, Avenida Complutense 40, Madrid 28040 (Spain); Polunovskiy, Eduard; Sabourin, Flavien [ITER Organization, Route de Vinon sur Verdon, 13067 Saint Paul lez Durance Cedex (France)

2016-11-01

Highlights: • Mitigation of the radiation environment inside the cryostat needed to reduce ITER coil heating and occupational exposure. • Cryopump and diagnostics lower ports are significant contributors, shielding options for both are explored. • Shielding performance studied in terms of neutron transmission and nuclear heating to coils for a range of options. • Benefits/constraints discussed together with other engineering parameters. - Abstract: Mitigation of the neutron environment inside the cryostat, and of the subsequent decay gamma dose field from activated materials, is necessary in order to reduce heating of coils and occupational exposure, thereby facilitating smooth operation and maintenance of ITER. Several lines of action are currently being explored to mitigate crucial contributions, such as the leakage through the lower ports. Results are presented here for the two types of lower ports in ITER: cryopump ports and remote-handling ports. Different shielding configurations and material options are investigated and compared in terms of neutron attenuation, coil heating and shutdown dose rate reduction, whilst also considering other engineering constraints such as weight or pumping power. Results enable informed decision-making of best compromise solutions for subsequent design and integration.

Comparative investigation of three dose rate meters for their viability in pulsed radiation fields

International Nuclear Information System (INIS)

Gotz, M; Karsch, L; Pawelke, J

2015-01-01

Pulsed radiation fields, characterized by microsecond pulse duration and correspondingly high pulse dose rates, are increasingly used in therapeutic, diagnostic and research applications. Yet, dose rate meters which are used to monitor radiation protection areas or to inspect radiation shielding are mostly designed, characterized and tested for continuous fields and show severe deficiencies in highly pulsed fields. Despite general awareness of the problem, knowledge of the specific limitations of individual instruments is very limited, complicating reliable measurements. We present here the results of testing three commercial dose rate meters, the RamION ionization chamber, the LB 1236-H proportional counter and the 6150AD-b scintillation counter, for their response in pulsed radiation fields of varied pulse dose and duration. Of these three the RamION proved reliable, operating in a pulsed radiation field within its specifications, while the other two instruments were only able to measure very limited pulse doses and pulse dose rates reliably. (paper)

Shielding calculational system for plutonium

International Nuclear Information System (INIS)

Zimmerman, M.G.; Thomsen, D.H.

1975-08-01

A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

Development of dose rate estimation system for FBR maintenance

Energy Technology Data Exchange (ETDEWEB)

Iizawa, Katsuyuki [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Takeuchi, Jun; Yoshikawa, Satoru [Hitachi Engineering Company, Ltd., Hitachi, Ibaraki (Japan); Urushihara, Hiroshi [Ibaraki Hitachi Information Service Co., Ltd., Omika, Ibaraki (Japan)

2001-09-01

During maintenance activities on the primary sodium cooling system by an FBR Personnel radiation exposure arises mainly from the presence of radioactive corrosion products (CP). A CP behavior analysis code, PSYCHE, and a radiation shielding calculation code, QAD-CG, have been developed and applied to investigate the possible reduction of radiation exposure of workers. In order to make these evaluation methods more accessible to plant engineers, the user interface of the codes has been improved and an integrated system, including visualization of the calculated gamma-ray radiation dose-rate map, has been developed. The system has been verified by evaluating the distribution of the radiation dose-rate within the Monju primary heat transport system cells from the estimated saturated CP deposition and distribution which would be present following about 20 cycles of full power operation. (author)

Development of dose rate estimation system for FBR maintenance

International Nuclear Information System (INIS)

Iizawa, Katsuyuki; Takeuchi, Jun; Yoshikawa, Satoru; Urushihara, Hiroshi

2001-01-01

During maintenance activities on the primary sodium cooling system by an FBR Personnel radiation exposure arises mainly from the presence of radioactive corrosion products (CP). A CP behavior analysis code, PSYCHE, and a radiation shielding calculation code, QAD-CG, have been developed and applied to investigate the possible reduction of radiation exposure of workers. In order to make these evaluation methods more accessible to plant engineers, the user interface of the codes has been improved and an integrated system, including visualization of the calculated gamma-ray radiation dose-rate map, has been developed. The system has been verified by evaluating the distribution of the radiation dose-rate within the Monju primary heat transport system cells from the estimated saturated CP deposition and distribution which would be present following about 20 cycles of full power operation. (author)

Shielding Studies for Reducing the associated Radiological Risks Due To Irradiated Low Enriched Uranium Foil

International Nuclear Information System (INIS)

Margeanu, C.A.

2011-01-01

Present work estimates the radiation dose rates corresponding to irradiated Low Enriched Uranium (20 wt % 235 U) foil as part of shielding studies for radiological risks reduction after irradiation inside TRIGA 14 MW Research Reactor in an investigation on 99 Mo production possibility. Post-Irradiation Examination Laboratory's cell shielding calculations have been performed; radiation source was obtained by using ORIGEN-S code with specific cross-sections libraries. Different post-irradiation cooling times have been considered, gamma dose rates being estimated by using MAVRIC module from Scale 6 programs package, for following exposure situations (relative to Pie cell): i) front side, ii) lateral side and iii) back side. Three different calculations were performed: a) without any protection shield between operator and cell, except for the cell stainless steel wall; b) with a Lead protection shield between operator and cell and c) with a depleted Uranium shield, located inside the cell in between the radiation source and cell window. Radiation dose rates to cell external wall surface and for other eight fixed distances from cell wall were estimated. To obtain a consistent set of solutions, the study was done for various Uranium foil weights and different Lead and depleted Uranium shields thicknesses. Calculations were focused to assure that the dose rate to an operator positioned at 60 cm working distance from the cell will not exceed 0.02 mSv/h, maximum allowed dose rate for professionally exposed personnel according to Romanian regulations.

Absorbed dose kernel and self-shielding calculations for a novel radiopaque glass microsphere for transarterial radioembolization.

Science.gov (United States)

Church, Cody; Mawko, George; Archambault, John Paul; Lewandowski, Robert; Liu, David; Kehoe, Sharon; Boyd, Daniel; Abraham, Robert; Syme, Alasdair

2018-02-01

Radiopaque microspheres may provide intraprocedural and postprocedural feedback during transarterial radioembolization (TARE). Furthermore, the potential to use higher resolution x-ray imaging techniques as opposed to nuclear medicine imaging suggests that significant improvements in the accuracy and precision of radiation dosimetry calculations could be realized for this type of therapy. This study investigates the absorbed dose kernel for novel radiopaque microspheres including contributions of both short and long-lived contaminant radionuclides while concurrently quantifying the self-shielding of the glass network. Monte Carlo simulations using EGSnrc were performed to determine the dose kernels for all monoenergetic electron emissions and all beta spectra for radionuclides reported in a neutron activation study of the microspheres. Simulations were benchmarked against an accepted 90 Y dose point kernel. Self-shielding was quantified for the microspheres by simulating an isotropically emitting, uniformly distributed source, in glass and in water. The ratio of the absorbed doses was scored as a function of distance from a microsphere. The absorbed dose kernel for the microspheres was calculated for (a) two bead formulations following (b) two different durations of neutron activation, at (c) various time points following activation. Self-shielding varies with time postremoval from the reactor. At early time points, it is less pronounced due to the higher energies of the emissions. It is on the order of 0.4-2.8% at a radial distance of 5.43 mm with increased size from 10 to 50 μm in diameter during the time that the microspheres would be administered to a patient. At long time points, self-shielding is more pronounced and can reach values in excess of 20% near the end of the range of the emissions. Absorbed dose kernels for 90 Y, 90m Y, 85m Sr, 85 Sr, 87m Sr, 89 Sr, 70 Ga, 72 Ga, and 31 Si are presented and used to determine an overall kernel for the

Radiation dose reduction by water shield

International Nuclear Information System (INIS)

Zeb, J.; Arshed, W.; Ahmad, S.S.

2007-06-01

This report is an operational manual of shielding software W-Shielder, developed at Health Physics Division (HPD), Pakistan Institute of Nuclear Science and Technology (PINSTECH), Pakistan Atomic Energy Commission. The software estimates shielding thickness for photons having their energy in the range 0.5 to 10 MeV. To compute the shield thickness, self absorption in the source has been neglected and the source has been assumed as a point source. Water is used as a shielding material in this software. The software is helpful in estimating the water thickness for safe handling, storage of gamma emitting radionuclide. (author)

Analysis of lenses absorbed dose in head CT scan with the use of bismuth shielding

Energy Technology Data Exchange (ETDEWEB)

Santos, F.S.; Santana, P.C., E-mail: fernanda.stephaniebh@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil); Mourão, A.P. [Centro de Educação Tecnológica de Minas Gerais (CEFET-MG), Belo Horizonte, MG (Brazil)

2017-07-01

Computed Tomography (CT) has become an important tool to diagnose cancer and to obtain additional information for different clinical questions. However, CT scan usually requires a higher radiation exposure than a conventional radiography examination. Head CT scans are used for diagnosis of traumatic head injuries, infections and other diseases with instability. Based on this information, it was studied the dose variation deposited in the lenses and in nearby organs, such as: pharynx, hypophysis and salivary gland with and without the use of bismuth shield. In this study a head CT scan was performed on anthropomorphic male phantom using a GE scanner. Dose measurements have been performed by using radiochromic film strips to register the individual doses in the organs of interest. The results show that the lenses had a reduction of 26% of the dose with the use of the bismuth shield. (author)

Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

International Nuclear Information System (INIS)

Sakamoto, Yukio

2001-01-01

The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

International Nuclear Information System (INIS)

Matijević, Mario; Pevec, Dubravko; Trontl, Krešimir

2015-01-01

Highlights: • Shielding analysis of the typical PWR primary loop components was performed. • FW-CADIS methodology was thoroughly investigated using SCALE6.0 code package. • Versatile ability of SCALE6.0/FW-CADIS for deep penetration models was proved. • The adjoint source with focus on specific material can improve MC modeling. - Abstract: The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the S N part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high S N /P N parameters. Detailed voxelization (homogenization) process in Denovo together with high S N /P N parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR

Analysis of neutron dose rates on RGTT200K core using MCNP5

International Nuclear Information System (INIS)

Suwoto; Zuhair

2016-01-01

The conceptual design of RGTT200K (High Temperature Gas-cooled Reactor of 200 MWth Cogeneration) is the non-annular cylindrical reactor core with TRISO kernel coated fuel particles in the form of balls called pebble and cooled by helium gas. The RGTT200K reactor core design adopts high temperature gas cooled reactor (HTGR) technology with inherent passive safety. The RGTT200K spherical fuel called pebble fuel containing thousand of TRISO-coated fuel particles of uranium oxide (UO 2 ) 10 % enriched. TRISO coating comprises four layers, namely: porous carbon buffer layer, inner pyrolytic carbon layer (IPyC, Inner Pyrolytic Carbon), silicon carbide layer (SiC) and a layer of pyrolytic carbon outer portion (OPyC, Outer Pyrolytic Carbon). Modeling and analysis of preliminary calculation of neutron dose rate on normal operating temperature (T kernel =1200K) and accident temperature (T kernel =1800K) of the RGTT200K core were performed using Monte Carlo MCNP5v1.2 code. The continuous energy nuclear data cross-sections was taken from ENDF/B-VII, JENDL-4 and JEFF-3.1 nuclear data files . Double heterogeneity model in TRISO-coated fuel particles kernel and the pebble of RGTT200K core. By utilizing EGS99304 code, the 640 amount of energy group structures (SAND-II neutron group structures) is used in the neutron fluxes and spectrum calculation in RGTT200K reactor. The RGTT200K reactor core is divided into 25 zones (5 zones in radial and 10 zones in axial directions), while the modeling of radiation and biological shielding reactor RGTT200K are used to determine of preliminary neutron dose rate emitted by the neutron source with tally cards are available in the MCNP5v1.2 code. The calculation result analyses of the neutron dose rate distributions are determined using a conversion factor of flux-to-dose taken from International Commission on Radiological Protection, ICRP. The preliminary calculations result show that the neutrons dose rate using ICRP-74 conversion factor for

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

Directory of Open Access Journals (Sweden)

Ersez Tunay

2017-01-01

Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

Science.gov (United States)

Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

2017-09-01

The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Interstitial rotating shield brachytherapy for prostate cancer

International Nuclear Information System (INIS)

Adams, Quentin E.; Xu, Jinghzu; Breitbach, Elizabeth K.; Li, Xing; Rockey, William R.; Kim, Yusung; Wu, Xiaodong; Flynn, Ryan T.; Enger, Shirin A.

2014-01-01

Purpose: To present a novel needle, catheter, and radiation source system for interstitial rotating shield brachytherapy (I-RSBT) of the prostate. I-RSBT is a promising technique for reducing urethra, rectum, and bladder dose relative to conventional interstitial high-dose-rate brachytherapy (HDR-BT). Methods: A wire-mounted 62 GBq 153 Gd source is proposed with an encapsulated diameter of 0.59 mm, active diameter of 0.44 mm, and active length of 10 mm. A concept model I-RSBT needle/catheter pair was constructed using concentric 50 and 75 μm thick nickel-titanium alloy (nitinol) tubes. The needle is 16-gauge (1.651 mm) in outer diameter and the catheter contains a 535 μm thick platinum shield. I-RSBT and conventional HDR-BT treatment plans for a prostate cancer patient were generated based on Monte Carlo dose calculations. In order to minimize urethral dose, urethral dose gradient volumes within 0–5 mm of the urethra surface were allowed to receive doses less than the prescribed dose of 100%. Results: The platinum shield reduced the dose rate on the shielded side of the source at 1 cm off-axis to 6.4% of the dose rate on the unshielded side. For the case considered, for the same minimum dose to the hottest 98% of the clinical target volume (D 98% ), I-RSBT reduced urethral D 0.1cc below that of conventional HDR-BT by 29%, 33%, 38%, and 44% for urethral dose gradient volumes within 0, 1, 3, and 5 mm of the urethra surface, respectively. Percentages are expressed relative to the prescription dose of 100%. For the case considered, for the same urethral dose gradient volumes, rectum D 1cc was reduced by 7%, 6%, 6%, and 6%, respectively, and bladder D 1cc was reduced by 4%, 5%, 5%, and 6%, respectively. Treatment time to deliver 20 Gy with I-RSBT was 154 min with ten 62 GBq 153 Gd sources. Conclusions: For the case considered, the proposed 153 Gd-based I-RSBT system has the potential to lower the urethral dose relative to HDR-BT by 29%–44% if the clinician allows

Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

International Nuclear Information System (INIS)

Turner, Andrew; Pampin, Raul; Loughlin, M.J.; Ghani, Zamir; Hurst, Gemma; Lo Bue, Alessandro; Mangham, Samuel; Puiu, Adrian; Zheng, Shanliang

2014-01-01

The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge

Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

Energy Technology Data Exchange (ETDEWEB)

Turner, Andrew, E-mail: andrew.turner@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pampin, Raul [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Loughlin, M.J. [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghani, Zamir; Hurst, Gemma [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lo Bue, Alessandro [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Mangham, Samuel [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Puiu, Adrian [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Zheng, Shanliang [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2014-10-15

The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge.

SU-E-T-556: Monte Carlo Generated Dose Distributions for Orbital Irradiation Using a Single Anterior-Posterior Electron Beam and a Hanging Lens Shield

International Nuclear Information System (INIS)

Duwel, D; Lamba, M; Elson, H; Kumar, N

2015-01-01

Purpose: Various cancers of the eye are successfully treated with radiotherapy utilizing one anterior-posterior (A/P) beam that encompasses the entire content of the orbit. In such cases, a hanging lens shield can be used to spare dose to the radiosensitive lens of the eye to prevent cataracts. Methods: This research focused on Monte Carlo characterization of dose distributions resulting from a single A-P field to the orbit with a hanging shield in place. Monte Carlo codes were developed which calculated dose distributions for various electron radiation energies, hanging lens shield radii, shield heights above the eye, and beam spoiler configurations. Film dosimetry was used to benchmark the coding to ensure it was calculating relative dose accurately. Results: The Monte Carlo dose calculations indicated that lateral and depth dose profiles are insensitive to changes in shield height and electron beam energy. Dose deposition was sensitive to shield radius and beam spoiler composition and height above the eye. Conclusion: The use of a single A/P electron beam to treat cancers of the eye while maintaining adequate lens sparing is feasible. Shield radius should be customized to have the same radius as the patient’s lens. A beam spoiler should be used if it is desired to substantially dose the eye tissues lying posterior to the lens in the shadow of the lens shield. The compromise between lens sparing and dose to diseased tissues surrounding the lens can be modulated by varying the beam spoiler thickness, spoiler material composition, and spoiler height above the eye. The sparing ratio is a metric that can be used to evaluate the compromise between lens sparing and dose to surrounding tissues. The higher the ratio, the more dose received by the tissues immediately posterior to the lens relative to the dose received by the lens

Shielding modification design of the N.S. Mutsu

International Nuclear Information System (INIS)

Yamaji, A.; Miyakoshi, J.; Kageyama, T.; Futamura, Y.

1983-01-01

Shielding modification design of the N.S. Mutsu was performed for reducing the radiation doses outside the primary and the secondary shields by providing shields for neutrons streaming through the air gap between the pressure vessel and the primary shield. This was accomplished by replacing parts of the shields and adding new shields in the upper and lower sections of both primary and secondary shields, and also replacing the thermal insulator in the gap. The shielding design calculations were made using one- and two-dimensional discrete ordinates codes and also a point kernel code. Special attention was paid to the calculations of, (1) the neutrons streaming through the gap between the pressure vessel and the primary shield, (2) the radiations transmitted through the radial shield of the core in the primary shield, (3) the radiations transmitted through the upper and lower sections of the secondary shield, and (4) the dose rate equivalent in the accommodation area. Their calculational accuracies were estimated by analyzing various experiments. To support the modification, a variety of experiments and tests were carried out, which were material tests, cooling test of the primary shield, mechanical strength test of the double bottom, trial fabrication tests of new shields, performance degradation test of heavy concrete and duct streaming experiment in the secondary shield. (author)

Shielding requirements for particle bed propulsion systems

Science.gov (United States)

Gruneisen, S. J.

1991-06-01

Nuclear Thermal Propulsion systems present unique challenges in reliability and safety. Due to the radiation incident upon all components of the propulsion system, shielding must be used to keep nuclear heating in the materials within limits; in addition, electronic control systems must be protected. This report analyzes the nuclear heating due to the radiation and the shielding required to meet the established criteria while also minimizing the shield mass. Heating rates were determined in a 2000 MWt Particle Bed Reactor (PBR) system for all materials in the interstage region, between the reactor vessel and the propellant tank, with special emphasis on meeting the silicon dose criteria. Using a Lithium Hydride/Tungsten shield, the optimum shield design was found to be: 50 cm LiH/2 cm W on the axial reflector in the reactor vessel and 50 cm LiH/2 cm W in a collar extension of the inside shield outside of the pressure vessel. Within these parameters, the radiation doses in all of the components in the interstage and lower tank regions would be within acceptable limits for mission requirements.

Assessment of axial gamma dose rate profile on irradiated fuel assembly using polycarbonate film and perspex dosimeters

International Nuclear Information System (INIS)

Joshi, V.B.; Janardhanan, S.; Pillai, P.R.; Somanathan, K.; Narayan, K.K.; John, J.; Kutty, K.N.; Deo, V.R.; Popli, O.L.

1986-01-01

The dose-rate profile of irradiated fuel rod is required for optimisation of radiation shielding from safety point of view during storage, handling and metallurgical examination. Since the dose-rates are in kilogray per hour, their determination requires special evaluation techniques. This paper illustrates the application of Makrofol-N and red perspex (AERE 4034B) for this purpose. They are compared with CaSO 4 :Dy thermoluminescence dosimeter. (author). 4 refs

A comparison of dose savings of lead and lightweight aprons for shielding of 99m-Technetium radiation

International Nuclear Information System (INIS)

Warren-Forward, H.; Cardew, P.; Smith, B.; Clack, L.; McWhirter, K.; Johnson, S.; Wessel, K.

2007-01-01

Nuclear medicine technologists (NMTs) have the highest effective doses of radiation among medical workers. With increase in the use of lightweight materials in diagnostic radiography, the aim was to compare the effectiveness of lead and lightweight aprons in shielding from 99m-Technetium ( 99m Tc) gamma rays. The doses received from a scattering phantom to the entrance, 9 cm depth and exit of a phantom were measured with LiF:Mg, Cu, P thermoluminescent dosemeters (TLDs). Doses and spectra were assessed without no shielding, with 0.5-mm lead and lightweight aprons. The lead and lightweight aprons decreased entrance surface doses by 76 and 59%, respectively. The spectral analysis showed that the lightweight apron provided better dose reduction at energies 99m Tc labelled radiopharmaceutical. (authors)

Shielding designs and tests of a new exclusive ship for transporting spent nuclear fuels

International Nuclear Information System (INIS)

Ono, M.; Ito, D.; Kitano, T.; Ueki, K.; Akiyama, H.; Obara, I.; Sanui, T.

2000-01-01

The Rokuei-Maru, a ship built specially for the transport of spent nuclear fuels in casks, was launched April in 1996. She is the first ship to comply with special Japanese regulations, KAISA 520, based on the INF code. DOT3.5 and MCNP-4A were used for the evaluation of dose equivalent rates of her shielding structures. On-board gamma-ray shielding tests were executed to confirm the effectiveness of the ship's shielding performance. The tests confirmed that effective shielding has been achieved and the dose equivalent rate in the accommodation and other inhabited spaces is sufficiently lower than the regulated limitations. This was achieved by employing the appropriate calculation methods and shielding materials. (author)

Gonad shielding in computerized tomography

International Nuclear Information System (INIS)

Rockstroh, G.

1984-01-01

The reduction of gonadal dose by shielding of the gonads was investigated for a Somatom 2 using an anthropomorphic phantom. For small distances from the slice examined the gonadal dose results from intracorporal secondary radiation and is only insignificantly reduced by shielding. For greater distances shielding is relatively more effective, the gonadal dose however is small because of the approximately exponential decay. Shielding of the gonads therefore does not seem adequate for the reduction of gonadal dose. From dose measurements in cylinder phantoms of several diameters it appears that no different results would be obtained for children and young adults. An effective reduction of gonadal dose is only possible with lead capsules for males. (author)

EBT-P gamma-ray-shielding analysis

International Nuclear Information System (INIS)

Gohar, Y.

1983-01-01

First, a one-dimensional scoping study was performed for the gamma-ray shield of the ELMO Bumpy Torus proof-of-principle device to define appropriate shielding material and determine the required shielding thickness. The dose-equivalent results are analyzed as a function of the radiation-shield thickness for different shielding options. A sensitivity analysis for the pessimistic case is given. The recommended shielding option based on the performance and cost is discussed. Next, a three-dimensional scoping study for the coil shield was performed for four different shielding options to define the heat load for each component and check the compliance with the design criterion of 10 watts maximum heat load per coil from the gamma-ray sources. Also, a detailed biological-dose survey was performed which included: (a) the dose equivalent inside and outside the building, (b) the dose equivalent from the two mazes of the building, and (c) the skyshine contribution to the dose equivalent

DOSKMF2, Dose Rate Distribution in Co60 Gamma Irradiation Plants

International Nuclear Information System (INIS)

Remer, M.

1988-01-01

1 - Description of program or function: The DOSKMF2 code calculates dose rate distributions in gamma irradiation facilities with 60 Co sources. It has been developed as a tool for the evaluation of research units for the characterization, modification and optimization as well as for the simulation, planning design and evaluation of pilot and industrial facilities. 2 - Method of solution: The basic model of DOSKMF2 contains the computation of the exposure rate, which is dependent on a system of radiation sources (line and/or point sources) at one point of the radiation field. The geometric conditions are described by two coordinate systems. The first is used to arrange the radiation sources and tubes; the second is used to describe the shielding layers in the form of concentric circles. The build-up factor is approximated by Taylor coefficients, also in the case of multi-layer configurations. Calculations of radiation fields in complex irradiation facilities are made by many organizational variants based on the basic model. The interpretation of the results is supported by some statistical calculations. It is possible to store the calculated dose rate values for further processing. 3 - Restrictions on the complexity of the problem: The DOSKMF2 code is presently limited to 40 irradiation tubes and 10 crossed shielding layers. This can be increased by changing dimension statements and input organization

Dose estimates in a loss of lead shielding truck accident.

Energy Technology Data Exchange (ETDEWEB)

Dennis, Matthew L.; Osborn, Douglas M.; Weiner, Ruth F.; Heames, Terence John (Alion Science & Technology Albuquerque, NM)

2009-08-01

The radiological transportation risk & consequence program, RADTRAN, has recently added an updated loss of lead shielding (LOS) model to it most recent version, RADTRAN 6.0. The LOS model was used to determine dose estimates to first-responders during a spent nuclear fuel transportation accident. Results varied according to the following: type of accident scenario, percent of lead slump, distance to shipment, and time spent in the area. This document presents a method of creating dose estimates for first-responders using RADTRAN with potential accident scenarios. This may be of particular interest in the event of high speed accidents or fires involving cask punctures.

Set of programs for determining exposure and dose rates from selected sources of gamma radiation

International Nuclear Information System (INIS)

Hep, J.; Kralovcova, E.; Smutny, V.; Valenta, V.

1982-01-01

The programs are described for the determination of exposure and dose rate of gamma radiation from point, surface, linear and volume sources with and without shielding. The computation is conducted using the classical method taking into consideration the buildup factor. For the computation of the buildup factor in heterogeneous shielding the Broder and Kitazuma formulas are used. Kitazuma's alpha coefficients were calculated recurrently using a new semi-empirical method. Taylor's approximation was used for the calculation of the buildup factor in a single layer

Radiological dose rate calculations for the International Thermonuclear Experimental Reactor (ITER)

International Nuclear Information System (INIS)

Khater, H.Y.; Santoro, R.T.

1996-01-01

Two-dimensional biological dose rates were calculated at different locations outside the International Thermonuclear Experimental Reactor (ITER) design. An 18 degree sector of the reactor was modeled in r-θ geometry. The calculations were performed for three different pulsing scenarios. This included a single pulse of 1000 s duration, 10 pulses of 1000 s duration with a 50% duty factor, and 9470 pulses of 1000 s duration with a 50% duty factor for a total fluence of 0.3 MW.a/m 2 . The dose rates were calculated as a function of toroidal angle at locations in the space between the toroidal field (TF) coils and cryostat, and in the space between the cryostat and the biological shield. The two-dimensional results clearly showed the toroidal effect, which is dominated by contribution from the activation of the cryostat and the biological shield. After one pulse, full access to the machine is possible within a few hours following shutdown. After 10 pulses, full access is also possible within the first day following shutdown. At the end of the Basic Performance Phase (BPP), full access is possible at any of the locations considered after one week following shutdown. 5 refs., 5 figs., 2 tabs

Normalization of shielding structure quality and the method of its studying

International Nuclear Information System (INIS)

Bychkov, Ya.A.; Lavdanskij, P.A.

1987-01-01

Method for evaluation of nuclear facility radiation shield quality is suggested. Indexes of shielding structure radiation efficiency and face efficiency are used as the shielding structure quality indexes. The first index is connected with radiation dose rate during personnel irradiation behind the shield, and the second one - with the stresses in shielding structure introduction of the indexes presented allows to evaluate objectively the quality of nuclear facility shielding structure quality design construction and operation and to economize labour and material resources

Location of radiosensitive organs, measurement of absorbed dose to radiosensitive organs and use of bismuth shields in paediatric anthropomorphic phantoms

International Nuclear Information System (INIS)

Inkoom, S.

2014-08-01

The aim of this study was to investigate: firstly, (i) location of radiosensitive organs in the interior of four (4) paediatric anthropomorphic phantoms, and, secondly, (ii) effectiveness of single and double bismuth thyroid shields, distance between shield and phantom surface, during paediatric multi-detector computed tomography (MDCT) using fixed tube current (FTC) and automatic exposure control (AEC) on dose reduction and image quality. Four (4) paediatric anthropomorphic phantoms representing the equivalent of a newborn, 1-, 5-, and 10-y-old child underwent head, thorax and abdomen computed tomography (CT) scans. CT and magnetic resonance imaging scans of all children aged 0-16 y-old performed during a 5-y-period at the University Hospital of Heraklion, Crete, Greece were reviewed, and five hundred and three (503) were found to be eligible for normal anatomy. Anterior-posterior and lateral dimensions of twelve (12) of the above children closely matched that of the phantoms' thoracic and abdominal region in each four (4) phantoms. The mid-sagittal plane (MSP) and mid-coronal plane (MCP) were drawn on selected matching axial images of patients and phantoms. Multiple points outlining large radiosensitive organs and centres of small organs in patient images were identified at each slice level and their orthogonal distances from the MSP and MCP were measured. The outlines and centres of all radiosensitive organs were reproduced using the coordinates of each organ on the corresponding phantom's transverse images. The four (4) phantoms were also subjected to routine head and neck, neck and thorax CT scans on a 16-slice CT system. Each phantom was first scanned with both FTC and AEC for with and without bismuth shields. Each scan was repeated ten (10) times to increase thermoluminescent dosimeters (TLDs) signal and reduce measurement statistical error. For neck CT, the effect of using single and double thickness of bismuth shields and 1-3 cm cotton spacers

Shielding calculations for the TFTR neutral beam injectors

International Nuclear Information System (INIS)

Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr.; Barnes, J.M.

1979-07-01

Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse

In-Plane Shielding for CT: Effect of Off-Centering, Automatic Exposure Control and Shield-to-Surface Distance

Energy Technology Data Exchange (ETDEWEB)

Kalra, Mannudeep K.; Dang, Pragya; Singh, Sarabjeet; Saini, Sanjay; Shepard, Jo Anne O. [Massachusetts General Hospital, Boston (United States)

2009-04-15

To assess effects of off-centering, automatic exposure control, and padding on attenuation values, noise, and radiation dose when using in-plane bismuth-based shields for CT scanning. A 30 cm anthropomorphic chest phantom was scanned on a 64-multidetector CT, with the center of the phantom aligned to the gantry isocenter. Scanning was repeated after placing a bismuth breast shield on the anterior surface with no gap and with 1, 2, and 6 cm of padding between the shield and the phantom surface. The 'shielded' phantom was also scanned with combined modulation and off-centering of the phantom at 2 cm, 4 cm and 6 cm below the gantry isocenter. CT numbers, noise, and surface radiation dose were measured. The data were analyzed using an analysis of variance. The in-plane shield was not associated with any significant increment for the surface dose or CT dose index volume, which was achieved by comparing the radiation dose measured by combined modulation technique to the fixed mAs (p > 0.05). Irrespective of the gap or the surface CT numbers, surface noise increased to a larger extent compared to Hounsfield unit (HU) (0-6 cm, 26-55%) and noise (0-6 cm, 30-40%) in the center. With off-centering, in-plane shielding devices are associated with less dose savings, although dose reduction was still higher than in the absence of shielding (0 cm off-center, 90% dose reduction; 2 cm, 61%) (p < 0.0001). Streak artifacts were noted at 0 cm and 1 cm gaps but not at 2 cm and 6 cm gaps of shielding to the surface distances. In-plane shields are associated with greater image noise, artificially increased attenuation values, and streak artifacts. However, shields reduce radiation dose regardless of the extent of off-centering. Automatic exposure control did not increase radiation dose when using a shield.

Radiation shielding in dental radiography

International Nuclear Information System (INIS)

Stenstroem, B.; Rehnmark-Larsson, S.; Julin, P.; Richter, S.

1983-01-01

The protective effect in the thyroid region from different types of radiation shieldings at intraoral radiography has been studied as well as the reduction of the absorbed dose to the sternal and the gonadal regions. The shieldings tested were five different types of leaded aprons, of which three had an attached leaded collar and the other two were used in combination with separate soft leaded collars. Furthermore one of the soft leaded collars and an unflexible horizontal leaded shield were tested separately. Two dental x-ray machines of 60 and 65 kVp with rectangular and circular tube collimators were used. The exposure time corresponded to speed group E film. The absorbed doses were measured with two ionization chambers. No significant difference in the protective effect in the thyroid gland could be found between the different types of radiation shieldings. There was a dose reduction by approximately a factor of 2 to the thyroid region down to 0.08 mGy per full survey using parallelling technique, and below 0.001 mGy per single bitewing exposure. The shieldings reduced the thyroid dose using bisecting-angle technique by a factor of 5 down to 0.15 mGy per full survey (20 exposures). In the sternal region the combinations of apron and collar reduced the absorbed dose from a full survey to below 2 μGy compared with 18 μGy (parallelling) and 31 μGy (biscting-angle) without any shielding. With the horizontal leaded shield a reduction by a factor of 6 was obtained but no significant sternal dose reduction could be detected from the soft collar alone. The gonadal dose could be reduced by a factor of 10 with the horizontal leaded shield, parallelling technique and circular collimator. Using leaded aprons the gonadal dose was approximately one per cent of the dose without any shielding, i.e. below 0.01 μGy per single intraoral exposure. (Authors)

Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

International Nuclear Information System (INIS)

Carbajo, J.J.; Lindner, C.N.

1992-01-01

The design of spent-fuel casks for storage and transport requires modeling the cask for criticality, shielding, thermal, and structural analyses. While some parts of the cask are homogeneous, other regions are heterogeneous with different materials intermixed. For simplicity, some of the heterogeneous regions may be modeled as homogeneous. This paper evaluates the effect of homogenizing some regions of a cask on calculating radiation dose rates outside the cask. The dose rate calculations were performed with the one-dimensional discrete ordinates shielding XSDRNPM code coupled with the XSDOSE code and with the three-dimensional QAD-CGGP code. Dose rates were calculated radially at the midplane of the cask at two locations, cask surface and 2.3 m from the radial surface. The last location corresponds to a point 2 m from the lateral sides of a transport railroad car

Monte Carlo simulation using MCNP4B for an optimal shielding design of a 252 Cf source

International Nuclear Information System (INIS)

Silva, Ademir X. da; Crispim, Verginia R.

2001-01-01

This study aim to investigate an optimum shielding design against neutrons and gamma-rays from a source of 252 Cf, using Monte Carlo simulation. The shielding materials studied were: borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP, version 4B, was used to design shielding for 252 Cf based neutron irradiator systems. By normalizing the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independents of the intensity of actual 252 Cf source. The results shown what the total dose equivalent rates were reduced significantly by the shielding system optimization. (author)

Calculations of the photon dose behind concrete shielding of high energy proton accelerators

International Nuclear Information System (INIS)

Dworak, D.; Tesch, K.; Zazula, J.M.

1992-02-01

The photon dose per primary beam proton behind lateral concrete shieldings was calculated by using an extension of the Monte Carlo particle shower code FLUKA. The following photon-producing processes were taken into account: capture of thermal neutrons, deexcitation of nuclei after nuclear evaporation, inelastic neutron scattering and nuclear reactions below 140 MeV, as well as photons from electromagnetic cascades. The obtained ratio of the photon dose to the neutron dose equivalent varies from 8% to 20% and it well compares with measurements performed recently at DESY giving a mean ratio of 14%. (orig.)

Shielding concerns at a spallation source

International Nuclear Information System (INIS)

Russell, G.J.; Robinson, H.; Legate, G.L.; Woods, R.

1989-01-01

Neutrons produced by 800-MeV proton reactions at the Los Alamos Neutron Scattering Center spallation neutron source cause a variety of challenging shielding problems. We identify several characteristics distinctly different from reactor shielding and compute the dose attenuation through an infinite slab/shield composed of iron (100 cm) and borated polyethylene (15 cm). Our calculations show that (for an incident spallation spectrum characteristic of neutrons leaking from a tungsten target at 90/degree/) the dose through the shield is a complex mixture of neutrons and gamma rays. High-energy (> 20 MeV) neutron production from the target is ≅5% of the total, yet causes ≅68% of the dose at the shield surface. Primary low-energy (< 20 MeV) neutrons from the target contribute negligibly (≅0.5%) to the dose at the shield surface yet cause gamma rays, which contribute ≅31% to the total dose at the shield surface. Low-energy neutrons from spallation reactions behave similarly to neutrons with a fission spectrum distribution. 6 refs., 8 figs., 1 tab

Survivor shielding. Part A. Nagasaki factory worker shielding

International Nuclear Information System (INIS)

Santoro, Robert T.; Barnes, John M.; Azmy, Yousry Y.; Kerr, George D.; Egbert, Stephen D.; Cullings, Harry M.

2005-01-01

Recent investigations based on conventional chromosome aberration data by the RERF suggest that the DS86 doses received by many Nagasaki factory workers may have been overestimated by as much as 40% relative to those for other survivors in Japanese-type houses and other shielding configurations (Kodama et al. 2001). Since the factory workers represent about 25% of the Nagasaki survivors with DS86 doses in excess of 0.5 Gy (50 rad), systematic errors in their dose estimates can have a major impact on the risk coefficients from RERF studies. The factory worker doses may have been overestimated for a number of reasons. The calculation techniques, including the factory building modeling, weapon source spectra and cross-section data used in the DS86 shielding calculations were not detailed enough to replicate actual conditions. The models used did not take into account local shielding provided by machinery, tools, and the internal structure in the buildings. In addition, changes in the disposition of shielding following collapse of the building by the blast wave were not considered. The location of large factory complexes may be uncertain, causing large numbers of factory survivors, correctly located relative to each other, to be uniformly too close to the hypocenter. Any or all of these reasons are sufficient to result in an overestimate of the factory worker doses. During the DS02 studies, factory worker doses have been reassessed by more carefully modeling the factory buildings, incorporating improved radiation transport methods and cross-section data and using the most recent bomb leakage spectra (Chapter 2). Two-dimensional discrete ordinates calculations were carried out initially to estimate the effects of workbenches and tools on worker doses to determine if the inclusion of these components would, in fact, reduce the dose by amounts consistent with the RERF observations (Kodama et al. 2001). (author)

Evaluation of a pyrex glass shield for the dose reduction in extremities to manipulate a 90 Sr- 90 Y generator

International Nuclear Information System (INIS)

Ayra P, F.E.; Xiques C, A.; Torres B, M.B.

2006-01-01

The production of Y-90 of high activity it specifies (free of payee) for their use in radioimmunotherapy uses the Strontium 90 as isotope source. Depending on the method employee for the separation of both isotopes several types of generators are described in different bibliographies. The column generator used in the facilities of the Center of Isotopes requires of a frequent manipulation causing significant dose in the skin of the extremities due to the exhibition to the radiation beta of high energy. The properties of the shieldings for this radiation type have been well studied Y they consist in several publications. To be in correspondence with requirements of radiological protection in the Cuban legislation, the column was covered with a tube of glass pyrex of 5 mm of thickness and it was monitored the exposure with an ionization chamber. At the own time, the shielding using the Monte Carlo method was evaluated. It was used the MCNP 4C code to simulate the absorption of the beta particles generated in the process of disintegration of the Sr-90 and Y-90 in the glass shielding. The column generator and the fluence of beta particles were modeled in different points inside the shielding to determine if the experimentally measured values correspond to electrons that were not absorbed or to the weak stopping radiation generated in the glass due to the deceleration of these particles. A cylinder of 4 mm of diameter simulates the source (it dilutes) and a tube of walls of 6 mm of thickness simulates the shielding more the wall of the column around the generator. This it was divided in cells of 1 mm of thickness and the energy deposited in them was evaluated. The results show that all the electrons generated in the source are absorbed in the shielding and the exposure rates decrease in more of 78 times using the 5 mm of pyrex glass. The doses in extremities to the operators of the generator don't surpass the 70 mSv by year that is the dose restriction imposed in the

SU-C-BRB-02: Symmetric and Asymmetric MLC Based Lung Shielding and Dose Optimization During Translating Bed TBI

Energy Technology Data Exchange (ETDEWEB)

Ahmed, S; Kakakhel, MB [Pakistan Institute of Engineering & Applied Sciences (PIEAS), Islamabad (Pakistan); Ahmed, SBS; Hussain, A [Aga Khan University Hospital (AKUH), Karachi (Pakistan)

2015-06-15

Purpose: The primary aim was to introduce a dose optimization method for translating bed total body irradiation technique that ensures lung shielding dynamically. Symmetric and asymmetric dynamic MLC apertures were employed for this purpose. Methods: The MLC aperture sizes were defined based on the radiological depth values along the divergent ray lines passing through the individual CT slices. Based on these RD values, asymmetrically shaped MLC apertures were defined every 9 mm of the phantom in superior-inferior direction. Individual MLC files were created with MATLAB™ and were imported into Eclipse™ treatment planning system for dose calculations. Lungs can be shielded to an optimum level by reducing the MLC aperture width over the lungs. The process was repeated with symmetrically shaped apertures. Results: Dose-volume histogram (DVH) analysis shows that the asymmetric MLC based technique provides better dose coverage to the body and optimum shielding of the lungs compared to symmetrically shaped beam apertures. Midline dose homogeneity is within ±3% with asymmetric MLC apertures whereas it remains within ±4.5% with symmetric ones (except head region where it drops down to −7%). The substantial over and under dosage of ±5% at tissue interfaces has been reduced to ±2% with asymmetric MLC technique. Lungs dose can be reduced to any desired limit. In this experiment lungs dose was reduced to 80% of the prescribed dose, as was desired. Conclusion: The novel asymmetric MLC based technique assures optimum shielding of OARs (e.g. lungs) and better 3-D dose homogeneity and body-dose coverage in comparison with the symmetric MLC aperture optimization. The authors acknowledge the financial and infrastructural support provided by Pakistan Institute of Engineering & Applied Sciences (PIEAS), Islamabad and Aga Khan University Hospital (AKUH), Karachi during the course of this research project. Authors have no conflict of interest with any national / international

Discussion on the standardization of concrete composition for radiation shielding design 2. Evaluation of the effect of the composition variance on the shielding property

International Nuclear Information System (INIS)

Ogata, Tomohiro; Kimura, Ken-ichi; Nakata, Mikihiro; Okuno, Koichi; Ishikawa, Tomoyuki

2017-01-01

Radiation Shielding Material Standardization Working Group of AESJ has been organized to establish Japanese standard concrete composition for radiation shielding design. We have collected concrete composition data to organize a representative concrete composition data. Neutron and Gamma dose rates penetrated through several concrete compositions are calculated by one dimensional discrete ordinate code ANISN. Effects of the variation of concrete composition on the neutron and gamma dose are evaluated. In this paper, recent standardization activity is summarized. (author)

Effect of dose rate on intestinal tolerance in mice. Implications in radiotherapy

International Nuclear Information System (INIS)

Wambersie, A.; Stienon-Smoes, M.R.; Octave-Prignot, M.

1978-01-01

Effect of dose rate on intestinal tolerance after 60 Co irradiation was studied in BALB/c mice. Intestinal tolerance was assessed from LD50, after selective abdominal irradiation and after total body irradiation. Three dose rates were compared, corresponding to irradiation times of about 15-20 minutes ('acute irradiation' taken as reference), 5-6 hours and 10-15 hours. Irradiations were performed simultaneously, with three telecobaltherapy units, the dose rates being adjusted with lead shields and by increasing the distances. Comparison of the experimental data already published indicates that, for some biological systems and effects, additional dose necessary to reach a given effect when passing from 'acute' to 'continuous low dose rate' irradiation is comparable to that expected when considering only repair of sublethal lesions. For other biological systems and effects, it is necessary to consider, besides repair of sublethal lesions, other mechanisms such as cell distribution and, for tumours, the oxygen effect. A differential effect then appears to be possible. However, as far as the clinical applications are concerned, a general agreement is not yet reached on the exact shape of the iso-effect curves as a function of irradiation time for the effects relevant to radiation therapy [fr

A model-based approach of scatter dose contributions and efficiency of apron shielding for radiation protection in CT.

Science.gov (United States)

Weber, N; Monnin, P; Elandoy, C; Ding, S

2015-12-01

Given the contribution of scattered radiations to patient dose in CT, apron shielding is often used for radiation protection. In this study the efficiency of apron was assessed with a model-based approach of the contributions of the four scatter sources in CT, i.e. external scattered radiations from the tube and table, internal scatter from the patient and backscatter from the shielding. For this purpose, CTDI phantoms filled with thermoluminescent dosimeters were scanned without apron, and then with an apron at 0, 2.5 and 5 cm from the primary field. Scatter from the tube was measured separately in air. The scatter contributions were separated and mathematically modelled. The protective efficiency of the apron was low, only 1.5% in scatter dose reduction on average. The apron at 0 cm from the beam lowered the dose by 7.5% at the phantom bottom but increased the dose by 2% at the top (backscatter) and did not affect the centre. When the apron was placed at 2.5 or 5 cm, the results were intermediate to the one obtained with the shielding at 0 cm and without shielding. The apron effectiveness is finally limited to the small fraction of external scattered radiation. Copyright © 2015 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Secondary gamma-ray data for shielding calculation

International Nuclear Information System (INIS)

Miyasaka, Sunichi

1979-01-01

In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

SU-E-T-548: Modeling of Breast IORT Using the Xoft 50 KV Brachytherapy Source and 316L Steel Rigid Shield

Energy Technology Data Exchange (ETDEWEB)

Burnside, W [Mountain View, CA (United States)

2015-06-15

Purpose: Xoft provides a set of 316L Stainless Steel Rigid Shields to be used with their 50 kV X-ray source for Breast IORT treatments. Modeling the different shield sizes in MCNP provides information to help make clinical decisions for selecting the appropriate shield size. Methods: The Xoft Axxent 50 kV Electronic Brachytherapy System has several applications in radiation therapy, one of which is treating cancer of the breast intraoperatively by placing the miniaturized X-ray tube inside an applicator balloon that is expanded to fill the lumpectomy bed immediately following tumor removal. The ribs, lung, and muscular chest wall are all regions at risk to receive undesired dose during the treatment. A Xoft 316L Stainless Steel Rigid Shield can be placed between the intracostal muscles of the chest wall and the remaining breast tissue near the balloon to attenuate the beam and protect these organs. These shields are provided in 5 different sizes, and the effects on dose to the surrounding tissues vary with shield size. MCNP was used to model this environment and tally dose rate to certain regions of interest. Results: The average rib dose rate calculated using 0cm (i.e., no shield), 3cm, and 5cm diameter shields were 26.89, 15.43, and 8.91 Gy/hr respectively. The maximum dose rates within the rib reached 94.74 Gy/hr, 53.56 Gy/hr, and 31.44 Gy/hr for the 0cm, 3cm, and 5cm cases respectively. The shadowing effect caused by the steel shields was seen in the 3-D meshes and line profiles. Conclusion: This model predicts a higher dose rate to the underlying rib region with the 3cm shield compared to the 5cm shield; it may be useful to select the largest possible diameter when choosing a shield size for a particular IORT patient. The ability to attenuate the beam to reduce rib dose was also confirmed. Research sponsored by Xoft Inc, a subsidiary of iCAD.

Quality control, mean glandular dose estimate and room shielding calculation in mammography

International Nuclear Information System (INIS)

Rakotomalala, H.M.

2014-01-01

This study focuses in the importance of Radiation Protection in mammography. A good control of the radiological risk depends on the dose optimization, room shielding calculation and the quality of equipment. The work was carried out in the three private medical centers called A, B, and C. Dosimetry estimates were made on the equipment of the three centers. Values has been compared with the Diagnostic Reference Levels established by the International Atomic Energy Agency (IAEA). Conformity control of the radiological devices has also been done with the Mammographic Quality Control Kit of the INSTN-Madagascar. Verifications of shields of the room containing the mammography equipment were done by theoretical calculations using the method provided by NCRP 147. [fr

Assessment of shielding integrity of Co-60 gamma irradiation-ray scanner at Aflao Border, Ghana

International Nuclear Information System (INIS)

Agbemafo, Edwin Capacity

2016-07-01

This study examines the current state of the shielding integrity of the 38.7 TBq Co-60 gamma ray scanner with an average energy of 1.25 MeV operated by NICK TC Scan Limited, which has been in use for destination inspection at Aflao Border of Ghana, for the past six years, (2010-2016). The facility uses a high energy ionizing radiation in its operation; therefore continuous adequacy of the installed biological shielding is critical to the protection and safety of the workers and the general public. The workload of the facility has increased since its commissioning, requiring the review of the status of the installed shielding. Theoretical calculations for dose rates and barrier thicknesses based on tenth – value layer (TVL) concept and NCRP 151, 2005 recommendations, were done around the scanning facility using the current operational data. The results were then compared with the measured dose rates and the shielding thickness constituted during the commissioning stage, and international standards. Calculated dose rate at commissioning state ranges from 0.6μSv/hr to 2.4μSv/hr with an average dose rate of 1.43μSv/hr and that of the current operational state ranges from 1.1μSv/hr to 2.6μSv/hr with an average dose rate of 1.54μSv/hr, indicating an increase of 7.9%. Even though the dose rates were all below the recommended dose limit of 20μSvh"-"1 by NCRP, there has been an increase in dose to the staff and the general public. It has been observed that, the workload has increased three-fold from the commissioning stage to current operational state over the past six years. The assessment done on the installed shielding using the current operational data indicates that the shielding is inadequate in providing protection for the general public and the workers against X-ray radiation source of energy of at least 6MeV, and therefore the facility in its current state cannot be used to house a linear accelerator of energy up to 10MeV. (au)

Simulation of Shielding Effects on the Total Dose Observed in TDE of KISAT-1

Directory of Open Access Journals (Sweden)

Sung-Joon Kim

2001-06-01

Full Text Available The threshold voltage shift observed in TDE (Total Dose Experiment on board the KITSAT-1 is converted into dose (rad(SiO2 usinsg the result of laboratory calibration with Co-60 gamma ray source in KAERI (Korea Atomic Energy Research Institute. Simulation using the NASA radiation model of geomagnetosphere verifies that the dose difference between RADFET1 and RADFET3 observed on KITSAT-1 comes from the difference in shielding thickness at the position of these RADFETs.

SU-E-T-400: Evaluation of Shielding and Activation at Two Pencil Beam Scanning Proton Facilities

International Nuclear Information System (INIS)

Remmes, N; Mundy, D; Classic, K; Beltran, C; Kruse, J; Herman, M; Stoker, J; Nelson, K; Bues, M

2015-01-01

Purpose: To verify acceptably low dose levels around two newly constructed identical pencil beam scanning proton therapy facilities and to evaluate accuracy of pre-construction shielding calculations. Methods: Dose measurements were taken at select points of interest using a WENDI-2 style wide-energy neutron detector. Measurements were compared to pre-construction shielding calculations. Radiation badges with neutron dose measurement capabilities were worn by personnel and also placed at points throughout the facilities. Seven neutron and gamma detectors were permanently installed throughout the facility, continuously logging data. Potential activation hazards have also been investigated. Dose rates near water tanks immediately after prolonged irradiation have been measured. Equipment inside the treatment room and accelerator vault has been surveyed and/or wipe tested. Air filters from air handling units, sticky mats placed outside of the accelerator vault, and water samples from the magnet cooling water loops have also been tested. Results: All radiation badges have been returned with readings below the reporting minimum. Measurements of mats, air filters, cooling water, wipe tests and surveys of equipment that has not been placed in the beam have all come back at background levels. All survey measurements show the analytical shielding calculations to be conservative by at least a factor of 2. No anomalous events have been identified by the building radiation monitoring system. Measurements of dose rates close to scanning water tanks have shown dose rates of approximately 10 mrem/hr with a half-life less than 5 minutes. Measurements around the accelerator show some areas with dose rates slightly higher than 10 mrem/hr. Conclusion: The shielding design is shown to be adequate. Measured dose rates are below those predicted by shielding calculations. Activation hazards are minimal except in certain very well defined areas within the accelerator vault and for objects

Shut-down dose rate analyses for the ITER electron cyclotron-heating upper launcher

Energy Technology Data Exchange (ETDEWEB)

Weinhorst, Bastian; Serikov, Arkady; Fischer, Ulrich; Lu, Lei [Institute for Neutron Physics and Reactor Technology INR (Germany); Karlsruhe Institute of Technology KIT (Germany); Spaeh, Peter; Strauss, Dirk [Institute for Applied Materials IAM (Germany); Karlsruhe Institute of Technology KIT (Germany)

2014-10-15

The electron cyclotron resonance heating upper launcher (ECHUL) is going to be installed in the upper port of the ITER tokamak thermonuclear fusion reactor for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). The paper reports the latest neutronic modeling and analyses which have been performed for the ITER reference front steering launcher design. It focuses on the port accessibility after reactor shut-down for which dose rate (SDDR) distributions on a fine regular mesh grid were calculated. The results are compared to those obtained for the ITER Dummy Upper Port. The calculations showed that the heterogeneous ECHUL design gives rise to enhanced radiation streaming as compared to the homogenous dummy upper port. Therefore the used launcher geometry was upgraded to a more recent development stage. The inter-comparison shows a significant improvement of the launchers shielding properties but also the necessity to further upgrade the shielding performance. Furthermore, the analysis for the homogenous dummy upper port, which represents optimal shielding inside the launcher, demonstrates that the shielding upgrade also needs to include the launcher's environment.

Application of a simple analytical model to estimate effectiveness of radiation shielding for neutrons

International Nuclear Information System (INIS)

Frankle, S.C.; Fitzgerald, D.H.; Hutson, R.L.; Macek, R.J.; Wilkinson, C.A.

1993-01-01

Neutron dose equivalent rates have been measured for 800-MeV proton beam spills at the Los Alamos Meson Physics Facility. Neutron detectors were used to measure the neutron dose levels at a number of locations for each beam-spill test, and neutron energy spectra were measured for several beam-spill tests. Estimates of expected levels for various detector locations were made using a simple analytical model developed for 800-MeV proton beam spills. A comparison of measurements and model estimates indicates that the model is reasonably accurate in estimating the neutron dose equivalent rate for simple shielding geometries. The model fails for more complicated shielding geometries, where indirect contributions to the dose equivalent rate can dominate

Radiation shielding for neutron guides

International Nuclear Information System (INIS)

Ersez, T.; Braoudakis, G.; Osborn, J.C.

2005-01-01

Full text: Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions. (authors)

Radiation shielding for neutron guides

International Nuclear Information System (INIS)

Ersez, T.; Braoudakis, G.; Osborn, J.C.

2006-01-01

Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions

Measurement of radiocesium concentration in trees using cumulative gamma radiation dose rate detection systems - A simple presumption for radiocesium concentration in living woods using glass-badge based gamma radiation dose rate detection system

Energy Technology Data Exchange (ETDEWEB)

Yoshihara, T.; Hashida, S.N. [Plant Molecular Biology, Laboratory of Environmental Science, Central Research Institute of Electric Power Industry (CRIEPI), 1646 Abiko, Chiba 270-1194 (Japan); Kawachi, N.; Suzui, N.; Yin, Y.G.; Fujimaki, S. [Radiotracer Imaging Gr., Quantum Beam Science Center, Japan Atomic Energy Agency (JAEA), 1233 Watanuki, Takasaki, Gunma 370-1292 (Japan); Nagao, Y.; Yamaguchi, M. [Takasaki Advanced Radiation Research Institute, Japan Atomic Energy Agency (JAEA), 1233 Watanuki, Takasaki, Gunma 370-1292 (Japan)

2014-07-01

Radiocesium from the severe accident at the Fukushima Dai-ichi Nuclear Power Plant on 11 March 2011 contaminates large areas. After this, a doubt for forest products, especially of mushroom, is indelible at the areas. Pruned woody parts and litters are containing a considerable amount of radiocesium, and generates a problem at incineration and composting. These mean that more attentive survey for each subject is expected; however, the present survey system is highly laborious/expensive and/or non-effective for this purpose. On the other hand, we can see a glass-badge based gamma radiation dose rate detection system. This system always utilized to detect a personal cumulative radiation dose, and thus, it is not suitable to separate a radiation from a specific object. However, if we can separate a radiation from a specific object and relate it with the own radiocesium concentration, it would enable us to presume the specific concentration with just an easy monitoring but without a destruction of the target nature and a complicated process including sampling, pre-treatment, and detection. Here, we present the concept of the measurement and results of the trials. First, we set glass-badges (type FS, Chiyoda Technol Corp., Japan) on a part of bough (approximately 10 cm in diameter) of Japanese flowering cherry trees (Prunus x yedoensis cv. Somei-Yoshino) with four different settings: A, a direct setting without any shield; B, a setting with an aluminum shield between bough and the glass-badge; C, a setting with a lead shield between bough and the glass-badge; D, a setting with a lead shield covering the glass-badge to shut the radiation from the surrounding but from bough. The deduction between the amount of each setting should separate a specific radiation of the bough from unlimited radiation from the surrounding. Even if the hourly dose rate is not enough to count the difference, a moderate cumulative dose would clear the difference. In fact, results demonstrated a

Shielding calculations using computer techniques; Calculo de blindajes mediante tecnicas de computacion

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Portilla, M. I.; Marquez, J.

2011-07-01

Radiological protection aims to limit the ionizing radiation received by people and equipment, which in numerous occasions requires of protection shields. Although, for certain configurations, there are analytical formulas, to characterize these shields, the design setup may be very intensive in numerical calculations, therefore the most efficient from to design the shields is by means of computer programs to calculate dose and dose rates. In the present article we review the codes most frequently used to perform these calculations, and the techniques used by such codes. (Author) 13 refs.

Survey of environmental radiation dose rates in Kyoto and Shiga prefectures, Japan

International Nuclear Information System (INIS)

Minamia, Kazuyuki; Shimo, Michikuni; Oka, Mitsuaki; Ejiri, Kazutaka; Sugino, Masato; Minato, Susumu; Hosoda, Masahiro; Yamada, Junya; Fukushi, Masahiro

2008-01-01

We have measured environmental radiation dose rates in several Prefectures, such as Ai chi Prefecture, Gifu Prefecture, and Mie Prefecture, in central Japan. Recently, we measured the environmental radiation dose rates in Kyoto and Shiga Prefectures that are also located in central Japan with a car-borne survey system. At the time of measurement, Kyoto Prefecture (area: 4,613 km 2 ) had a total of 36 districts, and Shiga Prefecture (area: 3,387 km 2 ) a total of 26. Terrestrial gamma ray dose rates and secondary cosmic ray dose rates were measured by a 2 inches ψ x 2 inches NaI(Tl) scintillation counter and a handy-type altimeter (GPS eTrex Legend by Gamin), respectively. The following factors were taken into consideration the shielding effect of the car body, the effect of the road pavement, radon progeny borne by precipitation, and increases in tunnels and near the walls. Terrestrial gamma ray dose rates in Kyoto and Shiga Prefectures were estimated to be 51.7 ± 6.0 n Gy/h (district average: 52.4 ± 4.7 n Gy/h), 52.2 ± 10.5 n Gy/h (district average: 51.9 ± 8.1 n Gy/h), respectively. Secondary cosmic ray dose rates in Kyoto and Shiga Prefectures were 30.0 ± 0.6 n Gy/h (district average: 29.9 ±0.3 n Gy/h), 30.1 ± 0.3 n Gy/h (district average: 30.0 ± 0.2 n Gy/h), respectively. The environmental radiation dose rates due to the sum dose rates of terrestrial gamma ray and secondary cosmic ray in Kyoto and Shiga Prefectures were 81.7 ± 6.2 n Gy/h (district average: 82.3 ± 4.8 n Gy/h), 82.3 ± 10.6 n Gy/h (district average: 82.0 ± 8.1 n Gy/h), respectively. We confirmed that the environmental radiation dose rates in Kyoto and Shiga Prefectures mainly depended on the change of the terrestrial gamma ray dose rates, since the secondary cosmic ray dose rates had little change. Therefore, radiation dose-rate maps of the terrestrial gamma rays as well as maps of the environmental radiation dose-rate were drawn. (author)

The status of low dose rate and future of high dose rate Cf-252 brachytherapy

International Nuclear Information System (INIS)

Rivard, M.J.; Wierzbicki, J.G.; Van den Heuvel, F.; Chuba, P.J.; Fontanesi, J.

1997-12-01

This work describes the current status of the US low dose rate (LDR) Cf-252 brachytherapy program. The efforts undertaken towards development of a high dose rate (HDR) remotely after loaded Cf-252 source, which can accommodate 1 mg or greater Cf-252, are also described. This HDR effort is a collaboration between Oak Ridge National Laboratory (ORNL), commercial remote after loader manufactures, the Gershenson Radiation Oncology Center (ROC), and Wayne State University. To achieve this goal, several advances in isotope chemistry and source preparation at ORNL must be achieved to yield a specific material source loading of greater than or equal 1 mg Cf-252 per mm3. Development work with both radioactive and non-radioactive stand-ins for Cf-252 have indicated the feasibility of fabricating such sources. As a result, the decreased catheter diameter and computer controlled source placement will permit additional sites (e.g. brain, breast, prostate, lung, parotid, etc.) to be treated effectively with Cf-252 sources. Additional work at the Radiochemical Engineering and Development Center (REDC) remains in source fabrication, after loader modification, and safe design. The current LDR Cf-252 Treatment Suite at the ROC is shielded and licensed to hold up to 1 mg of Cf-252. This was designed to maintain cumulative personnel exposure, both external to the room and in direct isotope handling, at less than 20 microSv/hr. However, cumulative exposure may be greatly decreased if a Cf-252 HDR unit is employed which would eliminate direct isotope handling and decrease treatment times from tilde 3 hours to an expected range of 3 to 15 minutes. Such a Cf-252 HDR source will also demonstrate improved dose distributions over current LDR treatments due to the ability to step the point-like source throughout the target volume and weight the dwell time accordingly

Shielding proposal to reduce cross-talk from ITER lower port to equatorial port

Energy Technology Data Exchange (ETDEWEB)

Juarez, Rafael, E-mail: rjuarez@ind.uned.es [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Pampin, Raul [F4E, Torres Diagonal Litoral B3, Josep Pla 2, Barcelona 08019 (Spain); Levesy, Bruno [ITER Organization, 13115 Route de Vinon sur Verdon, St Paul Lez Durance (France); Moro, Fabio [ENEA, Via Enrico Fermi 45, Frascati, Rome (Italy); Suarez, Alejandro [ITER Organization, 13115 Route de Vinon sur Verdon, St Paul Lez Durance (France); Catalan, J.P.; Sanz, Javier [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain)

2015-12-15

Radiation cross-talk from Torus Cryopump LP to EP was found to be a phenomenon driving Shutdown Dose Rates at EP Port Interspace after 12 days of cooling time, as relevant as neutron permeation through EP itself. In this work three different shields are proposed to mitigate the radiation cross-talk: two neutron shields placed inside LP, and a temporary gamma shield placed at EP PI during maintenance activities. Contributions from different reactor regions to Shutdown Dose Rates are computed, for the unshielded design, as long as the different shielded cases. The Rigorous-Two-Steps (R2S) method was used. The neutron shields inside TCP-LP are found to reduce SDR at EP PI 43 μSv/h and 99 μSv/h, while the gamma shield inside EP PI offers a reduction of 157 μSv/h in its heaviest configuration. Among these relevant reductions, the gamma shield inside the EP PI offers the best shielding option, as it reduces gamma cross-talk from TCP-LP and also protects EP PI from Port Duct and EP bellows activation, while it does not interfere with TCP performance.

Prediction of dose and field mapping around a shielded plutonium fuel fabrication glovebox

International Nuclear Information System (INIS)

Strode, J.N.; Soldat, K.L.; Brackenbush, L.W.

1984-01-01

Westinghouse Hanford Company, as the Department of Energy's (DOE) prime contractor for the operation of the Hanford Engineering Development Laboratory (HEDL), is responsible for the development of the Secure Automated Fabrication (SAF) Line which is to be installed in the recently constructed Fuels and Materials Examination Facility (FMEF). The SAF Line will fabricate mixed-oxide (MOX) fuel pins for the Fast Flux Test Facility (FFTF) at an annual throughput rate of six (6) metric tons (MT) of MOX. The SAF Line will also demonstrate the automated manufacture of fuel pins on a production-scale. This paper describes some of the techniques used to reduce personnel exposure on the SAF Line, as well as the prediction and field mapping of doses from a shielded fuel fabrication glovebox. Tables are also presented from which exposure rate estimates can be made for plutonium recovered from fuels having different isotopic compositions as a result of varied burnup

Novel hybrid Monte Carlo/deterministic technique for shutdown dose rate analyses of fusion energy systems

International Nuclear Information System (INIS)

Ibrahim, Ahmad M.; Peplow, Douglas E.; Peterson, Joshua L.; Grove, Robert E.

2014-01-01

Highlights: •Develop the novel Multi-Step CADIS (MS-CADIS) hybrid Monte Carlo/deterministic method for multi-step shielding analyses. •Accurately calculate shutdown dose rates using full-scale Monte Carlo models of fusion energy systems. •Demonstrate the dramatic efficiency improvement of the MS-CADIS method for the rigorous two step calculations of the shutdown dose rate in fusion reactors. -- Abstract: The rigorous 2-step (R2S) computational system uses three-dimensional Monte Carlo transport simulations to calculate the shutdown dose rate (SDDR) in fusion reactors. Accurate full-scale R2S calculations are impractical in fusion reactors because they require calculating space- and energy-dependent neutron fluxes everywhere inside the reactor. The use of global Monte Carlo variance reduction techniques was suggested for accelerating the R2S neutron transport calculation. However, the prohibitive computational costs of these approaches, which increase with the problem size and amount of shielding materials, inhibit their ability to accurately predict the SDDR in fusion energy systems using full-scale modeling of an entire fusion plant. This paper describes a novel hybrid Monte Carlo/deterministic methodology that uses the Consistent Adjoint Driven Importance Sampling (CADIS) method but focuses on multi-step shielding calculations. The Multi-Step CADIS (MS-CADIS) methodology speeds up the R2S neutron Monte Carlo calculation using an importance function that represents the neutron importance to the final SDDR. Using a simplified example, preliminary results showed that the use of MS-CADIS enhanced the efficiency of the neutron Monte Carlo simulation of an SDDR calculation by a factor of 550 compared to standard global variance reduction techniques, and that the efficiency enhancement compared to analog Monte Carlo is higher than a factor of 10,000

News from the Library: Facilitating access to a program for radiation shielding - the Library can help

CERN Multimedia

CERN Library

2013-01-01

MicroShield® is a comprehensive photon/gamma ray shielding and dose assessment programme. It is widely used for designing shields, estimating source strength from radiation measurements, minimising exposure to people, and teaching shielding principles.   Integrated tools allow the graphing of results, material and source file creation, source inference with decay (dose-to-Bq calculations accounting for decay and daughter buildup), the projection of exposure rate versus time as a result of decay, access to material and nuclide data, and decay heat calculations. The latest version is able to export results using Microsoft Office (formatted and colour-coded for readability). Sixteen geometries accommodate offset dose points and as many as ten standard shields plus source self-shielding and cylinder cladding are available. The library data (radionuclides, attenuation, build-up and dose conversion) reflect standard data from ICRP 38 and 107* as well as ANSI/ANS standards and RSICC publicat...

SU-F-I-71: Fetal Protection During Fluoroscopy: To Shield Or Not to Shield?

International Nuclear Information System (INIS)

Joshi, S; Vanderhoek, M

2016-01-01

Purpose: Lead aprons are routinely used to shield the fetus from radiation during fluoroscopically guided interventions (FGI) involving pregnant patients. When placed in the primary beam, lead aprons often reduce image quality and increase fluoroscopic radiation output, which can adversely affect fetal dose. The purpose of this work is to identify an effective and practical method to reduce fetal dose without affecting image quality. Methods: A pregnant patient equivalent abdominal phantom is set on the table along with an image quality test object (CIRS model 903) representing patient anatomy of interest. An ion chamber is positioned at the x-ray beam entrance to the phantom, which is used to estimate the relative fetal dose. For three protective methods, image quality and fetal dose measurements are compared to baseline (no protection):1. Lead apron shielding the entire abdomen; 2. Lead apron shielding part of the abdomen, including the fetus; 3. Narrow collimation such that fetus is excluded from the primary beam. Results: With lead shielding the entire abdomen, the dose is reduced by 80% relative to baseline along with a drastic deterioration of image quality. With lead shielding only the fetus, the dose is reduced by 65% along with complete preservation of image quality, since the image quality test object is not shielded. However, narrow collimation results in 90% dose reduction and a slight improvement of image quality relative to baseline. Conclusion: The use of narrow collimation to protect the fetus during FGI is a simple and highly effective method that simultaneously reduces fetal dose and maintains sufficient image quality. Lead aprons are not as effective at fetal dose reduction, and if placed improperly, they can severely degrade image quality. Future work aims to investigate a wider variety of fluoroscopy systems to confirm these results across many different system geometries.

SU-F-I-71: Fetal Protection During Fluoroscopy: To Shield Or Not to Shield?

Energy Technology Data Exchange (ETDEWEB)

Joshi, S; Vanderhoek, M [Henry Ford Health System, Detroit, MI (United States)

2016-06-15

Purpose: Lead aprons are routinely used to shield the fetus from radiation during fluoroscopically guided interventions (FGI) involving pregnant patients. When placed in the primary beam, lead aprons often reduce image quality and increase fluoroscopic radiation output, which can adversely affect fetal dose. The purpose of this work is to identify an effective and practical method to reduce fetal dose without affecting image quality. Methods: A pregnant patient equivalent abdominal phantom is set on the table along with an image quality test object (CIRS model 903) representing patient anatomy of interest. An ion chamber is positioned at the x-ray beam entrance to the phantom, which is used to estimate the relative fetal dose. For three protective methods, image quality and fetal dose measurements are compared to baseline (no protection):1. Lead apron shielding the entire abdomen; 2. Lead apron shielding part of the abdomen, including the fetus; 3. Narrow collimation such that fetus is excluded from the primary beam. Results: With lead shielding the entire abdomen, the dose is reduced by 80% relative to baseline along with a drastic deterioration of image quality. With lead shielding only the fetus, the dose is reduced by 65% along with complete preservation of image quality, since the image quality test object is not shielded. However, narrow collimation results in 90% dose reduction and a slight improvement of image quality relative to baseline. Conclusion: The use of narrow collimation to protect the fetus during FGI is a simple and highly effective method that simultaneously reduces fetal dose and maintains sufficient image quality. Lead aprons are not as effective at fetal dose reduction, and if placed improperly, they can severely degrade image quality. Future work aims to investigate a wider variety of fluoroscopy systems to confirm these results across many different system geometries.

Two-dimensional radiation shielding optimization analysis of spent fuel transport container

International Nuclear Information System (INIS)

Tian Yingnan; Chen Yixue; Yang Shouhai

2013-01-01

The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

Dose rate constants for new dose quantities

International Nuclear Information System (INIS)

Tschurlovits, M.; Daverda, G.; Leitner, A.

1992-01-01

Conceptual changes and new quantities made is necessary to reassess dose rate quantities. Calculations of the dose rate constant were done for air kerma, ambient dose equivalent and directional dose equivalent. The number of radionuclides is more than 200. The threshold energy is selected as 20 keV for the dose equivalent constants. The dose rate constant for the photon equivalent dose as used mainly in German speaking countries as a temporary quantity is also included. (Author)

Investigation of 1-cm dose equivalent for photons behind shielding materials

International Nuclear Information System (INIS)

Hirayama, Hideo; Tanaka, Shun-ichi

1991-03-01

The ambient dose equivalent at 1-cm depth, assumed equivalent to the 1-cm dose equivalent in practical dose estimations behind shielding slabs of water, concrete, iron or lead for normally incident photons having various energies was calculated by using conversion factors for a slab phantom. It was compared with the 1-cm depth dose calculated with the Monte Carlo code EGS4. It was concluded from this comparison that the ambient dose equivalent calculated by using the conversion factors for the ICRU sphere could be used for the evaluation of the 1-cm dose equivalent for the sphere phantom within 20% errors. Average and practical conversion factors are defined as the conversion factors from exposure to ambient dose equivalent in a finite slab or an infinite one, respectively. They were calculated with EGS4 and the discrete ordinates code PALLAS. The exposure calculated with simple estimation procedures such as point kernel methods can be easily converted to ambient dose equivalent by using these conversion factors. The maximum value between 1 and 30 mfp can be adopted as the conversion factor which depends only on material and incident photon energy. This gives the ambient dose equivalent on the safe side. 13 refs., 7 figs., 2 tabs

Induced radioactivity in the forward shielding and semiconductor tracker of the ATLAS detector.

Science.gov (United States)

Bĕdajánek, I; Linhart, V; Stekl, I; Pospísil, S; Kolros, A; Kovalenko, V

2005-01-01

The radioactivity induced in the forward shielding, copper collimator and semiconductor tracker modules of the ATLAS detector has been studied. The ATLAS detector is a long-term experiment which, during operation, will require to have service and access to all of its parts and components. The radioactivity induced in the forward shielding was calculated by Monte Carlo methods based on GEANT3 software tool. The results show that the equivalent dose rates on the outer surface of the forward shielding are very low (at most 0.038 microSv h(-1)). On the other hand, the equivalent dose rates are significantly higher on the inner surface of the forward shielding (up to 661 microSv h(-1)) and, especially, at the copper collimator close to the beampipe (up to 60 mSv h(-1)). The radioactivity induced in the semiconductor tracker modules was studied experimentally. The module was activated by neutrons in a training nuclear reactor and the delayed gamma ray spectra were measured. From these measurements, the equivalent dose rate on the surface of the semiconductor tracker module was estimated to be LHC) operation and 10 d of cooling.

Double-layer neutron shield design as neutron shielding application

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

Soil biological shield exposed to high energy neutrons; Zemlja kao bioloski stit od neutrona visokih energija

Energy Technology Data Exchange (ETDEWEB)

Simovic, R; Marinkovic, N [Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1993-04-15

Shielding efficiency of soil biological shield exposed to high energy neutrons was investigated. Dose rate equivalents for neutrons, secondary gamma and gamma radiation were computed on the surface of soil slabs having different thicknesses. Yields of primary and secondary nuclear radiation in the total dose were evaluated. Influence of the incident neutron spectrum, water content and chemical composition of the material on its shielding efficiency was examined. It was found that the soil density and the water content determine the quality of biological shield, the influence of other factors being less important. Comparison of shielding efficiencies for soil with sand, brick and ordinary concrete shields was done.

Measurement of dose rates and Monte Carlo analysis of neutrons in a spent-fuel shipping vessel

International Nuclear Information System (INIS)

Ueki, K.; Namito, Y.; Fuse, T.

1986-01-01

On-board experiments were carried out in a spent-fuel shipping vessel, the Pacific Swan, in which 13 casks of TN-12A and Excellox 3 were loaded in five holds, and neutron and gamma-ray dose rates were measured on the hatch covers of the holds. Before shipping those casks, dose rates were also measured on the cask surfaces, one by one, to eliminate radiation from other casks. The Monte Carlo coupling technique was employed successfully to analyze the measured neutron dose rate distributions in the spent-fuel shipping vessel. Through this study, the Monte Carlo coupling code system, MORSE-CG/CASK-VESSEL, on which the MORSE-CG code was based, was established. The agreement between the measured and the calculated neutron dose rates on the TN-12A cask surface was quite satisfactory. The calculated neutron dose rates agreed with the measured values within a factor of 1.5 on the hold 3 hatch cover and within a factor of 2 on the hold 5 hatch cover in which the concrete shield was fixed in the Pacific Swan

Post Three Mile Island shielding review - a case history

International Nuclear Information System (INIS)

Isakari, H.H.; Shaw, H.C.

1983-01-01

The radiation shielding review of the Diablo Canyon Nuclear Power Plant was performed in accordance with the requirement of the Three Mile Island Action Plan. The review covered plant shielding and environmental qualification of equipment for spaces and systems which may be used in post-accident operations. Radiation doses during postulated loss-of-coolant accident and high-energy-line-break accident were calculated for equipment located both inside and outside the containment. Vital areas, those requiring post-accident access and occupancy, were identified and their associated dose rates and integrated doses were calculated. It was found that all four of the vital areas (Control Room, Technical Support Center, Switchgear Room, and Emergency Sampling Compartment) are shielded from external sources of radiation sufficiently to permit personnel access and occupancy that would not be unduly limited by the radiation environment caused by the postulated accidents. (author)

The effect of high-scatter shielding geometries in validating the dose inferred by N-VisageTM - 16130

International Nuclear Information System (INIS)

Adams, Jamie C.; Joyce, Malcolm J.; Mellor, Matthew

2009-01-01

The aim of this paper is to further validate the physical capability of N-Visage TM under more challenging shielding geometries, when the number of mean free paths is greater than one. N-Visage TM is a recently established technique developed at REACT Engineering Ltd. The software locates radionuclide sources and contours radiation magnitude. The N-Visage TM software uses a geometric computer model combined with measured spectra. The software is able to estimate source locations through shielding materials by using mass attenuation coefficients to calculate the number of unscattered gamma photons arriving at the detector, and build-up factors to estimate scatter contribution to dose rate. The experiments described in this paper were carried out in a high-scatter environment using cobalt-60 and cesium- 137 sources, these two sources are the primary sources of radiological contamination found in the nuclear industry. It is hoped that this will further assist in the identification, characterisation and removal of buried radiologically contaminated waste. (authors)

Shield design of concrete wall between decay tank room and primary pump room in TRIGA facility

International Nuclear Information System (INIS)

Khan, M. J. H.; Rahman, M.; Haque, A.; Zulquarnain, A.; Ahmed, F. U.; Bhuiyan, S. I.

2007-01-01

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II research reactor facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit (10 μSv/hr). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete

Optimal selection for shielding materials by fuzzy linear programming

International Nuclear Information System (INIS)

Kanai, Y.; Miura, N.; Sugasawa, S.

1996-01-01

An application of fuzzy linear programming methods to optimization of a radiation shield is presented. The main purpose of the present study is the choice of materials and the search of the ratio of mixture-component as the first stage of the methodology on optimum shielding design according to individual requirements of nuclear reactor, reprocessing facility, shipping cask installing spent fuel, ect. The characteristic values for the shield optimization may be considered their cost, spatial space, weight and some shielding qualities such as activation rate and total dose rate for neutron and gamma ray (includes secondary gamma ray). This new approach can reduce huge combination calculations for conventional two-valued logic approaches to representative single shielding calculation by group-wised optimization parameters determined in advance. Using the fuzzy linear programming method, possibilities for reducing radiation effects attainable in optimal compositions hydrated, lead- and boron-contained materials are investigated

Shielding analysis of the IEM cell offset adapter plate

International Nuclear Information System (INIS)

Simons, R.L.

1995-01-01

The adapter plate for the Interim Examination and Maintenance (IEM) cell ten foot ceiling valve was modified so that the penetration through the valve is offset to the north side of the steel plate. The modifications required that the shielding effectiveness be evaluated for several operating conditions. The highest gamma ray dose rate (51 mrem/hr) occurs when a Core Component Container (CCC) with six high burn-up driver fuel assemblies is transferred into or out of Solid Waste Cask (SWC). The neutron dose rate at the same source location is 2.5 mrem/hr. The total dose rate during the transfer is less than the 200 mrem/hr limit. If the ten foot ceiling valve is closed, the dose rate with twelve DFA in the cell will be less than 0.1 mrem/hr. However, with the ceiling valve open the dose rate will be as high as 12 mrem/hr. The latter condition will require controlled access to the area around the offset adapter plate when the ceiling valve is open. It was found that gaps in the shield block around the SWC floor valve will allow contact dose rates as high as 350 mrem/hr during the transfer of a fully loaded CCC. Although this situation does not pertain to the offset adapter plate, it will require controlled access around the SWC valve during the transfer of a fully loaded CCC

Shielding of a neutron irradiator with {sup 241}Am-Be source

Energy Technology Data Exchange (ETDEWEB)

Oliveira, K.A.M. de; Crispim, V.R.; Silva, A.X., E-mail: koliveira@con.ufrj.b, E-mail: verginia@con.ufrj.b, E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear; Fonseca, E.S., E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2011-07-01

The equivalent dose rates at 1.0 cm from the outer surface of the shielding of a neutron irradiation system that uses {sup 241}Am-Be source with activity of 185 GBq (5 Ci) were determined. A theoretical-experimental approach including case studies, through computer simulations with MCNP code was employed to calculate the best shielding thickness. Following the construction of the neutron irradiator, dose measurements were conducted in order to validate data obtained from simulation. The neutron irradiator shielding was designed in such a way to allow transport of the neutron radiography system for in loco inspections ensuring workers' radiologic safety. (author)

Investigations on construction material and construction concepts in order to obtain dose-reducing effects in the dismantling of the biological shield of a 1300 MWe-PWR

International Nuclear Information System (INIS)

Bittner, A.; Jungwirth, D.; Knell, M.; Schnitzler, L.

1984-04-01

Numerical values of neutron fluxes, activations, dose rates etc. as a function of characteristic values of materials required for optimization purposes to reduce the radiation effect of the biological shield of a PWR are not available. Design concepts are presented for biological shields of PWRs made of concrete with respect to both the most suitable application of materials and the design principles aiming at reduced radiation exposure as compared to present designs during entering, waste disposal and ultimate storage. To evaluate the present-state design the above values have been calculated. Suggested alternative designs are biological shields with selective material application, built from precast elements with or without boron carbide layer arranged in front of it. (orig./HP) [de

Pulsed dose rate brachytherapy – is it the right way?

Directory of Open Access Journals (Sweden)

Janusz Skowronek

2010-10-01

Full Text Available Pulsed dose rate (PDR-BT treatment is a brachytherapy modality that combines physical advantages of high-doserate (HDR-BT technology (isodose optimization, radiation safety with the radiobiological advantages of low-dose-rate (LDR-BT brachytherapy. Pulsed brachytherapy consists of using stronger radiation source than for LDR-BT and producing series of short exposures of 10 to 30 minutes in every hour to approximately the same total dose in the sameoverall time as with the LDR-BT. Modern afterloading equipment offers certain advantages over interstitial or intracavitaryinsertion of separate needles, tubes, seeds or wires. Isodose volumes in tissues can be created flexibly by a combinationof careful placement of the catheter and the adjustment of the dwell times of the computerized stepping source.Automatic removal of the radiation sources into a shielded safe eliminates radiation exposures to staff and visitors.Radiation exposure is also eliminated to the staff who formerly loaded and unloaded multiplicity of radioactive sources into the catheters, ovoids, tubes etc. This review based on summarized clinical investigations, analyses the feasibility and the background to introduce this brachytherapy technique and chosen clinical applications of PDR-BT.

A novel removable shield attached to C-arm units against scattered X-rays from a patient's side

Energy Technology Data Exchange (ETDEWEB)

Mori, Hiroshige [Hokkaido Social Insurance Hospital, Department of Radiology, Sapporo, Hokkaido (Japan); Kanazawa University, Department of Quantum Medical Technology, Division of Health Sciences, Graduate School of Medical Science, Kanazawa, Ishikawa (Japan); Koshida, Kichiro; Matsubara, Kosuke [Kanazawa University, School of Health Sciences, College of Medical, Pharmaceutical and Health Sciences, Kanazawa, Ishikawa (Japan); Ishigamori, Osamu [Hokkaido Social Insurance Hospital, Department of Radiology, Sapporo, Hokkaido (Japan)

2014-08-15

We invented a drape-like shield against scattered X-rays that can safely come into contact with medical equipment or people during fluoroscopically guided procedures. The shield can be easily removed from a C-arm unit using one hand. We evaluated the use of the novel removable shield during the endoscopic retrograde cholangiopancreatography (ERCP) procedure. We measured the dose rate of scattered X-rays around endoscopists with and without this removable shield and surveyed the occupational doses to the ERCP staff. We also examined the endurance of the shield. The removable shield reduced the dose rate of scattered X-rays to one-tenth and reduced the monthly dose to an endoscopist by at least two-fifths. For 2.5 years, there was no damage to the shield and no loosening of the seam. The bonding of the hook-and-loop fasteners did not weaken, although the powerful double-sided tapes made especially for plastic did. The removable shield can reduce radiation exposure to the ERCP staff and may contribute to reducing the exposure to the eye lenses of operators. It would also be possible to expand its use to other fluoroscopically guided procedures besides ERCP because it is a light, simple, and useful device. (orig.)

Design optimization of radiation shielding structure for lead slowing-down spectrometer system

Energy Technology Data Exchange (ETDEWEB)

Kim, Jeong Dong; Ahn, Sang Joon; Lee, Yong Deok [Nonproliferation System Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Chang Je [Dept. of Nuclear Engineering, Sejong University, Seoul (Korea, Republic of)

2015-04-15

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux (>101{sup 2n}/cm{sup 2}·s) neutron source comprised of a high-energy (30 MeV)/high-current (∼2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h), a few shielding materials [high-density polyethylene (HDPE)–Borax, B{sub 4}C, and Li{sub 2}CO{sub 3}] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in

Design optimization of radiation shielding structure for lead slowing-down spectrometer system

International Nuclear Information System (INIS)

Kim, Jeong Dong; Ahn, Sang Joon; Lee, Yong Deok; Park, Chang Je

2015-01-01

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux (>101 2n /cm 2 ·s) neutron source comprised of a high-energy (30 MeV)/high-current (∼2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h), a few shielding materials [high-density polyethylene (HDPE)–Borax, B 4 C, and Li 2 CO 3 ] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

Design optimization of radiation shielding structure for lead slowing-down spectrometer system

Directory of Open Access Journals (Sweden)

Jeong Dong Kim

2015-04-01

Full Text Available A lead slowing-down spectrometer (LSDS system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea is planned to utilize a high-flux (>1012 n/cm2·s neutron source comprised of a high-energy (30 MeV/high-current (∼2 A electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h, a few shielding materials [high-density polyethylene (HDPE–Borax, B4C, and Li2CO3] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near

SHIELDOSE, Doses from Electron and Proton Irradiation in Space Vehicle Al Shields

International Nuclear Information System (INIS)

Seltzer, Stephen

1986-01-01

1 - Description of problem or function: The ability to predict absorbed dose within a spacecraft due to a specified radiation environment is important for design and planning considerations pertaining to the reliability of electronic components and to the radiological safety of on-board personnel. This computer code SHIELDOSE evaluates the absorbed dose as a function of depth in aluminum shielding material of spacecraft, given the electron and proton fluences encountered in orbit. 2 - Method of solution: It makes use of pre-calculated, monoenergetic depth-dose data for an isotropic, broad-beam fluence of radiation incident on uniform aluminum plane media. Such data are particularly suitable for routine dose predictions in situations where the geometrical and compositional complexities of the spacecraft are not known. Furthermore, restricting our consideration to these rather simple geometries has allowed for the development of accurate electron and electron-Bremsstrahlung data sets based on detailed transport calculations rather than on more approximate methods. The present version of SHIELDOSE calculates, for arbitrary proton and electron incident spectra, the dose absorbed in small volumes of the detector materials Al, H 2 O (tissue-equivalent detector), Si and SiO 2 , in the following aluminum shield geometries: (1) in a semi- infinite plane medium, as a function of depth; (2) at the transmission surface of a plane slab, as a function of slab thickness; and (3) at the center of a solid sphere, as a function of sphere radius. 3 - Restrictions on the complexity of the problem: - No. of depth Z for which dose calculation is desired (IMAX) ≤50; - No. of prints used in the numerical evaluation of the integral over the incident proton spectrum (NPTSP) ≤301; - No. of points used in the numerical evaluation of the internal over the incident electron spectrum (NPTSE) ≤101; - No. of energy for which the solar-flare-proton spectrum is read in (JSMAX), incident

Dose rate correction in medium dose rate brachytherapy for carcinoma cervix

International Nuclear Information System (INIS)

Patel, F.D.; Negi, P.S.; Sharma, S.C.; Kapoor, R.; Singh, D.P.; Ghoshal, S.

1998-01-01

Purpose: To establish the magnitude of brachytherapy dose reduction required for stage IIB and III carcinoma cervix patients treated by external radiation and medium dose rate (MDR) brachytherapy at a dose rate of 220±10 cGy/h at point A.Materials and methods: In study-I, at the time of MDR brachytherapy application at a dose rate of 220±10 cGy/h at point A, patients received either 3060 cGy, a 12.5% dose reduction (MDR-12.5), or 2450 cGy, a 30% dose reduction (MDR-30), to point A and they were compared to a group of previously treated LDR patients who received 3500 cGy to point A at a dose rate of 55-65 cGy/h. Study-II was a prospective randomized trial and patients received either 2450 cGy, a 30% dose reduction (MDR-II (30)) or 2800 cGy, a 20% dose reduction (MDR-II (20)), at point A. Patients were evaluated for local control of disease and morbidity. Results: In study-I the 5-year actuarial local control rate in the MDR-30 and MDR-12.5 groups was 71.7±10% and 70.5±10%, respectively, compared to 63.4±10% in the LDR group. However, the actuarial morbidity (all grades) in the MDR-12.5 group was 58.5±14% as against 34.9±9% in the LDR group (P 3 developed complication as against 62.5% of those receiving a rectal BED of (140 3 (χ 2 =46.43; P<0.001). Conclusion: We suggest that at a dose rate of 220±10 cGy/h at point A the brachytherapy dose reduction factor should be around 30%, as suggested by radiobiological data, to keep the morbidity as low as possible without compromising the local control rates. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)

Re-evaluation of the shielding adequacy of the brachytherapy treatment room at Korle-Bu teaching hospital, Ghana

International Nuclear Information System (INIS)

Arwui, C. C.

2009-06-01

Staff and the general public's safety during the operation of the 137 Cs brachytherapy unit at the Korle Bu teaching hospital depends on the adequacy of the shielding of the facility. Shielding design of the brachytherapy unit at the hospital was based on postulated workload and postulated occupancy factors to critical locations at the facility where the public and staff may occupy. This facility has been in existence for the past twelve (12) years and has accumulated operational workload data which differs from the postulated one. A study was carried out to re-evaluate the integrity of the biological shielding of the 137 Cs brachytherapy unit. This study analyzed the accumulated workload data and used the information to perform shielding calculations to verify the adequacy of the biological shielding thicknesses to provide sufficient protection of staff and the public. Dose rate calculations were verified by measurements with calibrated dose rate meters. This provided the basis for determining the current state of protection and safety for staff and the general public. The results show that despite the variation in actual and postulated workloads, the dose rates were below the reference values of 0.5μSv/h for public areas and 7.5μSv/h for controlled areas. It was confirmed that the present shielding thickness of 535 mm can accommodate a high dose rate (HDR) 192 Ir source with activity in the range 370 - 570 GBq with an operational workload of 30 patients per week and an average treatment time of 10 minutes.

Comparison between steel and lead shieldings for radiotherapy rooms regarding neutron doses to patients

International Nuclear Information System (INIS)

Silva, M.G.; Rebello, W.F.; Andrade, E.R.; Medeiros, M.P.C.; Mendes, R.M.S.; Braga, K.L.; Gomes, R.G.

2015-01-01

The NCRP Report No. 151, Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities, considers, in shielding calculations for radiotherapy rooms, the use of lead and/or steel to be applied on bunker walls. The NCRP Report calculations were performed foreseeing a better protection of people outside the radiotherapy room. However, contribution of lead and steel to patient dose should be taken into account for radioprotection purposes. This work presents calculations performed by MCNPX code in analyzing the Ambient Dose Equivalent due to neutron, H *(10) n , within a radiotherapy room, in the patients area, considering the use of additional shielding of 1 TVL of lead or 1 TVL of steel, positioned at the inner faces of walls and ceiling of a bunker. The head of the linear accelerator Varian 2100/2300 C/D was modeled working at 18MeV, with 5 x 5 cm 2 , 10 x 10 cm 2 , 20 x 20 cm 2 , 30 x 30 cm 2 and 40 x 40 cm 2 openings for jaws and MLC and operating in eight gantry's angles. This study shows that the use of lead generates an average value of H *(10) n at patients area, 8.02% higher than the expected when using steel. Further studies should be performed based on experimental data for comparison with those from MCNPX simulation. (author)

Radiation shielding analysis of medical cyclotron at Radiation Medicine Centre, Parel

International Nuclear Information System (INIS)

Gathibandhe, M.V.; Agrawal, R.A.; Utge, C.G.

2003-01-01

Full text: PET (Positron Emission Tomography) is a diagnostic method to obtain 3-D functional images of the distribution of radio-nuclides introduced in the human body as tracers for specific biological processes. Tracers are produced by bombardment of different target nuclides by protons and deuterons of high energy produced in the cyclotron. A Wipro-GE medical cyclotron was installed in the basement of RMC, Parel. Shielding around the cyclotron is provided in the form of borated concrete walls of required thickness to limit dose rates to design values as per AERB criteria. The roof of the cyclotron room is made of heavy concrete. Entry in to the room is through a maze. Shielding analysis for the cyclotron room has been carried out using computer code ANISN. The maze has been analyzed using code MCNP. Based on the analysis carried out additional shielding was recommended to meet the design requirements. The paper discusses the shielding analysis carried out for the cyclotron room and the maze. Dose rate estimated at various locations are highlighted

Biological shield around the neutral beam injector ducts in the ITER conceptual design

International Nuclear Information System (INIS)

Maki, Koichi; Takatsu, Hideyuki; Satoh, Satoshi; Seki, Yasushi

1994-01-01

There are gaps between the toroidal field coils and neutral beam injector (NBI) duct wall for the thermal insulator in tokamak reactors such as ITER (International Thermonuclear Experimental Reactor). Neutrons stream through the duct, and some of them penetrate the wall and stream through the gaps. These neutrons activate the materials composing the duct wall, toroidal field coil (TFC) case and cryostat wall surfaces. The dose rate is enhanced just outside the cryostat around the ducts in the reactor room after reactor operation by activation. We investigated the gamma-ray dose rate just outside the cryostat after shutdown due to gamma-rays from activity induced by the neutrons streaming through the gaps. By evaluating the difference between the dose rate in models with and without gaps, we decided whether the thickness of the cryostat as biological shielding is sufficient or not. From these investigations, we recommend a cryostat design suitable for radiation shielding. Dose rates after shutdown at a point just outside the cryostat around the NBI ducts in the model with gaps are two orders larger than those without gaps. The value at this point is approximately 400 mrem h -1 (4 mSv h -1 ), which is two orders larger than the design value for workers to enter the reactor room. In order to reduce the dose rate after shutdown, a method of providing the shielding function of the cryostat is suggested. ((orig.))

Determination of the absorbed dose rate to a person exposed to a spent source of {sup 60}Co for radiotherapy; Determinacion de la rapidez de dosis absorbida a una persona expuesta a una fuente gastada de {sup 60}Co para radioterapia

Energy Technology Data Exchange (ETDEWEB)

Garcia M, T.; Angeles C, A.; Benitez, J. A.; Ruiz C, M. A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Departamento de Proteccion Radiologica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

In this work the analysis of absorbed dose rate to a person in made due to the exposure to a spent source of {sup 60}Co of radiotherapy, which has been removed from its shielding clandestinely to sell the shielding as scrap. During the removal of the source of their shielding the people were necessarily exposed to the field of gamma radiation. The activity of the source is considered to be 2595 Ci at the exposure time and to determine the rate of absorbed dose to different organs and the velocity of effective absorbed dose to which the person (s) who manipulated the source of {sup 60}Co were considered three plausible scenarios of manipulation of the source , through modeling with MCNP5. For the execution of the scenarios and the determination of the absorbed doses, two different phantoms are considered. The results obtained for each scenario show that the dose rates to which the people who manipulated the source without the shielding were exposed are extremely high, and in short time the lethal dose is reached. (Author)

Impact of a flattening filter free linear accelerator on structural shielding design

International Nuclear Information System (INIS)

Jank, Julia; Kragl, Gabriele; Georg, Dietmar; Medical University of Vienna

2014-01-01

Purpose: The present study aimed to assess the effects of a flattening filter free medical accelerator on structural shielding demands of a treatment vault of a medical linear accelerator. We tried to answer the question, to what extent the required thickness of the shielding barriers can be reduced if instead of the standard flattened photon beams unflattened ones are used. Material and Methods: We chose both an experimental as well as a theoretical approach. On the one hand we measured photon dose rates at protected places outside the treatment room and compared the obtained results for flattened and unflattened beams. On the other hand we complied with international guidelines for adequate treatment vault design and calculated the shielding barriers according to the therein given specifications. Measurements were performed with an Elekta Precise trademark linac providing nominal photon energies of 6 and 10 MV. This machine underwent already earlier some modifications in order to be able to operate both with and without a flattening filter. Photon dose rates were measured with a LB133-1 dose rate meter manufactured by Berthold. To calculate the thickness of shielding barriers we referred to the Austrian standard OeNORM S 5216 and to the US American NCRP Report No. 151. Results: We determined a substantial photon dose rate reduction for all measurement points and photon energies. For unflattened 6 MV beams a reduction factor ranging from 1.4 to 1.8 was identified. The corresponding values for unflattened 10 MV beams were 2.1 and 3.2. The performed shielding calculations indicated the same tendency: For all relevant radiation components we found a reduction in shielding thickness when unflattened beams were used. The required thickness of primary barriers was reduced up to 8.0%, the thickness of secondary barriers up to 11.4%, respectively. Conclusions: For an adequate dimensioning of treatment vault shielding barriers it is by no means irrelevant if the

Impact of a flattening filter free linear accelerator on structural shielding design.

Science.gov (United States)

Jank, Julia; Kragl, Gabriele; Georg, Dietmar

2014-03-01

The present study aimed to assess the effects of a flattening filter free medical accelerator on structural shielding demands of a treatment vault of a medical linear accelerator. We tried to answer the question, to what extent the required thickness of the shielding barriers can be reduced if instead of the standard flattened photon beams unflattened ones are used. We chose both an experimental as well as a theoretical approach. On the one hand we measured photon dose rates at protected places outside the treatment room and compared the obtained results for flattened and unflattened beams. On the other hand we complied with international guidelines for adequate treatment vault design and calculated the shielding barriers according to the therein given specifications. Measurements were performed with an Elekta Precise™ linac providing nominal photon energies of 6 and 10 MV. This machine underwent already earlier some modifications in order to be able to operate both with and without a flattening filter. Photon dose rates were measured with a LB133-1 dose rate meter manufactured by Berthold. To calculate the thickness of shielding barriers we referred to the Austrian standard ÖNORM S 5216 and to the US American NCRP Report No. 151. We determined a substantial photon dose rate reduction for all measurement points and photon energies. For unflattened 6 MV beams a reduction factor ranging from 1.4 to 1.8 was identified. The corresponding values for unflattened 10 MV beams were 2.1 and 3.2. The performed shielding calculations indicated the same tendency: For all relevant radiation components we found a reduction in shielding thickness when unflattened beams were used. The required thickness of primary barriers was reduced up to 8.0%, the thickness of secondary barriers up to 11.4%, respectively. For an adequate dimensioning of treatment vault shielding barriers it is by no means irrelevant if the accommodated linac operates with or without a flattening filter. The

Impact of a flattening filter free linear accelerator on structural shielding design

Energy Technology Data Exchange (ETDEWEB)

Jank, Julia [Klinikum - Klagenfurt am Woerthersee (Austria). Inst. fuer Strahlentherapie und Radioonkologie; Kragl, Gabriele [Medical University of Vienna/AKH Vienna (Austria). Div. Medical Radiation Physics; Georg, Dietmar [Medical University of Vienna/AKH Vienna (Austria). Div. Medical Radiation Physics; Medical University of Vienna (Austria). Christian Doppler Lab. for Medical Radiation Research for Radiation Oncology

2014-04-01

Purpose: The present study aimed to assess the effects of a flattening filter free medical accelerator on structural shielding demands of a treatment vault of a medical linear accelerator. We tried to answer the question, to what extent the required thickness of the shielding barriers can be reduced if instead of the standard flattened photon beams unflattened ones are used. Material and Methods: We chose both an experimental as well as a theoretical approach. On the one hand we measured photon dose rates at protected places outside the treatment room and compared the obtained results for flattened and unflattened beams. On the other hand we complied with international guidelines for adequate treatment vault design and calculated the shielding barriers according to the therein given specifications. Measurements were performed with an Elekta Precise trademark linac providing nominal photon energies of 6 and 10 MV. This machine underwent already earlier some modifications in order to be able to operate both with and without a flattening filter. Photon dose rates were measured with a LB133-1 dose rate meter manufactured by Berthold. To calculate the thickness of shielding barriers we referred to the Austrian standard OeNORM S 5216 and to the US American NCRP Report No. 151. Results: We determined a substantial photon dose rate reduction for all measurement points and photon energies. For unflattened 6 MV beams a reduction factor ranging from 1.4 to 1.8 was identified. The corresponding values for unflattened 10 MV beams were 2.1 and 3.2. The performed shielding calculations indicated the same tendency: For all relevant radiation components we found a reduction in shielding thickness when unflattened beams were used. The required thickness of primary barriers was reduced up to 8.0%, the thickness of secondary barriers up to 11.4%, respectively. Conclusions: For an adequate dimensioning of treatment vault shielding barriers it is by no means irrelevant if the

MO-F-CAMPUS-I-04: Patient Eye-Lens Dose Reduction in Routine Brain CT Examinations Using Organ-Based Tube Current Modulation and In-Plane Bismuth Shielding

International Nuclear Information System (INIS)

Tsai, Hui-Yu; Liao, Ying-Lan; Lai, Nan-Ku; Chen, Tou-Rong; Chen, Jun-Rong

2015-01-01

Purpose: The purpose of this study is to assess eye-lens dose for patients who underwent brain CT examinations using two dose reduction Methods: organ-based tube current modulation (OBTCM) and in-plane bismuth shielding method. Methods: This study received institutional review board approval; written informed consent to participate was obtained from all patients. Ninety patients who underwent the routine brain CT examination were randomly assigned to three groups, ie. routine, OBTCM, and bismuth shield. The OBTCM technique reduced the tube current when the X-ray tube rotates in front of patients’ eye-lens region. The patients in the bismuth shield group were covered one-ply bismuth shield in the eyes’ region. Eye-lens doses were measured using TLD-100H chips and the total effective doses were calculated using CT-Expo according to the CT scanning parameters. The surface doses for patients at off-center positions were assessed to evaluate the off-centering effect. Results: Phantom measurements indicates that OBTCM technique could reduced by 26% to 28% of the surface dose to the eye lens, and increased by 25% of the surface dose at the opposed incident direction at the angle of 180°. Patients’ eye-lens doses were reduced 16.9% and 30.5% dose of bismuth shield scan and OBTCM scan, respectively compared to the routine scan. The eye-lens doses were apparently increased when the table position was lower than isocenter. Conclusion: Reducing the dose to the radiosensitive organs, such as eye lens, during routine brain CT examinations could lower the radiation risks. The OBTCM technique and in-plane bismuth shielding could be used to reduce the eye-lens dose. The eye-lens dose could be effectively reduced using OBTCM scan without interfering the diagnostic image quality. Patient position relative the CT gantry also affects the dose level of the eye lens. This study was supported by the grants from the Ministry of Science and Technology of Taiwan (MOST103-2314-B-182

Medium-dose-rate intracavitary brachytherapy for cervical cancer

International Nuclear Information System (INIS)

Tanaka, Eiichi; Isohashi, Fumiaki; Oh, Ryoong-Jin

2003-01-01

The purpose of this study was to evaluate the results of medium-dose-rate (MDR) intracavitary brachytherapy (ICRT) for cervical cancer. Between May 1991 and March 2001, 80 patients with cervical cancer were treated with external radiotherapy combined with MDR-ICRT. Two patients were excluded from this study. The median age of patients was 61 years (range: 30-87 years). Seventy-five patients had pathologically proved squamous cell carcinoma, and 3 had adenocarcinoma. The patients were staged by Union Internationale Contre le Cancer (UICC) classification as follows: Stage IA (2), Stage IB (4), Stage IIA (5), Stage IIB (22), Stage IIIA (1), Stage IIIB (32), Stage IVA (5), Stage IVB (7). Median follow-up for survivor was 68 months (range: 12-131 months). The radiation therapy was based on a combination of ICRT and external pelvic irradiation. Patients with stages II, III and IVA were treated with whole-pelvic irradiation with respective total doses of 20, 30, and 40 Gy. Doses of 40, 30, 20, and 20 Gy parametrial irradiation were added with central shield pelvic irradiation for stages IB, II, III and IVA lesions respectively. For MDR-ICRT, from May 1991 to December 1995, point A dose were 40 Gy/4 fractions for stages I and II, 38 Gy/4 fractions for stage III, and 28.5 Gy/3 fractions for stage IVA. And from January 1996 to March 2001, point A dose of 36 Gy/4 fractions for stages I and II, 34 Gy/4 fractions for stage III, and 25.5 Gy/3 fractions for stage IVA. The median dose rate at point A was 1.7 Gy/hour (range: 1.3-2.2 Gy/hour). The 5-year cause-specific survival rates were 100%, 76%, 51% and 40% for stages I, II, III and IVA respectively. All patients with stage IVB died from the tumor with a median survival time of 12 months. The 5-year pelvic control rates were 100%, 88%, 69% and 40% for stages I, II, III and IVA respectively. Major late complications occurred in 2 patients (3%). One patient developed vesico- and recto-vaginal fistulae, and died of pelvic infection

MOSFET Dosimetry for Evaluation of Gonad Shielding during Radiotherapy

International Nuclear Information System (INIS)

Kim, Hwi Young; Choi, Yun Seok; Park, So Yeon; Park, Yang Kyun; Ye, Sung Joon

2011-01-01

In order to confirm feasibility of MOSFET modality in use of in vivo dosimetry, evaluation of gonad shielding in order to minimize gonadal dose of patients undergoing radiotherapy by using MOSFET modality was performed. Gonadal dose of patients undergoing radiotherapy for rectal cancer in the department of radiation oncology of Seoul National University Hospital since 2009 was measured. 6 MV and 15 MV photon beams emitted from Varian 21EX LINAC were used for radiotherapy. In order to minimize exposed dose caused by the scattered ray not only from collimator of LINAC but also from treatment region inside radiation field, we used box.shaped lead shielding material. The shielding material was made of the lead block and consists of 7.5 cm x 9.5 cm x 5.5 cm sized case and 9 cm x 9.5 cm x 1 cm sized cover. Dosimetry for evaluation of gonad shielding was done with MOSFET modality. By protecting with gonad shielding material, average gonadal dose of patients was decreased by 23.07% compared with reference dose outside of the shielding material. Average delivered gonadal dose inside the shielding material was 0.01 Gy. By the result of MOSFET dosimetry, we verified that gonadal dose was decreased by using gonad shielding material. In compare with TLD dosimetry, we could measure the exposed dose easily and precisely with MOSFET modality

MOSFET Dosimetry for Evaluation of Gonad Shielding during Radiotherapy

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwi Young; Choi, Yun Seok; Park, So Yeon; Park, Yang Kyun [Seoul National University College of Medicine, Seoul (Korea, Republic of); Ye, Sung Joon [Seoul National University, Seoul (Korea, Republic of)

2011-03-15

In order to confirm feasibility of MOSFET modality in use of in vivo dosimetry, evaluation of gonad shielding in order to minimize gonadal dose of patients undergoing radiotherapy by using MOSFET modality was performed. Gonadal dose of patients undergoing radiotherapy for rectal cancer in the department of radiation oncology of Seoul National University Hospital since 2009 was measured. 6 MV and 15 MV photon beams emitted from Varian 21EX LINAC were used for radiotherapy. In order to minimize exposed dose caused by the scattered ray not only from collimator of LINAC but also from treatment region inside radiation field, we used box.shaped lead shielding material. The shielding material was made of the lead block and consists of 7.5 cm x 9.5 cm x 5.5 cm sized case and 9 cm x 9.5 cm x 1 cm sized cover. Dosimetry for evaluation of gonad shielding was done with MOSFET modality. By protecting with gonad shielding material, average gonadal dose of patients was decreased by 23.07% compared with reference dose outside of the shielding material. Average delivered gonadal dose inside the shielding material was 0.01 Gy. By the result of MOSFET dosimetry, we verified that gonadal dose was decreased by using gonad shielding material. In compare with TLD dosimetry, we could measure the exposed dose easily and precisely with MOSFET modality.

Quantitative analysis of biological responses to low dose-rate γ-radiation, including dose, irradiation time, and dose-rate

International Nuclear Information System (INIS)

Magae, J.; Furukawa, C.; Kawakami, Y.; Hoshi, Y.; Ogata, H.

2003-01-01

Full text: Because biological responses to radiation are complex processes dependent on irradiation time as well as total dose, it is necessary to include dose, dose-rate and irradiation time simultaneously to predict the risk of low dose-rate irradiation. In this study, we analyzed quantitative relationship among dose, irradiation time and dose-rate, using chromosomal breakage and proliferation inhibition of human cells. For evaluation of chromosome breakage we assessed micronuclei induced by radiation. U2OS cells, a human osteosarcoma cell line, were exposed to gamma-ray in irradiation room bearing 50,000 Ci 60 Co. After the irradiation, they were cultured for 24 h in the presence of cytochalasin B to block cytokinesis, cytoplasm and nucleus were stained with DAPI and propidium iodide, and the number of binuclear cells bearing micronuclei was determined by fluorescent microscopy. For proliferation inhibition, cells were cultured for 48 h after the irradiation and [3H] thymidine was pulsed for 4 h before harvesting. Dose-rate in the irradiation room was measured with photoluminescence dosimeter. While irradiation time less than 24 h did not affect dose-response curves for both biological responses, they were remarkably attenuated as exposure time increased to more than 7 days. These biological responses were dependent on dose-rate rather than dose when cells were irradiated for 30 days. Moreover, percentage of micronucleus-forming cells cultured continuously for more than 60 days at the constant dose-rate, was gradually decreased in spite of the total dose accumulation. These results suggest that biological responses at low dose-rate, are remarkably affected by exposure time, that they are dependent on dose-rate rather than total dose in the case of long-term irradiation, and that cells are getting resistant to radiation after the continuous irradiation for 2 months. It is necessary to include effect of irradiation time and dose-rate sufficiently to evaluate risk

MMW [multimegawatt] shielding design and analysis

International Nuclear Information System (INIS)

Olson, A.P.

1988-01-01

Reactor shielding for multimegawatt (MMW) space power must satisfy a mass constraint as well as performance specifications for neutron fluence and gamma dose. A minimum mass shield is helpful in attaining the launch mass goal for the entire vehicle, because the shield comprises about 1% to 2% of the total vehicle mass. In addition, the shield internal heating must produce tolerable temperatures. The analysis of shield performance for neutrons and gamma rays is emphasized. Topics addressed include cross section preparation for multigroup 2D S/sub n/-transport analyses, and the results of parametric design studies on shadow shield performance and mass versus key shield design variables such as cone angle, number, placement, and thickness of layers of tungsten, and shield top radius. Finally, adjoint methods are applied to the shield in order to spatially map its relative contribution to dose reduction, and to provide insight into further design optimization. 7 refs., 2 figs., 3 tabs

Repository Waste Package Transporter Shielding Weight Optimization

International Nuclear Information System (INIS)

C.E. Sanders; Shiaw-Der Su

2005-01-01

The Yucca Mountain repository requires the use of a waste package (WP) transporter to transport a WP from a process facility on the surface to the subsurface for underground emplacement. The transporter is a part of the waste emplacement transport systems, which includes a primary locomotive at the front end and a secondary locomotive at the rear end. The overall system with a WP on board weights over 350 metric tons (MT). With the shielding mass constituting approximately one-third of the total system weight, shielding optimization for minimal weight will benefit the overall transport system with reduced axle requirements and improved maneuverability. With a high contact dose rate on the WP external surface and minimal personnel shielding afforded by the WP, the transporter provides radiation shielding to workers during waste emplacement and retrieval operations. This paper presents the design approach and optimization method used in achieving a shielding configuration with minimal weight

Estimation of low-level neutron dose-equivalent rate by using extrapolation method for a curie level Am–Be neutron source

International Nuclear Information System (INIS)

Li, Gang; Xu, Jiayun; Zhang, Jie

2015-01-01

Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am–Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am–Be neutron source was investigated by measuring the count rates obtained through a 3 He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, 3 He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter. - Highlights: • The scope of the affected area for

Hot-cell shielding system for high power transmission in DUPIC fuel fabrication

International Nuclear Information System (INIS)

Kim, K.; Lee, J.; Park, J.; Yang, M.; Park, H.

2000-01-01

This paper presents a newly designed hot-cell shielding system for use in the development of DUPIC (Direct Use of spent PWR fuel In CANDU reactors) fuel at KAERI (Korea Atomic Energy Research Institute). This hot-cell shielding system that was designed to transmit high power to sintering furnace in-cell from the out-of-cell through a thick cell wall has three subsystems - a steel shield plug with embedded spiral cooling line, stepped copper bus bars, and a shielding lead block. The dose-equivalent rates of the hot-cell shielding system and of the apertures between this system and the hot-cell wall were calculated. Calculated results were compared with the allowable dose limit of the existing hot-cell. Experiments for examining the temperature changes of the shielding system developed during normal furnace operation were also carried out. Finally, gamma-ray radiation survey experiments were conducted by Co-60 source. It is demonstrated that, from both calculated and experimental results, the newly designed hot-cell shielding system meets all the shielding requirements of the existing hot-cell facility, enabling high power transmission to the in-cell sintering furnace. (author)

Radiobiological aspects of continuous low dose-rate irradiation and fractionated high dose-rate irradiation

International Nuclear Information System (INIS)

Turesson, I.

1990-01-01

The biological effects of continuous low dose-rate irradiation and fractionated high dose-rate irradiation in interstitial and intracavitary radiotherapy and total body irradiation are discussed in terms of dose-rate fractionation sensitivity for various tissues. A scaling between dose-rate and fraction size was established for acute and late normal-tissue effects which can serve as a guideline for local treatment in the range of dose rates between 0.02 and 0.005 Gy/min and fraction sizes between 8.5 and 2.5 Gy. This is valid provided cell-cycle progression and proliferation can be ignored. Assuming that the acute and late tissue responses are characterized by α/β values of about 10 and 3 Gy and a mono-exponential repair half-time of about 3 h, the same total doses given with either of the two methods are approximately equivalent. The equivalence for acute and late non-hemopoietic normal tissue damage is 0.02 Gy/min and 8.5 Gy per fraction; 0.01 Gy/min and 5.5 Gy per fraction; and 0.005 Gy/min and 2.5Gy per fraction. A very low dose rate, below 0.005 Gy/min, is thus necessary to simulate high dose-rate radiotherapy with fraction sizes of about 2Gy. The scaling factor is, however, dependent on the repair half-time of the tissue. A review of published data on dose-rate effects for normal tissue response showed a significantly stronger dose-rate dependence for late than for acute effects below 0.02 Gy/min. There was no significant difference in dose-rate dependence between various acute non-hemopoietic effects or between various late effects. The consistent dose-rate dependence, which justifies the use of a general scaling factor between fraction size and dose rate, contrasts with the wide range of values for repair half-time calculated for various normal-tissue effects. This indicates that the model currently used for repair kinetics is not satisfactory. There are also few experimental data in the clinical dose-rate range, below 0.02 Gy/min. It is therefore

Adaptive statistical iterative reconstruction and bismuth shielding for evaluation of dose reduction to the eye and image quality during head CT

Science.gov (United States)

Kim, Myeong Seong; Choi, Jiwon; Kim, Sun Young; Kweon, Dae Cheol

2014-03-01

There is a concern regarding the adverse effects of increasing radiation doses due to repeated computed tomography (CT) scans, especially in radiosensitive organs and portions thereof, such as the lenses of the eyes. Bismuth shielding with an adaptive statistical iterative reconstruction (ASIR) algorithm was recently introduced in our clinic as a method to reduce the absorbed radiation dose. This technique was applied to the lens of the eye during CT scans. The purpose of this study was to evaluate the reduction in the absorbed radiation dose and to determine the noise level when using bismuth shielding and the ASIR algorithm with the GE DC 750 HD 64-channel CT scanner for CT of the head of a humanoid phantom. With the use of bismuth shielding, the noise level was higher in the beam-hardening artifact areas than in the revealed artifact areas. However, with the use of ASIR, the noise level was lower than that with the use of bismuth alone; it was also lower in the artifact areas. The reduction in the radiation dose with the use of bismuth was greatest at the surface of the phantom to a limited depth. In conclusion, it is possible to reduce the radiation level and slightly decrease the bismuth-induced noise level by using a combination of ASIR as an algorithm process and bismuth as an in-plane hardware-type shielding method.

Adaptive statistical iterative reconstruction and bismuth shielding for evaluation of dose reduction to the eye and image quality during head CT

Energy Technology Data Exchange (ETDEWEB)

Kim, Myeong Seong [The Korean National Cancer Center, Goyang (Korea, Republic of); Seoul National University, Seoul (Korea, Republic of); Choi, Ji Won [Jeonju University, Jeonju (Korea, Republic of); Kim, Sun Young [Hallym University of Graduate Studies, Seoul (Korea, Republic of); The Korean National Cancer Center, Goyang (Korea, Republic of); Kweon, Dae Cheol [Shinhan University, Uijeongbu (Korea, Republic of)

2014-03-15

There is a concern regarding the adverse effects of increasing radiation doses due to repeated computed tomography (CT) scans, especially in radiosensitive organs and portions thereof, such as the lenses of the eyes. Bismuth shielding with an adaptive statistical iterative reconstruction (ASIR) algorithm was recently introduced in our clinic as a method to reduce the absorbed radiation dose. This technique was applied to the lens of the eye during CT scans. The purpose of this study was to evaluate the reduction in the absorbed radiation dose and to determine the noise level when using bismuth shielding and the ASIR algorithm with the GE DC 750 HD 64-channel CT scanner for CT of the head of a humanoid phantom. With the use of bismuth shielding, the noise level was higher in the beam-hardening artifact areas than in the revealed artifact areas. However, with the use of ASIR, the noise level was lower than that with the use of bismuth alone; it was also lower in the artifact areas. The reduction in the radiation dose with the use of bismuth was greatest at the surface of the phantom to a limited depth. In conclusion, it is possible to reduce the radiation level and slightly decrease the bismuth-induced noise level by using a combination of ASIR as an algorithm process and bismuth as an in-plane hardware-type shielding method.

The angular gamma flux in an iron slab shield

International Nuclear Information System (INIS)

Penkuhn, H.

1975-08-01

The angular distribution of the photon energy and dose rate flux in a plane iron shield is investigated assuming an isotropic volume source. Near the shield axis (cos phi approximately 1, with phi=angle between shield axis and gamma direction) the angular spectrum is strongly space-dependent. For large phi, space-independent fits are given. Source energies from 0.662 to 6 MeV and penetrations from 6 to 60 cm are treated and the results are compared with a similar investigation on normal concrete. The differences iron-concrete are appreciable only for the lowest source energy

Gonadal shield.

Science.gov (United States)

Purdy, J A; Stiteler, R D; Glasgow, G P; Mill, W B

1975-10-01

A secondary gonadal shield for use in the pelvic irradiation of males was designed and built using material and apparatus available with the Cerrobend blocking system. The gonadal dose was reduced to approximately 1.5 to 2.5% of the given dose.

Shielding calculations for the Intense Neutron Source Facility. Final report

International Nuclear Information System (INIS)

Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.

1978-06-01

Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

Comparison between steel and lead shieldings for radiotherapy rooms regarding neutron doses to patients

Energy Technology Data Exchange (ETDEWEB)

Silva, M.G.; Rebello, W.F.; Andrade, E.R.; Medeiros, M.P.C.; Mendes, R.M.S.; Braga, K.L.; Gomes, R.G., E-mail: eng.cavaliere@gmail.com, E-mail: ggrprojetos@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear; Silva, A.X., E-mail: ademir@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

2015-07-01

The NCRP Report No. 151, Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities, considers, in shielding calculations for radiotherapy rooms, the use of lead and/or steel to be applied on bunker walls. The NCRP Report calculations were performed foreseeing a better protection of people outside the radiotherapy room. However, contribution of lead and steel to patient dose should be taken into account for radioprotection purposes. This work presents calculations performed by MCNPX code in analyzing the Ambient Dose Equivalent due to neutron, H *(10){sub n}, within a radiotherapy room, in the patients area, considering the use of additional shielding of 1 TVL of lead or 1 TVL of steel, positioned at the inner faces of walls and ceiling of a bunker. The head of the linear accelerator Varian 2100/2300 C/D was modeled working at 18MeV, with 5 x 5 cm{sup 2}, 10 x 10 cm{sup 2}, 20 x 20 cm{sup 2}, 30 x 30 cm{sup 2} and 40 x 40 cm{sup 2} openings for jaws and MLC and operating in eight gantry's angles. This study shows that the use of lead generates an average value of H *(10){sub n} at patients area, 8.02% higher than the expected when using steel. Further studies should be performed based on experimental data for comparison with those from MCNPX simulation. (author)

Problems of the power plant shield optimization

International Nuclear Information System (INIS)

Abagyan, A.A.; Dubinin, A.A.; Zhuravlev, V.I.; Kurachenko, Yu.A.; Petrov, Eh.E.

1981-01-01

General approaches to the solution of problems on the nuclear power plant radiation shield optimization are considered. The requirements to the shield parameters are formulated in a form of restrictions on a number of functionals, determined by the solution of γ quantum and neutron transport equations or dimensional and weight characteristics of shield components. Functional determined by weight-dimensional parameters (shield cost, mass and thickness) and functionals, determined by radiation fields (equivalent dose rate, produced by neutrons and γ quanta, activation functional, radiation functional, heat flux, integral heat flux in a particular part of the shield volume, total energy flux through a particular shield surface are considered. The following methods of numerical solution of simplified optimization problems are discussed: semiempirical methods using radiation transport physical leaks, numerical solution of approximate transport equations, numerical solution of transport equations for the simplest configurations making possible to decrease essentially a number of variables in the problem. The conclusion is drawn that the attained level of investigations on the problem of nuclear power plant shield optimization gives the possibility to pass on at present to the solution of problems with a more detailed account of the real shield operating conditions (shield temperature field account, its strength and other characteristics) [ru

Shutdown dose rate contribution from diagnostics in ITER upper port 18

Energy Technology Data Exchange (ETDEWEB)

Cheon, M.S., E-mail: munseong@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Pak, S.; An, Y.H.; Seon, C.R.; Lee, H.G. [National Fusion Research Institute, Daejeon (Korea, Republic of); Bertalot, L.; Krasilnikov, V. [ITER Organization, St Paul-lez-Durance (France); Zvonkov, A. [Agency ITER-RF, Moscow (Russian Federation)

2016-11-01

Highlights: • The Shutdown Dose Rate in the interspace of ITER upper port 18 was evaluated. • VUV spectrometer is the dominant contributor to the average SDR. • The existence and size of the blanket cooling pipes impacts significantly on SDR. - Abstract: D-T operation of ITER plasma will produce high-energy fusion neutrons those can activate materials around the place where human-access is necessary. The interspace of the diagnostic port is one of the area where human-access is necessary for the maintenance of diagnostic systems installed at the port, so it is important to evaluate a dose rate of the interspace area in order to comply with ALARA principle. The shutdown dose rate (SDR) in the interspace of ITER upper port 18 was evaluated by the Direct 1-Step (D1S) method using MCNP5 code. This port contains three diagnostics: Vacuum Ultra-Violet (VUV) Spectrometer, Neutron Activation System (NAS), and Upper Vertical Neutron Camera (UVNC). The contribution of each diagnostic in the port was evaluated by running separate upper port MCNP models those contain individual diagnostic only, and the total dose rate contribution was evaluated with the model which was fully integrated with all the diagnostics. The effect of the opening around the upper port plug and of the other ports was also investigated. The purpose of this assessment is to provide the shielding design basis for the preliminary design of the diagnostic integration in the port. The method and result of the calculation will be presented in this paper.

Shield wall evaluation of hot cell facility for advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Cho, I. J.; Kuk, D. H.; Ko, J. H.; Jung, W. M.; Yoo, G. S.; Lee, E. P.; Park, S. W.

2002-01-01

The future hot cell is located in the Irradiated Material Experiment Facility (IMEF) at the Korea Atomic Energy Research Institute (KAERI). It is β-γ type hot cell that was constructed on the base floor in IMEF building for irradiated material testing. And this hot cell will be used for carrying out the Advanced spent fuel Conditioning Process (ACP). The radiation shielding capability of hot cell should be sufficient to meet the radiation dose requirements in the related regulations. Because the radioactive sources of ACP are expected to be higher than radioactive sources of IMEF design criteria, the future hot cell in current status is unsatisfactory to hot test of ACP. So the shielding analysis of the future hot cell is performed to evaluate shielding ability of concrete shield wall. The shielding analysis included (a) identification of ACP source term; (b) photon source spectrum; (c) shielding analysis by QADS and MCNP-4C; and (d) enhancement of concrete shield wall. In this research, dose rates are obtained according to ACP source, geometry and hot cell shield wall thickness. And the evaluation and reinforcement thickness of the shield wall about future hot cell are concluded

Modeling the effectiveness of shielding in the earth-moon-mars radiation environment using PREDICCS: five solar events in 2012

Science.gov (United States)

Quinn, Philip R.; Schwadron, Nathan A.; Townsend, Larry W.; Wimmer-Schweingruber, Robert F.; Case, Anthony W.; Spence, Harlan E.; Wilson, Jody K.; Joyce, Colin J.

2017-08-01

Radiation in the form of solar energetic particles (SEPs) presents a severe risk to the short-term health of astronauts and the success of human exploration missions beyond Earth's protective shielding. Modeling how shielding mitigates the dose accumulated by astronauts is an essential step toward reducing these risks. PREDICCS (Predictions of radiation from REleASE, EMMREM, and Data Incorporating the CRaTER, COSTEP, and other SEP measurements) is an online tool for the near real-time prediction of radiation exposure at Earth, the Moon, and Mars behind various levels of shielding. We compare shielded dose rates from PREDICCS with dose rates from the Cosmic Ray Telescope for the Effects of Radiation (CRaTER) onboard the Lunar Reconnaissance Orbiter (LRO) at the Moon and from the Radiation Assessment Detector (RAD) on the Mars Science Laboratory (MSL) during its cruise phase to Mars for five solar events in 2012 when Earth, MSL, and Mars were magnetically well connected. Calculations of the accumulated dose demonstrate a reasonable agreement between PREDICCS and RAD ranging from as little as 2% difference to 54%. We determine mathematical relationships between shielding levels and accumulated dose. Lastly, the gradient of accumulated dose between Earth and Mars shows that for the largest of the five solar events, lunar missions require aluminum shielding between 1.0 g cm-2 and 5.0 g cm-2 to prevent radiation exposure from exceeding the 30-day limits for lens and skin. The limits were not exceeded near Mars.

Gonad shielding in diagnostic radiology

International Nuclear Information System (INIS)

Anon.

1975-01-01

The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. The types of gonad shields in use are discussed as are the types of diagnostic examinations that should include gonad shielding. It was found that when properly used, most shields provided substantial gonad dose reductions

Female gonadal shielding with automatic exposure control increases radiation risks

Energy Technology Data Exchange (ETDEWEB)

Kaplan, Summer L.; Zhu, Xiaowei [Children' s Hospital of Philadelphia, Department of Radiology, Philadelphia, PA (United States); University of Pennsylvania, Perelman School of Medicine, Philadelphia, PA (United States); Magill, Dennise; Felice, Marc A. [University of Pennsylvania, Environmental Health and Radiation Safety, Philadelphia, PA (United States); Xiao, Rui [University of Pennsylvania, Department of Biostatistics and Epidemiology, Philadelphia, PA (United States); Ali, Sayed [Temple University Hospital, Department of Radiology, Philadelphia, PA (United States)

2018-02-15

Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation. (orig.)

Female gonadal shielding with automatic exposure control increases radiation risks

International Nuclear Information System (INIS)

Kaplan, Summer L.; Zhu, Xiaowei; Magill, Dennise; Felice, Marc A.; Xiao, Rui; Ali, Sayed

2018-01-01

Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation. (orig.)

Female gonadal shielding with automatic exposure control increases radiation risks.

Science.gov (United States)

Kaplan, Summer L; Magill, Dennise; Felice, Marc A; Xiao, Rui; Ali, Sayed; Zhu, Xiaowei

2018-02-01

Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation.

Radiation dose reduction at a price: the effectiveness of a male gonadal shield during helical CT scans

OpenAIRE

Erdi Yusuf E; Casciotta Kevin A; Dauer Lawrence T; Rothenberg Lawrence N

2007-01-01

Abstract Background It is estimated that 60 million computed tomography (CT) scans were performed during 2006, with approximately 11% of those performed on children age 0–15 years. Various types of gonadal shielding have been evaluated for reducing exposure to the gonads. The purpose of this study was to quantify the radiation dose reduction to the gonads and its effect on image quality when a wrap-around male pediatric gonad shield was used during CT scanning. This information is obtained to...

Preliminary Shielding Analysis for HCCB TBM Transport

Science.gov (United States)

Miao, Peng; Zhao, Fengchao; Cao, Qixiang; Zhang, Guoshu; Feng, Kaiming

2015-09-01

A preliminary shielding analysis on the transport of the Chinese helium cooled ceramic breeder test blanket module (HCCB TBM) from France back to China after being irradiated in ITER is presented in this contribution. Emphasis was placed on irradiation safety during transport. The dose rate calculated by MCNP/4C for the conceptual package design satisfies the relevant dose limits from IAEA that the dose rate 3 m away from the surface of the package containing low specific activity III materials should be less than 10 mSv/h. The change with location and the time evolution of dose rates after shutdown have also been studied. This will be helpful for devising the detailed transport plan of HCCB TBM back to China in the near future. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

Onboard radiation shielding estimates for interplanetary manned missions

International Nuclear Information System (INIS)

Totemeier, A.; Jevremovic, T.; Hounshel, D.

2004-01-01

The main focus of space related shielding design is to protect operating systems, personnel and key structural components from outer space and onboard radiation. This paper summarizes the feasibility of a lightweight neutron radiation shield design for a nuclear powered, manned space vehicle. The Monte Carlo code MCNP5 is used to determine radiation transport characteristics of the different materials and find the optimized shield configuration. A phantom torso encased in air is used to determine a dose rate for a crew member on the ship. Calculation results indicate that onboard shield against neutron radiation coming from nuclear engine can be achieved with very little addition of weight to the space vehicle. The selection of materials and neutron transport analysis as presented in this paper are useful starting data to design shield against neutrons generated when high-energy particles from outer space interact with matter on the space vehicle. (authors)

Analytic flux formulas and tables of shielding functions

International Nuclear Information System (INIS)

Wallace, O.J.

1981-06-01

Hand calculations of radiation flux and dose rates are often useful in evaluating radiation shielding and in determining the scope of a problem. The flux formulas appropriate to such calculations are almost always based on the point kernel and allow for at most the consideration of laminar slab shields. These formulas often require access to tables of values of integral functions for effective use. Flux formulas and function tables appropriate to calculations involving homogeneous source regions with the shapes of lines, disks, slabs, truncated cones, cylinders, and spheres are presented. Slab shields may be included in most of these calculations, and the effect of a cylindrical shield surrounding a cylindrical source may be estimated. Detector points may be located axially, laterally, or interior to a cylindrical source. Line sources may be tilted with respect to a slab shield. All function tables are given for a wide range of arguments

Evaluation of syringe shield effectiveness in handling radiopharmaceuticals

Directory of Open Access Journals (Sweden)

Cho Yong-In

2015-01-01

Full Text Available The purpose of this study was to evaluate the effectiveness of the radiation shield of radionuclide syringes and the personal dose equivalent by performing a simulation of radionuclides used in nuclear medicine diagnosis. In order to evaluate the dose depending on the distance between the radiation source and the ICRU sphere against the thickness of the shielding device, the distance at which a nuclear medicine worker may inadvertently come into contact with radiation from the radiation source was set at 0 cm to 30 cm according to the thickness of the shield, thus fixing the ICRU sphere. For a dose evaluation, Hp(10, Hp(3, and Hp(0.07 measurable in specific depth of the ICRU were evaluated. It was found that a dose measured on skin surface of nuclear medicine workers was relatively higher, that the dose varied in relation to the thickness of the radiation shield, and that the shielding effect decreased for some radiation sources such as 67Ga and 111In. It proved necessary to increase thickness of shielding device to the radiation sources such as 67Ga and 111In. It is also considered that a study of proper shielding thickness will be needed in future.

{sup 3}He detector analysis of some special shielding materials

Energy Technology Data Exchange (ETDEWEB)

Avdic, S; Pesic, M [Boris Kidric, Institute of Nuclear Sciences, Beograd (Yugoslavia); Marinkovic, P [ETF Belgrade Univ. (Yugoslavia)

1990-07-01

The shielding properties of commercial materials of reactor Experiments, Inc. (R/X) were analyzed at the facility which includes bare heavy water experimental reactor RB with external neutron converter ENC, The fast neutron spectrum measurements in energy range from 1 MeV to 10 MeV was performed using ORTEC semiconductor neutron detector with He{sup 3} in diode coincidence arrangement. The neutron spectra have been evaluated from measured pulse-height distribution using numerical code HE3 for computation of detector efficiency in a collimated neutron beam. The neutron dose rates behind ENC with and without sample R/X material were determined using cubic spline interpolation routine for calculating the corresponding flux-dose rate conversion factors. Satisfactory shielding properties of the examined material in a fast neutron field in measurements and calculations are demonstrated. (author)

Computed tomography shielding methods: a literature review.

Science.gov (United States)

Curtis, Jessica Ryann

2010-01-01

To investigate available shielding methods in an effort to further awareness and understanding of existing preventive measures related to patient exposure in computed tomography (CT) scanning. Searches were conducted to locate literature discussing the effectiveness of commercially available shields. Literature containing information regarding breast, gonad, eye and thyroid shielding was identified. Because of rapidly advancing technology, the selection of articles was limited to those published within the past 5 years. The selected studies were examined using the following topics as guidelines: the effectiveness of the shield (percentage of dose reduction), the shield's effect on image quality, arguments for or against its use (including practicality) and overall recommendation for its use in clinical practice. Only a limited number of studies have been performed on the use of shields for the eyes, thyroid and gonads, but the evidence shows an overall benefit to their use. Breast shielding has been the most studied shielding method, with consistent agreement throughout the literature on its effectiveness at reducing radiation dose. The effect of shielding on image quality was not remarkable in a majority of studies. Although it is noted that more studies need to be conducted regarding the impact on image quality, the currently published literature stresses the importance of shielding in reducing dose. Commercially available shields for the breast, thyroid, eyes and gonads should be implemented in clinical practice. Further research is needed to ascertain the prevalence of shielding in the clinical setting.

Potential gonadal dose from leakage radiation?

International Nuclear Information System (INIS)

Nicholson, R.A.

1995-01-01

The author draws attention to the potential dangers of leakage radiation from mobile image intensifier units, and points out that during interventional urological procedures, radiation from below the urologist's knees may irradiate male gonads without being intercepted by protective aprons. Results are presented for a Shimatzu WHA mobile II, phantom doses being measured with an ionization chamber. Dose rates measured in the male gonad position were compared with rates at waist level behind a 0.35 mm lead equivalent shielding and dose rates at collar level outside the lead apron. Results are also presented of a study on the effect on gonad dose of a) adding 0.7 mm lead shielding to the tube housing and b) adding 0.7 mm lead and removing the spacer cone to reduce scatter. Results show that it is possible for gonad doses to be comparable with those assumed for the eyes, rather than the body. (Author)

Comprehensive analysis of shielding effectiveness for HDPE, BPE and concrete as candidate materials for neutron shielding

International Nuclear Information System (INIS)

Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

2015-01-01

In the compact accelerator based DD neutron generator, the deuterium ions generated by the ion source are accelerated after the extraction and bombarded to a deuterated titanium target. The emitted neutrons have typical energy of ∼2.45MeV. Utilization of these compact accelerator based neutron generators of yield up to 10 9 neutron/second (DD) is under active consideration in many research laboratories for conducting active neutron interrogation experiments. Requirement of an adequately shielded laboratory is mandatory for the effective and safe utilization of these generators for intended applications. In this reference, we report the comprehensive analysis of shielding effectiveness for High Density Polyethylene (HDPE), Borated Polyethylene (BPE) and Concrete as candidate materials for neutron shielding. In shielding calculations, neutron induced scattering and absorption gamma dose has also been considered along with neutron dose. Contemporarily any material with higher hydrogenous concentration is best suited for neutron shielding. Choice of shielding material is also dominated by practical issues like economic viability and availability of space. Our computational analysis results reveal that utilization of BPE sheets results in minimum wall thickness requirement for attaining similar range of attenuation in neutron and gamma dose. The added advantage of using borated polyethylene is that it reduces the effect of both neutron and gamma dose by absorbing neutron and producing lithium and alpha particle. It has also been realized that for deciding upon optimum thickness determination of any shielding material, three important factors to be necessarily considered are: use factor, occupancy factor and work load factor. (author)

Dose and dose rate monitor

International Nuclear Information System (INIS)

Novakova, O.; Ryba, J.; Slezak, V.; Svobodova, B.; Viererbl, L.

1984-10-01

The methods are discussea of measuring dose rate or dose using a scintillation counte. A plastic scintillator based on polystyrene with PBD and POPOP activators and coated with ZnS(Ag) was chosen for the projected monitor. The scintillators were cylindrical and spherical in shape and of different sizes; black polypropylene tubes were chosen as the best case for the probs. For the counter with different plastic scintillators, the statistical error 2σ for natural background was determined. For determining the suitable thickness of the ZnS(Ag) layer the energy dependence of the counter was measured. Radioisotopes 137 Cs, 241 Am and 109 Cd were chosen as radiation sources. The best suited ZnS(Ag) thickness was found to be 0.5 μm. Experiments were carried out to determine the directional dependence of the detector response and the signal to noise ratio. The temperature dependence of the detector response and its compensation were studied, as were the time stability and fatigue manifestations of the photomultiplier. The design of a laboratory prototype of a dose rate and dose monitor is described. Block diagrams are given of the various functional parts of the instrument. The designed instrument is easiiy portable, battery powered, measures dose rates from natural background in the range of five orders, i.e., 10 -2 to 10 3 nGy/s, and allows to determine a dose of up to 10 mGy. Accouracy of measurement in the energy range of 50 keV to 1 MeV is better than +-20%. (E.S.)

CASKCODES, Program CAPSIZE Scope KWIKDOSE for Shipping Cask Shielding

International Nuclear Information System (INIS)

1988-01-01

1 - Description of program or function: CAPSIZE is an interactive program to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel casks designed to meet those objectives. 2 - Method of solution: Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the load cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium- shielded cask meeting those objectives. Using optimal packing arrangements and shielding requirements input by the user, SCOPE will design a cask to carry a single fuel assembly and then continue incrementing the number of assemblies until one or more of the design limits can no longer be met. KWIKDOSE queries the user for the number of PWR fuel assemblies in a cask, the type of cask and thickness of the shield. Upon getting the necessary input, KWIKDOSE prints out the total dose rate, 10 feet from the centerline of the cask, as a function of the burnup and cooling time of the spent fuel. 3 - Restrictions on the complexity of the problem: The restrictions are subject to the shielding requirements of the shipping cask

Results of radiation therapy for uterine cervical cancer using high dose rate remote after loading system

International Nuclear Information System (INIS)

Ogawa, Yoshihiro; Nemoto, Kenji

2003-01-01

In Japan, radiotherapy with high dose rate remote after loading system (HDR-RALS) for intracavitary brachytherapy is the standard treatment for more than 30 years. This report showed the usefulness of HDR-RALS for uterine cervical cancer. From 1980 through 1999, 442 patients with uterine cervical cancers (stage I: 66, stage II: 161, stage III: 165, stage IV: 50) were treated. Radiotherapy was performed both external teletherapy and HDR-RALS. Overall survival rate at 5 years was 60.2%. The 5-year actuarial incidence of all complications was 16.4%. The 5-year actuarial incidence of all complications in cases treated with the sum doses of whole pelvic irradiation (without central shield) and RALS up to 49 Gy, 50 to 59 Gy or larger doses were 7.5%, 11.0% and 25.2%, respectively. Radiation therapy using HDR-RALS was very effective. While the dose of whole pelvic irradiation was increased, the actuarial incidence of all complications was increased. (author)

The Spallation Neutron Source (SNS) conceptual design shielding analysis

International Nuclear Information System (INIS)

Johnson, J.O.; Odano, N.; Lillie, R.A.

1998-03-01

The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

Modeling the effectiveness of shielding in the earth-moon-mars radiation environment using PREDICCS: five solar events in 2012

Directory of Open Access Journals (Sweden)

Quinn Philip R.

2017-01-01

Full Text Available Radiation in the form of solar energetic particles (SEPs presents a severe risk to the short-term health of astronauts and the success of human exploration missions beyond Earth’s protective shielding. Modeling how shielding mitigates the dose accumulated by astronauts is an essential step toward reducing these risks. PREDICCS (Predictions of radiation from REleASE, EMMREM, and Data Incorporating the CRaTER, COSTEP, and other SEP measurements is an online tool for the near real-time prediction of radiation exposure at Earth, the Moon, and Mars behind various levels of shielding. We compare shielded dose rates from PREDICCS with dose rates from the Cosmic Ray Telescope for the Effects of Radiation (CRaTER onboard the Lunar Reconnaissance Orbiter (LRO at the Moon and from the Radiation Assessment Detector (RAD on the Mars Science Laboratory (MSL during its cruise phase to Mars for five solar events in 2012 when Earth, MSL, and Mars were magnetically well connected. Calculations of the accumulated dose demonstrate a reasonable agreement between PREDICCS and RAD ranging from as little as 2% difference to 54%. We determine mathematical relationships between shielding levels and accumulated dose. Lastly, the gradient of accumulated dose between Earth and Mars shows that for the largest of the five solar events, lunar missions require aluminum shielding between 1.0 g cm−2 and 5.0 g cm−2 to prevent radiation exposure from exceeding the 30-day limits for lens and skin. The limits were not exceeded near Mars.

Gonadal Shielding in Radiography: A Best Practice?

Science.gov (United States)

Fauber, Terri L

2016-11-01

To investigate radiation dose to phantom testes with and without shielding. A male anthropomorphic pelvis phantom was imaged with thermoluminescent dosimeters (TLDs) placed in the right and left detector holes corresponding to the testes. Ten exposures were made of the pelvis with and without shielding. The exposed TLDs were packaged securely and mailed to the University of Wisconsin Calibration Laboratory for reading and analysis. A t test was calculated for the 2 exposure groups (no shield and shielded) and found to be significant, F = 8.306, P shield was used during pelvic imaging. Using a flat contact shield during imaging of the adult male pelvis significantly reduces radiation dose to the testes. Regardless of the contradictions in the literature on gonadal shielding, the routine practice of shielding adult male gonads during radiographic imaging of the pelvis is a best practice. © 2016 American Society of Radiologic Technologists.

Investigation of the dose rate dependency of the PAGAT gel dosimeter at low dose rates

International Nuclear Information System (INIS)

Zehtabian, M.; Faghihi, R.; Zahmatkesh, M.H.; Meigooni, A.S.; Mosleh-Shirazi, M.A.; Mehdizadeh, S.; Sina, S.; Bagheri, S.

2012-01-01

Medical physicists need dosimeters such as gel dosimeters capable of determining three-dimensional dose distributions with high spatial resolution. To date, in combination with magnetic resonance imaging (MRI), polyacrylamide gel (PAG) polymers are the most promising gel dosimetry systems. The purpose of this work was to investigate the dose rate dependency of the PAGAT gel dosimeter at low dose rates. The gel dosimeter was used for measurement of the dose distribution around a Cs-137 source from a brachytherapy LDR source to have a range of dose rates from 0.97 Gy h −1 to 0.06 Gy h −1 . After irradiation of the PAGAT gel, it was observed that the dose measured by gel dosimetry was almost the same at different distances (different dose rates) from the source, although the points nearer the source had been expected to receive greater doses. Therefore, it was suspected that the PAGAT gel is dose rate dependent at low dose rates. To test this further, three other sets of measurements were performed by placing vials containing gel at different distances from a Cs-137 source. In the first two measurements, several plastic vials were exposed to equal doses at different dose rates. An ionization chamber was used to measure the dose rate at each distance. In addition, three TLD chips were simultaneously irradiated in order to verify the dose to each vial. In the third measurement, to test the oxygen diffusion through plastic vials, the experiment was repeated again using plastic vials in a nitrogen box and glass vials. The study indicates that oxygen diffusion through plastic vials for dose rates lower than 2 Gy h −1 would affect the gel dosimeter response and it is suggested that the plastic vials or (phantoms) in an oxygen free environment or glass vials should be used for the dosimetry of low dose rate sources using PAGAT gel to avoid oxygen diffusion through the vials.

Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

International Nuclear Information System (INIS)

Seki, Y.; Mori, S.

1984-01-01

The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

Effect of bedside shielding on air-kerma rates around gynecologic intracavitary brachytherapy patients containing 226Ra or 137Cs

International Nuclear Information System (INIS)

Papin, P.J.; Ramsey, M.J.; LaFontaine, R.L.; LePage, R.P.

1990-01-01

An anthropomorphic phantom was implanted with 226Ra or 137Cs gynecologic intracavitary brachytherapy sources. Air-kerma rate measurements were taken at 10-cm increments along a horizontal plane from the side of the bed at 50 cm, 87 cm, and 136 cm heights above the floor. Five portable lead shields were placed at the head, at the foot and along one side of the bed and readings were taken again at the corresponding heights above, below and behind the shields. The readings were normalized to 100-mg Ra equivalence, and air-kerma rate curves were drawn allowing for the comparison of 226Ra and 137Cs with and without lead shields. The data demonstrated that the air-kerma rates for 137Cs were reduced more than those for 226Ra with the use of the portable lead shields. There was four times the transmission with 226Ra than with 137Cs. The optimal placement was with the lateral bedside shields proximal to the head and foot closest to the bed, with the middle shield overlapping in back. The shields at the head and foot should extend out and overlap the bedside shields. The level of the sources should be positioned near the bottom of the shields. This information will provide the medical health physicist with an estimate of air-kerma rates for both 226Ra and 137Cs with and without shielding for evaluating personnel exposures as well as the effectiveness of current shielding in relation to radiation protection requirements in adjacent rooms or hallways

Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

International Nuclear Information System (INIS)

Newhauser, W.D.; Titt, U.; Dexheimer, D.; Yan, X.; Nill, S.

2002-01-01

The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent

Applications to shielding design and others of monte carlo method

Energy Technology Data Exchange (ETDEWEB)

Ito, Daiichiro [Mitsui Engineering and Shipbuiding Co., Ltd., Tokyo (Japan)

2001-01-01

One-dimensional or two-dimensional Sn computer code (ANISN, DOT3.5, etc.) and a point attenuation kernel integral code (QAD, etc.) have been used widely for shielding design. Application examples of monte carlo method which could follow precisely the three-dimensional configuration of shielding structure are shown as follow: (1) CASTER cask has a complex structure which consists of a large number of fuel baskets (stainless steel), neutron moderators (polyethylene rods), the body (cast iron), and cooling fin. The R-{theta} model of Sn code DOT3.5 cannot follow closely the complex form of polyethylene rods and fuel baskets. A monte carlo code MORSE is used to ascertain the calculation results of DOT3.5. The discrepancy between the calculation results of DOT3.5 and MORSE was in 10% for dose rate at distance of 1 m from the cask surface. (2) The dose rates of an iron cell at 10 cm above the floor are calculated by the code QAD and the MORSE. The reflected components of gamma ray caused by the auxiliary floor shield (lead) are analyzed by the MORSE. (3) A monte carlo code MCNP4A is used for skyshine evaluation of spent fuel carrier ship 'ROKUEIMARU'. The direct and skyshine components of gamma ray and neutron flux are estimated at each center of engine room and wheel house. The skyshine dose rate of neutron flux is 5-15 times larger than the gamma ray. (M. Suetake)

Bulk Shielding Calculation for 90 .deg. Bending Section of RISP

Energy Technology Data Exchange (ETDEWEB)

Oh, J. H.; Jung, N. S.; Lee, H. S. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Oranj, L. Mokhtari [POSTECH, Pohang (Korea, Republic of); Ko, S. K. [Univ. of Ulsan, Ulsan (Korea, Republic of)

2014-10-15

The charge state of {sup 238}U beams with maximum intensity was 79+ among multi-charge states of 70+ to 89+, which were estimated by using LISE++ code. The bending section consists of twenty four quadrupoles, two dipoles, two two-cell type superconducting RF cavities and eleven slits. The complicated radiation environment is caused by the beam losses occurred normally during the stripping process and when the produced {sup 238}U beams are transported along the beam line. Secondary radiations generated by {sup 238}U beams irradiation are very important for predicting the prompt and residual doses and the radiation damage at the component. The production characteristics of neutron and photon from thin carbon and thick iron were studied to set up the shielding strategy. The dose estimation was done to the pre-designed the tunnel structure. In these calculations, major Monte Carlo codes, PHITS and FLUKA, were used. The present study provided information of shielding analysis for the 90 .deg. bending section of RISP facility. The source term was evaluated to determine fundamental parameter of the shielding analysis using PHITS and FLUKA codes. And the distribution of the dose rate at the outside of thick shielding wall was presented.

Simplification of an MCNP model designed for dose rate estimation

Science.gov (United States)

Laptev, Alexander; Perry, Robert

2017-09-01

A study was made to investigate the methods of building a simplified MCNP model for radiological dose estimation. The research was done using an example of a complicated glovebox with extra shielding. The paper presents several different calculations for neutron and photon dose evaluations where glovebox elements were consecutively excluded from the MCNP model. The analysis indicated that to obtain a fast and reasonable estimation of dose, the model should be realistic in details that are close to the tally. Other details may be omitted.

A theoretical and experimental dose rate study at a multipurpose gamma irradiation facility in Ghana

International Nuclear Information System (INIS)

Sackey, Tracey A.

2015-01-01

Radiation dose rate monitoring out at the Radiation Technology Centre (RTC) of the Ghana Atomic Energy Commission (GAEC) to establish the safety or otherwise of staff at the occupied areas is presented. The facility operates a rectangular source of Co-60 gamma with an having activity of 27.4kCi as at March 2015 and has 14 workers. The aim of the research was determine by means of practical and theoretical evaluations shielding effectiveness of the irradiation chamber. This was to ensure that occupationally exposed workers are not over exposed or their exposures do not exceed the regulatory limits of 7.5μSv/h or 50mSv per annum. The study included dose rate measurements at controlled areas, evaluation of personnel dose history, comparison of experimental and theoretical values and determination of whether the shielding can support a. 18.5PBq (500kCi) Co-60 source. Practical dose rate measurements when the source was in the irradiation position was carried out using a Thermo Scientific Rad-Eye Gamma Survey Meter in the controlled areas of the facility which included the control room, electric room, deionizer room, on top of the roof of irradiation chamber (specifically above the roof plugs) and the two entrances to the irradiation chamber; the personnel door and the goods door. Background reading was found to be 0.08±0.01μSv/h whilst the average dose rates at the two entrances to the irradiation chamber (i e.,- the personnel door and the goods door) were measured to be 0.090μSv/h and 0.109μSv/h respectively. Practical measurements at the roof plugs produced average values of 0.135μSv/h. A particular point on the roof marked as plug-3 produced a relatively higher dose rate of 8.151μSv/h due probably to leakage along the cable to the drive motor. Measurements in the control room, electrical room and deionizer room had average readings of 0.116μSv/h, 0.089μSv/h and 0.614μSv/h respectively. All these average values were below the regulatory limits of 7.5�

An evaluation of in-plane shields during thoracic CT.

Science.gov (United States)

Foley, S J; McEntee, M F; Rainford, L A

2013-08-01

The object of this study was to compare organ dose and image quality effects of using bismuth and barium vinyl in-plane shields with standard and low tube current thoracic CT protocols. A RANDO phantom was scanned using a 64-slice CT scanner and three different thoracic protocols. Thermoluminescent dosemeters were positioned in six locations to record surface and absorbed breast and lung doses. Image quality was assessed quantitatively using region of interest measurements. Scanning was repeated using bismuth and barium vinyl in-plane shields to cover the breasts and the results were compared with standard and reduced dose protocols. Dose reductions were most evident in the breast, skin and anterior lung when shielding was used, with mean reductions of 34, 33 and 10 % for bismuth and 23, 18 and 11 % for barium, respectively. Bismuth was associated with significant increases in both noise and CT attenuation values for all the three protocols, especially anteriorly and centrally. Barium shielding had a reduced impact on image quality. Reducing the overall tube current reduced doses in all the locations by 20-27 % with similar increases in noise as shielding, without impacting on attenuation values. Reducing the overall tube current best optimises dose with minimal image quality impact. In-plane shields increase noise and attenuation values, while reducing anterior organ doses primarily. Shielding remains a useful optimisation tool in CT and barium is an effective alternative to bismuth especially when image quality is of concern.

Shielding effect of snow cover on indoor exposure due to terrestrial gamma radiation

International Nuclear Information System (INIS)

Fujimoto, Kenzo; Kobayashi, Sadayoshi

1988-01-01

Many people in the world live in high latitude region where it snows frequently in winter. When snow covers the ground, it considerably reduces the external exposure from the radiation sources in the ground. Therefore, the evaluation of snow effect on exposure due to terrestrial gamma radiation is necessary to obtain the population dose as well as the absorbed dose in air in snowy regions. Especially the shielding effect on indoor exposure is essentially important in the assessment of population dose since most individuals spend a large portion of their time indoors. The snow effect, however, has been rather neglected or assumed to be the same both indoors and outdoors in the population dose calculation. Snow has been recognized only as a cause of temporal variation of outdoor exposure rate due firstly to radon daughters deposition with snow fall and secondly to the shielding effect of snow cover. This paper describes an approach to the evaluation of shielding effect of snow cover on exposure and introduces population dose calculation as numerical example for the people who live in wooden houses in Japan

Mathematical modeling of the radiation dose received from photons passing over and through shielding walls in a PET/CT suite

DEFF Research Database (Denmark)

Fog, Lotte S; Cormack, John

2010-01-01

Given that the financial cost of shielding PET/CT suites can be substantial, it has become increasingly important to be able to accurately assess the thickness of shielding required for barriers and whether it is necessary to extend such shielding all the way to the ceiling. The overall shielding...... requirement for a PET/CT installation must take into account both 511 keV gamma ray emissions from PET scans and lower energy x-ray scatter from CT scans. This paper deals with the overall impact of emissions from both modalities. Radiation exposure from both scatter over shielding barriers as well...... as transmission through these barriers is taken into account. A series of simulations of the dose received by a person positioned behind a shielding barrier in a typical PET/CT scanning suite were carried out using both Monte Carlo and analytical models. The transmission through lead barriers was found to be very...

An investigation of dose changes for therapeutic kilovoltage x-ray beams with underlying lead shielding

International Nuclear Information System (INIS)

Hill, Robin; Healy, Brendan; Holloway, Lois; Baldock, Clive

2007-01-01

Kilovoltage x-ray beams are used to treat cancer on or close to the skin surface. Many clinical cases use high atomic number materials as shielding to reduce dose to underlying healthy tissues. In this work, we have investigated the effect on both the surface dose and depth doses in a water phantom with lead shielding at depth in the phantom. The EGSnrc Monte Carlo code was used to simulate the water phantom and to calculate the surface doses and depth doses using primary x-ray beam spectra derived from an analytical model. The x-ray beams were in the energy range of 75-135 kVp with field sizes of 2, 5 and 8 cm diameter. The lead sheet was located beneath the water surface at depths ranging from 0.5-7.5 cm. The surface dose decreased as the lead was positioned closer to the water surface and as the field size was increased. The variation in surface dose as a function of x-ray beam energy was only small but the maximum reduction occurred for the 100 kVp x-ray beam. For the 8 cm diameter field with the lead at 1 cm depth and using the 100 kVp x-ray beam, the surface dose was reduced to 0.898 of the surface dose in the water phantom only. Measured surface dose changes, using a Farmer-type ionization chamber, agreed with the Monte Carlo calculated doses. Calculated depth doses in water with a lead sheet positioned below the surface showed that the dose fall-off increased as the lead was positioned closer to the water surface as compared to the depth dose in the water phantom only. Monte Carlo calculations of the total x-ray beam spectrum at the water surface showed that the total fluence decreased due to a reduction in backscatter from within the water and very little backscatter from the lead. The mean energy of the x-ray spectrum varied less than 1 keV, with the lead at 1 cm beneath the water phantom surface. As the Monte Carlo calculations showed good agreement with the measured results, this method can be used to verify surface dose changes in clinical situations

Survivor shielding. Part C. Improvements in terrain shielding

International Nuclear Information System (INIS)

Egbert, Stephen D.; Kaul, Dean C.; Roberts, James A.; Kerr, George D.

2005-01-01

A number of atomic-bomb survivors were affected by shielding provided by terrain features. These terrain features can be a small hill, affecting one or two houses, or a high mountain that shields large neighborhoods. In the survivor dosimetry system, terrain shielding can be described by a transmission factor (TF), which is the ratio between the dose with and without the terrain present. The terrain TF typically ranges between 0.1 and 1.0. After DS86 was implemented at RERF, the terrain shielding categories were examined and found to either have a bias or an excessive uncertainty that could readily be removed. In 1989, an improvement in the terrain model was implemented at RERF in the revised DS86 code, but the documentation was not published. It is now presented in this section. The solution to the terrain shielding in front of a house is described in this section. The problem of terrain shielding of survivors behind Hijiyama mountain at Hiroshima and Konpirasan mountain at Nagasaki has also been recognized, and a solution to this problem has been included in DS02. (author)

SCALE6.1 Hybrid Shielding Methodology For The Spent Fuel Dry Storage

International Nuclear Information System (INIS)

Matijevic, M.; Pevec, D.; Trontl, K.

2015-01-01

The SCALE6.1/MAVRIC hybrid deterministic-stochastic shielding methodology was used for dose rates calculation of the generic spent fuel dry storage installation. The neutron-gamma dose rates around the cask array were calculated over a large problem domain in order to determine the boundary of the controlled area. The FW-CADIS methodology, based on the deterministic forward and adjoint solution over the phase - space, was used for optimized, global Monte Carlo results over the mesh tally. The cask inventory was modeled as homogenized material corresponding to 20 fuel assemblies from a standard mid - sized PWR reactor. The global simulation model was an array of 32 casks in 2 rows with concrete foundations and external air, which makes a large spatial domain for shielding calculations. The dose rates around the casks were determined using FW-CADIS method with weighted adjoint source and mesh tally covering a portion of spatial domain of interest. The conservatively obtained dose rates give the upper boundary, since the activation reduction of sources was not taken into account when sequential filling of the dry storage will start. The effective area of the dry storage installation can be additionally reduced with lowering concrete foundation under the ground, embankment raising, and with extra concrete walls, that would additionally lower the dominant gamma dose rates. (author).

Neutron beam-line shield design for the protein crystallography instrument at the Lujan Center

International Nuclear Information System (INIS)

Russell, G.J.; Pitcher, E.J.; Muhrer, G.; Ferguson, P.D.

2001-01-01

We have developed a very useful methodology for calculating absolute total (neutron plus gamma-ray) dose equivalent rates for use in the design of neutron beam line shields at a spallation neutron source. We have applied this technique to the design of beam line shields for several new materials science instruments being built at the Manuel Lujan Jr. Neutron Scattering Center. These instruments have a variety of collimation systems and different beam line shielding issues. We show here some specific beam line shield designs for the Protein Crystallography Instrument. (author)

Innovative analytical competence. Optimization of shielding components and lifetime activation calculations

Energy Technology Data Exchange (ETDEWEB)

Boehlke, Steffen; Wortmann, Birgit; Aguilar, Arturo Lizon [STEAG Energy Services GmbH, Essen (Germany)

2014-08-15

Shielding and activation calculations always require a high level of engineering competence and powerful hard- and software tools. With the application of current methods often certain limits were reached in the past. The engineering work for optimization efforts regarding complex components with high shielding requirements exceeded the savings in material. With regard to activation the challenges in size of the geometric model and considered operation time rises constantly and pushes computing time beyond reasonable time frames. These challenges require the application of new and faster methodologies. The application of new and innovative methods is presented for a shielding optimization project to decrease the radiation level, to keep the dose rate limits, and to reduce the amount of used shielding material. In a second case a prediction of the activated materials with it's dose distribution in the surrounding area and classification of waste quantities in the structural materials of a nuclear reactor is presented. For the shielding project the preliminary design CAD model was imported into the software tool, several iterations were run and a significantly reduced radiation exposure together with a significant reduction in shieling material were achieved. For the activation calculations it could be demonstrated that it is possible to determine the activation, waste quantities and dose distribution for the structural materials of a nuclear reactor based on lifetime operational data within reasonable time frames.

Simplification of an MCNP model designed for dose rate estimation

Directory of Open Access Journals (Sweden)

Laptev Alexander

2017-01-01

Full Text Available A study was made to investigate the methods of building a simplified MCNP model for radiological dose estimation. The research was done using an example of a complicated glovebox with extra shielding. The paper presents several different calculations for neutron and photon dose evaluations where glovebox elements were consecutively excluded from the MCNP model. The analysis indicated that to obtain a fast and reasonable estimation of dose, the model should be realistic in details that are close to the tally. Other details may be omitted.

Study of some health physics parameters of bismuth-ground granulated blast furnace slag shielding concretes

Energy Technology Data Exchange (ETDEWEB)

Kumar, Sandeep, E-mail: sandeep0078monu@gmail.com [Department of Physics, Punjabi University, Patiala (India); Singh, Sukhpal, E-mail: sukhpal-78@rediffmail.com [Department of Basic and Applied Sciences, Punjabi University, Patiala (India)

2016-05-06

The Bismuth-ground granulated blastfurnace slang (Bi-GGBFS) concrete samples were prepared. The weight percentage of different elements present in Bi-GGBFS Shielding concretewas evaluated by Energy Dispersive X-ray Microanalysis (EDX). The exposure rate and absorbed dose rate characteristics were calculated theoretically for radioactive sources namely {sup 241}Am and {sup 137}Cs. Our calculations reveal that the Bi-GGBFS concretes are effective in shielding material for gamma radiations.

Study of some health physics parameters of bismuth-ground granulated blast furnace slag shielding concretes

Science.gov (United States)

Kumar, Sandeep; Singh, Sukhpal

2016-05-01

The Bismuth-ground granulated blastfurnace slang (Bi-GGBFS) concrete samples were prepared. The weight percentage of different elements present inBi-GGBFS Shielding concretewas evaluated by Energy Dispersive X-ray Microanalysis (EDX). The exposure rate and absorbed dose rate characteristics were calculated theoretically for radioactive sources namely 241Am and 137Cs. Our calculations reveal that the Bi-GGBFS concretes are effective in shielding material for gamma radiations.

ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport

International Nuclear Information System (INIS)

1993-01-01

1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size

Shield evaluation and validation for design and operation of facility for treatment of legacy Intermediate Level Radioactive Liquid Waste (ILW)

International Nuclear Information System (INIS)

Deepa, A.; Jakhete, A.P.; Rathish, K.R.; Saroj, S.K.; Patel, H.S.; Gopalakrishnan, R.K.; Gangadharan, Anand; Singh, Neelima

2014-01-01

An ion exchange treatment facility has been commissioned at PRIX facility, for the treatment of legacy ILW generated at reprocessing plant, Trombay. The treatment system is based on the deployment of selective sorbents for removal of cesium and strontium from ILW. Activity concentration due to beta emitters likely to be processed is of the order of 111-1850 MBq/l. Dose rates in different areas of the facility were evaluated using shielding code and design input. Present work give details of the comparison of dose rates estimated and dose rates measured at various stages of the processing of ILW. At PRIX, the ILW treatment system comprises of shielded IX columns (two cesium and one strontium) housed in a MS cubicle the process lines inlet and outlet of IX treatment system and effluent storage tanks. The MS cubicle, prefilter and piping are housed in a process cell of 500 mm concrete shielding. Effluent storage tanks are outside processing building. Theoretical assessment of expected dose rates were carried out prior to installation of various systems in different areas of PRIX. Dose rate on IX column and MS cubicle for a maximum inventory of 3.7x10 7 MBq of 137 Cs and its contribution in operating gallery was estimated

Application of a dummy eye shield for electron treatment planning

International Nuclear Information System (INIS)

Kang, Sei-Kwon; Park, Soah; Hwang, Taejin; Cheong, Kwang-Ho; Han, Taejin; Kim, Haeyoung; Lee, Me-Yeon; Kim, Kyoung Ju; Oh, Do Hoon; Bae, Hoonsik

2013-01-01

Metallic eye shields have been widely used for near-eye treatments to protect critical regions, but have never been incorporated into treatment plans because of the unwanted appearance of the metal artifacts on CT images. The purpose of this work was to test the use of an acrylic dummy eye shield as a substitute for a metallic eye shield during CT scans. An acrylic dummy shield of the same size as the tungsten eye shield was machined and CT scanned. The BEAMnrc and the DOSXYZnrc were used for the Monte Carlo (MC) simulation, with the appropriate material information and density for the aluminum cover, steel knob and tungsten body of the eye shield. The Pinnacle adopting the Hogstrom electron pencil-beam algorithm was used for the one-port 6-MeV beam plan after delineation and density override of the metallic parts. The results were confirmed with the metal oxide semiconductor field effect transistor (MOSFET) detectors and the Gafchromic EBT2 film measurements. For both the maximum eyelid dose over the shield and the maximum dose under the shield, the MC results agreed with the EBT2 measurements within 1.7%. For the Pinnacle plan, the maximum dose under the shield agreed with the MC within 0.3%; however, the eyelid dose differed by -19.3%. The adoption of the acrylic dummy eye shield was successful for the treatment plan. However, the Pinnacle pencil-beam algorithm was not sufficient to predict the eyelid dose on the tungsten shield, and more accurate algorithms like MC should be considered for a treatment plan. (author)

Estimation of absorbed dose for poor shields under conditions of near-earth space flight

International Nuclear Information System (INIS)

Konyukov, V.V.; Krajnyukov, V.I.; Trufanov, A.I.

1995-01-01

Estimation of electron absorbed dose in materials of a space vehicle for poor shields under conditions of near-earth space flight is carried out. Impact of power and angular distribution of incidence electrons and radiation scattering processes under conditions of complex geometry and multitude of materials of flight vehicle elements and nodes is studied through simulator model by example of isolating layer of aluminium-polyethylene assembly. 3 refs.; 2 figs

Postoperative irradiation for the prevention of heterotopic bone: Analysis of different dose schedules and shielding considerations

International Nuclear Information System (INIS)

Blount, L.H.; Thomas, B.J.; Tran, L.; Selch, M.T.; Sylvester, J.E.; Parker, R.G.

1990-01-01

Ninety-seven high risk hips were irradiated postoperatively for prevention of heterotopic bone (HTB) in the UCLA Department of Radiation Oncology from 1980 to 1988. Ninety-two hips in 82 patients were eligible for analysis with a minimum follow-up of 2 months and a median follow-up of 10 months. Forty-nine of the hips had porous coated ingrowth prostheses. From 1980 to 1986, 2 Gy fractions were used to deliver 20 Gy (8 hips), 12 Gy (1 hip), and 10 Gy (27 hips). Since December of 1986, 38 hips received 8 Gy in two increments and 18 hips received a single 7 Gy fraction. All porous ingrowth components were shielded with custom blocks. Six out of 92 hips developed clinically significant. There was one clinically significant failure in 78 hips (1.3%) when irradiation was initiated before post-operative day (POD) No.6 and shielding was properly placed. One clinical failure occurred in 38 hips which received 8 Gy in two increments. One clinical failure occurred out of the 18 hips treated with 7 Gy in one fraction. This failure could be related to block malposition. There were four clinical failures in the 36 hips treated with 2 Gy fractions to total doses of 10 Gy, 12 Gy, or 20 Gy. Three of these failures were associated with initiation of treatment after POD No.5, and the fourth was related to block malposition. Unshielded trochanteric osteotomies resulted in five migrations and seven fibrous unions for a total non-osseous union rate of 12/36 (33%). Shielding of the remaining 28 trochanteric osteotomies resulted in a non-osseous union rate of 7%. There were no failures of union of components, and the only side effects noted in the series were the five trochanteric migrations. In conclusion, the use of 8 Gy in two increments or 7 Gy in one fraction was found to be as efficacious as conventional 2 Gy fractionation schemes with no increase in side effects

A perturbation technique for shield weight minimization

International Nuclear Information System (INIS)

Watkins, E.F.; Greenspan, E.

1993-01-01

The radiation shield optimization code SWAN (Ref. 1) was originally developed for minimizing the thickness of a shield that will meet a given dose (or another) constraint or for extremizing a performance parameter of interest (e.g., maximizing energy multiplication or minimizing dose) while maintaining the shield volume constraint. The SWAN optimization process proved to be highly effective (e.g., see Refs. 2, 3, and 4). The purpose of this work is to investigate the applicability of the SWAN methodology to problems in which the weight rather than the volume is the relevant shield characteristic. Such problems are encountered in shield design for space nuclear power systems. The investigation is carried out using SWAN with the coupled neutron-photon cross-section library FLUNG (Ref. 5)

OPAL shield design performance assessment. Comparison of measured dose rates against the corresponding design calculated values. A designer perspective

Energy Technology Data Exchange (ETDEWEB)

Brizuela, Martin; Albornoz, Felipe [INVAP SE, Av. Cmte. Piedrabuena, Bariloche (Argentina)

2012-03-15

A comparison of OPAL shielding calculations against measurements carried out during Commissioning, is presented for relevant structures such as the reactor block, primary shutters, neutron guide bunker, etc. All the results obtained agree very well with the measured values and contribute to establish the confidence on the calculation tools (MCNP4, DORT, etc.) and methodology used for shielding design. (author)

Performances of Kevlar and Polyethylene as radiation shielding on-board the International Space Station in high latitude radiation environment.

Science.gov (United States)

Narici, Livio; Casolino, Marco; Di Fino, Luca; Larosa, Marianna; Picozza, Piergiorgio; Rizzo, Alessandro; Zaconte, Veronica

2017-05-10

Passive radiation shielding is a mandatory element in the design of an integrated solution to mitigate the effects of radiation during long deep space voyages for human exploration. Understanding and exploiting the characteristics of materials suitable for radiation shielding in space flights is, therefore, of primary importance. We present here the results of the first space-test on Kevlar and Polyethylene radiation shielding capabilities including direct measurements of the background baseline (no shield). Measurements are performed on-board of the International Space Station (Columbus modulus) during the ALTEA-shield ESA sponsored program. For the first time the shielding capability of such materials has been tested in a radiation environment similar to the deep-space one, thanks to the feature of the ALTEA system, which allows to select only high latitude orbital tracts of the International Space Station. Polyethylene is widely used for radiation shielding in space and therefore it is an excellent benchmark material to be used in comparative investigations. In this work we show that Kevlar has radiation shielding performances comparable to the Polyethylene ones, reaching a dose rate reduction of 32 ± 2% and a dose equivalent rate reduction of 55 ± 4% (for a shield of 10 g/cm 2 ).

Tank Z-361 dose rate calculations

International Nuclear Information System (INIS)

Richard, R.F.

1998-01-01

Neutron and gamma ray dose rates were calculated above and around the 6-inch riser of tank Z-361 located at the Plutonium Finishing Plant. Dose rates were also determined off of one side of the tank. The largest dose rate 0.029 mrem/h was a gamma ray dose and occurred 76.2 cm (30 in.) directly above the open riser. All other dose rates were negligible. The ANSI/ANS 1991 flux to dose conversion factor for neutrons and photons were used in this analysis. Dose rates are reported in units of mrem/h with the calculated uncertainty shown within the parentheses

Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

Energy Technology Data Exchange (ETDEWEB)

SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

2000-07-01

Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

Bayesian estimation of dose rate effectiveness

International Nuclear Information System (INIS)

Arnish, J.J.; Groer, P.G.

2000-01-01

A Bayesian statistical method was used to quantify the effectiveness of high dose rate 137 Cs gamma radiation at inducing fatal mammary tumours and increasing the overall mortality rate in BALB/c female mice. The Bayesian approach considers both the temporal and dose dependence of radiation carcinogenesis and total mortality. This paper provides the first direct estimation of dose rate effectiveness using Bayesian statistics. This statistical approach provides a quantitative description of the uncertainty of the factor characterising the dose rate in terms of a probability density function. The results show that a fixed dose from 137 Cs gamma radiation delivered at a high dose rate is more effective at inducing fatal mammary tumours and increasing the overall mortality rate in BALB/c female mice than the same dose delivered at a low dose rate. (author)

Validation of SCALE code package on high performance neutron shields

International Nuclear Information System (INIS)

Bace, M.; Jecmenica, R.; Smuc, T.

1999-01-01

The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

The shielding calculation for the CN guide shielding assembly in HANARO

International Nuclear Information System (INIS)

Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.

2006-01-01

The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO

The measurement of exposure rate and absorbed dose for photons of 6 to 7 MeV

International Nuclear Information System (INIS)

Birchall, I.

1978-01-01

Using the 16 N source at the Universities Reactor Centre, which provides a source of high energy photons of 6 and 7 MeV, experiments have been carried out to provide a calibration point for secondary standard and general survey instruments. The source was designed primarily for shielding experiments and suffers from a number of disadvantages if precise dose or exposure data are required. In particular the exposure rate is low, the source is physically large and the energy spectrum is contaminated with lower energy photons. A summary of the basic techniques is given because of the inherent difficulties in dose measurements particularly for high energy photons. (author)

Study of some health physics parameters of bismuth-ground granulated blast furnace slag shielding concretes

International Nuclear Information System (INIS)

Kumar, Sandeep; Singh, Sukhpal

2016-01-01

The Bismuth-ground granulated blastfurnace slang (Bi-GGBFS) concrete samples were prepared. The weight percentage of different elements present in Bi-GGBFS Shielding concretewas evaluated by Energy Dispersive X-ray Microanalysis (EDX). The exposure rate and absorbed dose rate characteristics were calculated theoretically for radioactive sources namely "2"4"1Am and "1"3"7Cs. Our calculations reveal that the Bi-GGBFS concretes are effective in shielding material for gamma radiations.

Accelerator shield design of KIPT neutron source facility

International Nuclear Information System (INIS)

Zhong, Z.; Gohar, Y.

2013-01-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

MO-D-213-07: RadShield: Semi- Automated Calculation of Air Kerma Rate and Barrier Thickness

International Nuclear Information System (INIS)

DeLorenzo, M; Wu, D; Rutel, I; Yang, K

2015-01-01

Purpose: To develop the first Java-based semi-automated calculation program intended to aid professional radiation shielding design. Air-kerma rate and barrier thickness calculations are performed by implementing NCRP Report 147 formalism into a Graphical User Interface (GUI). The ultimate aim of this newly created software package is to reduce errors and improve radiographic and fluoroscopic room designs over manual approaches. Methods: Floor plans are first imported as images into the RadShield software program. These plans serve as templates for drawing barriers, occupied regions and x-ray tube locations. We have implemented sub-GUIs that allow the specification in regions and equipment for occupancy factors, design goals, number of patients, primary beam directions, source-to-patient distances and workload distributions. Once the user enters the above parameters, the program automatically calculates air-kerma rate at sampled points beyond all barriers. For each sample point, a corresponding minimum barrier thickness is calculated to meet the design goal. RadShield allows control over preshielding, sample point location and material types. Results: A functional GUI package was developed and tested. Examination of sample walls and source distributions yields a maximum percent difference of less than 0.1% between hand-calculated air-kerma rates and RadShield. Conclusion: The initial results demonstrated that RadShield calculates air-kerma rates and required barrier thicknesses with reliable accuracy and can be used to make radiation shielding design more efficient and accurate. This newly developed approach differs from conventional calculation methods in that it finds air-kerma rates and thickness requirements for many points outside the barriers, stores the information and selects the largest value needed to comply with NCRP Report 147 design goals. Floor plans, parameters, designs and reports can be saved and accessed later for modification and recalculation

MO-D-213-07: RadShield: Semi- Automated Calculation of Air Kerma Rate and Barrier Thickness

Energy Technology Data Exchange (ETDEWEB)

DeLorenzo, M [Oklahoma University Health Sciences Center, Oklahoma City, OK (United States); Wu, D [University of Oklahoma Health Sciences Center, Oklahoma City, Ok (United States); Rutel, I [University of Oklahoma Health Science Center, Oklahoma City, OK (United States); Yang, K [Massachusetts General Hospital, Boston, MA (United States)

2015-06-15

Purpose: To develop the first Java-based semi-automated calculation program intended to aid professional radiation shielding design. Air-kerma rate and barrier thickness calculations are performed by implementing NCRP Report 147 formalism into a Graphical User Interface (GUI). The ultimate aim of this newly created software package is to reduce errors and improve radiographic and fluoroscopic room designs over manual approaches. Methods: Floor plans are first imported as images into the RadShield software program. These plans serve as templates for drawing barriers, occupied regions and x-ray tube locations. We have implemented sub-GUIs that allow the specification in regions and equipment for occupancy factors, design goals, number of patients, primary beam directions, source-to-patient distances and workload distributions. Once the user enters the above parameters, the program automatically calculates air-kerma rate at sampled points beyond all barriers. For each sample point, a corresponding minimum barrier thickness is calculated to meet the design goal. RadShield allows control over preshielding, sample point location and material types. Results: A functional GUI package was developed and tested. Examination of sample walls and source distributions yields a maximum percent difference of less than 0.1% between hand-calculated air-kerma rates and RadShield. Conclusion: The initial results demonstrated that RadShield calculates air-kerma rates and required barrier thicknesses with reliable accuracy and can be used to make radiation shielding design more efficient and accurate. This newly developed approach differs from conventional calculation methods in that it finds air-kerma rates and thickness requirements for many points outside the barriers, stores the information and selects the largest value needed to comply with NCRP Report 147 design goals. Floor plans, parameters, designs and reports can be saved and accessed later for modification and recalculation

Radiation dose to technologists per nuclear medicine examination and estimation of annual dose.

Science.gov (United States)

Bayram, Tuncay; Yilmaz, A Hakan; Demir, Mustafa; Sonmez, Bircan

2011-03-01

Conventional diagnostic nuclear medicine applications have been continuously increasing in most nuclear medicine departments in Turkey, but to our knowledge no one has studied the doses to technologists who perform nuclear medicine procedures. Most nuclear medicine laboratories do not have separate control rooms for technologists, who are quite close to the patient during data acquisition. Technologists must therefore stay behind lead shields while performing their task if they are to reduce the radiation dose received. The aim of this study was to determine external radiation doses to technologists during nuclear medicine procedures with and without a lead shield. Another aim was to investigate the occupational annual external radiation doses to Turkish technologists. This study used a Geiger-Müller detector to measure dose rates to technologists at various distances from patients (0.25, 0.50, 1, and 2 m and behind a lead shield) and determined the average time spent by technologists at these distances. Deep-dose equivalents to technologists were obtained. The following conventional nuclear medicine procedures were considered: thyroid scintigraphy performed using (99m)Tc pertechnetate, whole-body bone scanning performed using (99m)Tc-methylene diphosphonate, myocardial perfusion scanning performed using (99m)Tc-methoxyisobutyl isonitrile, and (201)Tl (thallous chloride) and renal scanning performed using (99m)Tc-dimercaptosuccinic acid. The measured deep-dose equivalent to technologists per procedure was within the range of 0.13 ± 0.05 to 0.43 ± 0.17 μSv using a lead shield and 0.21 ± 0.07 to 1.01 ± 0.46 μSv without a lead shield. Also, the annual individual dose to a technologist performing only a particular scintigraphic procedure throughout a year was estimated. For a total of 95 clinical cases (71 patients), effective external radiation doses to technologists were found to be within the permissible levels. This study showed that a 2-mm lead shield

Shielding design for PWR in France

International Nuclear Information System (INIS)

Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

1983-05-01

Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

Reducing the radiation dose to the eye lens region during CT brain examination: the potential beneficial effect of the combined use of bolus and a bismuth shield

International Nuclear Information System (INIS)

Lai, C.W.K.; Chan, T.P.; Cheung, H.Y.; Wong, T.H.

2015-01-01

Objective: Computed Tomography (CT) is the leading contributor to medical exposure to ionizing radiation. Although the use of CT brain scans for patients with head injuries and convulsions has shown a tremendous growth, it has raised substantial concerns in the general public because of the risk of radiation-induced cataracts: the current available strategies to reduce the radiation dose to the eye lens region are limited. Therefore, the present research project was initiated with the aim of evaluating the potential benefit of the combined use of bolus and a bismuth shield on reducing the radiation dose to the eye lens region during CT brain examination. Materials and methods: We conducted a series of phantom studies to measure the entrance surface dose (ESD) that is delivered to the eye lens region during CT brain examination under the effect of different scanning and shielding setups. Results: Our results indicated, during CT brain examination: (1) a drastic reduction of 92.5% in the ESD to the eye lens region was found when the CT gantry was tilted from 0 deg. (overall ESD = 30.7 mGy) to 30 deg. cranially (overall ESD = 2.4 mGy), and (2) when the CT gantry was positioned at 0 deg. (the common practice in the clinical setting), the setups with the application of a) a bismuth shield, b) a bismuth shield with a face shield (air gap), c) a bismuth shield with bolus, and d) a bismuth shield with bolus and an air gap can result in an acceptable level of image quality with a smaller overall ESD delivered to the eye lens region (overall ESD = 23.2 mGy, 24 mGy, 21 mGy and 19.9 mGy, respectively) than the setup without the bismuth shield applied (overall ESD = 30.7 mGy). Conclusion: When the primary beam scanning through the eye lens region is unavoidable during CT brain examination, the combined use of a bismuth shield with bolus and a face shield is an easy-to-use and inexpensive shielding setup to reduce the radiation dose delivered to the eye lens region while

Evaluation of the efficiency of bismuth breast shield in CT chest paediatric examinations

International Nuclear Information System (INIS)

Sevillano, D.; Espana, M. L.; Castro, P.; Minguez, C.; Albi, G.; Garcia, E.; Lopez Franco, P.

2006-01-01

The aim of this study is to evaluate the efficiency of bismuth breast shield in CT chest paediatric examinations when automatic exposure control techniques are used. The influence in the radiation dose and the image quality has been evaluated with and without the breast shield in the scoutview. In addition the radiation dose in shielded and non-shielded areas has been compared. Measurements were made in a 16 cm diameter cylindrical PMMA phantom simulating a newborn, and older children were simulated controlling the maximum intensities allowed by the automatic exposure control system AutomA. The highest dose reduction (59%) was obtained with AutomA system and when the breast shield is not used in the scoutview. This reduction in the radiation dose does not mean a significant increase of noise level. The use of the breast shield in the scoutview yielded an increase in the radiation dose in non-shielded areas. The use of bismuth breast shield is recommended only after the scoutview in order to optimise the radiation dose in CT chest paediatric examination when using automatic exposure control AutomA. (Author)

Dose/dose-rate responses of shrimp larvae to UV-B radiation

International Nuclear Information System (INIS)

Damkaer, D.M.

1981-01-01

Previous work indicated dose-rate thresholds in the effects of UV-B on the near-surface larvae of three shrimp species. Additional observations suggest that the total dose response varies with dose-rate. Below 0.002 Wm -2 sub([DNA]) irradiance no significant effect is noted in activity, development, or survival. Beyond that dose-rate threshold, shrimp larvae are significantly affected if the total dose exceeds about 85 Jm -2 sub([DNA]). Predictions cannot be made without both the dose-rate and the dose. These dose/dose-rate thresholds are compared to four-year mean dose/dose-rate solar UV-B irradiances at the experimental site, measured at the surface and calculated for 1 m depth. The probability that the shrimp larvae would receive lethal irradiance is low for the first half of the season of surface occurrence, even with a 44% increase in damaging UV radiation. (orig.)

Dose/dose-rate responses of shrimp larvae to UV-B radiation

Energy Technology Data Exchange (ETDEWEB)

Damkaer, D.M.; Dey, D.B.; Heron, G.A.

1981-01-01

Previous work indicated dose-rate thresholds in the effects of UV-B on the near-surface larvae of three shrimp species. Additional observations suggest that the total dose response varies with dose-rate. Below 0.002 Wm/sup -2/sub((DNA)) irradiance no significant effect is noted in activity, development, or survival. Beyond that dose-rate threshold, shrimp larvae are significantly affected if the total dose exceeds about 85 Jm/sup -2/sub((DNA)). Predictions cannot be made without both the dose-rate and the dose. These dose/dose-rate thresholds are compared to four-year mean dose/dose-rate solar UV-B irradiances at the experimental site, measured at the surface and calculated for 1 m depth. The probability that the shrimp larvae would receive lethal irradiance is low for the first half of the season of surface occurrence, even with a 44% increase in damaging UV radiation.

Photon dose estimation from ultraintense laser–solid interactions and shielding calculation with Monte Carlo simulation

International Nuclear Information System (INIS)

Yang, Bo; Qiu, Rui; Li, JunLi; Lu, Wei; Wu, Zhen; Li, Chunyan

2017-01-01

When a strong laser beam irradiates a solid target, a hot plasma is produced and high-energy electrons are usually generated (the so-called “hot electrons”). These energetic electrons subsequently generate hard X-rays in the solid target through the Bremsstrahlung process. To date, only limited studies have been conducted on this laser-induced radiological protection issue. In this study, extensive literature reviews on the physics and properties of hot electrons have been conducted. On the basis of these information, the photon dose generated by the interaction between hot electrons and a solid target was simulated with the Monte Carlo code FLUKA. With some reasonable assumptions, the calculated dose can be regarded as the upper boundary of the experimental results over the laser intensity ranging from 10 19 to 10 21 W/cm 2 . Furthermore, an equation to estimate the photon dose generated from ultraintense laser–solid interactions based on the normalized laser intensity is derived. The shielding effects of common materials including concrete and lead were also studied for the laser-driven X-ray source. The dose transmission curves and tenth-value layers (TVLs) in concrete and lead were calculated through Monte Carlo simulations. These results could be used to perform a preliminary and fast radiation safety assessment for the X-rays generated from ultraintense laser–solid interactions. - Highlights: • The laser–driven X-ray ionizing radiation source was analyzed in this study. • An equation to estimate the photon dose based on the laser intensity is given. • The shielding effects of concrete and lead were studied for this new X-ray source. • The aim of this study is to analyze and mitigate the laser–driven X-ray hazard.

A Combined Shielding Design for a Neutron Generator and a Linear Accelerator at Soreq NRC

International Nuclear Information System (INIS)

Epstein, L.

2014-01-01

A new radiography facility is designed at Soreq NRC. The facility will hold a neutron generator that produces 1.73·109 n/s with an energy of 14 MeV and a linear accelerator that accelerates electrons to an energy of 9 MeV. The two radiation sources will be installed in 2 separate laboratories that will be built in an existing building. Each laboratory will have its own machine and control room. The dose rates around the sources were calculated using the FLUKA Monte Carlo code(1,2). The annual doses were calculated in several regions around the generator and the accelerator laboratories in accordance with the occupancy in each area. The calculated annual doses were compared with the dose limits specified in the Safety at Work Regulations(3) and the IAEC Standard for Protection against Ionizing Radiation. The shielding was designed to comply with the following dose constraints: 0.3 mSv/y for members of the public and 2 mSv/y for radiation workers. Each radiation source is planned to produce radiation for a maximum of 500 hours per year. The dose rate in the direct beam of the accelerator is 30 Gy/min at 1 m from the source and it will be surrounded by a collimator with an opening of 30N-tilde horizontally and 2 mm vertically, 3 m from the radiation source. The leakage radiation dose will not be greater than 1.5 mGy/min (0.005% of the direct beam, according to the manufacturer). The leakage radiation will be produced isotropically. The neutron generator will be surrounded by a shielding made of a 10 cm iron cylinder (density 7.87 g/cm3), surrounded by 50 cm of borated polyethylene (atomic percent: H (13.8%), C (82.2%), B (4%), density: 0.92 g/cm3) and 5 cm of lead (density 11.35 g/cm3). The neutron generator shielding was not designed or required in the present shielding design but was considered in the shielding calculations

Study of the radiation scattered and produced by concrete shielding of radiotherapy rooms and its effects on equivalent doses in patients' organs

International Nuclear Information System (INIS)

Braga, K.L.; Rebello, W.F.; Andrade, E.R.; Gavazza, S.; Medeiros, M.P.C.; Mendes, R.M.S.; Gomes, R.G.; Silva, M.G.; Thalhofer, J.L.; Silva, A.X.; Santos, R.F.G.

2015-01-01

Within a radiotherapy room, in addition to the primary beam, there is also secondary radiation due to the leakage of the accelerator head and the radiation scattering from room objects, patient and even the room's shielding itself, which is projected to protect external individuals disregarding its effects on the patient. This work aims to study the effect of concrete shielding wall over the patient, taking into account its contribution on equivalent doses. The MCNPX code was used to model the linear accelerator Varian 2100/2300 C/D operating at 18MeV, with MAX phantom representing the patient undergoing radiotherapy treatment for prostate cancer following Brazilian Institute of Cancer four-fields radiation application protocol (0°, 90°, 180° and 270°). Firstly, the treatment was patterned within a standard radiotherapy room, calculating the equivalent doses on patient's organs individually. In a second step, this treatment was modeled withdrawing the walls, floor and ceiling from the radiotherapy room, and then the equivalent doses calculated again. Comparing these results, it was found that the concrete has an average shielding contribution of around 20% in the equivalent dose on the patient's organs. (author)

The use of radiological characterisation in support of the design and build of a new facility in an area of elevated dose rate - 16009

International Nuclear Information System (INIS)

Hughes, Karl; Thornley, David; Pienkowski, Czeslaw

2009-01-01

A new building, a maintenance facility, is to be constructed in the 'separation area' of the Sellafield Site. Sellafield is a complex and busy nuclear facility covering about two square miles in the north-west of the United Kingdom. The facility, which is to provide necessary storage and maintenance functionality to support bulk waste retrievals from legacy silos, is being built in close proximity to spent fuel storage ponds, intermediate level legacy waste stores and a pipe-bridge carrying active materials. In addition, the site is adjacent to a main pedestrian and vehicle thoroughfare and a railway line used for the regular transfer of nuclear materials. The facility itself is to be built on the site of a recently demolished active facility. The ambient gamma dose rates in the construction area have been measured to be typically 5 to 50 μSv/hr. Although there are known specific sources of dose rate close to the construction site, the relative contributions of each within the envelope of the new facility are unknown. This paper describes how a series of gamma dose rate measurements, gamma spectroscopy measurements and gamma-ray imaging surveys, using RadScan R 800, have been used to better understand the origins of the dose rates at a number of key locations within the area. The results of the survey are being used to assess the requirements for shielding within the structure of the facility in order to reduce the dose to personnel that will work there when it is operational. The results are also being used to identify whether any localised shielding would prove beneficial for reduction of the dose both during construction and occupancy. The unique mix of facilities surrounding the site have meant that the contributions to the dose rate come not only from adjacent facilities in the form of unscattered, higher energy, penetrating radiation, but also in the form of lower energy scattered radiation, both from radiation that has passed through shielding and also in