Personnel Dose Research Articles

Modeling of the radiation doses during dismantling of RBMK-1500 reactor emergency core cooling system large diameter pipes

Personnel radiation safety is one of the most important issues during the dismantling of nuclear installations. In this paper, results of modeling radiation doses during the dismantling of the large diameter pipes from the emergency core cooling system of RBMK-1500 reactor at Ignalina NPP are presented. The effective doses to the workers are modeled for four dismantling alternatives in order to propose the optimal dismantling approach. The impacts of the cutting technology and individual respiration protection on effective doses are analyzed. The total effective personnel doses are obtained by summing the effective personnel doses from various sources of exposure, i.e., direct radiation from radioactive equipment, internal radiation due to inhalation and ingestion of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting. The collective effective doses and their uncertainties are analyzed using VISIPLAN and MATLAB codes.

Replacement of Tritiated Water from Irradiated Fuel Storage Bay

Recently, AECL developed a novel method to reduce tritium emissions (to groundwater) and personnel doses at the NRU (National Research Universal) reactor irradiated fuel storage bay (also known as rod or spent fuel bay) through a water swap process. The light water in the fuel bay had built up tritium that had been transferred from the heavy water moderator through normal fuel transfers. A demonstration of the method was done that involved Computational Fluid Dynamics (CFD) modeling of the swap process and a test program that showed excellent agreement with model prediction for the effective removal of almost all the tritium with a minimal water volume. Building on the successful demonstration, AECL fabricated, installed, commissioned and operated a full-scale system to perform a water swap. This paper describes the results of the successful full-scale water swap operation, which achieved a tritium removal efficiency of ~96%

Occupational Radiation Exposure During Endovascular Aortic Repair.

The aim of the study was to evaluate the radiation exposure to operating room personnel and to assess determinants for high personal doses during endovascular aortic repair. Occupational radiation exposure was prospectively evaluated during 22 infra-renal aortic repair procedures (EVAR), 11 thoracic aortic repair procedures (TEVAR), and 11 fenestrated or branched aortic repair procedures (FEVAR). Real-time over-lead dosimeters attached to the left breast pocket measured personal doses for the first operators (FO) and second operators (SO), radiology technicians (RT), scrub nurses (SN), anesthesiologists (AN), and non-sterile nurses (NSN). Besides protective apron and thyroid collar, no additional radiation shielding was used. Procedural dose area product (DAP), iodinated contrast volume, fluoroscopy time, patient's body weight, and C-arm angulation were documented. Average procedural FO dose was significantly higher during FEVAR (0.34 ± 0.28 mSv) compared to EVAR (0.11 ± 0.21 mSv) and TEVAR (0.06 ± 0.05 mSv; p = 0.003). Average personnel doses were 0.17 ± 0.21 mSv (FO), 0.042 ± 0.045 mSv (SO), 0.019 ± 0.042 mSv (RT), 0.017 ± 0.031 mSv (SN), 0.006 ± 0.007 mSv (AN), and 0.004 ± 0.009 mSv (NSN). SO and AN doses were strongly correlated with FO dose (p = 0.003 and p < 0.001). There was a significant correlation between FO dose and procedural DAP (R = 0.69, p < 0.001), iodinated contrast volume (R = 0.67, p < 0.001) and left-anterior C-arm projections >60° (p = 0.02), and a weak correlation with fluoroscopy time (R = 0.40, p = 0.049). Average FO dose was a factor four higher than SO dose. Predictors for high personal doses are procedural DAP, iodinated contrast volume, and left-anterior C-arm projections greater than 60°.

Potential hazard due to induced radioactivity secondary to radiotherapy: the report of task group 136 of the American Association of Physicists in Medicine.

External-beam radiation therapy mostly uses high-energy photons (x-rays) produced by medical accelerators, but many facilities now use proton beams, and a few use fast-neutron beams. High-energy photons offer several advantages over lower-energy photons in terms of better dose distributions for deep-seated tumors, lower skin dose, less sensitivity to tissue heterogeneities, etc. However, for beams operating at or above 10 MV, some of the materials in the accelerator room and the radiotherapy patient become radioactive due primarily to photonuclear reactions and neutron capture, exposing therapy staff and patients to unwanted radiation dose. Some recent advances in radiotherapy technology require treatments using a higher number of monitor units and monitor-unit rates for the same delivered dose, and compared to the conventional treatment techniques and fractionation schemes, the activation dose to personnel can be substantially higher. Radiotherapy treatments with proton and neutron beams all result in activated materials in the treatment room. In this report, the authors review critically the published literature on radiation exposures from induced radioactivity in radiotherapy. They conclude that the additional exposure to the patient due to induced radioactivity is negligible compared to the overall radiation exposure as a part of the treatment. The additional exposure to the staff due to induced activity from photon beams is small at an estimated level of about 1 to 2 mSv y. This is well below the allowed occupational exposure limits. Therefore, the potential hazard to staff from induced radioactivity in the use of high-energy x-rays is considered to be low, and no specific actions are considered necessary or mandatory. However, in the spirit of the "As Low as Reasonably Achievable (ALARA)" program, some reasonable steps are recommended that can be taken to reduce this small exposure to an even lower level. The dose reduction strategies suggested should be followed only if these actions are considered reasonable and practical in the individual clinics. Therapists working with proton beam and neutron beam units handle treatment devices that do become radioactive, and they should wear extremity monitors and make handling apertures and boluses their last task upon entering the room following treatment. Personnel doses from neutron-beam units can approach regulatory limits depending on the number of patients and beams, and strategies to reduce doses should be followed.

Portal monitor characterization for internally and externally deposited radionuclides.

In this evaluation, both the internal and external radionuclide detection efficiencies for a portal monitor were evaluated as a function of photon energy using an anthropomorphic phantom. Pass-through and static measurements were completed using 241Am, 57Co, 133Ba, 137Cs, 60Co, 109Cd, and 54Mn in various locations both external and internal to the phantom. Other parameters, such as single detector uniformity, total detector uniformity, background linearity, and activity linearity have been analyzed. It was found that the minimum detectable activity for internally deposited 137Cs in the abdomen was approximately ten times higher for pass-through versus static measurements. Additionally, it was found that the minimum detectable activity for 137Cs in the abdominal region for both internal and external pass-through scenarios are nearly equivalent. In general, if the expected radionuclide source term is primarily non-transuranic, the pass-through mode offers sufficient sensitivity to identify potential overexposures while providing much greater personnel throughput. However, minimum detectable committed effective doses for transuranics such as 241Am, show potential for personnel over exposure if the radionuclide mixture contains a significant fraction of transuranics. It is therefore recommended that nuclear facilities evaluate their radionuclide source term in order to bound potential personnel doses.

New scheme for dual readout of dose in polycrystalline Li2B4O7 irradiated with synchrotron X-rays from Indus-2

A new thermoluminescent (TL) dosimeter readout scheme was developed which allows for the measurement of both the single incident dose and the accumulated dose from multiple irradiations, through a novel dual readout procedure. Undoped, Mn doped and Cu doped polycrystalline Li2B4O7 pallets were irradiated by synchrotron X-rays from Indus-2 with different dose. Thermoluminescence glow curves were taken at various heating rates. Two glow peaks were observed in each sample. TL intensities of different samples were compared and it was found that TL counts are more in Cu doped Li2B4O7 sample. TL counts under each peak were calculated and a fixed ratio for two peaks was obtained. A new method of reading out the two peaks separately was formalized, allowing for greater flexibility in the use of TL based personnel dose monitoring devices.

Medical workers operating in nuclear medicine vs PET/CT: Radiation exposure comparison

s / Physica Medica 30 (2014) e16ee44 e31 Background: Long-lived radioactive impurities in radiopharmaceuticals administered at time of calibration usually have minor impact on image quality and the uncertainty of the activity measurement. However, such impurities may result to non-negligible additional radiation burden to patients, personnel, relatives, caregivers and the public, and may trigger innocent alarms in radiation portal monitors. In addition, the presence of certain levels of impurities may impose modifications in waste management policies. The aim of the present study was to identify the most important long-lived impurities, quantify their amount in the patients’ body and address radiation protection issues. Methods and materials: Initial analysis of the radiopharmaceuticals used at the Ioannina University Hospital indicated substantial amounts of longlived (T1/2> 14 d) photon emitting impurities only in 123I-ioflupene (DaTSCAN, GE) and 153Sm-EDMTP (Quadramet, IBA). Spectroscopic analysis of the residuals of at least ten administrations per agent was carried out at least five times per sample. A prototype shadow-shield whole body counter equipped with 16 NaI(Tl) detectors was used to measure the activity present in at least 7 seven patients per agent, 2 to 5 five weeks postadministration. The in vivo activity measurements were coupled with external dose rate measurements using a portable system with spectroscopic capabilities. Results: 121Te, 123Te and 125I were the main photon emitting impurities in 123I-ioflupene, resulting in a median 30 nSv/h external Hp(10) rate and a 3 kBq 121Te body burden. 152Eu, 152Eu and 156Eu were the main impurities in 153Sm-EDMTP, resulting in amedian 300 nSv/h external rate, mainly due to the 152Eu (T1/2:13.5 y) and 154Eu (T1/2:8.6 y) presence in the body, 25 and 33 kBq, respectively. Thus patient and waste management protocols were modified accordingly. Conclusions: Both patient and waste management have to take into account the long-lived impurities present in some of the agents used in daily clinical practice. MEDICAL WORKERS OPERATING IN NUCLEAR MEDICINE VS PET/CT: RADIATION EXPOSURE COMPARISON K. Dalianis , J. Malamitsi , K. Gogos , R. Efthimiadou , J. Andreou , V. Prassopoulos . a PET/CT Department Hygeia SA, Athens, Greece; Medical Physics Laboratory School of Medicine, University of Athens, Greece Purpose: Personnel monitoring results provide information on routine radiation exposure, assist in work planning and allow control of the workplace. The aim of this study was to compare the gamma dose received by dedicated medical workers operating in the first PET/CT department in Greece and also by dedicated medical workers operating in conventional Nuclear Medicine procedures in the same center. PET/CT studies are restricted to the use of 18F fluorodeoxyglucose (FDG). In addition Tc-99m, Tl-201, Ga-67 and I-131 are the radiotracers mostly used in our Nuclear Medicine department. Method: To estimate the effective dose from external exposure, all 9 members of the staff (2 nurses, 2 medical physicists, 5 technologists) had TLD badges worn at the upper pocket of their overall and digital dosimeters worn at the side pocket. Nurses and Medical Physicists also had TLD rings. The nurses and technologists 1,2 are working only in the PET/CT department, while technologists 3,4,5 are operating only in the Nuclear Medicine department covering the most common procedures. Medical Physicists 1,2 are operating in both departments. Results: In the period of January 2013 to December 2013 a total of 982 PET/ CT studies and 2157 conventional Nuclear Medicine procedures were performed. The collective effective and finger doses received by all 4 members of the PET/CT staff were the following: Nurse 1 received 2,94 mSv as a whole body dose and 6,84 mSv as a hand dose and Nurse 2 received 2,87 mSv whole body dose and 5,91 mSv hand dose respectively. Technologists 1 and 2 received 1,95 mSv and 1,56 mSv as the whole body dose respectively. Medical Physicist 1 received 1,75 mSv whole body dose and 7,77 mSv hand dose and Medical Physicist received 2 2,17 mSv and 4,68 mSv respectively. Technologists 3,4 and 5 received 1,85 mSv, 1,76 mSv and 1,82 mSv as whole body doses respectively Conclusion: The personnel dose results are significantly lower than the recommended annual dose by International Commission for Radiological Protection. The higher value of gamma dose for PET/CT workers by comparison with the staff operating conventional Nuclear Medicine procedures is attributable to the higher specific gamma constant of 18F, as well as the longer exposure time required for accurate positioning. THREE-DIMENSIONAL METRICS FOR QUANTITATIVE MONITORING OF TREATMENT EFFECTS WITH PET/CT N. Politis , A. Georgakopoulos , M. Metaxas , S. Chatziioannou , M. Kallergi . Department of Biomedical Engineering, Technological Educational Institute of Athens, Greece; Nuclear Medicine Division, Biomedical Research Foundation of the Academy of Athens, Greece; Nuclear Medicine Section, 2 Department of Radiology, “Attikon” University General Hospital, National and Kapodistrian University of

Prediction of radiation doses during the dismantling of a maintenance cooling reservoir of RBMK-1500 reactor

*Decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. Personnel radiation safety during the dismantling activities is one of them. In order to assess the optimal personnel radiation safety, the modelling is performed by means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium). In this paper it was used for the evaluation of direct gamma radiation from the radioactive equipment to personnel when carrying out dismantling activities and selection of gamma radiation shielding geometry (structure), shielding materials and positions. A modelling of radiation doses during the dismantling of the maintenance cooling reservoir (MCR) of RBMK-1500 reactor has been performed in this paper. The basic MCR model’s development was based on internal contamination and the existing geometrical data. The assessment of workers exposure was performed to comply with ALARA. The effective doses to the workers were calculated for different strategies of reservoir dismantling. The impacts of dismantling tools, shielding types, variation of parameters such as shielding distance, shielding installation time and exhaust ventilation flow rate during the dismantling of reservoir on effective doses were analyzed. The total effective personnel doses were obtained by summarizing the effective personnel doses from various sources of exposure, i.e., direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises.

Investigation of thermoluminescence characteristics of Li2B4O7:Mn (TLD-800)

Li2B4O7:Mn is a well-known TLD material with an excellent human tissue equivalence. It is used in practice for personnel dose monitoring in variable branches. In the given study, the additive dose (AD), variable heating rate (VHR), peak shape (PS), three points method (TPM), and computerized glow deconvolution (CGCD) methods were used to determine the kinetic parameters of the main glow peak (P3) of Li2B4O7:Mn (TLD-800). In addition, the fading characteristics and the effect of the heating rate on the dose dependence of TL glow curves have been investigated. The results of the analysis have shown that TLD-800 has a general order main glow peak with an activation energy of Ea=0.58eV at 170°C at a linear heating rate of 1°C/s.

Radiation exposure to operating room personnel and patients during endovascular procedures

To characterize radiation exposure to patients and operating room personnel during fluoroscopic procedures. Patient dose information was collected from the imaging equipment. Real-time dosimetry was used to measure doses to the operators, scrub nurse, radiologic technologist (RT), and anesthesiologist in 39 cases of endovascular thoracoabdominal aortic aneurysm repair using fenestrated endografts. Overall equivalent doses and dose rates at time points of interest were noted and compared with the corresponding patient doses. The dosimeter on the anesthesia equipment received 143 μSv (38-247) more radiation per case than the average operator, and the scrub nurse and RT received 106 μSv (66-146) and 100 μSv (55-145) less, respectively. Adjusting for protective lead aprons by the Webster methodology, the average operator received an effective dose of 38 μSv. Except for the RT, personnel doses were well correlated with patient dose as measured by kerma area product (KAP) (r = .82 for average operator, r = .85 for scrub nurse, and r = .86 for anesthesia; all P < .001) but less well correlated with fluoroscopy time or cumulative air kerma (CAK). When preoperative cone beam computed tomography was performed, the equivalent dose to the RT was 1.1 μSv (0.6-1.5) when using shielding and 37 μSv (22-53) when unshielded. Digital subtraction acquisitions accounted for a large fraction of all individuals' doses. Decreasing field size (and thus, increasing magnification) was associated with decreased KAP (r = .47; P < .001) and increased CAK (r = -.56; P < .001). The square of the field size correlated strongly with the KAP/CAK ratio (r = .99; P < .001). Increased lateral angulation of the C-arm increased both CAK and KAP (at field size, 22 cm; r = .54 and r = .44; both P < .001) and the average dose rate to an operator was 1.78 (1.37-2.31) times as high in a lateral projection as in a posterior-anterior projection. Personnel doses were best correlated with KAP and less well correlated with fluoroscopy time or CAK. The dosimeter on the anesthesia equipment recorded the highest doses attributable to ineffective shielding. Operators can reduce the effective dose to themselves, the patient, and other personnel by minimizing the use of digital subtraction acquisitions, avoiding lateral angulation, using higher magnification levels when possible, and being diligent about the use of shielding during fluoroscopy cases.

SU-C-144-03: Validation of Ceiling Shielding in CT/PET Room with RADEYE

Purpose: The purpose of this study was two-fold: (a) to measure the magnitude of scattered radiation levels in CT/PET room; and (b) to evaluate the effectiveness of shielding of the ceiling of a PET/CT suite. Methods: Six RadEye detectors were placed in the CT/PET room, four in the ceiling, and two at one meter above floor. A RadEye detector recorded data every 25 seconds throughout the day. The detector was turned on at the beginning of the day and the doses were transferred to a laptop for analysis at the end of the day. The dose to a non-radiation worker above the CT/PET room was estimated based on the ceiling data. The dose build up factor was used in the dose estimate. The magnitude of transmitted CT radiation in the room above was measured separately with RadEye. Results: The CT component of the scattered radiations showed approximately cone beam shape of the iso-dose curve. The CT dose contributed about 80% of the total dose while the PET contributed 20% within the scanning room. The combined dose to a person upstairs sitting in a chair at 0.5 meter high above the PET/CT suite was 4×10−6 mSv per week. In addition, the physical measurement at the interstitial space above the ceiling recorded only 0.12 mR per day, which was the background dose. Conclusion: This study has quantified the CT and PET doses contributions separately in the clinical CT/PET room. An analytical model was developed to calculate the non-occupational personnel dose above the CT/PET room and the calculated results were confirmed by physical measurements. The actual physical dose was 50 times smaller than the NCRP design goals of 0.02 mSv/wk.

Assessment of eye and body dose for interventional radiologists, cardiologists, and other interventional staff

A dose limit for the eye of 20 mSv, as proposed by the ICRP, could be exceeded by interventional clinicians. Data on eye dose levels for interventional radiologists and cardiologists provided by medical physicists from hospitals around the UK have been collated. The results indicate that most hospitals would require one or more interventional clinicians to be classified and several would have exceeded a 20 mSv limit. Dose data in the literature have been reviewed to derive factors that might be used to predict eye dose levels based on dose per procedure or kerma-area product workload. These could be used in prior risk assessments to establish monitoring practice. An alternative approach to personnel dose monitoring in radiology applications using a collar dosimeter worn outside the lead apron as the first dosimeter is proposed. The collar dosimeter would provide an assessment of eye dose in terms of Hp(3) and body dose in terms of Hp(10), which could be divided by ten to provide an assessment of effective dose. If Hp(3) exceeded 1 mSv per month, regular monitoring with a head dosimeter would be recommended, and if Hp(10) exceeded 2 mSv per month, then an under-apron dosimeter should also be worn.

Evaluation of the secondary radiation impact on personnel during the dismantling of contaminated nuclear equipment

The article contains a numerical analysis of the secondary radiation contribution to the total radiation affecting the operational personnel during the dismantling activities of the contaminated equipment at a nuclear power plant. This study considers a widely applicable Monte Carlo particle transport code MCNPX and real Ignalina nuclear power plant records. A simplified albedo method is investigated in order to analyze the selected geometrical design cases. Additionally, the impact of the secondary radiation on the personnel dose was analyzed. The numerical MCNPX simulation allowed ascertaining the optimal distance between the source and the wall for the working personnel in closed rooms with contaminated equipment. The developed dose rate maps of the secondary radiation showed cross-sectional distribution of the dose rate inside the enclosed area.

Transportation of radioactive wastes to the site «Vector» and site «Buryakovka» for radioactive wastes

В статті представлено вивчення альтернативних маршрутів безпечного та надійного транспортування радіоактивних відходів на комплекс «Вектор» та ПЗРВ «Буряківка».

Modeling of the Radiation Doses during Dismantling of RBMK-1500 Reactor Pressurized Tanks from Emergency Core Cooling System

Decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. One of them is personnel radiation safety during the performance of dismantling activities. In this paper, modeling results of radiation doses during the dismantling of the pressurized tank from the emergency core cooling system (ECCS PT) of RBMK-1500 reactor are presented. The radiological surveys indicate that the inner surface of the ECCS PT is contaminated with radioactive products of corrosion and sediments due to the radioactive water. The effective doses to the workers have been modeled for different strategies of ECCS PT dismantling. In order to select the optimal personnel radiation safety, the modeling has been performed by the means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium). The impacts of dismantling tools, shielding types, and extract ventilation flow rate on effective doses during the dismantling of ECCS PT have been analyzed. The total effective personnel doses have been obtained by summarizing the effective personnel doses from various sources of exposure, that is, direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises. The uncertainty of the collective doses is also presented in this paper.

Induced Radioactivity in Medical Accelerators

Radiation-induced activation of patients and materials in radiotherapy treatment rooms may result in occupational radiation hazards due to in-room induced activity. Induced activation following 18 MV X-ray irradiations was assessed during commissioning and use of a medical linear accelerator at Ioannina University Hospital. Measurements were carried out using hand-held instruments at twenty locations in the accelerator bunker (three portable dose rate meters, a contamination meter were used). Temporal ambient dose rate data fitted with exponential functions and coupled with in situ spectroscopic data were used to characterize the post-irradiation field and develop detailed and practical site-specific work-plans to keep personnel doses secondary to radiotherapy as low as reasonably practical. The application of the suggested work-plans kept occupational exposure at very low levels during the first 2 y of clinical use of such beams.

Excitation functions for 7Be, 22,24Na production in Mg and Al by deuteron irradiations up to 50 MeV

New experimental data for production of 7Be and 22,24Na in deuteron irradiation of natMg and Al up to 50MeV are presented. The induced activity, measured with HPGe spectroscopy, allows us to determine excitation functions of natMg(d,x) and 27Al(d,x) reactions involved in the activation process with reference to natTi(d,x)48V monitor cross sections. A comparison with experimental literature values and results from updated theoretical codes is discussed.Thick target yields were derived from fits to our cross-sections and integrated personnel dose was calculated for different irradiation cycles and exposure scenarios around high power deuteron accelerator facilities.

Impact of the Fukushima Nuclear Accident on Background Radiation Doses Measured by Control Dosimeters in Japan

After the 9.0 magnitude earthquake and subsequent massive tsunami on 11 March 2011 in Japan, several reactors at the Fukushima Daiichi Nuclear Power Plant suffered severe damage. There was immediate participation of U.S. Navy vessels and other United States Department of Defense (DoD) teams that were already in the area at the time of the disaster or arrived shortly thereafter. The correct determination of occupational dose equivalent requires estimation of the background dose component measured by control dosimeters, which is subsequently subtracted from the total dose equivalent measured by personal dosimeters. The purpose of the control dosimeters is to determine the amount of radiation dose equivalent that has accumulated on the dosimeter from background or other non-occupational sources while they are in transit or being stored. Given the release of radioactive material and potential exposure to radiation from the Fukushima Daiichi Nuclear Power Plant and the process by which the U.S. Navy calculates occupational exposure to ionizing radiation, analysis of pre- and post-event control dosimeters is warranted. Several hundred historical dose records from the Naval Dosimetry Center (NDC) database were analyzed and compared with the post-accident dose equivalent data of control dosimeters. As result, it was shown that the dose contribution of the radiation and released radiological materials from the Fukushima nuclear accident to background radiation doses is less than 0.375 μSv d for shallow and deep photon dose equivalent. There is no measurable effect on neutron background exposure. The latter has at least two important conclusions. First, the NDC can use doses measured by control dosimeters at issuing sites in Japan for determination of personnel dose equivalents; second, the dose data from control dosimeters prior to and after the Fukushima accident may be used to assist in dose reconstruction of non-radiological (non-badged) personnel at these locations.

Excitation function for 7Be production in carbon by deuteron irradiations up to 50 MeV

In order to check and complement available experimental data for production of 7Be in deuteron irradiation of natC, new measurements with incident particle energy up to 50MeV were performed. The 7Be content, measured with HPGe spectroscopy, allows determination of the excitation function of the natC(d,x)7Be reactions. A comparison with experimental literature values and results from updated theoretical codes (ALICE-D, EMPIRE-D, EAF2007 and TENDL2010 on-line libraries) is discussed.Thick target yields were derived from fits to our cross-sections and integrated personnel dose was calculated for different irradiation cycles and exposure scenarios around the IFMIF facility.

Occupational exposures in industrial application of radiation during 1999-2008

Radiation sources are used in various industrial applications like industrial radiography, industrial irradiation, industrial fluoroscopy, nucleonic gauges, well logging etc.. Gamma, beta X-ray as well as neutron sources are used for various applications. Number of radiation workers in this field has increased over the years. Due to operating conditions prevailing during the exposure as well as the strength of the sources used in some of the applications, radiation protection plays an important role in this field. Analysis of doses received by radiation workers in industry provides information on trends of doses as well as adequateness of radiation protection practices followed in this sector. In India, National Occupational Dose Registry System (NODRS) of Radiological Physics and Advisory Division (RPAD), Bhabha Atomic Research Centre (BARC) maintains personnel dose information of monitored radiation workers in the country. Analysis of occupational dose data of industrial radiation workers for last 10 years, i.e., 1999-2008 has been presented in this paper. It is observed that even though there is an increase in monitored radiation workers, percentage of persons receiving radiation exposure has come down during this period. There is also a decrease in the average annual dose as well as the collective dose. Further analysis of sub-categories shows that industrial radiography operations are the main contributor for collective dose (about 77%) followed by well logging and industrial X-ray operations (about 8% each). Thus, in addition to industrial radiography, attention is also to be given to operations in these areas.