I. Surface Contamination 1
JAERI-Conf--97-003 JAERI-Conf 97-008 JP9706044
PROCEEDINGS OF THE IAEA/RCA TRAINING WORKSHOP ON CONTAMINATION MONITORING OCTOBER 21-25,1996, JAERI, TOKAI, JAPAN
June 1997
(Eds.) Akira ENDO, Tetsuya OISHI, Fumiaki TAKAHASHI and Hiroyuki MURAKAMI
Japan Atomic Energy Research Institute 28 1 9 Mi. HLTU&.
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This report is issued irregularly. Inquiries about availability of the reports should be addressed to Research Information Division, Department of Intellectual Resources, Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken 319-11, Japan.
© Japan Atomic Energy Research Institute, 1997 JAERI-Conf 97-008
Proceedings of the IAEA/RCA Training Workshop on Contamination Monitoring October 21-25, 1996, JAERI, Tokai, Japan
(Eds.) Akira ENDO, Tetsuya OISHI, Fumiaki TAKAHASHI and Hiroyuki MURAKAMI
Department of Health Physics Tokai Research Establishment Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, Ibaraki-ken
(Received May 7, 1997)
The International Atomic Energy Agency (IAEA) has been conducting the Regional Cooperative Agreement (RCA) for prevailing the nuclear-related techniques to Asia and Oceania countries. The RCA project for strengthening radiation protec tion infrastructures, which started in 1988 as one of the RCA activities, has been conducted with performing such programs as training courses, workshops, etc.. The present proceedings comprise of papers presented at “IAEA/RCA Training Workshop on Contamination Monitoring ”, held at Tokai Research Establishment, JAERI, from October 21 to 25, 1996 as a program of the RCA radiation protection project. The purpose of the workshop is to familiarize participants, who are engaged in operational contamination monitoring in their countries, with the latest techniques on measurement and calibration of instruments for contamination monitoring through lectures, country reports, technical demonstrations and exercises. The workshop has proved successful fruits on understanding the latest contamination monitoring techniques and grasping future subjects. The present proceedings will provide useful information for engineers and scientists in this field.
Keywords : Radiation Protection, Contamination Monitoring, Surface Contamination, Calibration, Intercomparison. JAERI —Conf 97-008
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Contents
1. Country Reports on Status of Contamination Monitoring ...... 1 1.1 Status of Contamination Monitoring in Bangladesh ...... 3 Aleya Begum : BANGLADESH 1.2 Status on Contamination Monitoring in China ...... 7 Gou Quanlu : CHINA 1.3 Monitoring of Surface and Airborne Contamination ...... 11 Pradeep Kumar K.S. and Natarajan G. : INDIA 1.4 Status of Contamination Monitoring in Radiation Protection Activities of National Atomic Energy Agency (NAEA) in Indonesia ...... 18 Gatot Suhariyono : INDONESIA 1.5 The Status of Contamination Monitoring in Radiation Protection Activities of Korea ...... 32 Jong-11 Lee : KOREA 1.6 Status of Nuclear Technology in Malaysia ...... 47 Bustami bin Abu : MALAYSIA 1.7 Country Report on Contamination Monitoring ...... 50 Navaangalsan Oyuntulkhuur : MONGOLIA 1.8 Contamination Monitoring in Radiation Protection Activities in Myanmar ...... 55 Kay Thi Thin and Semi Htoon : MYANMAR 1.9 Contamination Monitoring Activities in KANUPP ...... 59 S. Sarwar Naqvi : PAKISTAN 1.10 Contamination Monitoring ...... 65 Arlean L. Alamares : PHILIPINES 1.11 Country Report for the Regional (RCA) Training Workshop on Contamination Monitoring ...... 69 T.H.S. Shantha : SRI LANKA 1.12 The Status on Contamination Monitoring in Thailand ...... 78 Fookiat Sinakhom : THAILAND 1.13 Present Status of Contamination Monitoring at the Dalat Nuclear Research Institute (DNRI) ...... 87 Hoang Van Nguyen : VIETNAM 2. Special Lectures and Statement ...... 91 JAERI-Conf 97-008
2.1 Radioactive Standards and Calibration Methods for Contamination Monitoring Instruments ...... 93 Makoto Yoshida 2.2 Monitoring and Evaluation Techniques for Airborne Contamination ...... 102 Xia Yihua 2.3 Monitoring and Evaluation Techniques for Surface Contamination ...... 120 David Woods 2.4 Features and Properties of Instruments for Contamination Monitoring ...... 141 Richard V. Griffith 2.5 Basic Techniques of Decontamination ...... 147 Naoki Yokosawa 2.6 Review of the IAEA Program in Radiation Protection ...... 149 Richard V. Griffith 3. Intercomparison Exercise for Surface Contamination Monitoring ...... 167 Appendix Workshop Agenda and Participants ’ List ...... 175
IV JAERI-Conf 97-008
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Participants of the IAEA/RCA Training Workshop on Contamination Monitoring (October 21-25, 1996, Tokai, Japan) JAERl-Conf 97-008
1. Country Reports on Status of Contamination Monitoring
NEXT PAGE(6) !eft BLANK JAERI—Conf 97 —008
l. l STATUS OF CONTAMINATION MONITORING IN BANGLADESH
Aleya Begum
Health Physics & Radiation Protection Dixision Institute of Nuclear Science & Technology Atomic Energy Research Establishment Ganakbari. Savar. GPO Box 3787 Dhaka-1000. Bangladesh.
The applications of radioisotopes and radiation sources in research and development activities in medicine, l'ood agriculture, industries, sterilization of medical products and appliances, etc. are rapidly increasing in Bangladesh. Bangladesh Atomic Energy Commission (BAEC) has a viable programme on the peaceful application of radioisotopes in medicine, industry and agriculture. The existing major nuclear facilities and allied laboratories of the country include: 3 MW TRIGA Mark-II research reactor which is being utilized for training . research and radioisotope production: 14 MeY Neutron Generator (SANIES) is used for nuclear data measurements and elemental analysis via neutron activation process: 3 MeY Yan de Graff Accelerator. Atomic Energy Center. Dhaka (ACED) is mainly utilized as an important tool of research and application of nuclear physics and related field: nominal 5O.OOOCi ""Co and 5.000 Ci 60Co irradiators are installed at the Institute of Food and Radiation Biology (IFRB). AERE. Savar. Bangladesh Institute of Nuclear Agriculture (BINA) respectively for preservation and processing of several important food items, industrial products, sterilization of medical products and supplies as well as promoting R&D activities m the field of agriculture and radiobiology including evolution of improved and high y ield variety of agricultural mutants. About ten teletherapy units G Co and ;' Cs therapy units) are operational in several hospitals of the country. In addition, many unsealed sources such as 1' Cs. wCo. "Zn. etc. are being used for research works.
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Radioactive contamination of the working areas, equipment, protective clothing and skin may result from normal operations involving unsealed radioactive materials or from accidental releases. Contamination monitoring and decontamination is an essential part of a radiation protection programme. Radioactive contamination results from contact between radioactive materials and any surface and occurs to some extent whenever radioactive isotopes are handled. The presence of contamination is an indication of the inadequacy of the programme of containment of unwanted sources of radioactivity. Therefore, it is always most useful to evaluate the basic and proximate causes of contamination so as to prevent similar contamination later.
Surface contamination was monitored by Berthold survey meter (LB 1200. window open). FRG. The monitor displays both counts per second (cps) and dose rate (mr.'h) on the scale of the ratemeter. The survey meter was calibrated periodically using the SSDL facilities.
Hand and Foot monitor has been used for checking contamination, if any. of the occupational workers .
Routine systematic searches have been done w here the possibility of surface contamination exists. All openings in source shields or contamination barriers and frequently handled items such as source manipulators, control switches and knobs have been checked for contamination. Walls and horizontal surfaces such as floors, work benches and shelves have been sun eyed.
Small pieces of paper, such as discs of filter paper, are rubbed over the surface often with appropriate wetting or solvent material on the wiping surface
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and examined at a remote location. This procedure only removes part of the contamination. Wipes have been examined for different types of radiation by using simply counting techniques and can also be used to determine the radionuclides present by pulse-height analysis. The equipment used in contaminated areas have been checked for contamination. The contaminated areas have been clearly marked until the contamination has been removed, decayed or immobilized.
Continuous air monitoring system has been used for detection of air-bome contamination. Air samples have been collected in the filter paper using high volume air sampler. Filter papers have been counted by gamma spectrometry system. The gamma spectrometry system consists of HPGe detector. MCA and associated electronics. Using this technique we could quantitatively measure the radioactivity and determine the nature and degree of hazard arising from contamination and also can be identify the source. The analysis of the filter paper did not show any contamination.
In order to determine the pathways of silt movement in the Chittagong harbour about 10.38 Ci of 46Sc was used. During injection of 46Sc to the sea-bed the injecting system (injector and reflector) became contaminated with 4o Sc. The injecting system was decontaminated by washing with detergent and water and checked-up by frequent monitoring until release (discharge) limit of 46Sc into
environment was reached as per recommendation (14.8 kBq litre) of ICRP. Injecting system was placed into a container of 11 litre capacity in which detergent and water were added with constant brushing. One liter contaminated w ater sample was collected and analyzed by HPGe detector. Smear tests were performed before and after the decontamination of isotope injecting system. The surface dose rate of the system was determined by beta-gamma survey meter. Evaluation of the data of smear paper, water samples and surface dose rates of the injecting system show that
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the system is free from any contamination due to 46Sc. So the injected system could be reused for further related experiment.
Measurement of tritium contamination: The tritium levels have been measured in and around the 14 MeV (D-T reaction) neutron generator facility at INST. AERE. Savar. Tritium can be present in the work place as 3H itself or as a
vapour or tritiated dust or aerosol; in contaminated liquids (oil. water) or on contaminated equipment and materials. Tritium concentrations in the form of tritiated water vapour (HTO) in cooling water, pump oil and other samples collected from the neutron generator facility have been measured. Urine and blood samples were collected from personnel involved in the operation maintenance of the neutron generator and measured. Surface contaminations due to tritium on smear and filter paper have also been detected and analysed using a low background liquid scintillation counter (Model Packard Tri-carb. 1000, USA) based on the sample to channel ratio (SCR) method. From the preliminary results, it is observed that the air vapour in the neutron generator facility and control rooms are significantly contaminated with tritium. Some urine samples from personnel involved in the operation/maintenance of 14 MeV neutron generator were found to be contaminated with JH. However, no contamination due to tritium was found in blood samples. Necessary advice were imparted to the relevant personnel in the facility in order to reduce health risk. Further works are in progress to estimate the radiation doses for evaluation of health hazard.
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l. 2 Status on Contamination Monitoring in China
Gou Quanlu
(China Institute for Radiation Protection, Taiyuan, China)
1. Introduction
In nuclear enterprises and some radioactive workplaces, the air may be contaminated by radioactive materials, and therefore radioactive aerosol is formed. Some of radioactive aerosol will be detained inside their bodies and cause internal exposure damage when the workers inhaled them. On the other hand, leakage of radioactive materials in operation of radioactivity will lead to contamination of the bodies, clothes, ground or equipments. These radioactive materials might be transferred into body through mouth or skin permeation, and they also might be resuspended into air, and enter in bodies through respiratory tracts and give rise to internal exposure. The environment will be contaminated if the contaminated equipments or articles have been transferred to clean area, It is necessary and important to monitor the air and surface contamination for the health of the public and workers and for protecting environment.
2. Status on surface contamination monitoring in China
At the present, there are many institutes which are engaged in studies on the surface contamination monitoring in China. Our government also have paid attention to it for a long time, and have formulated control limits of surface contamination in the Regulations on Radiation Protection (see table 1).
Table 1. Control levels of surface radioactivity contamination (Bq/cnT)
kinds of surface alpha radioactive matenal beta radioactive material ultra-toxicity others
working pate controlled area facilities supervised area 4 4x10 4x10 wall ground uncontrolled area 4x10 ' 4 4
ciothes controlled area gloves supervised area 4x10' 4x10 ' 4
-"I snces uncontrolled ansa 4: 1C"' 4x’C 4x'C
hands. SKin. underwear. sccxs 4x10 " 4X10"' 4x*.C' ‘
Monitors which are used in radioactive of surface contamination monitoring are almost home-made. Methods and monitors should be chosen according to different monitoring
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purposes. Methods can be divided into two types roughly; direct measurement methods and indirect measurement methods. Direct measurement is measurement of surface activity by means of a contamination monitor, and indirect measurement is evaluation of the removable activity on the surface by means of sampling. There are many kinds of indirect measurement methods, but often used methods are smear test and placing surface sample test. Scintillation counters(ZnS screen) .semiconductor detectors and G-M counters with thin window or gas flow proportional counter have been used to detect a surface contamination . Common a surface contamination monitors and their properties show in table 2. The scintillation meters which use plastic scintillators as their detectors (for example FJ-364, FJ-335D using (3 detector)and thin wall / window G-M counter are used to measure p surface contamination and seme meters such as FJ-2203, FJ-22C6, FJ-2207 can measure a, n or a-|3 surface contamination. In general, These meters can meet the need of measurement on the site , and measurement results of them can be comparable with control limits. It is necessary that special detectors should be designed for the surface contamination monitoring of low energy p nuclides such as 3H,241Up,1JC and u7Pm etc. Some special meters have been designed for the others low energy nuclides except for 3H and can meet the need for surface contamination monitoring of these nuclides. Although some specialized devices have been designed for measuring 3H surface contamination, there are some problems for 3H because of its too low radiation energy. At the present, indirect method is usually used for the measurement of 3H surface contamination .In general, scrap materials can be reused through some proper treatment such as melting and decontamination, but some deep studies should be undertaken. In the future, except for the application and improvement of the existing technology some surface contamination monitor which use to measure the surface contamination of defined nuclides on the special workplace should be developed at the same time.
Table 2. Common alpha surface contamination monitors and their properties
k.% ';,i effective detecting background :4i minimum detectable 'imit models areaiem") (cpm) natural :39 Pu Am counts/100 cm" 2mmm natural Pu Am
FJ-335B 1CC 2 10 22 24 20 9 8 FJ-355AGf u 155 10 11 25 23 59 25 FJ-335AG1S 17 3 11 23 27 161 77 66 I.PA.92C 28 5 12 2C 24 149 89 75 PCM-4 too 1 5 18 23 20 6 5 FJ-348 353 3 1 6 7 35 14 12
Note:- the cistance between contaminated area and protective net before detector ,s 5 mm.
3. Status cr airborne ccrtarriraticn monitoring in China
Seme explicit stipulations have been provided in the Regulations cn Radiation Protection for radioactive aerosol monitoring .The annual average concentration of airborne radioactive nuclides in the environment in which the public live resulting from the
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emission of gases or aerosols should not exceed fifty percent of DAC (derived air concentration) in Annex E in the Regulations on Radiation Protection. There are two steps including air sampling and activity measurement in the monitoring technology of aerosol. In general, there are three types of aerosol sampling including total concentration, particle size selection and particle size distribution according to the request in some related regulations. The qualities of these devices and technology which often used in the monitoring of airborne contamination have reached the advanced level in the world . At the present, multi-stage impactors, cascade centripters and cyclone samplers have been developed successfully and used in different purpose sampling mentioned above. Devices for measuring radioactivity in samples are almost Chinese-made goods. Except for a few of them in which decay method has been taken to eliminate the influence of radon progeny .the technology of energy discrimination and alpha-beta anti- coincidence have been taken to give measuring results. The in-situ aerosol continuous monitor which has been successfully developed and put into use is basically the sign of the level of radioactive aerosol monitoring technology in our country, and its characteristics and reliability can meet the need for in-situ monitoring of radioactive aerosol. The monitoring information of concentration and particle size distribution of radioactive aerosol have been used in the evaluation of the internal dose. The monitoring results of aerosol have not only been used in the assessment of air contamination, but also in the evaluation of the internal dose of workers in some cases, and some satisfied achievement have been gained in this field. So far, studies on the calibration methods for different kinds of air contamination monitor are being undertaken widely, and have gained some practical achievement. The sensitivity of the aerosol continuous monitor is a very impartant factor. Only when have the calibration method been established the sensitivity of the monitor could be evaluated correctly. More studies should be taken in this fields such as studies on the calibration method and the development of calibration devices of in-situ radioactive aerosol continuous monitor. It is hoped that we can learn some advanced technology and experience in the world through this workshop.
4. Problems and expectation
For many years, studies on contamination monitoring both of the airborne and surface contamination have been taken under the effort of our science and technique workers, and gained some practical achievements, and it took an important effect to the nuclear power enterprise of our country. But more studies should be in progress in the future such as: (1 development of high quality surface contamination monitor for monitoring of low energy ; nuclides so as to solve the problem about surface contamination monitoring for lew energy r< nuclides, especially for 3H. (2)Studies on surface contamination monitoring methods used in some facilities which have special shape of surface such as chimney, the inner surface of pipeline.
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(3) Studies on the technology for the decontamination, disposal and recycling of scrap materials. (4) Studies on the calibration methods and devices for on-line radioactive aerosol continuous monitor. It is hoped that we could find some cooperatcrs to cooperate in above fields and make a contribution for protecting the of the general public and the environment.
Acknowledgment
The author gratefully thanks professor Zhang Yansheng and Lu Zhengyong for their help I
References
1, GE-8703-88 ,Regulations for Radiation Protection, 1988,
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1.3 MONITORING OF SURFACE AND AIRBORNE CONTAMINATION
PRADEEP KUMAR K.S. Bhabha Atomic Research Centre Mumbai - 400 085, INDIA and NATARAJAN G. Atomic Energy Regulatory Board Mumbai - 400 094, INDIA
1. Introduction Indian nuclear energy programme aims at total safety in all activities involved in the entire fuel cycle for the occupational workers, members of the public and the environment as a whole. Routine radiation monitoring with clearly laid out procedures are followed for ensuring the safety of workers and public. Radiation monitoring carried out for the nuclear installations comprises of process monitoring, monitoring of effluent releases and also of the radiation protection monitoring of the individuals, work place and environment. Details of the radiation protection monitoring programme are shown in Figure- 1. Inspite of good design, operating practices and precautions followed in nuclear installations, occurrence of radioactive contamination in working areas cannot be ruled out. Control of radioactive contamination is achieved by source containment, provision of adequate air changes, system of zoning of the plant area based on the potential for contamination, interzonal barriers, wearing of protective clothing, monitoring and change room procedures. Choice of appropriate equipment for the handling of radioactive materials, correct operating procedures along with strict management are also aimed at reducing the risk of radioactive contamination. Arrangements for preventing spillage of radioactive liquids, use of double rubber stations and procedures for monitoring and change of dresses by workers at the rubber station itself have prevented the spread of contamination during special jobs. Regulations like banning of smoking and consumption of food and drink etc. reduces the risk of direct ingestion even if inadvertent spread of contamination takes place. Though limit of transportable surface contamination is prescribed, the health physicists always follow a "clean on swipe" philosophy which compensates any error in the measurement of surface contamination.
2. Necessity of Contamination Monitoring Presence of airborne radioactivity in working area or contamination on surface of equipment, floor and similarly on skin, hands and on protective clothing of workers can result in external exposure and/or internal exposure by inhalation of radioactive materials or absorption through skin. Hence proper assessment of radioactive contamination (on surface as well as airborne) is very essential for radiological safety. During the operation of the nuclear installations, various areas/equipments/clothing etc. are likely to get contaminated. Thus the need for contamination monitoring arises in these areas on routine basis as well as when specific jobs are undertaken. Surface contamination on large scale can also lead to airborne contamination. As airborne contamination can lead to surface contamination and viceversa, monitoring for air activity and surface contamination are carried out when one of them are detected.
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During the routine operation of nuclear reactors, Xenons, Kryptons and Iodines may get released into the air in addition to Ar-41 in some air cooled reactors. Due to release of noble gases like Xenons and Kryptons from the cover gas system/heat transport systems or during the transfer of defective fuel, radioactive particles (daughter products like Rb-88 [T.^18 min, DAC=3.3E5 Bq/m 3] and Cs-138 [T1/4 =32.2min, DAC=2.5E5 Bq/m 3] were detected on few occasions in reactor building, fuel storage bay and also during maintenance jobs in reactor cover gas system. Similarly during transfer of fuels, Iodines were detected in reactor building and fuel storage bay atmosphere. In the event of failed fuels, mixed fission products, Np-239, Uranium, Plutonium etc. can also get into the coolant system and may result in contamination of working environment. Tritium is also found to become a source of airborne contamination where heavy water is used as moderator/and coolant. During fuel reprocessing of fuel elements, gaseous and volatile radioactive materials like Kr-85, Xe-133 may get released in small quantities along with particulates in the form of Cs-137 and plutonium. Release of radioactivity in radiological laboratories during leakages in glove boxes or failure of ventilation also cannot be ruled out. Working levels for surface contamination are estimated based on all likely exposures and for control purposes, most restrictive values are recommended (Table-1). For beta emitters, external dose as well as internal exposure resulting from transfer of contamination are considered. For alpha emitters with low DAC values, airborne contamination arising from resuspension is found to be more restrictive than ingestion route. Regulations on equipments used in active areas which are sent to outside agencies for minor repair are also given in Table-1. In case of airborne activity, appropriate respiratory protection is recommended whenever the air activity levels exceeds the recommended limits. Table-2 gives the limits at which respiratory area is declared.
3. Accuracy Required in the Calibration of Surface Contamination Monitors Routine monitoring is carried out to confirm whether contamination is removable and within recommended levels. Hence these type of measurements which involve variety of radiations and different types of detectors do not require very high accuracy as they are only comparative. While the monitor is calibrated using large area sources which are made by uniformly depositing a thin coating of radionuclide, the health physicist normally faces non uniform contamination made up of a mixture of radionuclides deposited on irregular surfaces with dust covering the radioactivity. The uncertainties created by these factors overweigh those arising from calibration. Accurate measurement of surface contamination is difficult due to its nature, and may be unnecessary as derived limits are generally based on very conservative assumptions. Sufficient conservatism is built in the models used in arriving at the DWLs such that further safety factors are deemed unnecessary 111. It can be accepted that degree of accuracy of +20% for operational field work is sufficient121. There is a feeling that calibration of surface contamination monitors within 30% is quite adequate 131. As performance can vary significantly, it is very essential to have the monitors calibrated and tested regularly. Instruments such as hand and foot monitors, portal monitors and large area portable monitors are calibrated using standard contaminated surface sources of large area.
4. Methodology for Contamination Monitoring Methodology adopted for monitoring of surface contamination can be divided into
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'direct method' where radiation detector is kept directly near the surface to be monitored and ' indirect methods' where transferrable contamination is detected by swipe test followed by checking with contamination monitors. Since very few radionuclides emit only gamma rays, better evaluation of surface contamination can be made by monitoring for beta particles. The health physicists prefer detection of beta rays over gamma rays as they have low penetration and hence is easy to localise the contamination. Also, the low efficiency of the detector and high background make detection of surface contamination by gamma ray measurement difficult. The basic requirements sought by a health physicist for surface contamination monitoring instruments are good reliability, sufficiently low detection limit, adequate range, quick response, audible output and also rugged nature. The methodology/monitors used for monitoring of surface and air contamination are shown in figure-2. The surface contamination monitors generally use scintillation detectors (ZnS or Plastic scintillator) or gas detectors (end window and pancake type GMs, gas flow counters). Multichannel analysers using solid state detectors are used for analysis of samples for identifying the radionuclides which help to identify the source of contamination . Alpha contamination is usually detected with ZnS screen covered with 1 mg/cm 2 aluminised window (window area= 100 cm2) and having a background count rate of »2cpm and high efficiency of —25%. Gas flow counters with window area of 160cm2, and thickness of ~0.7mg/cm 2 are used for alpha monitoring as well as beta monitoring. Geiger tubes (window area = 5cm2) and pan cake detectors (window area= 16cm2) both having window thickness of 1.5-2mg/cm 2and beta efficiency of =45% for Sr-90/Y-90 are used for beta contamination monitoring. Installed counting units using gas flow type beta counter having high efficiency are also in use141. This system uses a gas flow type main counter and another gas flow type guard counter operating in GM region in anticoincidence mode along with 3" lead shielding. The system has the background of the order of 1 cpm and efficiency of 40% for Sr90/Y90. The lower side cathode of the counter which also acts as window (area= 16 cm2), consists of aluminised mylar film of thickness 0.9 mg/cm 2. In case of high level beta radiation, some of the survey meters having removable beta shield which can expose the thin window of the detector are also used for the detection of high level of beta radiation though the measured dose rate may vary depending on the energy of the beta rays.
5. Air Monitoring The airborne radioactivity encountered consists of a) Radioactive noble gases b) Radioactive particulates c) Tritium as oxide and d) Radio iodine. In case of noble gases, as they cause only external exposure, radiation survey is found to be adequate. But for the detection of leakages like that of Ar-41, sample collected in chambers are analysed by gamma spectrometry. The assessment of airborne activity is normally done by analysing the sample collected from the air sampler whereas in continuous air monitors (fixed filter or moving filter type), the detector is close to the filter paper. Air to be sampled is drawn through a 4" dia filter paper at the rate of Inf/min by a staplex high volume air sampler. Glass fibre filters are normally used in India for airsamplers which is reported to have good collection efficiency (99.97%), low pressure drop, and low alpha burial losses. Normally 5m3 of air is sampled. The filter paper is counted in beta and
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alpha counting setup. The Iodine activity in air is assessed by activated charcoal impregnated filter paper used in the air sampler. Personal airsampler usually samples 5-7 litres per minute. Identification of radionuclides is done by gamma or alpha spectrometry or chemical analysis. Since fraction of activity likely to be deposited in various compartments of lung depends on the particle size (AMAD), the aerosol size is determined by Anderson multistage sampler . The detection of shortlived activity even in low level has helped as indicator of unusual system disturbance like defective fuel, leakage of moderator cover gas, leakage of heavy water etc. and calls for survey of radiation field, assessment of surface contamination and tritium in air activity. The level of tritium airborne activity depend on the level of leakage, and the tritium activity level in the coolant/moderator system. The airborne concentrations are reported to vary upto 100s of DAC (1.0E8 Bq/m 3) during large heavy water spills. As tritium is reported to contribute around 30% of the annual whole body radiation dose received by field personnel in PHWRs, measurement of tritium in air concentration is important. Two Ion chambers of equal volume capable of estimating the ion current due to tritiated air pumped through one of them is used in the assessment of tritium. It is capable of measuring concentration of the order of 1.0E4Bq/m 3. Where high radiation field or noble gases exists, tritiated water vapour is collected by bubbling the air through water and subjecting the sample for tritium counting in liquid scintillation counting technique. HTO samples are also collected by condensing the water vapour from the air with dry ice or by cold strip method.
6. Guidelines for Unrestricted Release of Scrap Materials At present all the contaminated scrap materials are disposed as active waste. All the inactive scrap originating from the nuclear industry is sold after proper monitoring and clearance certification. As per the guideline scrap material can be released without any restriction if the total activity or concentration of the radionuclide present does not exceed the exemption levels specified in International Basic Safety Standards151. Loose contamination should further be restricted to 0.37Bq/cm 2 for beta activity and 0.037Bq/cm 2 for alpha activity when averaged over 300 cm2 for accessible areas.
7. Problems in Contamination Monitoring a) Monitoring of surface contamination is unlikely to provide very good accuracy. Also swipe sampling cannot guarantee that the area is totally clear of loose contamination. The swipe sampling method has the uncertainty of the level of surface contamination removed during the swipe as the removal factor can vary significantly depending on the pressure applied, the area cleared in each swipe, type of surface and physical form of contaminant. b) As drastic decontamination methods may affect the integrity of the skin, and absorption to the blood stream may turn out be more dangerous, health physicists may face the problem as to what level the decontamination of skin has to be stopped. Natural process of sloughing of the outer layer of the skin is expected to remove the contaminant in 15 days 161. Even if the level of contamination is 700 Bq/cm 2 over an area of 100cm2 and the maximum dose rate at initial stage is 1.43 mSv/hr, the total dose for skin is 0.25Sv due to turnover half-life 4 days to the epidermal layer 171. c) Chances of probe getting contaminated due to accidental touching on surface in the process of getting better efficiency.
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d) Thin wall detectors are more prone for damage e) Contamination monitors with large area detectors may underestimate hot spot contamination. Hot spot particles can have high specific activity resulting in high surface dose rates. Dose rates to lcm2 of skin from particulate contamination on clothing from 1.0 ^Ci Sr/Y-90 or two year old fission products are reported to be 2.5 rad/h and 4R/h respectively' 21. f) Clearance for welding of pipelines suspected for fixed contamination by measuring low level fixed contamination may be difficult in view of the location of the spot. Moving a shielded contamination monitor to the spot of the equipment may not be feasible. g) On many occasions both direct and indirect methods may have to be adopted as a direct method will not be able to predict about the transferable contamination and indirect method may miss a fixed contamination. h) When the background in certain area is very high it will not be possible to monitor levels of contamination of the order of few Bq/cm 2. Gamma radiations will interfere with all beta detectors and when window is =2mg/cm 2 alpha radiations can also interfere. Using shielded geiger tubes in high gamma background area are not found practical. i) In air sampling, the question remains "how representative is the sample collected in comparison with the airborne contamination of the working area". In case of higher airborne activity, there is a likelihood of air sampler getting contaminated. Also there is a chance of resuspension of radioactive material by the discharged air from the sampler.
8. Conclusion Inspite of all precautions, occurrence of radioactive contamination in nuclear installations cannot be ruled out. Due to the necessity and importance of assessment of radioactive contamination, monitoring of working area is carried out by the health physicists/technicians before permitting work and also during work which has got the potential for generating airborne activity or surface contamination. The contamination monitors and monitoring methodology are found to be adequate to ensure detection of radioactive contamination below the recommended limits.
References 1. Krishnamony, S., Derived working levels for surface contamination, BARC/I/1072 /1990 2. IAEA, Technical reports series No: 120, Monitoring of radiation contamination on surfaces, 1970. 3. Burns, P.A., Australian radiation laboratory, Calibration of surface contamination monitors, 1990. 4. Abani, M.C., etal Low background gas flow beta counting system, SAINR, BARC, Jan 27-29, 1993. 5. International Basic Safety Standards for Protection Against Ionizing Radiation and for the Safety of Radiation Sources, Safety series No: 115-1, IAEA, 1994 6. NCRP, Report 65, Management of persons accidentally contaminated with radionuclides, 1980. 7. 1CRP-23,Report of the task group and reference Man, 1973.
- 15- JAERI-Conf 97-008
Table-1 Derived working levels for radioactive contamination
Surface Beta emitters Alpha emitters (Bq/cm2) (Bq/cm2) Skin 1.5 1.0 Hands 2.5 (Total 350Bq) 250(Total) Clothes Plant 6 2 Personal 2 0.5 Shoes Plant 37 3.7 Personal 0.37 0.037 Floor 3 . 7 0.37 Transportable Containers 3.7 0.37 Note: 1. The contamination on the inside surfaces of respirator face-pieces, air hoods, and all other breathing apparatus should not exceed the derived limit for skin. 2. Regulations on equipments used in active areas which are sent to outside agencies for minor repair are: a) Radiation level of 5xl0~3 mSv/h (0.5 mrem/h) at 10cm from any accessible surface (to ensure exposure will not exceed lmSv/year for an exposure duration of 200 hrs) b) Transferable contamination should not exceed 3.7Bq/cm2(beta) and 0.37Bq/cm2 (alpha) averaged over 300cm2 of the surface. c) Work scope assigned to the agencies should not involve operations such as machining/grinding etc which can generate airborne dust.
Table-2 Limits on air activity for declaration of respiratory area Limits Restricting DAC (Bq/m3) Nuclide (Bq/m3) Particulates Short lived beta emitters 1.0E5 Cs-138 2.50E5 Long lived Beta emitters 3 . 0E1 Sr- 90 2.50E1 Alpha emitters 0.125 Pu-239 0.125 Tritiated air 3.0E5 Tritium 2.67E5 Note: 1. Those with less than half lives of 30 minutes are Ci sidered as short lived and those with more than 30 minutes as long lived. 2. In reactors most of Strontium compounds are class D and hence limit for long lived particulates is specified as 1.5E2 Bq/m3, based on Sr-90 class D (DAC=1.66E2 Bq/m3). 3. DAG for 1-131 (T^=8.ldays) is 4.1E2Bq/m . Respirator for Iodine protection is recommended if air sampling with charcoal impregnated filter paper showes more than IDAC of Iodine activity in air.
16 RADIATION PROTECTION MONITORING
Monitoring of Monitoring of Monitoring of Individuals work place Environment
External Internal Radiation Activity Surface radiation activity environmental
field in air contamination field in air samples ERl-Conf JA
Figure - 1
CONTAMINATION MONITORING 97-008
Surface Air
Installed(alpha, beta) Transportable Portable Installed Portable Personal (Portal Monitors, (Trolley monitors continuous air samplers air samplers Hand & Foot Monitors mounted (Of, 6) air monitors (gross Of, 8, Qf, 8 Counters for monitors) (gross Of, 8, Iodine & swipe counting) Iodine) spectrometry)
Figure - 2 JAERI-Conf 97-008
Country Report for Training Workshop on Contamination Monitoring
l. 4 Status of Contamination Monitoring in Radiation Protection Activities of National Atomic Energy Agency (NAEA) in Indonesia
by
Gatot Suhariyono Center for Standardization and Radiation Safety Research (CSRSR) National Atomic Energy Agency (NAEA) PSPKR-BATAN, P O BOX 7043 JKSKL Jakarta 12440 Indonesia
Paper presented at IAEA / RCA Training Workshop on Contamination Monitoring October 21 - 25, 1996 JAERI, Tokai, Japan
18- JAERI —Conf 97 — 008
INTRODUCTION
National Atomic Energy Agency (NAEA) or Badan Tenaga Atom Nasional (BATAN) is a Non Departmental Governmental Agency, headed by a Director General who is directly responsible to the President.
In carrying out the task, the Director General of BATAN is assisted by four Deputies Director General, one Director of Education and Training Center and two Special Staffs.
The Deputies Director General are coordinating part of main tasks and functions of BAT AN and directly responsible to the Director General. Special Staffs have responsibility to treat and solve the problems upon direction of Director General.
Center for Standardization and Radiation Safety Research (CSRSR) is one of the research centers within the Deputy for the assessment of Nuclear Science and Technology of the NAEA. CSRSR was established in 1981 as Dosimetry and Standardization Center. The name of Center for Standardization and Radiation Safety Research was officially declared in 1986 as a result of the expansion of NAEA organization in the same year.
The main task of the CSRSR is to implement research and development program, development and services in the field of radiation safety, standardization, dosimetry, radiation health as well as the application of nuclear techniques in medicine, according to the policy confirmed by the Director General of BAT AN.
Task of radiation protection division is to set up programs and to develop radiation protection, personal monitoring system and radiation level of the working areas and their surroundings as well as dose limitation system, to carry out technical up grading of radiation protection officials skill and to help coping with radiation accident.
- 19 JAERI-Conf 97-008
The Structure of organization of CSRSR is as follows :
Center For Standardization and Radiological Safety Research Administration
Division
Personel Finance Logistics Scientific Documentation
Sub Div. Sub Div. Sub Div. Sub Div.
Security Standardization Dosimetry Radiation Nuclear Radiation
Radiation Protection Technology Effect Assessment Health Inst for Medicine
Calibration and Standardization Radiation Instrumentation
IMPORTANCE OF CONTAMINATION MONITORING
There are two types of radioactive contamination, i.e. : a) , surface contamination b) . airborne contamination
The surface contamination leads to various hazard or troubles. In a working environment where radioactive substance is handled, therefore, the surface contamination must be adequately controlled
Surface contamination monitoring is made for following purposes :
-20 JAEEI —Conf 97-008
a) . To detect contamination so as to determine its existence or spread and so as to control its movement from areas of higher contamination to those of lower contamination or to inactive areas. b) . To evaluate activity per unit area in order to verify that the permissible limits (derived limits) are not exceeded. c) . To detect failures of the containment and departure from the proper operating procedure. d) . To keep contamination of the workplace and of the skin below the control standard level. e) . To provide information for planning the individual monitoring and the air contamination monitoring and determining the operating procedure.
Surface contamination causes the following :
1) . Suspending radioactivity to the air, resulted is air contamination which leading to internal exposure by inhalation.
2) . Attachment contaminant to body surfaces, such as the hands, leading to internal exposure by oral intake.
3) . Contaminant on the wound enters in to body, leading to internal exposure.
4) . Body contamination such as on the hands and the skin, leads to external exposure.
Normal route of the workers taking radioactivity in to the body is via the inhalation. Therefore, air contamination monitoring is important in the working environment.
Air contamination monitoring is made for the following purposes :
1) . To estimate maximum quantities of radioactivities the workers may inhale.
2) . To prevent the intake of radioactivity at the occurrence of an unexpected air contamination.
-21- JAERl-Conf 97-008
3) . To choose suitable protective measures depending on the level of air contami nation.
4) . To provide information for planning the individual internal exposure monitoring.
REGULATION
Based on the BAT AN Director General decree on the safe Transport of Radioactive Material of 1994, surface contamination limits are 0,4 Bq / cm2 (=10"5 pCi / cm2) for beta and gamma emitters or 0,04 Bq / cm2 (=10"6 pCi / cm2) for alpha emitters.
According to the BATAN Director General decree on the Working Safety provisions against radiation, the working area is divided as : a) . Supervised area ; area that a radiation worker might receive radiation dose of less than 15 mSv (1500 mrem) a year and free from contamination. b) . Controlled area ; area that a radiation worker might receive radiation dose of 15 mSv (1500 mrem) or more a year.
Distribution of the working area in N.AEA is shown in table 1 and derived limit of maximum concentration of some radionuclides in inhaled air is shown in table 2.
- 22 JAERI —Conf 97-008
Table 1. Distribution of the working area in NAEA
The Working Area Remissible Limits Note
Radiation Area (R.A) Medium R.A 15 5 R.D. <50 for a whole body
High R.A R.D. 2 50 or a partial body
Low C A C.L. < 0.37 for surface ( for alpha emitters) C.L. < 3.7 contamination ( for beta emitters ) Controlled Area 0.37 5 C.L. <3.7 for surface
{ for alpha emitters) (R. D > 15 ) Contamination Area Medium C A 3.7 5 C L. <37 contamination ( for beta emitters ) (C A.) no more than 0.1 of for Airborne DLMCRIA contamination High C A. C.L. >3.7 for surface ( for alpha emitters) C.L. >37 contamination ( for beta emitters ) more than 0 .1 of for Airborne DLMCRIA contamination Supervised Area Low Radiation Area 5 5 R.D. < 15 for a whole body
(R. D < 15) Very Low Radiation Area 1 5 R.D. < 5 or a partial body
R.D = Radiation Dose (mSv / year) . C.L. = Contamination Level (Bq/cm 2 ) DLMCRIA = Derived Limit of Maximum Concentration of Radionuclides in Inhaled Air (Bq/cm 2)
Derived limit of maximum concentration of radionuclides in inhaled air (DLMCRIA) had been determined by NAEA, for example, i.e.:
Table 2. Derived limit of maximum concentration of some radionuclides in inhaled air
No. Radionuclide DLMCRIA (Bq/cm 3) 1. 1-131 7.03 ,10"4 2. 1-125 9.99. 10"4 3. P-32 5.92.1 0"3 4. Tc-99m 4.07.10"1
23- JAERI-Conf 97 — 008
METHODS OF MEASURING SURFACE CONTAMINATION
Methods of measuring surface contmination in NAEA have followed the International Organization for Standardization (ISO) 7503-1 procedures. Surface contamination is measured by survey method or by smear test.
Measurement procedure of survey method in NAEA
Before making a measurement, the background count rate is determined at the place of measurement. The background count rate is checked from time to time, otherwise before each use. The geometry conditions during a measurement are as close as practicable to those used during instrument calibration; removable spacers may be used for this purpose.
For accurate measurements, the detector is held stationary for three times the response time. The distance between the detector and the surface is kept as small as practicable. Spacers may be used for this purpose.
The beta or alpha activity per unit area (As) of fixed and loose contamination on the surface being checked, expressed in Bq.cm" 2, in relation to the measured count rate is given by the following equation :
where : n is the measured total count rate in reciprocal seconds.
nB is the background count rate in reciprocal seconds.
Ej is the instrument efficiency for beta or alpha radiation in cps / dps.
W is active surface area of detector in cm2,
t is counting time (average 10 seconds).
- 24 JAERI-Conf 97-008
Measurement procedure of smear test in NAEA
The area of sampling by smear test is generally about 100 cm2 for averaging purpose. But in order to only detect surface contamination, a larger area will be useful. The smear method is capable of sampling in large areas around 300 cm2 in NAEA. The smear method is suitable for measurement of low level surface contamination.
The smear is pressed moderately against the surface to be checked by fingertips of some people.
The smear material is immersed in to the wetting agent (as HN03 1%) in the case of alpha-emitters. After sampling, the smear material is carefully dried in such a way that loss of activity is prevented. Then the wetting agent is counted with the radiation measurement instrument.
The beta or alpha activity per unit area (Asr) of loose contamination of the surface being smeared, expressed in Bq cm"2, in relation to the measured count rate is given by the following equation
Ei.F.S
where : n is the measured total count rate in reciprocal seconds.
nB is the background count rate in reciprocal seconds.
E, is the instrument efficiency for beta or alpha radiation in cps / dps.
F is removal factor, if F is not determined experimentally, a conservative value of F = 0.1 is used.
S is the area smeared in centimetres squared.
-25- JAERI-Conf 97-008
Measurement procedure of radioactive concentration in air
Sampling air from working areas or sampling air at the point where there may occur local contamination are taken and it is collected in a filter paper as glass fiber filter papers (Whatman GF/A) or High Efficiency Particle Air (HEPA) filter with average sampling flow rate 30 1 / min during one hour, and radioactivity trapped in the filter paper is counted with the radiation measurement instrument. So, the airborne radioactive concentration is evaluated.
Flow meter airborne inlet
Air sampler Pump Whatman filter
fig. 1. a sampling system for airborne radioactive
The airborne radioactive concentration (Ca) is obtained by the following equation :
n~nB Ca = (Bq / cm1 ) E.F.T. 1000 where : n is counting rate of the sample (cps)
nBis background counting rate of the counter (cps)
E is counting efficiency of the counter (cps / dps)
F is average sampling flow rate (1 / min)
T is sampling time (second)
- 26- JAERI-Conf 97-008
If radioactive substances in air exhausted is radio-iodine (as I-131), in order to improve this collection efficiency, the activated carbon cartridge as charcoal filter impregnated with TEDA is used sampling during one hour. Radioactivity trapped in the charcoal filter is counted using Gamma Spektrometry with HPGe detector during 30 minutes. Therefore, equation of radio-Iodin radioactive concentration (Ca) becomes the following equation :
n~nB Ca = {Bq/cmi) E.F. 7.1000. where : t is time of ( sampling time + counting time) / 2
X is decay constant of 1-131 radioactive
MONITORING FOR SURFACE AND AIR CONTAMINATION IN NAEA
Periodical surface and air contamination monitoring is made once a week or once a month at representative locations in the working area. Measurements are then made in the place where the spread of contamination is detected most efficiently, the place where the contamination occurs most frequently, the place around of the entry / exit of a controlled area and the place where the contamination must be kept at low levels.
Contamination check of the following may be also convenient to detect any rise of contamination level in the working area ; cleaning mops in the working area, mats at the entry / exit, bottoms of working shoes, etc. Especially, the hand and foot monitor is useful to detect the occurrence of the contamination check when leaving the controlled area.
-27- JAERI-Conf 97-008
Table 3. Result of average surface and airborne contamination monitoring in some places ofNAEA every weeks on August 1996
No. The organization The place a Cont. p-y Cont. Airborne Cont. in NAEA (Bq / cm 2) (Bq / cm 2) (Bq / cm3) 1. CSRSR Laboratory of 0.87 to 2.89 — Standardization Laboratory of environment 0.435 to 1.737 -- - chemistry 2. RWTTC Installation of Radioactive 5.5E-4 to 5.9E- a) 2E-9 to 3E-9 # Waste Treatment 4 b) 2.3E-8 to 4E-8* Interim Storage 5.5E-4 to 5.7E- a) 2E-9 to 2.5E-9 4 # b) 3E-8 to 3.2E-8* 3. RPC Hot Cell Service Area -- - 0.208-0.641 1.049E-5 (1-131)
Process of invent Uranium 0.019 to 0.111 8E-7 to 8.5E-7 # Capsule 4. NFEC Operating Area 1.98E-4 to 5.95E- 3.8E-8 to 4E-7# 4 Service Area 1.09E-4to 3.13E- 4.6E-8 to 1.2E-7# 4
Note: CSRSR = Center for Standardization and Radiation Safety Research ; -— sign is not detectable RW’l l C = Radioactive Waste Treatment Technology Center ; # sign is Alpha gross RPC = Radioisotope Production Center ; * sign is Beta gross NFEC = Nuclear Fuel Element Center ; cont. = contamination Average p-y surface contamination of shoes covers and laboratory coats is 0.27 Bq / cm2.
DECONTAMINATION
Purpose of decontamination is to let disappear radioactive contamination, at working area, equipments and instruments.
The equipments decontamination is done immediately to prevent the contaminant object to cling to equipments surface. If the contaminant object has short half life time, then the contaminated object is saved earlier up to reach safety activity level. Decontamination of contaminated protective clothing and footwear is
28- JAERI —Conf 97-008
done at the special cleaning place that contamination level can be monitored. The equipment surface that has been decontaminated is covered by two paint layers with different colours.
Decontamination is done by wet method, because dry method makes possible to rise dust that contains radioactive substances..
Decontamination methods of contaminated surface in NAEA are usually done by chemical methods, although the other decontamination method is used too, i.e. electrochemical method, mechanical method, etc. Objects of chemical decontamina-tion are among as follows :
- Destilation water
- Radiacwash
- Detergent
- Solution of HNO3 3 - 5 %
- Solution of H2SO4 3 - 5 % etc.
Decontamination is efforted in such a way that decontaminated surface is not broken and is done carefully, so that the contaminant object doesn ’t spread.
If decontamination can’t be done up to reach safety limits, then the contaminated instruments and the work equipments along with contaminated clothes are behaved as radioactive waste
CONTAMINATED SCRAP MATERIALS HANDLING
Contaminated scrap materials that are used in radiation working and those can’t be decontaminated up to reach safety limits, then these materials is behaved as radioactive waste.
Radioactive waste storage is done suitable with the groups, i.e. based on :
29- JAER1 —Conf 97-008
a. Solid or liquid phase b. High or low activity
c. Long or short half-life time of radioactive waste
d. Inflammable or not
Radioactive waste storage container is covered by a plastic bag and a absorbtion paper. Radioactive waste storage is done in such a way that radioactive waste does not spread to the environment.
If radioactive waste storage container has been full, it is taken to Radioactive Waste Treatment Technology Center.
CONCLUSION
The key factor on contamination monitoring is to reduce human error and mechanical failures. These problems can be achieved to the highest degree by develop knowledge and skill of staffs via trainings or courses on contamination and decontamination, so that they are hoped to become trained and qualified staffs. These trained people will work according to observed procedures for radiation protection and to control normal and emergency operations. In NAEA, we are trying to achieved these objectives with limited financial resources.
REFERENCES
1. Yoshio Ikezawa, “ Radiation Control in Facilities”, IAEA / RCA Training Course on the Basic Techniques of Radiation Protection, JAERI, Tokai, 1991.
2. International Organization for Standardization (ISO), “Evaluation of Surface Contamination” , ISO 7503-1, 1988.
-30 JAERI-Conf 97-008
3. Berger, J.D., “Manual for Conducting Radiological Surveys in Support of License Termination” , NUREG / CR-5849, US Nuclear Regulatory Commission, Washington DC, December 1993.
4. Badan Tenaga Atom Nasional (BATAN), ’Working Safety Provisions Against Radiation ”, Jakarta, 1989 (in Indonesia).
5. Badan Tenaga Atom Nasional (BATAN), "Provisions for the safe Transport of Radioactive Material”, Jakarta, 1994 (in Indonesia).
6. Severa, J., and Bar, J.,"Handbook of Radioactive Contamination and Decontamination ”, New York-Tokyo, 1991.
31 JAKRl-Conf 97-008
1.5 The Status of Contamination Monitoring in Radiation Protection Activities of Korea ( Surface & Airborne )
Oct. 1996
Presented at the Regional (RCA) Training Workshop on Contamination Monitoring (both surface and airborne) JAERI, Tokai, Japan, October 21 - 25, 1996
Jong-Il Lee
Korea Atomic Energy Research Institute ( KAERI )
-32- JAERI-Con f 97-008
Contents
I. Surface Contamination 1. Contamination Monitoring 2. Instruments 3. Trends of Contamination in Facilities 4. Problems and Experiences of Monitoring
II. Airborne Contamination 1. Contamination Monitoring 2. Instruments 3. Trends of Contamination in Facilities 4. Problems and Experiences of Monitoring
HI. Attachment - Radioactive Waste Management Program in Korea
33- JAERI —Conf 97-008
I. Surface Contamination
1. Contamination Monitoring
1) Purposes (1) Prevention of diffusion for contaminated material (2) Offering Information related to exposure of workers (3) Confirmation not to exceed control level of contamination
2) Type of monitoring (1) Monitoring for the body, clothes and things to be taken outside - contamination limit : a-emit nuclide : 0.37 kBq/m 2 non a-emit nuclide '• 3.7 kBq/m 2 (2) Special monitoring (for working condition) - subject = working tools and workplaces - evaluation : • working method • estimation of contamination in air • determination of protecting ways and means (3) Routine monitoring for controlled area - contamination limit a-emit nuclide : 37 kBq/m 2 non a-emit nuclide : 370 kBq/m 2 - main survey point • entrance of controlled area • entrance of each working room • passage • around hood - survey frequency • often contaminated area : weekly • rarely contaminated area : monthly
3) Control system for the surface contamination
worker health physics staff inspector / KINS self control survey & periodic inspection restrict or recommend
34- JAERI-Conf 97-008
2. Instruments
1) Indirect Method : Smear Test (1) filter : whatman smear filter paper ( dia. 4 cm ) (2) detection : - a/3 Low Background Counting System ( gas flow type proportional counter ) - frisker ( GM ) - portable contamination counter ( GM )
2) Direct Method (1) whole-body contamination monitor ( gas flow type ) (2) hand-foot contamination monitor ( gas flow type ) (3) portable floor monitor ( gas flow type ) (4) frisker ( GM ) (5) Nal scintillation monitor
3. Trends of Contamination in Facilities of KAERI
'95 unit : kBq/m 2 Detection Annual Annual Facility Subject Average Maximum Convesion Facility a 0.25 4.07 CANDU Fuel a 0.21 3.26 Fabrication Facility Research Reactor Facility 3 0.48 4.91 (TRIGA Mk-III) i Radio Isotope 3 0.30 9.66 Production Facility(Seoul) HANARO Facility 3 BKG BKG (30 MW Rx & RIPF) Post Irradiation a BKG 0.43 Examination Facility 3 0.83 22.6 Radioactive Waste a BKG 0.08 Treatment Facility 3 0.78 6.75 Irradiation Material 3 0.01 0.04 Examination Facility Radioisotope Waste 3 0.02 0.10 Treatment Facility
35 JAERl-Conf 97 - 008
4. Problems and Experiences of Monitoring
1) Sampling Error (1) exactly scrubed area ( It is actually difficult to scrub 100 cm2 exactly ) (2) difference of scrubed press ( Text book says that we must scrub with normal press, but what is the normal press ? ) (3) difference of scrubing method ( zigzag, circular, straight )
2) Optimization of sampling point and number
-36- JAERI-Conf 97-008
II. Airborne Contamination
1. Contamination Monitoring
1) Purposes (1) Observation of occurrence for air contamination (2) Offering Information related to internal exposure of workers (3) Confirmation not to exceed the control level of contamination
2) Type of monitoring (1) Continuous monitoring - purpose to observe change of air contamination value - giving a alert or alarm (2) Special momitoring (for working condition) - evaluation of more exact air contamination value - analysis of contaminated radionuclide - determination of protecting ways and means (3) Routine monitoring for controlled area - contamination limit •' MFC or DAC - main survey point : • main workplaces • around hood - survey frequency : • often contaminated area : weekly • rarely contaminated area : monthly
3) Control system of the airborne contamination
worker health physics staff inspector / KINS self control survey & periodic inspection restrict or recommend
37 JAERI-Conf 97-008
2. Instruments
1) RMS (Radiation Monitoring System) (1) remote control at II.P. & automatic detection and storage (2) the established place : IIANARO Facility
Stack Duct Reactor Hall RIPF IMEF Monitor Monitor Monitor Monitor Monitor
H P. Room Computer Screen
(3) the special quality of monitor unit : Bq/m 3
Channel ...... Gaseous Gaseous Particulate Iodine Type (normal) (abnormal) Detector plastic Nal plastic plastic Type scintillator scintillator scintillator scintillator Detection 3 Y 3 3 Subject Calibration Cs-137 1-131 Kr-85 Kr-85 Nuclide Sensitivity 1.95 x 10' 2.93 x 10' 8.84 x 10" 5.07 x 10'
Max. range 1.81 x 10' 9.84 x 10" 1.24 x 10'" 2.42 x 10'°
2) Air sampler (1) portable air sampler - flow rate : 20 - 30 1/min - filter : glass fiber filter!particulate) & charcoal filter(Iodine) - sampling : 1 hr. - detecting • by low background counting system after 4 hr. & by gamma spectrometer (2) personal air sampler ( wearing it during working time ) (3) special air sampler ( tritum )
38 JAERI-Conf 97-008
3. Trends of Contamination in Facilities of KAERI
'95 unit : Bq/m 3 Detection Annual Annual Facility i Subject Average Maximum j Convesion Facility a 0.13 0.37 CANDU Fuel a 0.11 0.95 Fabrication Facility Research Reactor Facility 3 0.54 1.97 (TRIGA Mk-III) ...... J Radio Isotope 3 81.6 444 Production Facility (Seoul) I HANARO Facility 3 BKG BKG j (30 MW Rx & RIPF) ! i Post Irradiation a 8.99 x 10" 1.74 x 10" | Examination Facility 3 1.41 x 10~3 2.37 x 10'3 Radioactive Waste a 0.12 0.35 Treatment Facility ... 3 0.43 2.33 Irradiation Material a, 3 BKG BKG Examination Facility 1 ...... , Radioisotope Waste a, 3 BGK BKG Treatment Facility ______
4. Problems and Experiences of Monitoring
1) Difficulty of quantitative analysis for noble gas as nuclide.
2) Difficulty of determination for sampling position & representative value in case that worker's active area is large.
3) proper exchange time of filter in case of continuous monitoring
-39- JAERl-Conf 97-008
Attachment
Radioactive Waste Management Program in Korea
-40- JAERI —Conf 97 — 008
Contents
I . Legal System of the Atomic Energy Laws
II. Regulatory Organization for Radioactive Waste Management
HI. Policy and Strategy for Radioactive Waste Management
IV. Radioactive Waste Treatment and Storage
V. Radioactive Disposal Site and Facility
VI. Number of Radioactive Material Users
VC. Annual Collection of Radioactive Waste
VE. Record Keeping and Reporting
IX. Number of Radium Sources
X. Exemption Rule
- 41 JAERI-Conf 97-008
I . Legal System of the Atomic Energy Laws
1. Atomic Energy Act : Law The Act defines the fundamental matters concerning development and utilization of atomic energy and safety regulation.
- Enacted in March 1958
- Revised in May 1986
2. Enforcement Decree of the Act : Presidential Decree
The Decree defines systematic, technical and administrative matters necessary for enforcement of the Atomic Energy Act.
- Enacted in Sep. 1982
- Revised in Oct. 1989
3. Enforcement Regulation of the Act : Prime Ministerial Ordinance
The regulation provides for the matters regarding the enforcement of the Atomic Energy Act and the Enforcement Decree of the Act.
- Enacted in April 1983
- Revised in Jan. 1990
4. Notice by Minister of Science and Technology
The Notice provides for detailed administrative procedures, technical standards and guides based on the Atomic Energy Act.
The matters related to Radiation Protection and Waste Management are described in the section of the above mentioned Law, Decree and Regulation, respectively.
- 42 JAERI-Conf 97-008
II. Regulatory Organization for Radioactive Waste Management
Prime Minister
Atomic Energy Commission
MOST
KINS Radiation & Environment Div. - Health Physics Dept. - RI Regulation Dept. - Radioactive Waste Regulation Dept.
User of Radioactive Materials
* MOST : Ministry of Science & Technology KINS : Korea Institute of Nuclear Safety
IH. Policy and Strategy for Radioactive Waste Management
1. Low- and Intermediate-Level Radioactive Waste - Disposal Method : Rock Cavern Type - Capacity of Repository : . 250,000 drums(first phase) . 1,000,000 drums (final) 2. Spent Fuel Interim Storage - Centralization Concept - Storage Method : Wet or Dry Type - Capacity of Storage : 3,000 MTU 3. Radioactive waste and spent fuel from the power plant are being stored at each plant site until disposal facility is in operation 4. Waste producers must bear the Expense of Waste Management
-43- JAERI-Conf 97-008
IV. Radioactive Waste Treatment and Storage
I Flow of Radioactive Waste Collection & Storage
MOST
Report
KAERI Treatment & Storage
Transports tion
KRIA Licensee of Collection
Consignment Inspection of Package
Waste Producer
KRIA : Korea Radioisotope Association
Nuclear Environment Management Center(NEMAC), KAERI
subsidiary, has treated the radioactive wastes collected by KRIA since
1993, and the treated wastes are being stored in KAERI's storage
facility.
V. Radioactive Waste Disposal Site and Facility
Two years ago, a small island(about 2 km2) was selected as a disposal facility site which is located in West Sea of Korea. But the plan was canceled by objection of inhabitants in and around the island.
44- JAERI-Conf 97-008
VI. Number of Radioactive Material Users
(MOST : As of Dec. 1993) Radiation Gas Classification RI Total Generator Chromatography General 414 284 57 755 Industrial NDT 24 25 - 49 Firms Sales 23 - - 23
Hospitals 100 32 - 132 Educational & 85 125 4 214 Research Institute
Total 646 466 61 1.173
VH. Annual Collection of Radioactive Waste
(KAERI/NEMAC : As of Dec. 1994)
Unsealed Source Sealed Source Unit : drum
Year Non Non- Combustible Hepatitis number Major Nuclide combustible compactable «°Co. '37Cs. '"Am. 1991 134 19 12 22 83 3H. “Ni *Co. *'lr. '37Cs. 1992 743 120 27 131 763 =H.*Ni
*Co. 3H. “Ni. 1993 1.004 151 20 173 81 '"Am. 147 Pm
^Co, 24lAm. 85 Kt. 1994 576 70 22 64 127 ' - ^ MNi. "®Ra
Total 2,457 360 81 390 1,054
45 JAERI-Conf 97-008
Vi. Record Keeping & Reporting
All users of radioactive materials must keep their inventory records
including the amount of acquisition, use and waste, and they must
report them to KINS quarterly.
DC. Number of Radium Sources
(MOST : As of Dec. 1994)
Activity 101- 201- 301- 401 - 501- (mCi) < 100 Total 400 User 200 300 500 1,000 Hospitals 40 2 1 3 1 47 Industrial 36 36 Firms i Research 1 10 I 10 Institutes Wastes Stored ; i 10 30 1 i 40 in KAERI 1 Total 96 32 1 3 1 133
* The collected radium wastes are stored in lead shield container at the
KAERI/NEMAC storage facility, and these will be treated with cementation method.
X. Exemption Rule
In 1994, The Ministerial Nctice(No. 94-17), an exemption rule based on the Atomic Energy Act. was enacted, and has been put into effect. •
• Exemption Level
- Effective Dose : < 10 uSv/y
- Collective Dose : < 1 man-Sv/y
- Specific Nuclides : < 100 Bq/g
46 JAERI —Conf 97-008
1.6 Status of Nuclear Technology in Malaysia
Bustami bin Abu Malaysian Institute for Nuclear Technology Research
MINT(Malaysian Institute for Nuclear Technology Research) was first established in 1972 by the name of "Tun Ismail Atomic Research Centre" . Its infrastructural development on the 27-hectare site at Bang! commenced in Jan.1979i reaching its full operation in June 1982 with the commisioning of its nuclear research reactor. A second 81-hectare site, Dengkil Complex, was acquired in 1984. The first name(PUSPATl) was later renamed the "Nuclear Energy Unit"(UTN) in June 1983 on being placed under the auspices of the
Prime Minister's Department. In Oct.1990, UTN was retrahsferred again to the Ministry of Science,Technology and the Zhvironment, and finally assumed its new identity as "MINT" on 10**1.Aug. 1994 .
In order to have a distinct separation of roles between promotional and regulatory functions, UTN formulated Act 304, the Atomic Zhergy Licensing Act of 1984, paving the way for the establishment of the "Atomic Zhergy - Licensing Board"(LPTA) as a separate entity, in Feb.1985 .
MINT is an R & D organisation, established with the responsibility for introducing and promoting the use of nuclear science and technology in the national development. MINT has its own mission, to enhance national develop ment and economic competetiveness through excellence in nuclear and related technologies. Dedicated to the introduction and promotion of a totally new field of technology in the nation, MINT thrives on innovation and creativity. MINT products and services range from standardised items, such as the steri lisation of medical products by gamma irradiation, repair and calibration of radiation measuring devices, neutron activation and radiochemical analytical services. Through customised services suah as the production of radiopharma ceutical, industrial and agricultural radiotracers, column scanning for oil refineries, and design and installation of nucleonic control systems, to training and consultancy in areas pertaining to nuclear and related technology. In addition, MINT has developed a reputation in policy planning and analysis, including nuclear weapons non-proliferation policy advisory, the formulation of national food irradiation guidelines, and energy planning with specific emphasis on nuclear power planning.
-47 JAERI-Conf 97-008
Aa a knowledge-base organisation, MINT is developing a strong CL foundation aa a research contractor. One major contact involves the provision of integrated agronomic management consultancy services to an agricultural development authority. Efforts are being made to develop similar integrated consultancy services in the industrial fields. MINT also conducts active RAD programmes for the generation and development of new productsand processes in 4 out of 5 IRPA-defined sectors (Intensification of Research in Priority Areas). MINT has served over 800 corporate clients, ranging from public and private hospitals procuring MINT-produced radiopharmeutical and repair and maintenance services for nuclear medical equipments, through medical product manufacturers procuring gamma sterilisation services, palm oil exporters procuring radiochemical analytical services, oil exploration and industrial testing companies also procuring non-destructive evaluation services, metal industry and electronic product manufacturers procuring waste management and dosimetry calibration services respectively, to process industry and forest product companies also procuring radiological safety advisory services. Truly, MINT is a broadly diversified, client-oriented, R&D organisa tion. The overall services given by MINT include : 1 ) Standardised Products. 2) Customised Products. 3) Training Services. 4) Consultancy and Advisory Services.
At present, MINT has several facilities such as j 1) Nuclear Research Reactor (neutron irradiation) 2) Cobalt-60 Irradiation Facility (gamma irradiation) 3) Electron Beam Machine (electron irradiation) 4) Radioactive Waste Treatment Centre 5) Secondary Standard Dosimetry Laboratory
48 JAERI-Conf 97-008
ORGANIZATION CHART IN MINT
DIRECTOR GENERAL
J PLANNING UNIT
DEPUTY DIRECTOR GENERAL
r~ RESEARCH DIVISION SUPPORT SERVICE & COORDINATION DIVISION
NON-DESTRIVT! \ 'E TRA/MNG EVALIAIION
HEALTH & RADIATION CONTROL L (BUSTAMI BIN ABU
ISOTOP & RADIATION 1 ENGINEERING I IN AGRICULTURE I J
MATERIAL SCIENCE I PHYSICAL 1 & TECHNOLOGY 1 DEVELOPMENT 1 x Jr l ... J
— . ( X INSTRUMENTATION 1 IRRADIATION 1 ADMINISTRATION & CONTROL 1 FACILITY I A SERVICE ^•WnnnnnwwgBBSBgBfaaiMWiiflaBBBB^^ COMPUTER J FINANCE CUSTOMERS SERVICE J
-49- JAERI-Conf 97-008
l. 7 Country Report on Contamination Monitoring
Navaangalsan OYUNTULKHUUR Central Radiological Laboratory National Centre for Hygiene, Epidemiology and Microbiology, Mongolia
Mongolia is landlocked country in the east Central Asia with large area of approximately 1565 thousand square kilometres and relatively small population of 2.4 million people. Mongolia is a non-nuclear country and we have currently neither nuclear power plants nor research reactors.
Our country joined the Regional Co-operation Agreement (RCA) for the Asia Pacific Region for Research, Development and Training Related to Nuclear Science and Technology in 1993. During this period our country has taken part all the activities organized under this agreement.
Our country has taken various measures for strengthening of radiation protection, cooperated with several international organizations mainly with IAEA and carried out radiation protection policy. "Strengthening of radiation protection", technical co-operation project (MON/9/004) has been implemented since 1991 with the assist ance of the IAEA.
In Mongolia radioactive substances and sources are used for the following purposes: for research work; medical radiotherapy and diagnostic radiology in hospitals; indus trial enterprises for technological processes and for non-destruction testing in indus try.
The largest use of radiation sources in Mongolia is in medicine, where radioactive isotopes are used in the State Oncological Centre. Approximately 130 medical estab lishments employ X-ray machines for diagnoses. As to research, 8 scientific estab lishments use radioactive substances and radioisotope sources in addition one elec tron accelerator microtron (22MeV/ electron).
-50- JAERI-Conf 97-008
The National Centre for Hygiene, Epidemiology and Microbiology (NCHEM) is the defacto regulatory body in Mongolia. The Central Radiological Laboratory (CRL) of the NCHEM is entrusted with the regulatory function. The present regulatory activi ties cover radiation monitoring for the personnel at the various establishments, super vising sectoral radiation safety services for medical personnel, inspections physically inventory of radiation sources, performance a radiation monitoring of foodstuff and water samples, issues authorization for work with ionizing radiation sources and a radiation protection situation in institutions working with radiation and radioactive materials.
Radiation safety inspection in conducted by the CRL. The inspections consist of verification of compliance with radiation safety requirements and with any additional requirements specified in the authorization of accounting records and a physical check on the presence of radiation sources; check on the work carried out by the radiation safety services to monitor radioactive contamination of the environment; and measurements and sampling.
The major thrust of a regulatory control programme should be on the protection of radiation workers, patients and the public involved with medical diagnostic X-rays, nuclear medicine, radiography, nucleonic gauges, consumer goods, sealed and un sealed sources and research and related investigation.
The NCHEM carries out licensing and inspection activities on regular basis in institu tions using radiation sources and radiation generating equipment including hospitals.
We use in Mongolia approximately 850 sources of radiation in about 25 organisa tions. A diagnostic X-ray units about 350 quantities operate in all the biggest hospi tals of the capital city, regions (aimaks), provinces and districts in whole Mongolia.
Routine work of the CRL involves development a personnel dosimetry system (based on TL dosimeters), maintaining radiation safety on premises, and supervising and consulting other institutions and laboratories on matters pertaining to radiation safety. Already 200 radiation workers which mainly for staff at X-ray devices are being monitored routinely.
— 51 — JAERI-Conf 97-008
The CRL provides expert advice on national radiation protection regulations to other laboratories and at a national level. We had organised seven National Radiation Protection workshops, training and seminars covered about 100 participants in the last 2 years.
The laboratories operating directly under the Nuclear Energy Commission (NEC) Mongolian Government, which acts as the chief governmental body for radiation safety. The NEC's main activities are a regulatory body for waste management and supervision of radiation safety.
Emergency preparedness comes under the Mongolian Civil Defence and includes the tasks of a rescue, a cleanup, etc.
Monitoring of internal contaminations is practically not carried out. Radiochemical analyses are used for determination of Cs-137 and Sr-90 in food, environmental and biological samples for global fallout monitoring.
Environmental radioactivity monitoring is carried out by the Environmental Radi ation Laboratory (ERL) of the Ministry of Nature and Environment. This work in cludes gamma monitoring, airborne and precipitation radioactivity measurements. Fall-out radiation monitoring started in 1984 and now sampling from 16 points using sedimentation method is done regularly on a weekly basis. The ERL is responsible for monitoring airborne contamination. There are, in total, 4 high volume sampling stations (gamma-dose ratemeters) used for airborne contamination and three of them installed in the regional centres in the country side and last one in Ulaanbaatar working for 24 hours a day. The stations are working since 1992. Now the daily, seasonal and yearly fluctuation of atmospheric beta-activity in Ulaanbaatar has been estimated.
The CRL is participating in the activity of Global Environmental Radiation Monitor ing Network (GERMON) programme organised by WHO/UNEP.
A routine surface contamination monitoring for working places is carried out at the institutions which using unsealed radiation sources and CRL is responsible to inspect their activities.
52- JAERI-Conf 97-008
Nuclear medicine equipment is lying only at Nuclear Medicine Department, First State Clinical Hospital. The Nuclear Medicine Department of the hospital carries out various diagnoses ( liver, bone, kidney, thyroid, lung etc. ) for patients using Tc-99 generators, 1-131 and 1-125 isotopes compound from Amersham, UK. The laboratory spends approximately lOGBq of Tc-99 generator, 200MBq of 1-131 and lOOMBq of 1-131 with hippuran per month and uses the SIEMENS Gamma camera with com puter system controller and micro dot imager. Also, use an NP-360, NP-354 (Hun gary), HP (USA) counters for research a function of thyroid and kidney.
There is a Radioimmunoassay Laboratory of the hospital use 150uCi of the 1-125 isotope per month. Radioactive nuclides calibrated by Calibrator CAP INTEC CRC- 15R (USA).
Before 1990, the laboratories (State Oncological Centre, First State Clinical Hospital) had used many of above-mentioned radionuclide, which from Democratic Republic of Germany and those nuclides used not only for diagnostic also in treatment. The treatment of therapy patient has come to a stop since 1990 depending from financial difficulties.
The quality control for nuclear medicine equipment not carried out completely at the Nuclear Medicine Department and the department has not separate water channel for radioactive liquid disposal. There are just small tanks ( total volume about 5 m3) of the disposal for temporary storage.
Nowadays, two more institutions are using 1-125 for research, but activity of the preparat is not high. Because of limited use of unsealed nuclides in Mongolia at the CRL we have not been take more attention in this field during last time.
The major instruments for surface monitoring at CRL are following: Universal radiometer model RUST 3S, Polish made, with the beta, gamma detection zonds; Silena Nuclear Information Processor SNIP 204G, Italian made with GM tube model SWGM B468 N for beta, gamma detection. We have ordered alpha, beta (gamma) contamination workplace Monitor model CONTAMAT FHT 111 M at CRL and will be received in 1997.
-53- JAERI-Conf 97-008
Calibration service for the dosimeters and contamination monitors is not available in Mongolia currently. Dosimetry calibration Laboratory of the National Institute for Standardization and Metrology is responsible for these activities, but laboratory building is under the construction now. Therefore, the laboratory is going to send some instruments to SSDL in China for calibration. This is one of the problems for Radiation Protection service in our country.
National Environmental Monitoring Network is required for environmental monitor ing and more airborne sampling stations are needed for monitoring whole country.
CRL should take much attention for improvement and development of the activities in the field of surface contamination monitoring concerning a wide use of radionu clides in different field of economy. I hope this training workshop will be very fruitful for our laboratory activity and it will be give a fine opportunity to acquire measure ment techniques and contamination level evaluation methods which we needed.
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1.8 CONTAMINATION MONITORING IN RADIATION PROD-CHON ACTIVITIES IN MYANMAR
KAY TH1 THIN & SE1N HT'OON NUCLEAR PHYSICS LABORATORY PHYSICS DEPARTMENT YANGON UNIVERSITY MYANMAR
Abstract The radioactive contamination in rainwater, seawater, air. milk powder and other eatahles were measured with low level counter assembly. The measured activities are found to he very low and well within the maximum permissible level.
Introduction The Union of Myanmar is not a country where nuclear weapons are tested and there exists no nuclear reactors. The existence of radioactive fallout can only reach our country from the air. the atmosphere and the rainwater due to contamination from other countries. In the Union of Myanmar, measurements on rainwater, air. seawater, milk and milkpowder, some edible vegetables, beans and oil have been carried out since 1966 in the Physics Department of Yangon University and also in the Myanma Scientific and Techno logical Research Department. Rainwater and air samples were collected from the various locations in Myanmar. Seawater were collected from the different locations of southern coast of Myanmar.
Beta Counting System For measuring the beta activity low-level beta counter is used. A low-level beta counter has been constructed employing the principle of anticocidence shielding. It is composed of two counters. They are guard counter and the Geiger Muller counter. Guard counter measures the radioactivity in the atmosphere and GM counter measures the radiation that is produced from the sample. A lead castle is used for shielding the whole assembly. ( Fig . I ) The background level of the system is 1.7 counts per minute. The efficiency of the counter was calculated by using the energy value of the standard sources. It was done by drawing a graph of energy versus efficiency. The efficiency of the counter was taken by the reference energy of ^^Sr. The activities of the samples were accumulated and then the strength of the samples were determined by diecl comparison with ^ Sr standard. The absolute activity is calculated from the spectrum. In this calculation, the necessary errors and the efficiency of the detector are taken into account.
-55- JAERI-Conf 97-008
Gamma Counting System Nal (T1) detector with Canberra Series 40 MCA is used for measuring the gamma activity. The experimental set-up is shown in Fig. 2. The efficiency of the detector for proper energy is obtained from the efficiency versus energy curves. To obtain these curves the standard gamma sources are employed.
Results From the measurements the activities of the samples are found to he very low and well within the maximum permissible concentration level recommended by 1CR1\ FAO and WHO.
References 1. Derived Intervention Levels for Radionuclides in Food. WHO. Geneva, 1988. 2. Manual of provisional Instructions Anticoincidence Low Background Tube Mount and Sample Holder Type D 4144, Lahgear Ltd.. Cambridge. Hngland. 3. Safe Handling of Radionuclides, Code of Tractive, IAEA Safety Standard. Safety Series NO. 1, 1973. 4. Hla Pe and Chit Hhaing, Radioactive Fallout in Air During 1966-67. Union of Burma, Jour, of Science and Tech., 1968. 5. Kay Thi Thin, MSc Thesis, Studies on the Fnvironmenlal Radioactivity in Myanmar. 1991. ( Unpublished ) 6. The Radiation Measurmenl of the Atomic Hnergy Department ( 1995-96 ). Myanma Scientific and Technological Reserch Department, 1996. ( Unpublished )
56 Low Voltage Power Supply- JAERI-Conf
Anti-coi njridence Scaler Circuit
97-008 Counter
Guard Counter
Fig . 1 Block diagram of the low level beta detection system. JAERI-Conf 97-008
Lead cover
Lead shield
Container Sample Nal(Tl) Scintil lation detecto
Pre-ampl i f ier
High Voltage
Power Supply
MCA Teletype/ Series 40 Cassette
Fig . 2 principle of the experimental set-un.
-58- JAERI-Conf 97-008
1.9 CONTAMINATION MONITORING ACTIVITIES IN KANUPP S.SARWAR NAQVI KARACHI NUCLEAR POWER PLANT KARACHI - PAKISTAN
1.INTRODUCTION
The Karachi Nuclear Power Plant (Kanupp) is a 137 MWe pressurized heavy water reactor,designed and erected by the Canadian General Electric Company (CGE) as a turn key project. The plant is in operation since it was commissioned in the year 1972, It belongs to the first generation of the nuclear power plants constructed during 1960's.It is a natural uranium, heavy water moderated and cooled, generating station located at the Arabian Sea coast about 15 miles to the west of Karachi. The plant is operating as an intergral part of Karachi Electric Supply corporation (KESC) grid system. During its more than two decade of operation, the plant has generated about 8 billion units of electricity with an average life time availability factor of 60%. The plant, with a designed live of 30 years, has comp 1eated 25 years of its successful operation.
2.MAIN ASPECTS OF CONTAMINATION IN KANUPP
In Kanupp radioactive contamination may exit due to the release of fission product,activation products etc, which may somehow escape from its confinment and may contaminate surface or other media such as air, water etc. The following contamination is expected at KANUPP.
i. Surface Contamination Radioactive Material found on surfaces of pipes, valves, floors etc.
i i . Air-borne contamination. Radioactive Vapours, Particulates or gases floating around in the air Some of the forms in which the air—borne contamination may t in Kanupp are:
a Radioactive Vapours:- The two most important radioactive vapours are iodine and tritium.
U Radioactive Particulate;- Such as radioactive dusts or smokes.
c Radioactive gases — Suvh as K—88, Xe-135 & A-41 are commonly found in anupp'reactor.
59- JAERI-Conf 97-008
.3.STATUS OF CONTAMINATION MONITORING IN KANUPP
A number of radiation monitors have been installed at various points throughout the plant to provide a complete scan and warning of radiation hazards to the plant personnel, the amount of radioactivity which is being injected into the environment and any other abnormal radiation condition. These monitors are in service continuously and are set to annunciate various alarms at pre—adjusted levels of radiation or radioactivity.
Following Monitoring Systems are working continously in KANUPP.
i. Personnel Contamination Monitoring System. ii. Area Gamma Monitoring System. iii. Tritium Monitoring System. iv. Stack Radiation Monitoring System. v. Drainage Monitoring System.
4. NEED OF CONTAMINATION MONITORING
In Kanu.pp the radioactivity is on account of the fission products which are present on the components of the active reactor systems, such as Primary Heat Transport piping, valves, pumps etc. Once it is out of the system such as for carrying out any maintenance job on the component, the radioactivity can spread out to those surfaces where it is not wanted. e.g. personnel's hands and clothes working tools, floors, job location etc. To stop this radioactive contamination from spreading out to clean areas, the station has been divided into four contamination zones 5. RADIATION PROTECTION ACTIVITY IN KANUPP 5.1 Station Access Status of each individual's access to various zones is indicated on his Access Control Card . A ~ed strip indicates no qualification and such personnel must stay in zone-1 areas as their un—escorted entry to active zones is not permitted. They can however accompany a qualified person (Green stripe holder) or proceed on their own it they obtain necessary sponsorship. The ne t category is a Yellow stripe which signifies that the person has had a basic R-T briefing, can proceed to zones 2 and 3 without restriction and can carry out his assignment. However he car ret proceed to zone-4 unless accompanied by a qualified person under whose supervision he is permitted to perform his task. A green stripe indicates full access to all the plant areas. This is obtained after cleaning an RPT—checkout, briefing on demonstration fami 1arisation with the protective and monitoring equipment. 60- JAERI-Conf 97-008 5.2 Radiat ion/Contamination Monitoring Equipment. In the active zones, equipment is readily available for monitoring radiation and contamination level in different locations. In addition to Fixed Radiat ion/Contarninat ion monitoring system, Portable equipments are also available , They included radiation survey meters , contamination monitors, air particulate monitors, air—borne tritium manitors,Swipe monitoring locations etc. All areas where radiation level or airborne radioactivity may exist are routinely surveyed by Radiation Surveyor on duty (part of Health Physics) and the existing dose/ contamination levels are duly posted in the respective areas so that all personnel can be aware of the existing hazards. 5.3 Protective Clothing, For carrying out work in radioactive zones, personnel are required to change from their personal clothings into the 'Whites" to prevent from the spread of contamination to Zone—1. This also protects personnel from being contaminated. Such personnel will enter into active zones through change room where they leave their personal clothes in their lockers and change into "Whites” and safety shoes. At the end of the job they return to the change room, remove "Whites"etc (and leave the whites in cotton 1aundary for washing ) take a shower and put on their street cloths from the locker. White dungarees are required for access to Zone 3&4. For any job which may invove splashing of radioactive material, a plastic apron or disposable plastic suit must be worn over the dungrees. Flastic suit and hood, with provision for fresh air supply must be worn for all work in tritium —contaminated atmosphere or with wet radioactive materials. Plastic suits and breathing-air—cylinders are available at the Main Air-lock where the field station of Radiation Control is located. In addition to the protective clothing, air-line respirators are also available for tritiated atmosphere or where there max be air—borne radioactive dust. Rubher Area In KANUFP entire Reactor Building is a Rubber Area".(Rubber Station at the Main Air lock ) and therefore any rubber area et up within it is designated as "Rubber Change Area. Rubber A^e a is setup, even within active zones. Such as Zone 3 & 4 in order to contain the loose contamination which is 1 i ke1y to be caused by the job being performed such as over- hauling of a piece of equipment from some active system. 61 JAER1 —Conf 97 — 008 5.5 Maximum Permisible Contamination/Radiat ion Levels Regarding surface contamination, the levels accepted to be safe as adopted by KANUPP arei — i. For contamination of skin and personnel clothing. 10—* uCi/cma over a 300 cm3 area. ii. Por protective clothing, after adequate laundring. 5x10—* uCi/cm2 of beta—gamma contamination. These levels are in continuity with the prescribed level of the international commission of radiological protection.C1CRP). For air-borne contamination Kanupp has also maintains the same contamination level as prescribed by ICRP. 6. INSTRUMENTS 6.1 Measurement of Surface Contamination. The beta—gamma surface contamination are measured by two different methods. i. Direct Method: ii. Indirect (Swipe) Method. 6.2 Measurement of Air-borne Contamination. The different kinds of air-borne radioactive contamination are measured separately, on account of the differences in their collection and detection techniques. In KANUPP one of the most frequently necessary assessment is that of the amount of tritium in air for which two different methods are in use. i. Tritium Bubbler Method. ii. Ionisation Chamber Method. The monitors employed for the measurement of the concentration of particulate radioactive material in air have generally two distinct parts. i. Air Sampler. ii. Radiation Detector With Counter.. -62- JAERI-Conf 97-008 DECONTAMINATION I terns which have bacome contaminated with radioactive material are cleaned before they can be re-used. Washing or sc cabbing with soap and water and mild chemical, such as "Radiac wash "are used for soaking contaminated equipment before scrubbing it with brush. Decontamination facilities like ultrasonic cleaning and steam jet cleaning are also available in KANUPP . After carrying out any radioactive job personnel are required to remove whites and gloves etc and leave these in Cot ton/Rubber 1aundary for washing and decontamination. 8. CURRENT STATUS OF CONTAMINATION SURVEY MATERIALS AMD THEIR DISPOSAL A wide variety of radioactive wastes e.g. solids ,gases and liquids of various activity levels and half—lives result from the operation and maintainance of KANUPP. All radioactive waste that can net be immediately disposed off may be placed in suitable storage. Their natural decay will lessen the degree of hazard. Long term storage facilities are also available in KANUPP,for the more hazardous long lived wastes. A relatively "permanent" storage may be done by putting the high level waste containers in concrete lined trenches in a suitable area marked off for this purpose. 9. ENVIRONMENTAL MONITORING So far various methods of controlling the release rates and radiation doses within the recomended limits of DNSRF and ICF.P have been outlined. The controlled release may result in accumulation of activity in certain regions and may effect the public. Therefore, following aspects of environmental monitoring are also covered at KANUPP. i. Fence post dosimetery ii. Monitors in the city. iii. Biological sampling monitoring. 63 JAERI-Conf 97-008 Table 1 Derived Maximum Permissible Release Limits (Gaseous Radioactivity) Maximum Permissible Annual Release Limit Radionuclide Continuous Release Rates BqXlO' 4 (Curies) Bq/s (Ci/s) Iodine-131 1.33 X105 ( 3.6 X10-6) 0.042 ( 114) Tritium 2.59X10* (0.7X10:) 81.4 (220,000) Noble Gases 2.59X10* (0.7X10:) 81.4 (220,000) Particulate (Unidentified) 2.59X10* (0.7X10 2) 0.014 (38) Particulate (Not containing Sr-90) 4.44X10* (1.2X10") 1,406 (3,800) Table 2 Environmental Biological Sampling at KANUPP Sample Location Frequency Milk Vicinity of the Plant and from Malir Quarterly Vegetable Sampling Seasonal Vegetables from Malir Quarterly Grass Vicinity of the Plant, KANUPP Colony, Malir and Clifton Quarterly Soil Vicinity Quarterly Water Taps and Sea Quarterly 64 JAERI-Conf 97-008 l.io CONTAMINATION MONITORING by ARLEAN L . AL AM ARES Science Research Specialist Philippine Nuclear Research Institute INTRODUCTION By virtue of Republic Act 2067 as amended and Republic Act 5207, as amended the Philippine Atomic Energy Commission (PAEC), now renamed Philippine Nuclear Research Institute (PNRI) is the government agency charged with the regulations and control of radioactive materials in the Philippines. Since its functions through the licensing of all activities involving radioactive materials and Atomic Energy Facilities in the country. Directly responsible for the implementation of these regulatory and licensing function is the Nuclear Regulations Licensing and Safeguards Division (NRLSD). The NRLSD is organized into six (6) sections with distinct functions and responsibility. These are the following: 1) Standard Development Section (SDS) - responsible for development and standardization of the relevant rules, regulations, criteria, and standards. 2) Licensing Review and Evaluation Section (LRES) - responsible for the review and evaluation of license applications and the subsequent issuance of appropriate PNRI license. 3) Inspection and Enforcement Section (IES) - deals with inspections, audits and enforcement of the PNRI regulations. 4) Safeguards Section (SGS) - responsible for the safeguards of special nuclear materials including the fuels of the erstwhile mothballed PNPP-1. 5) Radiation Protection Section (RPS) - responsible for the radiation protection services. 6) Radiological Impact Assessment Section (RIAS) - responsible for the radiological impact assessment including environmental radiological monitoring and evaluation. The protection against the hazards of non-ionizing radiation is being monitored by the Radiological Health Service (RHS) of the Department of Health pursuant to the provision of Presidential Decree 480. The RHS issues licenses for the possession, 65- JAERI-Conf 97-008 handling, and use of X-ray machines and equipment, both industrial and medical, and provide radiation protection training to X-ray technologists and technicians. RADIATION PROTECTION PROGRAM As part of its regulatory function, the PNRI is charged with the responsibility of assuring that the radiation workers and the public are protected from the hazards associated with the possession, handling, production, manufacturing, and the use of radioactive materials and atomic energy facilities in the Philippines. The PNRI controls and regulates the use of radioactive materials by individuals or organizations through implementation of the rules, regulations and orders of the Institute. Such individuals are deemed qualified by education, training, experience and have submitted appropriate programs for the safe handling and use of radioactive materials and energy facilities as well as the adequate design and construction of the relevant facilities where the radioactive materials are to be handled and used. The protection of radiation workers from the hazards of ionizing radiation has always been a primary concern of PNRI and by limiting the exposure of radiation workers, the risk to population is kept to within acceptable level. RADIATION PROTECTION SERVICES Radiation Control - leak testing of sealed sources - area monitoring - air monitoring - contamination check of hot laboratories - transport clearance of shipment of radioactive materials - RP operational support of multi-purpose Co-60 Irradiation Facility and TRIGA Research Reactor Personal Monitoring - film badge services - TLD personal dosimetry services - bioassay services SSDL Calibration - calibration of radiation monitoring instruments * survey meters * pendosimeters * contamination meters - maintenance of dosemeters — 66 — JAERI-Conf 97-008 Radioactive Waste Management - solid and liquid waste * collection * segregation * treatment /conditioning * storage * disposal - spent sealed sources * collection * immobilization * storage Special Services - radiation hazards evaluation survey of facilities - output calibration of therapy sources - decommissioning of radiation facilities - quality control test of nuclear medicine instruments * dose calibrator * uptake probe * gamma counter * rectilinear scanner * gamma camera RADIATION CONTROL Air monitoring To determine the level of radioactive concentrations in the air inside the facility, air monitoring is done twice a day, morning and afternoon. Instrument used is a Portable Air Monitor, MISCO Brand with GFA filter. At the research reactor building the instrument used is a Portable Dust Sampler PDS Type L-100 with GFA filter also. The airborne activity is then counted and computed to determine if the average concentration does not exceed the control standard level. The normal route of radiation workers taking radioactivity in the body is via inhalation. Therefore, air monitoring is important to determine: maximum quantity of radioactivity the workers may inhale; prevent intake of radioactivity at the occurrence of unexpected air contamination; select suitable protective measures depending the level of air contamination; and provide information for planing individual internal exposure monitoring. 67 JAERI —Conf 97-008 Surface Contamination Check Surface contamination may exist in the working environment as a result of handling radioactive materials and treatment of radwastes. Surface contamination on the floor, laboratory tables, instruments and equipment in the controlled areas are regularly monitored to keep it below the control level. Contamination of surfaces is generally measured by two methods: In survey method, surfaces are scanned with the portable surface contamination monitor. Levels of contamination found to be above background are subjected to the second method. The specified area is wiped with filter paper and then counted to determine the radionuclide content and activity. Decontamination of surfaces follow if needed. Surface contamination monitoring/check is done to prevent spread of radioactivre contamination; to keep the contamination of the work place below the control standard level and to provide information for planning the individual monitoring. PROBLEMS ENCOUNTERED/RECOMMENDATION: Acquisition of additional equipments for air and surface contamination monitoring to cater to the increasing number of clients. The radiation control staff needs further training and experiences in the field of air and surface contamination monitoring and maintenance and repair of equipments. -68 JAERI-Conf 97-008 l.li Country Report for the Regional (RCA) Training Workshop on Contamination Monitoring ( 1 ) . Name: T.H.S. Shantha. (2) . Name of the organization: Atomic Energy Authority, Sri Lanka. (3) . Introduction: The Atomic Energy Authority (AEA) is the organization in Sri Lanka responsible for the regulatory and development of applications in Nuclear technology. AEA was established in 1969 by the Atomic Energy Authority Act No. 19 of 1969 and at present it is under the Ministry of Science, technology and Human Resources Development. The Minister can appoint upto seven members on the Board of directors including the chairman of the Authority. The chairman is the chief Executive of the Authority. As a developing country, Sri Lanka still need technical assistance to support the development of application in Nuclear Technology. Participation in regional cooperation activities make it possible for Sri Lanka to develop the Nuclear Technology. Without such assistance the technology transfer to a developing country like Sri Lanka development of Nuclear Technology would be extremely difficult. -69- JAERI-Conf 97-008 Major Activities of the Authority. (a) . Protection of the radiation workers arid general public from unwanted exposure to Ionizing radiation. (b) . coordinate with other organizations to enhance the utilization of nuclear technology for national development. (4). Radiation protection Activities in Sri Lanka. According to our organizational chart, the regulatory and promotional activities are separated. This is very useful when one organization is engaged in both protection and regulatory work. The main activities under the radiation protection is as follows. (a) . Licenling and inspection of places where radioactive materials and irradiating apparatus are used. (b) . Personnel monitoring service using TLDs and Film Badges. (c) . Advice and arrangements for transport of radioactive materials. (d) . Advice the radioactive waste disposal activities. (e) . Organization of Radiation Protection training programs. (f) . Provide Industrial and medical advisory service. (g) . Quality assurance in diagnostic Radiology - 70 JAERr-Conf 97-008 (. h) . calibrations using the S.S.D.L. Facility. (Amershan OB85-BA) Co-60 - 37 GBq 87 April Cs-137 - 740 GBq 87 May (5) . My official duties: I am working at the Radiation Protection and Regulatory Division. My main field of work is licensing and inspection of facilities where radioisotopes and X-Ray machines are used. This inspection include irradiator facilities, research stablishment, X-Ray facilities, NUT facilities etc. Contamination monitoring and decontamination procedures of these facilities and equipment are also carried out as a major part of the inspection. I engage the quality control procedures related to a wide varity of X-Ray applications (therapeutic and diagnostic) and calibration of health Physics instruments and personnel monitoring devices. (6) . Emergency Preparaedness and Planning: So far our country doesn't have any type of nuclear reactors. Most of activities in our country are in health sector using in X-ray and radioisotopes for diagnostic and therapeutic purpose. In addition to that few number of radioisotopes are used in industrial and research institutes. There are 04 Co-60 Gamma Irradiators available in our country. - 71 - JAERl-Conf 97-008 INITIAL PLACE SOURCE PURPOSE ACTIVFITY Ansell Lanka Ltd, Co-60 Sterilization of Surgical 01 1 - 4 MCi Biyagama (Nordien) Gloves Tissue Bank, Co-60 Sterilization of Human 02 10,000 Ci Colombo 07 (Shepherd) Tissues Radioisotopes Centre, Co-60 Radiation Processing 03 5,000 Ci University of Colombo (Shepherd) and Research Central Agricultural Research Institute, Co-60 04 2,500 Ci Agricultural Research Gannoruwa, (Bare India) Peradeniya - 72 JAERI-Conf 97-008 The ironiztion radiation used in Sri Lanka can be olnsified as follows: (a) . Medical Sector: (for the purpose of radiology,radiotherapy, radioimmunoassy, sterlization of human tissues etc. ) (b) . Industrial Sector: (for the purpose of NOT,Tracer Technics, Unclear gauges) (c) . Agricultural and Research Sector: (for the purpose of agri cultural and reseach activities.) The Atomic Energy Authority has appointed a committee recently to prepare and emergency plann for most important areas in which radiological emergencies are likely to occur. Emergency plans are now being develop in the following areas: (a) . Emergency respons for lost or stolen sources or materials. (b) . Emergency respons in the event of stuck or expose sources. (c) . Emergency respons to contamination incidents. (d) . Arrangements for medical assistanence. (7). Food Testing Programme: After the Chernobyl accident, the Atomic Energy Authority started testing of all food items imported to Sri Lanka with the assistance of the Radioisotopes Centre, University of Colombo. At the begining of the testing severa Is samples -73- JAERI —Conf 97 — 008 which were contaminated with radio Ceasium were detected.Due to lack of adequate facilities at present testing restrict only for milk food. The Atomic Energy Authority has recently purchased all the equipments required for food testing and intend# comments testing all the food items imported to Sri Lanka with the collaboration of the Radioisotopes Centre, University of Colombo. (8). Nuclear Technics in Monitoring Insustrial Pollution: (a) . Pollution of water bodies: This has become of major problem in Sri Lanka as large quantities of effluents containing metals are discharge into water bodies from industries causing harmful effect on aquatic organizm . Studies being carried out at the Radioisotopes Centre, University of Colombo, to determine bio-accumilation of matel in species of fresh water fish in selected inland water bodies in Colombo areas. (b) . Measurement of Tracer Eliment in Air: Atomic Energy Authority officers carry out multi-eliment analysis using EDXRF in aerosol samples collected from different locations in the country. (c) . Measurement of Natural Radioactivity in Air: Atomic Energy Authority Officers measure the radioactivity in several parts of the Country. The maximum levels are between 5-10 jiSv/hr. which is due to diposition of Monosite in coastal areas of the country. - 74 JAERI-Conf 97-008 Soil and air samples are being measured for radioactivity and have not shown detectable levels of artificial radioactivities in the samples. These activities are being carried out under Interanat ional Atomic Energy Agency cord in a terl technical corporation project jointly by the Atomic Energy Authority and the Radioisotopes Centre. -75- JAERI-Conf 97-008 ORGAN(ZATIONAL CHART OF THE I NFR A-STRUCTUR F FOR RAO ATIOM PROTECTION [ministry OF SCIENCE TECHNOLOGY j |& HUMAN RESOURCES DEVELOPMENT j I j BOARD OF j------1 CHAIRMAN j ADVISORY | | MANAGEMENT | | | COMMITTEE I 1 | HEAD,RADIATION RADIATION PROTECTION ] j PROTECTION AND SUB COMMITTEE OF THE j j REGULATORY BOARD I I DIVISION 1 [licensing & | I PERSONAL | | EMERGENCY | I INSPECTION I DOSIMETRY I I TRAINING PREPAREDNESS j (UNDER DEVELOPMENT)| L - 76- Advisory Coeel ttee ! Chairman i 1 Board of i iPersonal Assistant: : Hanagement I •(Board Secretary I Internal Auditor: 1Scientific i Scientific I 1 Secretary : Secretary I Administration iDivision of Secretary iladlation !Protection ! (See fig. JAERI-Conf :Division of .i : IAEA IDivision of ! IGeneral :Nuclear Hedlclne : I Dlvlson! iIndustrial i !Scientific :and Agriculture ------!ApplicationsI !Division •Division of IDlv. of -j Coordination I : !Finance 4 1General A A; iLicence 4 Environ I Establishment Iministr • Dosimeters mental : Section Honltorlng 97-008 4 Waste Disposal Nuclear : :Nuclear! :T c : :bca 4 : IHDT iladlation Section Rsdiclne! lAgrl. i •Section! iTralmg : ! urn •Processing SO,TO : £ action 1 ! Section! : ! :Section: :4 ih : id Analyti- ical : : i 1 I Services : so, TO iSection SO i ! SO I C/7-1 ! ------:so, to : iso. TO : : Library 4 i lComputer: :Nuclear iPublications: !Section : 1Instnimen ------I -tatlon KlDT - Non-destructive Testing I/NTA - Industrial Tracer Applications XH - Isotope Hydrology I CT/2 i ISO I ------: TO : : so ------I TO JAERI-Conf 97-008 1.12 The Status on Contamination Monitoring in Thailand. Fookiat Sinakhom Office of Atomic Energy for Peace Bangkok, Thailand October 22, 1996 Introduction Thailand has embarked upon the development of nuclear energy for peaceful utilization ’s since 1961 when the Atomic Energy for Peace Act was enacted. The Atomic Energy Commission “Thai AEC” was established under section 5 of this Act having power and duty of carrying out matters concerning atomic energy for peace. The Act contains under section 19 a provision that shall be instituted of the Office of Atomic Energy for Peace “OAEP” given the duties among with are the regulation of activities involving ionizing radiation every matters in accordance with the resolution of Thai AEC and executing all other administrative affairs. Organization Chart of OAEP is shown in Annex 1. The applications of nuclear energy in Thailand, at present are exclusively in medicine education research and industry. The main radiation facilities are tabulated in the following table. Radiation facilities in Thailand. Facility______Competent authority ______Number of facilities Nuclear research reactor OAEP 1 Isotope production laboratory OAEP 1 Neutron generator OAEP 1 Research gamma irradiators OAEP 3 Industrial gamma irradiator OAEP 2 - 78 JAERI-Conf 97-008 Radiation barchytheraphy DRPS 16,(2 lunits) 226 Ra, 137 Cs, 198 Au, 192,Ir, I82-.Ta Radiation teletheraphy DRPS 15,(25units) Linear accelerators photon energy 4 MV-10 MV, electron energy 4-20MeV. Co-60 irradiators photon energy 1.17 MeV. And 1.33 MeV Ortho voltage X-rays 150 kV-350 k Superficial X-ray 50 kV-100 kV Radiation diagnosis DRPS 3900 x-rayapparatus40kv-150kv:Computerized tomography, Digital subtraction angiography, Fluoroscopic x-rays, General radiographic x-rays, Mobile x- rays, Dental x-rays, Magnetic resonance imaging. Manufacture using nucleonic techniques DRPS 40 Laboratory handling radionuclides OAEP 150 Rare earth processing plant OAEP 1 Contamination Monitoring The responsibilities on contamination monitoring and related activities involve three divisions within OAEP; Health Physic Division, Radiation Measurement Division and Waste Management Division. Their main duties are summarized in Annex 2. The main activities on contamination monitoring comprise 4 categories, o Controllable monitoring o Uncontrollable monitoring o Standardization of monitoring instruments o Decontamination and waste management Controllable Monitoring : included in this category are. 1.Research Reactor Building Exposure dose rate measurements are recorded daily in ten specific positions inside the reactor building. The value in 1994 were: 79- JAERI —Conf 97-008 exposure dose rate ranges 0.05-10.50 mR/hr. airborne contamination 905 cpm. measurement of reactor coolant activity below detection limit. 2. Thai Irradiation Center Surface contamination measurements are routinely performed by indirect method, measurement of 56°Co coolant activity below detection limit 3. Spent fuel storage Twenty eight of spent fuel rods of MTR Type and 140 rods of TRIGA MARK III type are arranged safely in the pool constructed for temporary storage. 4. Isotope Production Laboratory Production of ""'Tc Fifteen measurements for exposure dose rate are performed nearby the hot cell, prior production 0.10-70 mR/hr post production 0.10-50 mR/hr Production of ,3'l Continues monitoring for air borne contamination during the production are recorded. Annually average 4.21 Bq.m 3 Maximum value 82.5 Bq.m' Minimum value 0.10 Bq.m 3 5. Monitoring resulting from radioisotope users number of licenses issued in 1994 1,507 total activity 68,263 Ci. For the safety of radiation workers in the working places, routine monitoring are performed by OAEP inspectors. number of inspections in Medicine 11 number of inspections in Education 43 -80 - JAERI—Conf 97-008 number of inspections in Industry 42 6. Control release of waste water Release limit for beta activities 37 Bq.l Release limit for alpha activities 3.7 Bq.l Release limit for U7Cs 37 Bq.l ' Uncontrollable monitoring : Contamination monitoring were prepared specifically to each level of abnormal situation and including appropriate management for decontamination. The detailed description of emergency planning procedures was published in 1994 by OAEP. Staffs training and emergency drill was demonstrated once ever since. Standardization of Monitoring Instruments OAEP has established a calibration laboratory for ionizing radiation, so called Secondary Standard Dosimetry Laboratory, SSDL-OAEP since 1979. Calibration and services irradiation on high dose gamma dosimetry, radiation protection and environmental dosimetry calibration were then initiated. SSDL-DRPS is responsible in radiation therapy and radiation diagnosis standardization including service of personal dosimeters in hospitals. These services largely contributed to improve and promote safe application of ionizing radiation countrywide Calibration equipment’s (1.) Irradiation room ( 6x4x3 m3) consists of : standard calibration irradiators . Co-60 100 mCi. standard calibration irradiators . Cs-137 2 mCi. equipped with attenuations of 1:10 and 1:100 for lower range of calibration intensities of 2 mR.hr 1 and less. 81 - JAERI-Conf 97-008 (2.) Irradiation bunker ( 11 x 4.5 x 4 m3) consists of : OB85 panoramic irradiator (low level y-source) OB34 panoramic irradiator (low level y-source) OB26 neutron calibrator Pantak superficial x-ray machine The shielding properties of the walls, ceiling and floor are adequate, no restricted working conditions applied. The ambient dose rate is 2.5 p.Sv/hr. SSDL provided calibration with photon radiation of 119 survey meters, 600 TLDS, 448 films and 32 others. The personnel monitoring services include 1721 TLDS and 522 films in 1991. Average internal collective dose equipvalent of OAEP radiation workers was 0.097 man-Sv. Decontamination and Waste Management Decontamination practices have been mainly from small contamination area in the work place. Control standard for surface contamination are shown in the following table. Surface Alpha activity,Bq.cm 3 Beta activity.Bq.cm 3 Floor covers, bench and 0.37 3.7 other surfaces Hand 1.1x10" Bq/hand l.lxlO3 Bq/hand Whole body 0.185 3.7 Safety shoe 3.7 37 Lab coat 1.85 37 Personal clothes 0.37 18.5 82- JAERI-Conf 97-008 Solid radioactive Waste Processing : ♦ Collecting of Solid Radioactive Waste ♦ Documenting of incoming waste ♦ Segregation of solid waste to be processed ( 1995 ) combustible 1.2 ton ,14 m3 noncombustible-compressible 0.4 ton , 3 m3 noncombustible- incompressible 0.2 ton , 2 m3 ♦ Volume reduction incinerator 15 kg/hr capacity vrf ~ 50-80 compactor 38.5 ton compression force vrf ~ 3-5 ♦ Conditioning cementation research for optimization formulation ♦ Interim Storage 250 drums ( 200 1. drum ) treated waste. 116 drums ( 200 1. drum ) untreated waste. References : 1. The Atomic Energy for Peace Act. B E. 2504. Thai Royal Gazette, April, 1951. 2. Annual Report 1991, Office of Atomic Energy for Peace ( OAEP 3-14 ). 3. Annual Report 1995, Waste Management Division. 4. Annual Report 1994, Health Physic Division -83 - JAERI-Conf 97-008 Annex 1 ATOMIC ENERGY COMMISSION FOR PEACE Policy MINISTRY OF SCIENCE TECHNOLOGY AND ENERGY Administration SUBCOMMITTEE - REACTOR SAFETY - NUCLEAR LAW - LICENSING OF RADIOISOTPES AND NUCLEAR - MEDICAL APPLICATION - AGRICULTURAL APPLICATION - INDUSTRIAL APPLICATION - FOOD PROCESSING TECHNOLOGY - NUCLEAR POWER PLANT SAFETY - CONSIDERATION AND SCREENING IN COOPERATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY BETWEEN SCIENCE AND TECHNOLOGY AGENCY IN IAPAN AND THE INSTITUTIONS CONCERNED IN THAILAND ORGANIZATION CHART OFOAEP(1996) -84 JAERI —Conf 97 — 008 Annex 2 Health Physics Division o Issue licenses in accordance with the Atomic Energy for Peace Act as well as establish regulation and practices related to radiation and radioactivity to interested and qualified parties, o Undertake radiation protection tasks such as control of radiation hazards. o Undertake the radiation measurement of radiation hazard exposure and record keeping of the radiation workers. □ Compile information study research and develop techniques about radiation safety, o Support or cooperate with other related divisions on their tasks. Radioactive Waste Management Division o Research on technology for management of LEW related to shallow land disposal, o Services on collection of RAW and decontamination, o Research on safety assessment for shallow land disposal, o Control and environment impact assessment from routine releases, o Public information on safe RAW management, o Support or cooperate with other related divisions on their tasks o Radioactive Waste Management Division o Research on technology for management of LEW related to shallow land disposal, o Services on collection of RAW and decontamination, o Research on safety assessment for shallow land disposal, o Control and environment impact assessment from routine releases, o Public information on safe RAW management, o Support or cooperate with other related divisions on their tasks -85 JAERI-Conf 97-008 Radiation Measurement Division o Assessment of any release of radioactivity to the environment, control and evaluate the radioactivity in the environment. o Carry out standards of ionizing radiation metrology and standardize radioactivity materials, and also provide calibration and certification of radiation measurement devices and system for irradiation processing and protection purposes. o Undertake research and development to improve method for radiation measurements, research and develop on the application of radiation measurement and promote them for both private and government sectors. o Study and develop radiation measurement procedures for the beneficial of radiation application and protection. o Provide measurement and certification of radioactivity level in report foodstuffs, o Support and cooperate with other divisions on their related tasks. - 86 JAERI-Conf 97-008 1.13 PRESENT STATUS OF CONTAMINATION MONITORING AT THE DALAT NUCLEAR RESEARCH INSTITUTE (DNRI). Hoang Van Nguyen Department of Radiation Protection Dalat Nuclear Research Institute , Vietnam. 1. INTRODUCTION The Dalat nuclear research reactor was renovated and upgraded from the previous TRIG A reactor . Now , its nominal power is 500 kW . In Vietnam , it is a unique nuclear device having suitable neutron flux for the radioisotope production and neutron activation analysis . Soon after the reactor reached its initial criticality in November 1983 , a programme has been formed to develop the application of nuclear techniques in various fields . In addition , the use of radioisotopes for diagnostic, therapeutic and other research purposes has been in progress . Now , the radioisotope production programme is concentrated on the following radionuclides: P-32, Cr-51 , 1-131, Tc-99m ,etc. Every year the DNRI produces radioisotopes and radiopharmaceuticals with the total activity about 150 Ci and provides them for hospitals in country . In order to support these activities , the radiation protection , especially the radiation contamination monitoring has been properly paid attention to . In DNRI , the Radiation Protection Department is responsible for controlling and supervising radiation & working safety for all activities . 2. RADIATION CONTAMINATION MONITORING CARRIED OUT IN DNRI 2.1. Controlled area According to the Vietnamese Regulation & safety Standards TCVN4397-87, the “controlled area” means an area where the external exposure may exceed 15 mSv per year or where the average airborne radioactive concentration may exceed 3/10 of the concentration limit or where the surface contamination may exceed 3/10 of the surface contamination limit. At the DNRI there are three controlled areas : Laboratories of radioisotope production , the reactor hall and the radwaste treatment building . In controlled areas , radiation monitoring is strictly carried out in order to keep the radiation safety for workers and to prevent the leakage of radiation and the spread of radioactive contamination . 2.2. Surface contamination monitoring In DNRI, the main causes of surface contamination are : - Radioactive materials are being handled in the area . - Incidents in chemical radioactive laboratories , such as accidental spills , 87- JAERI-Conf 97-008 contamination in the process of posting materials into and out of glove boxes , breakage of glass apparatuses , etc. - Leakage of the primary water system . - Neglect during the radwaste treatment. The surface contamination causes various hazards , especially the increase in both internal and external exposure . So it must be strictly controlled . Surface contamination monitoring is carried out for the following purposes : - To detect failures of the containment and departure from the proper operating procedures . - To prevent spread of radioactive contamination . - To keep workplace contamination and body contamination below radiation control standards . - To provide neccessary information for planning individual & air contamination monitoring and determining operating procedures . 2.2.1. Method of surface contamination monitoring There are three types of surface contamination monitoring in controlled areas : - Periodical monitoring . It is carried out once a month at representative locations such as places around hot boxes , entry/exit,equipments and area for radwaste treatment,etc. - Operational monitoring . - Special monitoring . 2.2.2. Radiation control standards . According to The Vietnamese Regulation & safety Standards TCVN 4397-87 , the surface contamination limits are given in Tab.l . In DNRI, the control standard values are set at 1/5 of the limits . 2.2.3. Methods of measuring surface contamination. Two methods are used for measuring surface contamination : survey method and wipping-off method ( smear method ). In the survey method , measurements are taken by scanning over surfaces of an object with a portable surface contamination survey meter . Characteristics of some surface contamination survey meters are illustrated in Tab.2 The survey method is simple, fast and suitable for detecting spot contamination and estimating the contamination extent. But its sensitivity is low due to external radiation . In the wipping-off method , a specific area ( generally about 200 cm2 ) of an interested object is wipped with a filter paper ( or tampon ) and then its radioactivity is measured by a G-M counting system or by a surface contamination survey meter .The filter paper / tampon may be dry or wet and it usually is burnt before each measuring . The wipping -off method is suitable for detecting low -level surface contamination densities . 2.3. Airborne concentration monitoring -88- JAERI-Conf 97-008 2.3.1. Purpose Airborne concentration monitoring is made for the following purposes : - To detect abnormal situation when the airborne concentration in the working environment exceeds the control standards. - To estimate maximum quantities of radioactivities the workers may inhalate. - To provide information for planning the individual internal exposure moni - toring . 2.3.2. Control standards In Vietnamese Regulation & safety Standars , there are airborne concentration limits for all radionuclides. In DNRI, control standard values are set at 1/5 of the limits.When the external exposure is involved , the radioactive concentration is such that the sum of the ratio of the effective dose equivalent (due to external exposure ) to 50 mSv and the ratio of the average airborne concentration to the corresponding DAC is 1 . 2.3.3. Measurement of airborne contamination a/ Forms of airborne contamination In DNRI, there are three forms of airborne contamination : - Particulate contaminants ,such as P-32 ,Cr-51 ,etc. - Volatile form : 1-131 - Gaseous form : Ar, Xe , Kr b/ Methods of airborne monitoring - Radioactive gas concentration is continuously controlled by a stack insert type monitor with four Nal detectors . - Particulate airborne contaminants are collected by a portable air sampler with flow-rate nearly 30 m3/h . The cellulose filter papers are used for the collection. - Radio-iodine contaminant is collected by the above mentioned portable air sampler using activated carbon filter papers . There are also three monitoring types : periodical , operational and special . After sampling , the radioactivity of the filter papers are measured by a G-M counting system . 3.CONCLUSION Despite the fact that our reactor has small thermal power and the radioisotope production extent is small, we also have to pay more attention to improve the conta - ruination monitoring procedures and techniques , especially calibration techniques . - 89 JAERI-Conf 97-008 Tab.l Surface contamination limits ( Bq/cm2 ) Alpha emitter OBJECT Beta emitter Special nuclei Other Skin , clothes 0.05 0.05 4 Floor,individual protective 0.1 0.3 : 30 : means and equipments in controlled areas . Container , transport means : 0.2 0.2 : 2 : — Special nucleus : nucleus with concentration limit in air less than 10 Ci/1. Tab.2 Portable surface contamination survey meters Measurement: Type of detector : Window Thick . : Window area : Detection limit (Bq/cm2): : : (mg/cm2) : (cm2) : : a, (3 : butan flow : 0.9 : 166 0.04 : a : : counter : 0.08 :P P.Y : G-M counter : 20 : 100 1 :P REFERENCE 1. Qui pham an loan buc xa ion hoa TCVN 4397-87 2. Ten years of edification and maturation of the Nuclear Research Instutute ,DNRI ,1994 3. Radiation control in facilities,Y.Ikezawa,LAEA/RCA training course,Tokai Japan, 1991 4. Basic safety Standards for Radiation protection,Safety series No.9,IAEA,Vienna, 1982. -90 JAERI-Conf 97-008 2. Special Lectures and Statement NEXT PAOKSl left BLANK JAERl-Conf 97-008 2.1 RADIOACTIVE STANDARDS AND CALIBRATION METHODS FOR CONTAMINATION MONITORING INSTRUMENTS Makoto YOSHIDA Radiation Dosimetry Division Department of Health Physics Japan Atomic Energy Research Institute 1. Introduction Contamination monitoring in the facilities for handling unsealed radioactive materials is one of the most important procedures for radiation protection as well as radiation dose monitoring. For implementation of the proper contamination monitoring, radiation measuring instruments should not only be suitable to the purpose of monitoring, but also be well calibrated for the objective quantities of measurement. In the calibration of contamination monitoring instruments ( hereafter, contamination monitors ), quality reference activities need to be used. They are supplied in different forms such as extended sources, radioactive solutions or radioactive gases. These reference activities must be traceable to the national standards or equivalent standards. On the other hand, the appropriate calibration methods must be applied for each type of contamination monitor. This paper presents the concepts of calibration for contamination monitors, reference sources, determination methods of reference quantities and practical calibration methods of contamination monitors, including the procedures carried out in JAERI and some relevant experimental data. 2. Concepts of calibration, standards and calibration techniques In this paper, two types of contamination monitoring are discussed, surface contamination and air contamination monitorings. They use various types of detectors such as GM tube, proportional counter, scintillation detector, semiconductor detector and ionization chamber. And also the objective contamination emerges in such different physical forms as radioactive powder, residue, gas and dust. Hence the radioactivity standards and techniques used for their calibration fairly depend on the kind of contamination monitors. 2.1 Surface contamination monitor Different types of surface contamination monitors are used such as surface contamination surveymeters, hand-foot monitors, cloth monitors and articles surface contamination monitors. By these monitors, the surface contamination, that is, a superficial density of activity on the surface of material, is measured in a unit of Bq/cm 2. The relation of surface contamination, Ay and the measured count rates with a surface contamination monitor is given by ISO ( International Standards Organization ) report 11 and expressed with the following form. Tim ~ Tib A, = (1) Ey EjTF Where is the measured count rate including background count rate ( s'1), n*, is the background count rate (s"1), 93- JAERI —Conf 97-008 W is the sensitive window area of the measuring instrument ( cm2), £ , is the efficiency of the measuring instrument, and £ , is the efficiency of the contamination source. The two efficiencies, £ t and £ s, were introduced as a new concept by the ISO and their estimation is very important for reliable surface contamination measurements. The meanings of the two efficiencies are illustrated in Fig.2.1-1. The efficiency, £ ,, depends on objective nuclides, measuring geometry and the detectors used for measuring surface contaminations, and is defined by the following formula. 11m lib E, = (2) eTw Where Es is the surface emission rate per unit area of calibration source ( s'cm"2). The calibration of surface contamination monitors, in general, means the experimental determination of this efficiency. On the determination of this quantity, some extended sources are needed, which are uniform in activity distribution and prepared with proper nuclide. Various types of extended sources can be prepared such as dropped source, anodized aluminum source, plastic resin source, ion-exchange membrane source etc. and some of them are commercially available. Appropriate radionuclides used for the calibration are 14C, 147Pm, 204T1, 36C1, *Sr(**Y), 106Ru(106Rh) and 241 Am( Table 2.1-1 )2). In Japan, natural uranium plane sources have been so far used for the purpose. These calibration sources must be standardized concerning surface emission rates. Absolute measurement of the surface emission rate is commonly made with a window-less 2 tt gas-flow counter. It has no detector-window causing the count-loss of radiations and its threshold detection energy is low enough. Figure 2.1-2 shows the structure of the box-type 2 tt gas-flow counter used in our laboratory, which has plural anode wires and a sensitive area of 180mmx240mm. Its plateau characteristics and uniformity in distribution of counting efficiency are good enough in the area of 140mmx220mm . The source efficiency, £ s, is defined by the ratio of the surface emission rate of radiation from a contamination source to the activity in the source. They depend on self-absorption of contamination source and backscattering due to its backing material. In principle, the value of £ s should be estimated for each contamination source. In the absence of more precisely known values, however, the following values for £ s are recommended. For beta-emitters (400keV < EPmax), £ s = 0.5, and for beta-emitters (150keV < Ep max < 400keV) and alpha-emitters, £ s = 0.25. The source efficiency for each objective material can be experimentally determined using radioactive standard solutions. The standard solution is quantitatively dropped upon the material by the gravimetrical method with a pycnometer and then dried up naturally. And the surface emission rate is measured with the window-less 2 tt counter. Figure 2.1-3 shows the results of source efficiency for some materials used in the facilities for handling unsealed radioactive mate rials3'. Since most of the materials are unpermeable for radioactive solutions, they show much larger efficiency than 0.5 because of backscattering. However, a permeable filter paper gives a small source efficiency due to large self-absorption. The practical calibration method is quite simple if proper calibration plane sources are provided. The surface contamination monitor to be calibrated is kept in parallel at a distance of 5 mm or 10 mm from the calibration source having the same active area as the monitor window. And the instrument efficiency is obtained as the ratio of the readings of the monitor to the surface emission rate of the calibration source. The final conversion factor for the surface -94- JAERI-Conf 97-008 contamination monitor is given in the form of reciprocal product of the obtained instrument efficiency and the optimum source efficiency. In the case of tritium surface contamination monitors 41, their instrument efficiency, in general, arc extremely small in comparison with those of normal surface contamination monitors. For example, it may be several % in direct contact with source surface even though a thin film less than 100 p g/cnf is used as a detector window. Useful sources for the calibration are restricted to such types as mono-molecular layer, labeled acrylic resin and anodized aluminum sources because of chemical and physical behaviors of tritium. The beta-ray energy spectrum depends on the type of tritium plane source 51. The bremsstrahlung and characteristic X-rays sometimes influence on the calibration of monitors with a very small instrument efficiency. The source efficiency is changeable depending on the conditions of tritium contamination. Hence, it is very difficult to select an optimum source efficiency for tritium surface contamination monitoring. 2.2 Air contamination monitors Air contamination monitors are usually used for the aim of measuring radioactive gases and dust (or aerosols) in working areas or outlets of ventilation from nuclear facilities. The radioactive gases are measured with the gas monitors for which detectors are mostly gas-flow ionization chambers or scintillation detectors. In the gas monitor with an ionization chamber, the relation between the output current, 1(A), and the activity concentration in the air, C ( Bq/cm 3 ), is given as follows. w ■ I C = K1 (3) r\-e-V ■ E Where E is the mean beta-ray energy ( eV ), w is the w-value for air ( eV ), V is the effective volume of ionization chamber ( cm3 ), V is the ionization efficiency, e is the charge of electron ( C ), and K is the conversion factor ( Bq cm 3 A '). In the gas monitor with scintillation detectors, the relation between the count rate, R( s'1), and the activity concentration in the air, C ( Bq/cm 3) , is written as well. C =------R = KR (4) E, E.E,K Where Ed, £aand Eg are the efficiencies concerning the detector, absorption and geometry, V is the volume of sampling vessel ( cm3), and K is the conversion factor ( Bq cm"3-s ). Calibration of gas monitors means the estimation of the Ks of Eqs.(3) and (4) with radioactive standard gases. The activities of the standard gases are determined on the basis of absolute radioactivity measurements using an internal gas counter. In the internal gas counting methods 61, there are two considerable corrections of end effect and wall effect, among which the correction of end effect is the more important. Hence, most of the institutes supplying radioactive standard gases adopt the so-called length compensation method for the correction of end effect, explained in Fig.2.2-1. In JAERI, another internal gas counting method has been developed and applied, - 95 - JAKRI —Conf 97-008 which is called the diffusion-in long proportional counter method^. The radioactive gases used for gas monitor calibration are listed in Table 2.2-1. Among them, 3H, 14C and 85 Kr can be stored anytime because of their long half-lives. However, 41Ar, mXe, 135Xe etc. must be produced timely on the calibration by using thermal neutron fields of reactors. For calibration of gas monitors, two methods( in Fig.2.2-2) are used in general. One method is of introducing a known amount of gaseous activity to the gas-loop with a known volume including a gas monitor to be calibrated. The other is of making use of the reference gas measuring instrument which is connected with the gas monitor in a gas-loop. In the former, it is not easy to determine the exact volume of gas-loop for calibration and handle the radioactive standard gases stored in glass ampoules or gas cylinders. The latter one is very convenient because the activity and volume of gas-loop are not necessarily known. A 1.5/ gas-flow ionization chamber is popularly used in Japan as a reference gas measuring instrument because of its simplicity and stability. Its ionization efficiency has been precisely evaluated for each radioactive gas used for the calibration purposed Figure 2.2-3 gives the relation between the ionization current and mean beta-ray energy of each radionuclide. In the actual procedures of calibration, attention should be paid to the gas leakage from the gas-loop, the sufficient mixing of radioactive gas and the remained contamination after calibration. A dust monitor is another type of air contamination monitor and detects radioactive dust or aerosols other than radioactive gases. In this monitor, the dust in the working environments is collected on a specially designed filter paper such as HEPA filter and the activity is measured with a GM tube, a scintillation detector or a semiconductor detector. There is no essential difference between the calibration methods for dust monitors and surface contamination monitors, excepting for the difference in thickness of the material in which radioactivity distributes. And also the same type of reference plane source as used for surface contamination monitors is applied for the calibration of this monitor. Then the source efficiency can be usually assumed to be 0.5. 3. Conclusion In the calibration of contamination monitors, the procedures are different corresponding to each type of monitors. The surface contamination monitors are calibrated with reference plane sources on the basis of the measurement concepts recommended by the ISO. For air contamination monitors, on the other hand, the sophisticated calibration method with gaseous activities has to be applied as mentioned above. The radionuclide, radiation energy and precision to be required depend on the purposes of calibration, that is, whether it is for type-tests or for routine calibrations. Through these calibrations, however, it is commonly required to supply the reliable standards. In order to carry out quality contamination monitoring, it is impor tant to recognize sufficiently the necessity for these calibrations and establish the rational calibration system. References 1) International Organization for Standardization, Evaluation of surface contamination- Part 1, ISO 7503-1 (1988). 2) International Organization for Standardization, Reference sources for the calibration of surface contamination monitors- Beta-emitters ( maximum beta energy greater than 0.15 MeV ) and alpha-emitters, ISO 8769 (1988). 3) M.Yoshida, M.Yoshizawa and K.Minami, Efficiencies of contamination source for flooring and some materials used in unencapsulated radioactivity handling facilities, RADIOISOTOPES, -96 - JAERI-Conf 97-008 39(9), 396(1990) (in Japanese). 4) International Organization for Standardization, Evaluation of surface contamination- Part 2, ISO 7503-2 (1988). 5) M.Yoshida, K.Takahashi and S.Shimizu, Radiation characteristics of tritium plane sources for calibration of surface contamination monitoring instruments, RADIOISOTOPES, 43(12), 741(1994). 6) National Council on Radiation Protection and Measurements, NCRP Report No.58, 368 (1985). 7) C.Mori et al., Radioactivity measurement of 85 Kr with position-sensitive proportional counter, Nucl.Instr.Meth., A312,189(1992). 8) M.Yoshida, Y.Wu, Y.Ohi and T.Chida, Calibration method based on direct radioactivity measurement for radioactive gas monitoring instruments, RADIOISOTOPES, 42(8), 452(1993) (in Japanese). -97- JAERI-Conf 97-008 Table 2.1-1 Radionuclides used for calibration of surface contamination monitoring instruments. Radionuclide Haif-life(years) Maximum energy Minimum backing thickness (mm) (keV) Aluminum Stainless steel ' 4 c 5730 156 0.08 003 ' 4 7 Pm 2 62 225 0.13 0.04 2°'TI 3 78 763 0.7 0.23 3 6 Cl 3.01 x105 710 0.6 0 20 »°Sr+'°Y 28.5 2274 3.1 1.1 ,06 Ru + , 0 6 Rh 1.01 3540 4.8 1.7 2 4 ' Am 432 5544 (a ) 0.02 0.01 Un,, 4.47x10^ 4196 (a) 002 0.01 2292(g) 3.1 1.1 Table 2.2-1 Radioactive gases used for calibration Long half-life Half-life Maximum (Tray fray energy energy (keV) (keV) 3h 12.3y 18.6 ,4c 5730y 156 ^Kr 10.7y 688 514(0.4%) Short half-life Half-life Maximum 0-ray 7-ray energy energy (keV) (keV) ^Ar 35d 2.2-2.8(81%) 2.6(9%) 41Ar 1.83h 1199 1294 5 24d 346 81(37%) ,35Xe 9.09h 909(96%) 249.8(90%) -98 - Upper solid angle Detector 180mm ^j /%7 'Hi1 ' % Source JAERI-Conf Lower solid angle of 2%sr. Anode wire Activity of a source(Bq): To pre-amplifier 97-008 q, +q2 +q3 +q» +q= +q6 Counting gas Counting gas Surface emission rate of a sourcefs' 1 ): outlet inlet q« +q2 +q3 +qs Efficiency of instrument £j: n -7- (surface emission rate) Efficiency of source £s : (surface emission rate) -r (activity) Sample tray Fig.2.1-1 Two efficiencies for determination of surface contamination. Fig.2.1 -2 Structure of box-type 2% gas-flow counter. Source efficiency Fig. Coated Smear-test Smear-test Stainless Stainless Polyethylene 2.1 (drop) Lonleum (wipe-off) (rough) (smooth) Material P-tile -3 concrete steel steel paper paper Experimental Mean Code CC PE PT SS-R SS-S LL FP-D FP-W beta-ray Wiping Dropping SUS304, SUS304, SheetfO. PVC PVC Coated results flooring, flooring, off with 1 radioactiove sandblast-treared mirror-like mm energy radioactivity Specification epoxy of commercially commercially thick) source resin finished (keV) solution -▼-PE — — — efficiency. O x i — available available — — PT ss-s SS-R FP-W FP-D (£ s ) Counter-2 Fig.2.2-1 Counter- Counter- ] Proportional ; Subtraction different Length 1 between Mixing pump Counter-2 compensation lengths two counters of counters count Counter-3 with method. rate 97-008 JAERI-Conf (1) Calibration with radioactive standard gases Mixing Gas pump monitor Radioactive standard gas efficiency K =______(Indication of monitor) A : Activity of standard gas(Bq) JAERI-Conf V: Whole volume of calibration systemjcm3) Ionization • Experimental (2) Calibration with standard instruments A Calculated 97-008 10 100 1 Gas Standard Mean electron energy ( keV ) monitor instrument The mean electron energy is defined by; (Activity cone, by standard instrument) (whole released energy by electrons per decay) (Indication of monitor) (total number of electrons emitted per decay) Fig.2.2-2 Two calibration methods of gas Fig. 2.2-3 Experimental and calculated ionization monitoring instruments. efficiencies for the gas-flow ionization chamber. JAERI-Conf 97-008 2.2 Monitoring and Evaluation Techniques for Airborne Contamination Xia Yihua, prof. China Institute of Atomic Energy Beijing, People ’s Republic of China (Lecture on Regional (RCA) Training Workshop on Contamination Monitoring. Tokai,Japan 21 — 25 October 1996) -102 JAERI-Conf 97-008 Monitoring and Evaluation Techniques for Airborne Contamination I . Introduction Monitoring and evaluation of airborne contamination are of great importance for the purpose of protection of health and safety of workers in nuclear installa tions. Because airborne contamination is one of the key sources to cause expo sure to individuals by inhalation and ingestion, and to cause diffusion of contami nants in the environment. The main objectives of monitoring and evaluation of airborne contamination are : (a) To detect promptly a loss of control of airborne material, also long term trends in the concentrations of airborne radioactive materials ; (b) To help identify those individuals who may have been exposed to an ex tent which warrants assessment of intake, and to predict exposure levels so that proper protection can be provided for the workers? (c) To assess the intake and dose commitment to the individuals exposed to airborne ? (d) To provide sufficient documentation of airborne radioactivity to satisfy the varied requirements of regulatory agencies. From the viewpoint of radiation protection, the radioactive contaminants in air can be classified into the following types; (a) Airborne aerosol; (b) Gas (as Td/'CCh) and noble gas (as"Ar,»3Kr, '"Xe, '"Xe^Rn) (c) Volatile gas (as 131I,129 I, 125 1 , 32 P, 35S) and vapour (as HTO, DTO, TzO). n . Sampling methods and techniques Different types of sampling can be used which take into account the operat ing and release conditions, the nature of the materials to be measured, the re quirement of monitoring etc. The sampling can be continuous, periodic, special or self-actuated. 2. 1 Sampling of airborne aerosol 103 JAERI —Conf 97-008 The total airborne particle concentration is most frequently determined by using filter papers which are commercially available with a wide variety of char acteristics. The choice of the filter depends upon the methods used for subse quent analysis and operational requirements. A sufficient variety of filters is available with a range of these characteristics. There are mainly different filter papers , they are: cellulose fibre filters, cellulose—asbestos filters, glass fibre filters , synthetic filters (membrane filters , nucleopore filters) (Table 1). Electrostatic and thermal precipatation techniques can also be used in col lecting airborne radioactive materials , but they have very low flow rate, and are difficult to maintain. 2. 2 Sampling of noble gases When the activity concentration of noble gases is sufficiently high, sampling of the adequately filtered air stream in a evacuated container of appropriate vol ume , may be sufficient. If the concentration is not high enough, noble gas can be sampled by using compressors , a technique which permits filling a suitable sampling container at a high pressure, and also collecting a large quantity of gas in a small volume , to provide greater sensitivity. After sampling the container then be taken elsewhere for analysis and measurement of the radioactive con tent. Of course noble gas can also be collected in appropriate counting chamber directly. When necessary, the sampling of noble gases can be performed by using a cold trap, for example , activated charcoal which has been maintained at low temperature by liquid nitrogen. 2. 3 Sampling of iodine Radioactive isotopes of iodine may be generated in nuclear facilities in differ ent physical and chemical forms , some of them can penetrate the aerosol filters. These different species may, however , be trapped by drawing the sample stream, after the aerosol filter, through special iodine — collecting materials, such as activated charcoal , silver zeolites or other suitable substances. To im prove the collection of different physical and chemical forms of iodine the char coal may be treated with chemical products such as potassium iodide (KI) or tri ethylene diamine (TEDA). Nevertheless, the collection efficiency might be af fected by physical parameters such as the temperature of the air sampled or the presence of organic or water vapours. The effeciency may also decrease during a prolonged sampling period. Several forms of iodine and the suitable collectors - 104- JAERI-Conf 97-008 and absorbers as follows : Form of iodine Collectors used (a) Elemental — copper or silver screens (freshly treated) — activated charcoal impregnated paper — scrubbers with a solution of AgN03 ,NaI (basic), Na2S203 — activated charcoal — molecular sieves (b) CH3I — scrubbers with solution of AgN03 — impregnated charcoal — silicic acid impregnated with AgNOa or Ag — molecular sieves impregnated with Ag (c) HOI — aluminium or silicic acid impregnated with AgN03 — activated charcoal — Ag — X13 (silver zeolite ) when humidity is low — alumina impregnated with 4 — iodophenol (a special absorber for special investiga tions) (d) Lil or other basic iodide — these are particles and can be collected on any high efficiency filter papers, as previously described in detail By arranging in the proper sequence adsorbers which are specific collectors for the different chemical forms of iodine , estimates can be made of the quanti ties of the different chemical forms present. Heating of air stream to decrease the content of water in charcoal is also a effective method of collecting organic iodides. 2. 4 Sampling of tritium and tritium oxides The forms of tritium most likely to be encountered are tritiated water vapour as HTO, DTO, T20 and elemental tritium gas. Selection of the sampling techniques is dependent on the forms of tritium of concern 2. 4. 1 Tritiated water vapour For collecting of tritiated water vapour the currently used methods are fol 105 JAERI-Conf 97-008 lowing: a) Freezing method. Passing the air through a cold trap permitting the col lection of tritium oxide vapour by condensation or freezing with dry ice or liquid nitrogen. b) Desiccant method. Passing the air through some appropriate desiccant (such as silica gel or molecular sieve (alumino-silicates) ) to retain the tritiated water vapour. Desiccants have high capacity for moisture and will retain it until it is intentionally desorbed. The collection efficiency can be greater than 99% in a sampler of adequate capacity and may be determined by employing collectors in series. c) Bubbling method. Passing the air through non-tritiated water or other ap propriate solvents (distilled water or ethylene glycol). By isotopic exchange, the tritium oxide of the air stream is retained in the water. The bubbling method is simple and effective, but it has the disadvantage that the collected water vapour containing the 3H originally will be diluted by the water in the bubbler to a factor 10 and 100. By this, the detection limit of the subsequent measurement is re duced accordingly. 2. 4. 2 Tritium gas For tritium gas cannot be trapped as easily as tritium oxide and it must first be converted into tritium oxide by passing the air stream over heated oxide cop per wool or a palladium catalyst (~300"C) supported on a molecular sieve. The oxide copper is eloper but needs higher temperature (~700"C). 2. 4. 3 Mixture of tritium Collection of mixture of tritium must combine the methods both for tritium gas and tritium oxide vapour , that is to collect vapour first, then collect gas af ter converting them into vapour (figure 1(b)). 2 2. 5 Sampling of carbon- 14 Carbon-14 may be present in air in the form of gaseous compounds such as C02, CO and CH< or other hydrocarbons. Sampling of 14C can be done by taking samples in bottles or pressurized containers. Large volume (~ 800L ) of alu minized plastic bags may also be used. However, continuous sampling over weeks or a month may be preferable. As we can see below the carbone-14 can be sampled in a suitable solvent (NaOH). 106 JAERI-Conf 97-008 H . Measurement and evaluation When a single , known radionuclide is present in working environment, the collected sample is measured with a radiation counter for quantitative estimates of the radionuclide concentration in air. When the collected sample contains mix tures of radionuclides in varying proportions, analysis must be both quantitative and qualitative. Radionuclides may be identified by their half-lives .types of radi ations emitted, or emission energies. Samples collected on filter paper are easy to handle, store and measure. Gaseous samples can be filled in a counter directly or collected in flasks then measured in a light-tight sheilded photomultiplier assembly, while those collect ed on absorbed are transferred to analytical equipment in the laboratory for mea surement. Measurement of airborne radioactive contaminants poses several problems besides the usual determination of activity. The major ones are: (a) To adapt the detectors to the mechanical requirements of the various sampling devices (size and diameter of the filter, varying air pressure,etc. ); (b) To measure low levels of radioactivity in the presence of natural ra dioactivity, particularly radon and thoron, and their daughter products; (c) To seperate radioisotopes of interest from other , less dangerous nu clides, if present at the same time. 3. 1 Measuring of airborne aerosol 3. 1. 1 Measuring after collection The samples collected can be measured in a laboratory where counting e- quipment or if necessary, radiochemical facilities are available. 3. 1. 2 Continuous measurements during sampling The most common method of continuous mesurement of airborne aerosols is to collect the samples by drawing air through a filter paper (fixed or moving) and then to measure the radioactivity with a detector mounted just above the sampling window. Usually the stationary filters are changed either daily at the end of working hours or weekly. Stationary monitors with fixed filters are best suited for measurements in rooms with relatively clean air, a typical data is that sampling 40 hours at 12L/min, the mean radioactive concentration is as low as 3. 7 X 10_5Bq/m3. To avoid long-time build-up of radioactivity and dust on the filter, instru ments with moving filter band are used. They are suitable at locations where - 107- JAERI—Conf 97 — 008 high radioactive concentrations are to be assessed and registered. The filter tape can be moved in a stepwise fashion or continuously. It is fitted with one instanta neous and one delayed measuring head. The delay is adjusted such that the natu ral radioactivity on the tape is reduced to negligible amounts. The choice of detectors for measurement depends on the type of activity to be mesured and on the size of the filters. 3. 1. 3 Discrimination against natural radioactivity The most difficult problem in monitoring low-level radioactivity in air with instantaneous read-out instruments arises from (a) the presence of natural air borne radionuclides like radon and thoron and their daughters , and (b) the varia tion of their concentrations in air with day , time of the day , weather, ventilation condition etc. Radon and thoron emanate respectively from 226Ra and 224Ra which are pre sent in variable amounts in the soil and in most building materials. Radon and thoron are both gases , but their short-lived daughters are solids and remain air borne either along or by attaching themselves to the particles in the air. They emit both a and (3 radiations. The combined half-life of radon daughters is about 30 minutes and that for thoron daughters is about 10 hours. As the DAC for alpha-active radionuclides are rather low , the interference of natural radioactivity becomes most troublesome when monitoring particle plu tonium and transpultonium elements. The discrimination methods against natural radioactivity in aerosol monitor ing depend upon the difference of properties between natural and artificial aerosol : (a) Discrimination by sampling Radon or thoron decay products are found to be associated preferentially with small size airborne pacticles where as man-made aerosols are usually larger. For example , a single stage annular kinetic impactor may have a collection effi ciency of 70% for typical plutonium aerosol particles and 10% for radon or thoron decay products. (b) Discrimination by difference of half-lives of radionuclides (i) Delayed measurements : Measurement will be done after collection delaying a long enough time (i. e. several hours even days) to allow nearly a complete decay of natural radioactivi ty- (ii) Accumulation measurements : In the mode of accumulation sampling using fixed paper , the total natural - 108 JAERI —Conf 97 — 008 activities on the paper will be approximately saturated in about 2 hours , but the man-made activities will be increased with the collection time. Thus contamina tion of artificial aerosol can be found after enough sampling time on a roughly stable background. (c) Discrimination by difference of energy of radiation The alpha energies of radon and thoron daughters are all greater than 5. 99MeV , but that for 238 u , 239 Pu , 210Po ,241 Am etc are all below 5. 5MeV. There fore it is possible to discriminate against the natural activity by alpha spectrome try (semiconductor ,grid ionization chamber). (d) Discrimination by pseudo-coincidence method The natural radioactivity is measured specifically by assessing the two disin tegrations of RaC (£D and RaC' (a) , which follows the former with a half-life of 160ps (pseudo-coincidence). The pseudo-coincident count rate is then weighted by an appropriate factor and subtracted from the total alpha count rate. (e) Discrimination by ratio of a counts/(3 counts The ratio of a counts to (3 counts of a natural activity sample will approxi mately be constant , because they all are emanated from a same decay chain and the chain is under some equilibrium condition. The artificial a-counts in the sample can be obtained by subtracting the nat ural a-counts in sample from the total a-counts in the sample : Naa = Nat —N„n = N„, —NPn/K„ where : N,,,----- total a-counts in the sample (natural-)- artificial, if any) Nan------natural a-counts in the sample Npn----- natural (3-counts in the sample K„----- ratio of natural a-counts to ^-counts in the sample under cen- tain sampling and measuring condition , it can be determined by experiments. For the pseado-coincidence method , the formula for calculation of N„„is sim ilar as that above , but in this method the constant K„ equals the ratio of natural a-counts to natural (3,a pseudo-coincidence counts. 3 3. 2 Measurement of noble gases Mostly beta radiation are used for continuous measurements. This is neces sary for 85 Kr which is virtually a pure beta emitter (only 0. 7% gamma rays at 520KeV) and cannot reasonably be detected otherwise. Three main forms of measuring chambers are commonly used: cylindrical chambers with GM or scintillation counters along the central axis,"cubic" cham- 109 JAERI—Conf 97 —008 bers with one wall housing a large area proportional counter , and flat rectangu lar chambers with the same dimensions as a large area counter and a thickness of ~10cm(;Table I'). GM counters are simple and cheap, yet they have adequate sensitivity. They have however several disadvantages. The wall thickness limits detection to beta rays of energies greater than 300KeV. The long dead-time of GM counters limits the maximum measurable concentration and their characteristics change with age. Plastic phosphors can be used with thin windows and are therefore suitable of measuring low-energy beta rays. Their response is more than one hundred times higher than for GM counters. A large surface area of the phosphor is re quired to obtain high sensitivity and a small thickness (~ 1mm) is necessary to minimize the background due to gamma radiation. Variations in external back ground can be compensated by using a pair of detectors , one of which in rendered virtually insensitive to beta radiation by covering it with a plastic sheild of thick ness of the order of lg/cm 2. Large-area proportional counters (typically~ 700cm2) are the most sensitive detectors for monitoring radioactive gases. 3. 3 Measurement of air-containing tritium The low energy of tritium beta (Emax=18Kev) makes it necessary to intro duce air containing the tritium, mostly in the form of tritiated water vapour, into a suitable detector. Three types of detectors are suitable for continuous air moni toring : ionization flow chambers, proportional flow counters, and scintillation flow chambers. Under ideal conditions modern ionization chambers with a vol ume of about 10 litres will detect ~3. 7 X 1034Bq/m 3. Proportional counters will measure less than 3. 7 X 102Bq/m 3, while the detection limits with scintillation counters having active areas of several thousand square centimetres will fall somewhere in between. If long delay times (up to 1 hour) and a considerable amount of chemical engineering hardware are acceptable , then water vapour in air can be condensed continuously and the tritiated water measured. This method largely eliminates interferences from other radionuclides and background radia tion. After collecting ,the tritiated water samples can be measured by liquid sci- entillation method. In the .freezing method , the concentration of tritiated water vapour in air can be obtained by the formula : Cair — M ♦ Cwaur/ (Ec * C) where :C8 ir----- concentration of tritiated water vapour in air (Bq/m 3) ; -110- JAERI-Conf 97 —008 Cwater—— concentration of condensed water (Bq/g) ; Ec----- collective efficiency of condenser (%); V------total air volume through the bubbler(m 3) M----- amount of condensed water (g) In the desiccant method , the concentration of tritium in air can be calculated by : Cair = M • Cwater / (E * V * d) where :E----- captive efficiency of the desiccant (%); d----- desorption coefficient , which can be experimentally determined by ratio of the tritium desorbed from to the absorbed in the desiccant (%). In the bubbling method, the concentration of tritium in air is : Calr = M0 • Cwater/^ ‘ Eb) where:M0----amount of water in the bubbler at beginning of sampling (g) ; Eb----- sturation coefficient of bubbler for HTO in the air (%). If mixture of tritium gas (HT) and tritium oxide (HTO) is present , the HT must first be converted on to HTO. A total tritium monitor based on the catalytic oxidation-bubblers for stack tritium monitoring at nuclear facilities has been developed by China Institute for Radiation Protection. Now the model ACT-100 3H/14C sampler is available com mercially. The main component is a stainless steel tube heated in an electric fur nace to about 550 C , in which the catalytic oxidation occures converting the species from sampling air to carbon dioxide and water vapour , the subsequent calcium sulphate desiccant removes all water vapour containing tritium. The de tection limit for tritium is .0. 04 Bq/L for a week sampling at flow rate of 8 0~ 200mL/min by directly liquid scintillation counting of solid suspension sample. (Table IE ,Figure 1(b)) 3. 4 Measurement of iodines The radioactive iodine isotopes 131I and 125I belong to the most important ra dioisotopes in nuclear medicine. To measure 131I the use of scintillation or semi conductor methods is much more preferable to any other method. When collect ed in an iodine cartridge, the measurement of gamma radiation is the only method because beta rays are almost totally absorbed in the volume of the car tridge itself. But if collected in a charcoal paper , beta method can also be used. 3. 5 Measurement of carbon-14 14C is a pure beta emitter with a maximum energy of 148KeV. Special sam- 111 JAERI-Conf 97-008 pling and measuring methods are necessary. Liquid scientillation counting is commonly used for 14C. The technique pro vides for good detection efficiency and normally involves a minimum of chemical preparation. The 14C02 gas can be collected in a suitable solvent and then count ed in a liquid scintillation counter. For example, the carbone-14 can be sampled as carbon dioxide in sodium hydroxide solutions : 2Na0H+C02-----«-Na2C03+H20 Na2C03+CaCl2---- *-CaC0 3 i +2NaCl It is precipitated from these solutions as CaC03. Then mix CaC03 with scintillation liquid for counting. The activity concentration of 14C in air is then calculated with the formula = Kc • Rn 2. 22 - V . ' % T)ch = E/A where :C----- 14C activity concentration (if C is to be in Ci/m3, Kc=10~12,if in Bq/m 3, Kc —3. 7X10-2) Rn----- net pulse rate (counts/min) V-----air sample volume (m3) t)2-----counter efficiency r)ch------chemical yield E----- amount of CaC03 sampe used for the measurement(g) A----- amount of calcium barium carbonated precipitated. The detection limit L is: 3Kc \jRo/t m ”2. 22 • V • t)2 • 7)ch Where ;Ro ------background pusle rate(counts/min) tm----- measuring time(min). For a measurement of 100 min, detection limit is 7. 4X10-2Bq/m 3. — 112 — JAERI-Conf 97-008 Table Ri Summary of sampling and measuring of airborne contamination type Sampling method Measuring method aerosol a) filter paper (common used) a)scintillation detector b) electrostative b ) semiconductor, counting tube c) thermal precipatation (in-situ ,in laboratory) a) container (evacuated or with a ) measuring chamber with noble gases pressure) counter (GM, scintillation , pro b ) cold trap (activated charcoal portional) at low temperature) b)plastic phosphors a ) freezing and condensing by a) ionization flow chamber tritiated water cold trap b) proportional flow chamber vapour b) desiccation c) scintillation flow chamber c) bubbling d) liquid scintillation counter converted into tritium oxide by tritium gas over heated catalyst first, then the same as above sampled as oxide a) charcoal paper b) activated charcoal scintillation counter (Nal^semi- iodine c) molecular sieves conductor d) impregnated charcoal a) container similar as that for a) liquid scintillation counter noble gases carbon-14 b) flow ionization chamber or b) sampled in suitable solvent proportional counter (such as NaOH) IV. Particle size Analysis Data obtained from the total aerosol sampling may not be necessary suffi cient for the assessment of inhalation hazard to workers. It is important to know the particle size distribution in order to estimate the deposition, retention or translocation pattern in the pulmonary regions of the lung. Particle size analysis also provides information on particle transport and deposition in the environ ment. 4. 1 Impactors - 113 JAERI-Conf 97-008 The cascade impactor is probably the preferred field instrument for deter mining aerosal size characteristics in the working environment. Collection is based on the relative inertial properties of particles in an air stream changing in flow direction from perpendicular to parallel to the impaction surface. Particles with sufficient inertia due to size and density will not follow the direction of air flow , but will impact upon and be retained by the collector surface. By a suitable choice of the separation distance between the jet and the impaction surface, of width of the jet, and of flow rate, it is possible to collect particles with specific inertial parameters. If several impaction stages of decreasing jet widths (there fore higher flow velocities) are arranged in series, successive stages collect pro gressively smaller particles. The last stage is usually followed by a filter to col lect all the particles passing the impaction stages. Impactor calibration is usually performed using a set of monodisperse aerosols of spherical shape and of known densities. The manufacturer's data is generally accepted. 4. 2 Microscopy Optical or electron microscopy is also used for particle size measurements. Such a technique is useful in research and is not common in radiation protection. It is time consuming in terms of analysis and is likely to provide results unrelated to the physical parameters of concern. 4. 3 Autoradiography and autoscintigraphy It is useful for particle size analysis of both alpha and beta emitters. In this method the sample can be exposed to a nuclear emulsion film or kept in contact with a scintillator (such as a ZnS screen for alpha detector), which in turn is placed in contact with a light sensitive photographic film) for a known pe riod of time. Then the number of tracks is a measure of the particle activity. Of ten this is the only method of determing the particle size distribution by a filter sample after an accident. - 114- JAERI-Conf 97-008 References [1] IAEA, Radiological Surveillance of Airborne Contaminants in the Working Environment, Safety Series No. 49,1979. [2] IAEA, Objectives and Design of Environmental Monitoring Programmes for Radioactive Contaminants , Safety Series No. 41,1975. [3] Xia Yihua, Measurement of a-Radioactivity in Environmental Samples, Regional Training Course for Asia and the Pacific Region on Environmen tal Monitoring and Assessment of Nuclear Facility , 1991. [4] Yang Huaiyuan et al. Environmental Tritium Measurement Techniques , see above. [5] Xia Yihua et al. Study of Collection Efficiency of Charcoal-paper for Mixed 131I in off-gas , Atomic Energy Science And Technology, No. 3,1978. [6] Xia Yihua, Equipment and Technique for Aerosol Sampling and the Activi ty Determination by Decay, Health Protection , No. 2,1963 -115- TABLE L COLLECTION EFFICIENCY AND FLOW RESISTANCE OF SELECTED AIR SAMPLING FILTER MEDIA Collection efficiency (%) for retaining Flow resistance (mmllg) 0.3 #rm DOP Filter Trade Country of origin type name Velocity (cm/s) Velocity (cm/s) 10.7 26.7 53.0 106 35 53 71 106 Cellulose Whatman 41 64 72 84 98 24 36 48 72 UK SS-589/1 46 56 66 80 18 27 37 56 FRG TFA-41 62 74 86 98 23 40 48 81 USA JAERI-Conf Cellulose- S-P Rose 99.18 99.28 99.52 99.75 38 57 75 112 France asbestos HV-70 96.6 98.2 99.2 99.8 44 64 87 127 USA Glass MSA 1106 B 99.968 99.932 99.952 99.978 20 30 40 61 USA fibre Gelman A - 98.1 98.2 98.9 - 33 - - USA 97-008 Gelman E 99.974 99.964 99.970 99.986 19 28 38 57 USA Schleicher & Schiill No.6 > 99.9 for all flow rates FRG Whatman GF/A 99.989 99.982 99.985 99.992 20 29 40 60 UK Reeve Angel 934 AH - 90.00 99.4 99.7 - 37 - - UK Membrane Millipore AA 99.992 99.985 99.980 - 98 142 195 285 USA (pore size 0.8 ftm) Gelman AM-1 88 88 92 95 56 84 117 190 USA (pore size 5 pm) Nucleopore Nucleopore 3, 0.8 pirn - - - - - 99 - - USA pore diameter a Available in a variety of compositions. TABLE S CHARACTERISTICS OF VARIOUS SYSTEMS USED FOR MONITORING INERT GASES IN AIR Type Extent of yse Detection Advantages Disadvantages sensitivity Ionization flow Earlier often, now Bad to adequate Detects radiation of Constant calibration chamber rarely used in the FRG; (approx. all energies (but without checks required, low common abroad 7.4 X I03 Bq/m3) discrimination) detection sensitivity JAERI-Conf Proportional In reactors and laboratories, Good (approx. Simultaneous and Gas consumption, flow counter only in conjunction with 3H 1.85 X 103 Bq/m3) separate detection of limited air flow measurement (FRG only) inert gases and 3II rate GM counter in Often in accelerators, Good (approx. Cheap and simple No |4C measurement, measuring chamber sometimes in reactors 7.4 X I03 Bq/m3) electronics, stable limited counting range, 97-008 SO dia.X 50 cm due to long plateau counter tube ageing Large-area plastic Until now only at HMI Good (approx. Pressure resistant, Relatively expensive, scintillation Berlin 1.85 X I03 Bq/m3) long life, wide range. highly sensitive counter in I4C can still be measuring chamber detected 50 ilia.X 50 cm Large area Frequently in reactors Good (approx. Maximum detection Gas consumption proportional counter (FRG only) 74 Bq/m3) sensitivity for in measuring chamber all energies (except 3li) 100 X 100 X 50 cm JAERI-Conf 97-008 air and a average can gas un volume for susceptible pollution air Specific disadvantages Frequent pollution produce supply Very recalibration interference Counting required to contamination Extensive hardware, maintenance Measurement represents requires reading relatively defined of good which to minutes is with (minutes) good (hours) AIR minutes) ...... ) Speed reading Very Good obtained Bad to (seconds Adequate (several to IN H 3 bad good memory against defects Bad Very Good Adequate Very with interference only MONITORING good adequate increased good good compen extensive with background Inadequate, FOR gamma None, compensation shielding adequate with sation Only Very with Very against only with with with differen only inert Protection good and radio good SYSTEMS & measurement with active None, gases drier measurement adequate differential Adequate, compensation good Very Very Inadequate, drier tial adequate ) ) ) 3 ) 3 3 3 VARIOUS Bq/m Bq/m Bq/m Bq/m s 4 2 4 OF adequate 10 adequate 10 10 10 good good X to to X X X Detection Bad sensitivity 7.4 Very 1.85 (approx. 7.4 3.7 (approx. Adequate Bad Very (approx. (approx. for of only) many FRG common Most Extent Rare use Most in countries Rare common in random mental Usually checks (Canada) system (experi CHARACTERISTICS in in and Iff out, phase phase TABLE Type Ionization chambers Proportional counters Freezing measurement Plastic scintillator chambers measurement liquid liquid Conversion (discontinuous) (continuous) 118 JAERI-Conf 97-008 Figure 1 (a) Radon daughter (RaC') a-spectrum sampled by millipore filter and measured by semiconductor □ T, Figure 1 (b) Stack HTO-HT Mixture Sampler AF——- aerosol filter, F------flow ratemeter , P, , Pi------pump, ET------catalyst and temperature control, B,B,------HTO bubbling collector, B,B,------HTO bubbling collection after HT oxidation TCGA-Series TCAL-Series TC-Series TC-08 Carbon Cartridge - 119- JAERI-Conf 97-008 IAEA/RCA Training Workshop on Contamination Monitoring Tokai, Japan, 21-25 October 1996 2.3 Monitoring and Evaluation Techniques for Surface Contamination David Woods Australian Nuclear Science and Technology Organisation Abstract To the extent that surface contamination is possible it is necessary to monitor for its presence using appropriate contamination monitoring equipment and techniques. The frequency of monitoring and the type of detection equipment used will depend on the nature and type of radioactive contamination, the level of hazard it presents, and its physical location. This paper discusses the requirements of surface contamination monitoring and the techniques to achieve this. CONTENTS 1. Definitions 2. Introduction 3. Surface contamination exposure pathways 4. Surface contamination - Derived Limits (DLs) 5. Examples of reference levels 6. Surface contamination monitoring program objectives 7. Design of a monitoring program 8. Routine monitoring 9. Recording of surface monitoring data 10. Examples of records 11. Direct surface contamination monitoring 12. Indirect methods of surface monitoring 13. Monitoring for skin contamination -120 DEFINITIONS DEFINITIONS CONTROLLED AREA SURFACE CONTAMINATION IAEA 70 definitions ______1 Active areas. • Surface contamination is unsealed radioactive material Areas in which persons may be exposed to radioactive that has undesirably deposited, spilled or otherwise contamination, radiation, or the risk of inhaling or transferred from its containment to working surfaces, ingesting radioactive material from airborne or surface tools, equipment, outer containment surfaces, personal radioactive contamination at levels in excess of those protective equipment, clothing, and skin (usually hands recommended by the 1CRP or the IAEA; these may and face). include change rooms and offices as well as laboratories JAERI-Conf and process areas. • IAEA 1996 defines contamination as the presence of radioactive substances in or on a material or the human Inactive areas Areas other than those classified as active areas. body or other place where they are undesirable or could be harmful. 97-008 IAEA 96 definitions ______• IAEA 1970 defined surface contamination as unwanted Controlled area radioactive material deposited in an uncontrolled Any area in which specific protection measures and safety manner in or on animate or inanimate objects, provisions arc or could be required for: irrespective of their situations, in such concentrations (a) controlling normal exposures or preventing the spread that either operational inconvenience or radiological of contamination during normal working conditions; and hazard is caused. (b) preventing or limiting the extent of potential exposures. Supervised area Any area not designated as a controlled area but for which IAEA 1996 IAEA Basic Safety Standards occupational exposure conditions are kept under review IAEA 1970 Monitoring of Radioactive Contamination on Surfaces. even though specific protective measures and safety provisions are not normally needed. ______JAERI-Conf 97-008 to is with skin skin to skin the the rise and be important of of contamination surface workplace measurement is internal give also of pathway It to the measurements can in probe ingestion may lead exposure. irradiation radiation irradiation activity. can Exposure beta/gamma direct PATHWAYS tritium) pathway resuspension by Pathways and loose inadvertent for external external external ingestion inhalation via contamination radiation CONTAMINATION contamination Main and alpha circumstances. exposure contamination surface activity fixed monitoring although internal surface of some fixed) between EXPOSURE (particularly and in both levels. internal Source surface surface (not SURFACE contamination compare to main detect fixed contamination loose surface contamination skin inhalation external exposure. alpha beta/gamma Contamination will distinguish reference and • by contamination, absorption • The significant is it or achieve required material. to detection whether the or possible appropriate is techniques. of contamination, surface is then of assessment and and type using cleaning. yes the techniques radioactive radiation the question If by to or of the it radioactive and on not. presents, main presence of and equipment contamination it related requirements its or the remove CONTAMINATION type radiation for the depend to INTRODUCTION to location. MONITORING hazard surface and measurement monitoring present reasons will of monitoring is of practice, monitoring likely the that for monitor in is discusses used nature level to exposure physical most of SURFACE its the the extent is Often, paper • • • frequency the this. action contamination contamination This equipment contamination The radioactivity necessary Monitoring control To 122 REFERENCE LEVELS EXAMPLES OF REFERENCE LEVELS SURFACE CONTAMINATION • For general beta/gamma surface contamination limits values obtained using the most restrictive beta emitters, DERIVED LIMITS (DLs) 9 °Sr and/or 2l0Pb, have been used. • For general alpha surface contamination limits values A reference level or derived limit or derived working limit obtained using the most restrictive alpha emitters, for surface contamination can be derived either 226 Ra and 239 Pu, have been used. • in respect to a specific contaminant on a defined surface to certain defined disturbances, or Skin and inanimate surface contamination DLs commonly used • it can be derived to apply generally to any contamination. general beta/gamma emitters and 4Bq/cm 2 JAERI-Conf less toxic alpha emitters, and The latter case is more widely used but is subject to a greater range of uncertainties and is therefore set to be for more toxic alpha emitters 0.4 Bq/cm 2 more restrictive. 97-008 In specifying a DL, its use should imply compliance with These DLs apply to workplaces only and measurements primary dose limits. may he averaged as follows: floors, ceilings, walls 1000cm2 If a surface contamination DL is exceeded this implies skin 100cm2, hands 300cm2 or finger tips 30cm2 action is required but also when the measurement is other surfaces 300cm2 or 100cm2 reported in terms of number of DLs, conveys a concept of the degree of hazard Use of these most restrictive DLs in hospitals, where only a few low toxicity radionuclides are used, would be ______e.g., <1DL or 10DL or >1Q0DL etc.,______unnecessarily cautious. Limits for fixed contamination are based on external radiation derived limits. EXAMPLES OF REFERENCE LEVELS EXAMPLES OF REFERENCE LEVELS DERIVED WORKING LIMITS FOR National Radiation Protection Board SURFACE CONTAMINATION United Kingdom, NRPB-DL2 1979 United Kingdom ore-1979 (IAEA 1970) Derived Limits for Surface Contamination Principal Low- Beta Low energy Within Outside Personal Skin Type of alpha toxicity emitters beta controlled controlled clothing Area emitters alpha emitters areas, areas emitters protective HCi/ciV Bq/cm 4 gCi/cm 4 Bq/cm 4 pCi/cm 4 Bq/cm 4 ^iCi/cm4 Bq/cm 4 JAERI-Conf clothing Inactive (Bq/cm 2) (Bq/cm 2) (Bq/cm 2) (Bq/cm 2) and low 10 s 0.4 10"4 4 to 4 4 10"3 40 Class I 0.3-3 0.3 activity 0.3 0.3 areas Class II 3-30 0.3 0.3 0.3 Class III Active 10"4 4 10'3 40 10J 40 10"2 400 30 3 3 0.3 -3 97-008 areas Class IV 300 30 30 30 Personal nr5 0.4 to 4 4 to 4 4 to 3 40 Class V 3000 300 300 300 clothing Clothing nr4 luf3^ 10’3 4 40 40 to 3 400 Clans Radionuclide not normally Th-^t, I worn in J330, "V 236U, alpha mitten vith Z > 92 active areas '"s., "«Fb. "So. u_^t. II Skin nL 0.4 10"3 0.4 to 4 4 to 3 40 O-erir, ^ Pu Other nucliden except those In Classes 17 and m V "'c. %S. "c. %S«. ",r. rv "sr. )9"r., '^cd. '"i. '"i. '"c. V 3s. S'cr. "Ml. "3C. JAERI-Conf 97 - 008 EXAMPLES OF REFERENCE LEVELS National Radiation Protection Board United Kingdom. NRPB-DL2 1979 Limiting DLs for active area surfaces and skin (Bq/cm 2) tfuclida Surface* of ectivo Skin 10^ 3 X 3 x icr % * Surface* of active craaa 3 X io 2 3 x Z2Na 3 X Up * 3 x 33p ♦ 3 X io 2 3 x 10 35s 2 3 X 10 3 x 102 1j3c 1 101 3 10 3 x 3 x 10 ^Oa 3 AO2 3 x 101 L 3 x 10 S'r X io u Up 1 *' sL 3 3 x 3 x 10 uP * 102 1 — 3 X 3 x 3 X 10 SZ?o to' * 3 X 3 SSpe 10* 4 * 3 X 3 x Up 3 x 10 56Co . * 3 X io 2 3 x .101 1 »♦ 57c= uP • 3 3 x 102 3 x io 5=oo ’ 3 102 3 x 6Ui io u 1 ** 3 3 x 3 x io ^Ou io 2 - 3 3 x 10 3 x io 67Cu ♦ 3 101 3 x 65zn 10^ • 3 3 x 10 3 x 10 67Ca 102 * 3 3 x 10 68 3 x Ca io 1 3 10 3 x io 66„ 3 x Co 101 3 3 x 3 r 10 75Se 10^ ♦ 3 X 3 x 10 3 X 10 77Br io- 3 102 2 - 3 X 3 x 3 z 10 S,Hb 102 • 3 X 3 x 10 3 x 10 ®5sx 10^ ♦ 3 X 3 x S7n_ 2 a? 3 X 10* 3 x 10,2 ** 90, 3 x Sr 1 3 X 10 3 x 10 3 x io' S?Y 2 T 3 X 10* 3 x up U-car 3 x 10 90 v 1 Y 3 X 10 3 x 3 x 10 101 3 X 3 x 3 x 10 * ,1 «■ 3 X 103 3 x 10 3 x 10 109 Cd 3- 103 3 103 3 x 3 x 10 mi= 103 * 3 X 3 x 10 3 x 10 1l5la 1 3 X io 3 3 x 101 123I 2 3 X 10* 3 x 10 3 x io 12Sl IO3 „2 ^ 3 X 3 x 129 C, ' 3 X 10^ 3 x 10 10^ * 3 X 3 x 10^ external radiation liciting 102 3 X 3 x 10 ix^cetioa linitin^ 3 3 x 102 16Nb * 3 X 101 3 x 10 - 125- EXAMPLES OF REFERENCE LEVELS EXAMPLES OF REFERENCE LEVELS Health Physics Society Australian Code of Practice draft standard ANSI N13.12 1987 Recommended limits on radioactive contamination on surfaces in laboratories (1995), NHMRC, Radiation Health Series No.38, June 1995 Guidelines for Radioactive Surface Contamination SURFACE CONTAMINATION LIMITS Limit Nuclide Group Activity Guide No. Group Description (dpm /100 cm2) Bq/cm 2 103 JH ,4C 24Na J!,S J,’CI 48 Ca 47Ca 4<’Sc 4,Sc MCr S4Mn Removable Total 1 5SFe 59 Fc 57Co 5SCo 63Ni 67Ga 68 Ge 77Br 81 Rb 8S Sr AJI alpha emitters (except those in Group II) plus 20 100 JAERI-Conf CD 87m Sr 87 Y 88 Y "Mo "n'Tc "Tc m3Ru ulAg '"in 2,0Pb(2loPo) and 223Ra(-28 Th) "3Sn 125Sb ,23I l29 Cs l3lCs l33Ba ,39 Cc 14lCe 147Nd (4) Uranium (natural, depleted, enriched [< 10 %]), Thorium 200 1,000 l53Gd lf,nTb l69 Eb l,0Tm ir,9 Yb l77Lu ,8, Hf I8S W (natural) l8fi Re l92 Ir 198 Au 197 Hg 203Hg 20'I'! 204T1 (2) *Sr, 12SI, 126I, 129 I, 131i 200 5,000 10z 22Na 32I* 5,’Co wlCo MCu 67Cu A5Zn "8 Ga 75Se 8A Rb All p,Y-emittcrs not in Groups 1,000 5,000 ,mRu (3) 89 Sr 90 Y ,IOmAg ,09 Cd "5,nCd ,13"’In ,24Sb l2SI 97-008 1, II, or III except (5-cmiltcrs with EmJX s 0.15 McVb ,3lI l34Cs l37Cs l4nLa ,47Pm l52Eu l54Eu 2,0Bi plus all other non-alpha emitting nuclides. 4 "ToW includes removable and fixed coniaminaiion. 10' ar 223Ra 224Ra b Pure bela-cmiuing radionuclides wilh maximum bcla energies less lhan 0.15 MeV musi be handled on a casc-by-casc basis because ate not delectable by conventionaj monitoring instruments. 10u l47Sm l53Sm 2l"Pb 2'8 Po 22,'Ra 227Th 228 T3i 234U 235U 236y 238 y 10 1 iJUTh 2J2Th 23'Pa 2J2U 238 Pu 239 Pu "'Am 244Cm plus all other alpha emitters with a half life greater than three months. • Applies to all exposed surfaces in designated areas, including protective clothing and skin of workers. • Multiply by 10, for interiors of glove boxes and fume cupboards. • Divide by 20, for areas which are not designated and for personal clothing. JAERl-Conf 97 - 008 EXAMPLES OF REFERENCE LEVELS Australian Code of Practice SURFACE CONTAMINATION LIMITS NUCLIDE LIMIT NUCLIDE LIMIT NUCLIDE LIMIT Bq/cm2 Bq/cm2 Bq/cm 3H 103 ”Mo 103 ’*'Hf 103 ,4C 102 ”mTc 103 ,45W 103 22Na 102 ”Tc 103 '“Re 103 24Na 103 ,03Ru 103 ,«ir 103 32P 102 '“Ru 102 ,S4Au 103 31S to 3 "0mAg 102 "'Hg 103 MCI 103 ”’Ag 103 M3Hg to 3 45Ca 103 ,MCd 107 ”'TI to 3 4,Ca 103 nlmCd 107 204TI 103 "‘Sc 103 "'In 103 2,0Pb 10° "Sc 103 "3mln 102 2,0Bi 102 3,Cr 103 "3Sn 103 2,0Po 10° 14Mn 103 '2'Sb 102 223Ra 10' “Fe 103 ,25Sb 103 224 R a 10' 19 Fe 103 ,23l 103 226Ra 10° “Co 102 ,25l 102 227Th 10° U o 103 ,3,l 107 224Th 10° s o o 103 '"Cs 103 23°Th 10" 8 n o 102 ,2'Cs to 3 232Th 10" “Ni 103 ,34Cs 102 231Pa 10" ‘“Cu 102 ,37Cs 102 232u 10" 67Cu 102 ,33Ba 103 234u 10° “Zn 102 "°La 102 231u 10° 67Ga 103 ,3,Ce 103 23SU 10° “Ga 102 u'Ce 103 2«u 10° “Ge 103 ,47Nd 103 234Pu 10" "Se 102 ,47Pm 102 23aPu 10" 77Br 103 "7Sm 10° 24'Am 10" “Rb 103 ,S3Sm 10° 2< 127 EXAMPLES OF REFERENCE LEVELS EXAMPLES OF REFERENCE LEVELS Australian Code of Practice Regulations for the Safe Transport of SURFACE CONTAMINATION LIMITS Radioactive Material IAEA Safety Series No 6 ASSUMPTIONS 1985 Edition (as amended 1990) Surface contamination limits based on the committed effective dose limit of 20 mSv per year. LIMITS OF NON-FIXED CONTAMINATION ON SURFACES Dosimetric models used for the determination of the equivalent dose per unit intake of radioactive material JAERi-Conf for inhalation and ingestion are based on the models Applicable limit Applicable limit prescribed in ICRP 30 and modified in ICRP60 and of beta and of all other ICRJP61. gamma emitters alpha emitters Type of Surface and low toxicity For the calculation of the contamination limit for work alpha emitters 97-008 surfaces in designated areas via the inhalation pathway, pCi/cm 2 Bq/cm 2 pCi/cin 2 Bq/cm 2 a resuspension factor of 5x10 s m"1 is assumed and a External surfaces of 10 s 0.4 10* 0.04 worker breathing rate of 1.2 m3 per hour and 2000 excepted packages working hours per year. External surfaces of 10" 4 10 s 0.4 other than excepted For the ingestion pathway, it is assumed that a worker packages may ingest the entire contents of 10 cm2 of a contaminated surface each working day of the year and The level of non-fixed radioactive contamination may be that there are 250 working days in a year. determined by wiping an area of 300 cm2 of the surface concerned by hand with a dry filter paper, a wad of dry For external irradiation of the skin, equivalent dose cotton wool or any other material of this nature. The rates were taken from NRPB DL-2 and it was assumed above limits are applicable when averaged over any area continuous skin contact for 200h per year and of 300 cm2 of any part of the surface. contamination on the skin for 876h per year. SURFACE CONTAMINATION DESIGN OF A MONITORING PROGRAM MONITORING PROGRAM When there is likelihood of radioactive contamination of OBJECTIVES the workplace, a surface contamination monitoring program should be established. . TO ENSURE SATISFACTORY WORKING CONDITIONS ARE MAINTAINED; This program should include; • TO PROVIDE INFORMATION FOR ESTIMATING • routine monitoring to control normal operations and EXPOSURE (using models to interpret results. Beware confirm the adequacy of routine control procedures, to of the limitations of the model); ensure no changes have taken place and to ensure a satisfactory working environment. • TO ASSIST IN PREVENTING THE SPREAD OF JAERI-Conf CONTAMINATION; • task related (operational) monitoring focused on specific non-routine tasks aimed at ensuring local control • TO DETECT FAILURES OF CONTAINMENT OR against the spread of contamination, and to assist in DEPARTURES FROM GOOD OPERATING immediate decision making in the conduct of an 97 PROCEDURES; AND• operation. This applies to short term procedures and is not acceptable for continued long term use. • TO PROVIDE INFORMATION TO ASSIST IN THE PLANNING OF PROGRAMS FOR INDIVIDUAL • special monitoring in support of commissioning INTERNAL DOSIMETRY ASSESSMENT AND FOR activities, to develop contamination reference levels or DEFINING OPERATIONAL PROCEDURES AND to cover operations carried out under abnormal FACILITIES. circumstances e.g., accidents/incidents. Note: CONTAMINATION CONTROLLED ZONES • Experience has shown that there is not necessarily a Where there is a likelihood of significant contamination direct correlation between surface contamination in the controlled areas should be set up. This may require a workplace and the exposure of workers. graded approach to control with areas more susceptible to • The absence of surface contamination above defined contamination being specifically identified and delineated. reference levels usually indicates high standards of primary containment and management control. ROUTINE MONITORING RECORDING OF SURFACE MONITORING DATA MONITOR A REPRESENTATIVE FRACTION OF THE SURFACES IN AN AREA AT A FREQUENCY DICTATED BY EXPERIENCE. Records of monitoring arc required: THE MOST FREQUENT MONITORING SHOULD BE AT KEY INDICATOR POINTS SUCH AS CONTROL • to provide a check that monitoring is being done BARRIERS AND AREAS OF HIGH CONTAMINATION correctly and at the required frequency; POTENTIAL. • to enable the responsible organization to maintain a close MONITORING WITH A LOWER FREQUENCY surveillance of hazardous operations so that the SHOULD REPRESENTATIVELY COVER THE WHOLE effectiveness of control measures can be assessed. JAERI-Conf OF THE DESIGNATED CONTAMINATION AREA TOGETHER WITH SOME ADDITIONAL • to provide a basis from which a radiation protection CONFIRMATORY MEASUREMENTS OUTSIDE THE officer may make recommendations for altering control DESIGNATED AREA. procedures. 97-008 SPECIAL ATTENTION SHOULD BE GIVEN TO CLEANING MOPS, VACUUM CLEANER BAGS, • to provide in the event of llitigation,records of conditions SPECIAL MATS AND OTHER SURFACES AT THE during operations may be related to reported or alleged EXITS FROM AREAS, SHOES, GLOVES AND exposures to radiological hazard. LABCOATS WORN IN THE AREA. IF THERE IS A HIGH POTENTIAL FOR CONTAMINATION TO OCCUR, ROUTINE MONITORING SHOULD BE SUPPLEMENTED BY THE USE OF INSTALLED EXIT MONITORS SUCH AS FRISKING PROBES, HAND AND FOOT OR PORTAL MONITORS. DISTINGUISH BETWEEN FIXED AND LOOSE SURFACE CONTAMINATION. HEALTH PHYSICS SURVEY REPORT - BUILDING 23: MAIN "A" BLOCK RADIOCHEMICAL To:______TYPE: CONTAMINATION (CPS/DUL)* / RADIATION P / y / Py*(jxSv/h) DATE: / / TIME:______*delete whichever is not applicable INSTRUMENT(S): RECOMMENDATIONS/REMARKS Installed t-monitors (AtSv/li) LABORATORY (!) r I# lb) SURVEY INSTALLED AIR SAMPLERS (CPS) FORM HEALTH PHYSICS SURVEYOR: DATE: / AREA HEALTH PHYSICIST DATE: / JAERI —Conf 97 — 008 EXAMPLES OF RECORDS CONTAMINATION CLEARANCE CERTIFICATE RADIOACTIVE CONTAMINATION CLEARANCE CERTIFICATE A 29437 1. DESCRIPTION OF ITEM ______ ______Plant/Serial No. ______ Place of origin: Building ______. _ LRoom ______Classification ------ Destination: Building ______Room Classification ------—------— Signed ______(Originator) Date 2. HEALTH PHYSICS REPORT Surface dose rale...... #*Sv h" 1 Exterior fixed contamination ...... DWL Exterior removable conlominalion ...... DWL Cleared (or movement to: White/Blue/Red area' Interior may be contaminated Yes/No* Signed (Health physics surveyor) Date kMfll No. 7S30-7QS '•wtvettevet «ol *oo<<.»bic - 132- JAERI —Conf 97 —008 EXAMPLES OF RECORDS OCCUPATIONAL HEALTH AND SAFETY FORM HP,'PC/. HEALTH PHYSICS PERSONAL CONTAMINATION REPORT SECTION A: Details of contamination incident Name (s ) of contaminated person (s) :______. Programme area Date and time (approx.) of incident:______ Date and time reported:______ Type of contamination and radionuclide (if known):. Was HP Incident Report completed: Yes/No* Number of report Was Accident/Potential Hazard Report completed: Yes/No* SECTION B: Levels of contamination contam. level Beta Gamma Alpha cps DWL cps DWL cps DWL Areas of the body (as shown overleaf): Action taken to remove contamination Decontamination techniques * (i) used: ______ on ______(part of body) . Are all areas clear :Y/N‘ If not describe levels/locations remaining :______ Did contam. incident involve contamination of (i) wounds/cuts/abrasions Y/N * (ii) personal clothing or effects Y/N * If yes to i or ii then detail actions taken: HPS Name (print)______Signature______Date. Area Health Physicist comments: Signature______Date--- Distribution: (l)Head Radn. Prot. Services (2) Contaminated Person(s) (3) Area Health Physicist (4) H.P.file (5) Site Medical Officer (6) Dosimetry Lab * Delete whichever is not applicable 133 JAERI-Conf 97-008 EXAMPLES OF RECORDS Reverse side of Personal Contamination Report - 134- JAERI-Conf 97-008 a a has beta is A in for thin It and of with probe inch a surface low in plastic probe tube sulphide zinc thin If type tube dual the two to tight. a thin The used to GM zinc or The beta-gamma a a present, monitoring be monitoring. light window geiger of form very he circuit close it Another pulses. a with tube. probe extremely a to to may beta as thin is monitoring. employ adequate. beta means make GM surface used. with work is and for very to is screen These by probe and Melincx, probe glajs-wallcd leaks) used usually window direct most suspected the with alpha a tube is is window with mg/cm opened GM out alpha For tight (light photomultiplier sulphide detected keep 30 be damage. end tube a the is aluminium to to phosphors. counter monitors to zinc light walled thin probe. covered containing a DETECTION with carrying about should is flow the or simultaneous liable of detectors detectors plastic inch glass between Geigcr-Muller with As contamination DETECTORS______ 2 coupled probe probe when very coated which a two uses DETECTION holder screen beta permits contamination photomultiplier or contamination mg/cm contamination thickness possible. pancake 30 scintillation proportional scintillation one shutter • • • • • beta/gamma COMMON alpha which detector differentiate wall energy end-window conjunction single therefore emitters, a material, important alpha BETA/GAMMA Beta as suitable ALPHA sulphide Alpha scintillator be the or The to a to bench of watched. as establish surface level various suspected to be either the the to of listening out method to which surfaces by contamination over loudspeaker). to have or such consists carried not contamination is attention connected. on convenient surface detected detector does the CONTAMINATION be ratemeter pay for and the monitor to most can (earphones easily allow meter . limits 2 and skin. head easier the operated moving more MONITORING useful if contamination is monitoring is and also Bq/cm of SURFACE derived is in simplest slowly HINTS level contamination battery as it detecting signal by in the measurements or of clothing is and monitored presence typical aural mains types A reported calculated Direct the tops, This contamination DIRECT clicks being increase HELPFUL Monitor surfaces. An • • • 135 DIRECT METHODS DIRECT METHODS Alpha contamination monitoring Beta contamination monitoring Alpha monitoring raises special problems because of the • The detection of all but the lowest-encrgy beta radiation is short range of alpha particles in air and their complete easier than that of alpha contamination because of its longer attenuation by thin films of liquids or solids; 5-MeV alpha range and greater penetration. particles are completely attenuated in material with a thickness of about 5 mg/cm 2. • The response of a beta probe is dependent on the energy of the beta radiation. Probes should therefore be calibrated so The transient response of a probe passing over a source is that the count-rate from contamination by the relevant considerably less than the response with the probe radionuclide at its derived working level is known. stationary above the same source. The time constant of JAERI-Conf the ratemeter is often longer than the transit time of the • During routine monitoring the area under examination is probe past the source, therefore the meter may not usually ‘swept’ by the probe. The response of the probe indicate the presence of the contamination. Although a varies with the distance between the surface and the probe and with variable transit velocities. The probe should be loudspeaker has a quicker response, the ear may not used at a distance between 2 to 5 cm from the surface and 97-008 detect small variations in the loud-speaker sound pulse- the transit velocity should not exceed about 15 cm/s. rate. • When high levels of contamination are being monitored the To ensure detection of alpha contamination by direct dead time of the detector will affect the count-rate and may monitoring methods, the probe should not be more than lead to underestimation of the activity present. about 0.5 cm from the surface under examination and the probe transit velocity across the surface should not exceed j • Radiation from very high levels of contamination may about 15 cm/s. ‘saturate ’ a Geiger tube, resulting in an erroneous nil reading being obtained on the ratemeter. Most modern Although a variety of detectors is available for alpha instruments are designed to maintain full-scale deflection on monitoring, the technique for using them is always similar. the ratemeter. • Some probes arc fitted with a shield, which when closed prevents beta radiation from reaching the detector, enabling a gamma-only measurement to be made. DIRECT METHODS INDIRECT METHODS OF SURFACE MONITORING Low energy beta contamination monitoring Indirect methods of surface contamination monitoring should be used when direct monitoring is inappropriate. • Because of the short range of low-energy beta radiations Examples of such situations are when; special problems similar to those encountered in alpha monitoring are raised. • interference from other radiations prevent direct monitoring or a high radiation background is present; • Geiger tubes or scintillators with thin windows (about 2 mg/cm 2) or gas-flow or air-filled proportional counters • direct monitoring has shown the presence of radioactive are most suitable. contamination; JAERI-Conf • Indirect methods are usually employed to monitor for • the measurement geometry is unsuitable; tritium. • monitoring low energy beta surface contamination; • the surface to be monitored is inaccessible to the 97-008 contamination monitor probe; • direct monitoring underestimates the degree of contamination because of self absorption effects; • the degree of removable contamination is to he estimated. A wide range of techniques is available, but all depend upon the removal of radioactive material from the surface being examined. It may be necessary to take samples from the area being monitored to another area for counting. INDIRECT METHODS OF SURFACE MONITORING INDIRECT METHODS OF SURFACE MONITORING • The fraction of material removed during routine indirect • Most surfaces in radioactive laboratories and process monitoring is not known accurately. A smear test may areas can be smear-tested satisfactorily. Outside these remove anything from 0 to 100% of the contamination area where rough brickwork, tarmacadam roadways, present, but generally a removal factor of about 10 to concrete paths and hardstandings, areas of gravel, soil or 20% is assumed. grassland may be encountered, the ability to obtain a satisfactory smear sample varies from the difficult to the • The variations in the removal factor make quantitative almost impossible. estimation of the amount of contamination present on the surface uncertain. • The validity of indirect methods in which the amount of activity transferred to rinse solutions or wash-down JAERI-Conf • The removal factor is affected by varying the pressure liquors is estimated depends upon the solubility of the applied during smearing, by the type of surface or contaminant in the liquid in question, or on the smear, and by the form of the contamination. homogeneity of particles dispersed in the liquor. • Devices to ensure uniformity in pressures applied and • An article declared free from radioactive contamination 97-008 area usually include some form of spring loading of a by indirect monitoring, may still have fixed pad to which the smear paper is attached. contamination. Care is needed as this may become dislodged if the surface is more vigorously attacked • When a smear paper is used manually, most of the during work upon it, e.g. by grinding, filing or polishing. contamination picked up on the paper is concentrated in an area of about 5 cm2 coincident with the area of the finger tips. • • Because this area may be small compared to the area of the detector used to measure the smear activity, erroneous results may be obtained unless the probe is specially calibrated. INDIRECT METHODS OF SURFACE MONITORING INDIRECT METHODS OF SURFACE MONITORING WET SMEARS DRY SMEARS • As above, but moisten the smear paper with a suitable • In such a test the surface is wiped with absorbent liquid. Alternatively a moist cloth may be used. For material such as a filter paper or paper tissue which is alpha activity and low energy beta activity the wet smear then presented to the appropriate counter for must be dried before measuring. measurement in a region of reasonably low background. LARGE AREA SWABS Normally about 10% of removable contamination is • Floor mops or similar cleaning equipment can be used assumed to be transferred to the paper. Contamination here. levels may be averaged over 1000 cm2 for floors, ceilings ADHESIVE TAPE and walls, or 300 cm2 for other surfaces. For parts of • Adhesive tape pressed against a surface will collect a the body, levels may be averaged over an area not sample of removable contamination. JAERI-Conf exceeding 100 cm2. VACUUM CLEANERS • The contents of active area vacuum cleaners can give an indication levels of surface contamination. EXAMPLE OF ASSUMPTIONS FOR A SMEAR TEST OVERSHOES • Checking for contamination on used overshoes can give a 97-008 Assumptions quick indication of the presence of floor contamination • surface contamination is uniform and removable in a laboratory. • 10% removable contamination is transferred to paper LABORATORY COATS • 20% of surface is smeared • The presence of contamination on lab coats can indicate • averaged over 1000 cm2 for floors, ceilings, walls possible spillage requiring further investigation. • averaged over 300 cm2 for other surfaces EXAMINATION OF RINSE/WASHING SOLUTIONS • This is another check that can be useful to confirm If A = activity on smear successful decontamination of an item of equipment. TRITIUM SMEARS surface activity -- A x 10 x 5 • A filter paper, moistened with glycerol, can be used as a surface area smeared wipe, which is then checked in a gas flow proportional counter or a liquid scintillation counter. = A x 10 x 5 Bq/cm 2 1000 MONITORING FOR SKIN CONTAMINATION SKIN CONTAMINATION IS NEVER UNIFORM AND OCCURS PREFERENTIALLY ON CERTAIN PARTS OF THE BODY, NOTABLY THE HANDS. FOR ROUTINE PURPOSES REGARD THE CONTAMINATION AS BEING AVERAGED OVER AN AREA OF ABOUT 100cm2. JAERI-Conf DOSE ESTIMATES FOR HAND CONTAMINATION ARE USUALLY IMPRECISE (UNCERTAINTIES OF TWO ORDERS OF MAGNITUDE ARE POSSIBLE). 97 MEASUEMENTS ARE GENERALLY QUALITATIVE IF DOSE ESTIMATE > 0.1 DOSE LIMIT IT SHOULD BE INCLUDED IN THE INDIVIDUAL’S PERSONAL RECORD. SOME CONTAMINATION MAY BE TRANSFERRED INTO THE BODY, REQUIRING INTERNAL DOSIMETRY ASSESSMENT. Commercial instruments are available for various 2.4Commercial Contamination Monitoring applications Instrumentation • General surface and airborne contamination • Frisking • Floor monitoring • Portal (whole body and extremity) monitoring e Laundry monitoring IAEA/RCA Contamination Monitoring Training Workshop • Waste monitoring 21-25 October, 1996 • Road monitoring Tokai-Mura Japan • Vehicle and large object monitoring JAERI-Conf Richard Griffith s Training 97-008 Survey instruments have 5 basic components instruments can have a variety of readout styles Visual Amplifier Processor Readout Readout Alarm DETECTOR r— ► 50 i ili 77 cpm 0 50 100 Analogue Digital METER Audio Power Supply Audible clicks to indicate count rate Audible alarm Commercial Instruments use various configuration Commercially available detectors are based on systems 1 of 3 detection principles Single Instrument with selection of meters Motor Detector | (Readout) -* —{ Detector A e Gas filled counters Detector B ■ Ionization Single meter, sele ction of det e ctor? Detector C ■ Proportional " Geiger-Mueiier Detector A ] Meter A (Readout) Meter Detector B | Meter B « Scintillation (Readout) (Readout) Detector C | e Solid state - semiconductors JAERI-Conf 4^ eo 97-008 Principal components of an ion chamber Positive Negative electrode electrode GAS MULTIPLICATION Power Readout supply Gas amplification occurs In proportional and G-M counters Gas amplification increases with voltage 1,000,000 10,000 Alpha , Proportional JAERI-Conf Detector voltage 97-008 A variety of fill gases are used in commercial detectors G-M detectors are good for general monitoring Ionization • Electronic requirements are simple and not demanding - stable performance Air Proportional o Detectors are readily available and easily Air replaceable Butane P10 (90% Argon, 10% methane) e Particle selection or discrimination through window thickness - low energy Xenon betas and alphas require windows < 1 Xenon + Carbon dioxide mg/cm2 GM Halogen quenced o Gamma sensitive (background) G-M defectors have different configurations Proportional detectors offer particle discrimination Side window End window Pulse Discriminator height Detector voltage JAERI-Conf A. ■to. 97-008 Typical beta contamination monitor efficiency curve SCINTILLATION " Scintillation detectors Choice of scintillators depends on the application f------Output PulSC Alpha Zinc sulphide High Beta voltage Plastic - NE102, NE110 Photoeleclfon Gamma Sodium iodide Focuilng electrode Ionizing event 5 Crystal Caesium iodide JAERI-Conf Incident photon Bismuth germanate cn 97-008 Air sampling techniques depend radioactive material form Air sampling techniques depend radioactive material form Particulate Particulate Continuous air Filter Particulates are collected on a thin filter (Millipore, etc.) monitor (CAM) Filters may be recovered for laboratory analysis, or counted in place using appropriate detectors Detector Detector Gases Air flow Iodine - Collected on charcoal cartridges, or impregnated filter paper iodine Tritium Tritium Gas (HT) - Collected and counted in an ion chamber Ion HTO - Collected in a media (or subsequent laboratory chamber analysis Air flow Charcoal Detector JAERI-Conf 97-008 contamination Mev - alpha Energy monitoring Alpha for critical Is Distance alphas and monoenergetic contamination betas beta Alphas mev for - Energy monitoring different for Beta is Energy Betas important continuum - Is distribution 1500- 1000 2000 Distance Energy 146 JAERI—Conf 97 — 008 2.5 Basic Techniques of Decontamination PNC Naoki YOKOSAWA. Before going to decontamination, let’s look also at key items of procedures of radiation control. 1 Early detection of contamination In view of radiation protection, safe operations depend on monitoring of contamination. Early detection of contamination prevents diffusion of contamination and inhalation of RI. 2 Reporting contamination to neighbors After identification of contamination or possible contamination, it is reported to neighbors. The most important matter is personal health. Some cases may need instant evacuation from the site. The contaminated person should not walk around after move to another place. The neighbors help to put a half face mask on, to arrange every thing necessary for the decontamination, to communicate radiation control officers and so on. Contamination of cloth and shoes are surveyed. The contaminated part of cloth is covered by a wide tape to prevent inhalation of RI. 3 Body decontamination 3.1 Skin decontamination If the part of contamination is injured, some blood should be squeezed out so as not to intake RI in blood and see doctors. Following cleaning methods are taken at PNC in a case of no injury. 1) brushing with soft brush under running water 2) neutral soap 3) Ti02 paste 3.2 Inhalation 1) nostril cleaning and throat cleaning 2) vomiting 4 Floor decontamination In case of airborne contamination of a room, every body evacuates from the room. The extent of surface contamination and airborne contamination are monitored. The data are reflected to a planning of decontamination activities and selection of protective mask and kits so on. Normally ceiling and wall of a room are not much contaminated. Therefore a procedure for decontamination of a floor is discussed here. There are many measures for a decontamination. This is a one of possible measures. In most cases decontamination of a floor can be done by using half face masks. It is better to clean first an upper side of wind flow in a room. A moist paper towel with water is normally used for the - 147- JAERI-Conf 97-008 decontamination. The floor is wiped from a relatively cleaner corner to a dirty center. On the half way to the center the paper towel is folded in half to go to the center. When he goes back to, he take the route he cleaned, and on the half way he takes away outer layer of shoes covers. If the decontamination is not enough, some detergent is used. A RI buried into wax of floor polisher can be easily removed by acetone. A RI some times bonds to a surface too strong to be removed. In that case the surface or RI is removed by a chemical agent, electronic polishing or mechanical polishing. If some contamination is left over, the surface can be protected by painting , and the surface is periodically checked by smear check. A paint for decontamination is also helpful for reducing airborne concentration. One of the product name is strippable paint. After the painting the surface is scrubbed by a bush which gives a high decontamination factor. The paint is easily removed after two days drying and it is inflammable. - 148- 2.6 IAEA IAEA REORGANIZATION -1 JANUARY, 1996 Radiation Safety Programme DEPT. DEPT. NUCLEAR NUCLEAR SAFETY POWER z DEPT. > NUCLEAR Richard Griffith ENERGY AND Radiation Safety Section < SAFETY y JAERI-Conf 97-008 DEPARTMENT OF NUCLEAR SAFETY Deputy Director General Mr. Zygmund DOMARATZKI RADIATION SAFETY RADIATION SAFETY SECTION SERVICES SECTION Section Head Section Head SUUX CO-OflOiHlIOM i SECTION I Mr. Geoffrey WEBB Mr. Robert OUVRARD ActUifl S*cllefl Held j Mr rUwud OJTSASSlOON DIVISION OF NUCLEAR INSTALLATION DIVISION OF RADIATION AND Occupational and Laboratory SAFETY WASTE SAFETY Public Unit Services Director Director Richard GRIFFITH Robert OUVRARD Ms. Annlk CARNINO Mr. Abel Julio GONZALEZ j (OPERATIONAL SAFETY r RADIATION SAFETY ] r Sources and Medical Emergency UiCiNCtniMQ SAFETY RADIATION SAFETY ’ SECTION j SECTION SERVICES SECTION Unit Response Unit j i t*ctlen Hied Section Held I Sicilon Hoad Sicilon H*id j Ur CioHnyWCDB j Pedro QRTIZjLOPEZ Ian THOMPSON M, t eOhlJ KABANOV | i Ur. Kefih HOE Mr. MoVert OUVI1AHD SAftlT ASSESSMENT WASTE SAFETY SECTION Transport Unit Sldtoft Hlld Acting Section Heed Ur Rkk HlEMAUS Ur Cordon UMStEt Rick RAWL Efficffv* 4 Auguil IPS6 RADIATION SAFETY PROGRAMME MAIN ACTIVITIES • SAFETY STANDARDS • Regular Budget • The development and production of a • Safety Series Documents (60) comprehensive set of International ■ Co-ordinated Research Programmes (15-20) Consensus Radiation Safety Standards Conferences/Seminars, etc. (1-2) • ASSISTANCE • Technical Co-operation » Direct assistance to Member States to bring ■ National Projects (250) their radiation safety infrastructure to an » Regional/Interregional Projects (20-30) appropriate level for their usage of radiation, and to enable development ■ Country Profiles/Action Plans (95) ■ Model Projects (53) JAERI-Conf cn O 97-008 1996 Programme project areas H1. STRENGTHENING OF RADIATION SAFETY H1. STRENGTHENING OF RADIATION SAFETY • H.1.01 Harmonization of Radiation Safety H2. OCCUPATIONAL RADIATION PROTECTION Standards and Practices H3. RADIATION PROTECTION OF THE PUBLIC AND • H.1.02 Strengthening Radiation and Nuclear THE ENVIRONMENT Safety Infrastructure in Member States, Including Education and Training H4. SAFE TRANSPORT OF RADIOACTIVE MATERIAL • H.1.03 Joint UNDP/IAEA Assistance Programme H5. EMERGENCY PREPAREDNESS to Strengthen Radiation and Nuclear Safety Infrastructures in the Former USSR (New H6. SAFETY OF RADIATION SOURCES Project) H7. RADIATION SAFETY SERVICES H2. OCCUPATIONAL RADIATION PROTECTION H3. RADIATION PROTECTION OF THE PUBLIC AND THE ENVIRONMENT______ • H.2.01 Optimization of Occupational Radiation • H.3.01 Optimization of Radiation Protection of the Protection Public • H.2.02 Monitoring of Individual Workers and Controlled Areas • H.3.02 Radiation Modeling and Monitoring of the Environment • H.2.03 Guidance on Assessment and Treatment of Radiation Effects • H.3.03 Follow-up of the Radiological • H.2.04 Providing Guidance on Radiation Consequences of the Chernobyl Accident Protection in Mining • H.3.04 Assessment and Control of Radiation Exposure JAERI-Conf 97-008 H4. SAFE TRANSPORT OF RADIOACTIVE H6. SAFETY OF RADIATION SOURCES MATERIAL • H.4.01 Maintenance and Implementation of • H.6.01 Design, Control and Safe Use of the IAEA Transport Regulations Radiation Sources H7. RADIATION SAFETY SERVICES H5. EMERGENCY PREPAREDNESS • H.7.01 Radiation Protection Advisory Teams (RAPAT) Services • H.5.01 Radiological Emergency Strategies • H.7.02 Laboratory Services, Additional high • H.5.02 Emergency Assistance Services priority activities JAERl —Conf 97 — 008 Safety of Protection Material Implication Exposure 50,000 120,000 140,000 540,000 Radiation 1,700,000 3,380,000 1.0 $ $ $ 6.2 $ $ $ Radioactive for Public Society Medical of Protection for the Perspective: Basis of from in Safety Protection Budget: staff: Budget: staff: Transport Cooperation: Cooperation: Standards Safe Regular Protection Protection Radiation Occupational Conceptual Regular Transport Radiation Extrabudgetary Professional 1997 1.3.01 Technical Extrabudgetary Professional 1997 1.1.03 1.1.03 1.1.04 1.1.05 Technical 1.1.02 1.1.02 1.1.01 3 1. 1.1 Sources Accidents Interventions areas and 10,000 Source ______Radiation 730,000 Services y of 2,380,000 $ 3.9 $ $ Safety project Safet Radiation Security of Safety Protection Protection Source Emergencies Source Budget: and staff: Cooperation: Prevention Control Regular Programme Radiation Radiation Radiation Radiation Transport Radiation 2.02 5 3 4 2 1997 Extrabudgetary Professional Technical 1.2.01 1. __ 1. 1. 1. 1. u .2 1 1996 152 I.4 Radiation Emergenci es and Interventions 1.5 Radiation Protection Services 1.4.01 Preparation for Radiation Emergencies 1.5.01 Laboratory Services 1.4.02 Assessment of Radiological Situations Professional staff: 1.3 1997 Regular Budget: $ 550,000 Professional staff: 3.0 Extrabudgetary: $ 940,000 1997 Regular Budget: $ 190,000 Technical Cooperation: $ 1,300,000 JAERI — Conf CJ1 co 97-008 OTHER TOPICS PUBLICATIONS • Comprehensive Test Ban Treaty - verification • Mururoa Tests Radiological Assessment - current - long term SAFETY SERIES DOCUMENT STRUCTURE SAFETY SERIES DOCUMENT STRUCTURE RADIATION SAFETY • FUNDAMENTALS (Silver) Basic objectives, concepts Radiation Safety Fundamentals • STANDARDS (Red) Basic requirements and obligations International Basic Transport Safety Standards Regulations • GUIDES (Green) SS 115 SS Recommendations and explanation on how to fulfill requirements in the Standards Application Areas - Safety Guides and Safety Practices • PRACTICES (Blue) Specific examples of areas of application or Occupational Medical Chronic Transport techniques to be used in implementing Standards or Guides Public Emergency General JAERI-Conf 97-008 PUBLICATIONS PUBLICATIONS • Safety Fundamentals • Safety Guides Primary category in the hierarchy-lead document Guidance as recommendations to fulfill basic present basic safety principles, concepts & requirements set out in SF/SS. Less formal in style - objectives for radioactive waste management. "should" statements - shall statements for Require approval by the Board of Governors to mandatory requirements. Explanatory ensure full international consensus • Safety Practices • Safety Standards Practical examples a' .. 'ail methods for the Specify basic requirements to be fulfilled to ensure application of SS anu SG ampliation of data - how safety-governed by Principles In SF - termed safety to perform certain calcr .uons - illustrations - codes - require to be self standing - mandatory criteria to be used for fulfilling specific requirements requirements are styled as "shall" statements and - examples in annexes complement main text require approval by Board of Governors SAFETY FUNDAMENTALS International BASIC SAFETY STANDARDS • The Safety of Nuclear Installations - Safety Series 110 - Published 1993 • The Principles of Radioactive Waste Management - Safety Series II3-F - Published Sept 1995 • Radiation Protection and the Safety of Radiation Sources - Safety Series - Being considered for co-sponsorship by FAO, PROTECTION AGAINST IONIZING RADIATION ILO, OECD/NEA, PAHO and WHO AND FOR THE SAFETY OF RADIATION SOURCES JAERI-Conf cn cn 97-008 LAYOUT SAFETY STANDARDS PREPARATION PROCESS • PREAMBLE: — Principles and Fundamental Objectives ACSS ADVISORY COMMISSION ON # PRINCIPAL REQUIREMENTS: SAFETY STANDARDS — General - responsibilities and organizational — Practices — Interventions # APPENDICES: DETAILED REQUIREMENTS I ! — Occupational exposure NUSSAC RASSAC WASSAC |i TRANSSAC WASTE SAFETY : I TRANSPORT SAFETY — Medical exposure NUCLEAR SAFETY RADIATION SAFETY ' standards advisory STANDARDS ADVISORY STANDARDS ADVISORY l| STANDARDS ADVISORY — Public exposure COMMITTEE COMMITTEE COMMITTEE j COMMITTEE — Potential exposure: safety of sources — Emergency exposure situations — Chronic exposure situations Expert Expert Expert Expert Groups Groups Groups Groups • ANNEXES — Numerical JAERI-Conf 97-008 (in or the (in and levels in areas of design excluded to experience monitoring optimization protection incurred RADIATION exposure RADIATION planning on thinking" unless of and feedback operating workers optimized "way sources for work, individual classification an to a of constraint design to to a as occupational data in their as of of OCCUPATIONAL OCCUPATIONAL of reference analyzing exempted used ON ON Constraints exposures by by be relationship all course relationship from collection reality) operation set theory) Definition Optimization -prospective -operationally - Dose -set - - -to - - • • GUIDE PROTECTION • PROTECTION GUIDE with years; 5 but and year averaging year in Manual a in year RADIATION Mining mSv Health: single 5 available SS(new) SS(new) the SS(new) (revised) is mSv 100 in any 20 SS83 authorities? SS83 of BSS, mSv Exposure, Protection, Exposure, Workers Occupational in Ores, 50 of In limit application national Physicians, External Internal than to Radiation OCCUPATIONAL circumstances"; of of facie" Protection Protection Radioactive specified ON limitation of more practical Occupational helpful? recourse needed? "prima not limits Radiation Milling Dose for -for - -"special - Occupational - Guides -Assessment -Radiation -Assessment - • PROTECTION GUIDE • OCCUPATIONAL 156 GUIDE ON OCCUPATIONAL RADIATION PUBLIC PROTECTION______• Control of Exposure • Guides - classification of areas, firmer criteria being considered - Environmental and Effluent Monitoring, • Monitoring of Workers SS(new) - practical criteria - Regulatory Requirements for the Use of - Health surveillance Consumer Products Containing - comparable to other occupational health Radioactive Materials, SS(new) programmes • Management of overexposed workers JAERI-Conf cn 97-008 MEDICAL______EMERGENCY • Guides • Guides - Radiation Protection in the Medical Exposure of Patients, SSfnew] - Intervention Criteria in a Nuclear or Radiation Emergancy, SS109 - Medical Handling of Accidentally Exposed Individuals SS88 (revised) - Planning tor Response to Nuclear or Radiation Emergencies, SS(new) CHRONIC GENERAL TOPICS • Guides • Guides - Extension of the Principles of Radiation Protection to Sources of Potential Exposure, SS104 - Safety of Radiation Sources, SS (new) - Prevention of Illicit Trafficking of Radioactive Sources, - Application of the Principles of Radiation SS(new) Protection to Chronic Exposure - Principles for the Exemption of Radiation Sources and Situations, SS(new) Practices from Regulatory Control SS89, (revised) -Operational Radiaon Protection: A Guide to Optimization, SS101 -Training Requirements and Programmes for Post Graduate Education, SS(new) JAERI-Conf 97-008 TRANSPORT CONCLUSION • Guides • Objective is a comprehensive, internally - Advisory and Explanatory Material for the consistent and up-to-date set of Safety IAEA Regulations for the Safe Transport of Series documents (in a reasonable time) Radioactive Material, SS37, SS7 (revised) - Emergency Response Planning and • Occupational exposure Guides and Preparedness for Transport Accidents Practices are a major subset of this Involving Radioactive Material, SS87 series (revised) Other_Publicatjons ______ • Technical Report Series - Calibration of Instruments and Dosimeters for Radiation Protection Purposes (Revision of TRS 133) - Compendium of Neutron Neutron Spectra and Detector TRAINING Responses for radiation Protection Purposes: Supplement to TRS 318 •TECDOCS - Reference Asian Man (Phase I) • Practical Radiation Safety Modules • Practical Radiation Technical Modules JAERI-Conf cn to 97-008 INTERREGIONAL TRAINING COURSES RADIATION PROTECTION ism# TochhlqLes'foriExlSrl Uoi'o'AagTesmant™ JAERI-Conf 97-008 a data of the of in all radiation maintain the to up-to-date infrastructure on COOPERATION and system safety Agency information and the location to copy Profile TECHNICAL information Hard Database Questionnaire Includes: known protection country An specified - - - • • Country to Member Member 8 needed technical Profiles a MANAGEMENT infrastructures actions from developing COURSES Country WASTE of ASIA develop AND officer i-cis a o TRAINING n group WEST MM# implement aa a maintain nJM further Ul to n and 8 Agency and PROTECTION infrastructures REGIONAL progress and sPan an h Strategies f.i % n o ll c • RADIATION ot Sustain Monitor improve Formulate Establish department department States gemafel j Assign P i • • » • n • Agency a n^i5l35yniMsWm# MWinrtdswtRtoSBlHgSiBBI M •rid TjchnlquewwEwwnSTOfflmsmalDaM® 160 The Model Project Steps in the Model Project • Decide what are necessary components of the • To upgrade the radiation and waste safety infrastructure for countries with low, intermediate and infrastructure in selected countries to an high useage of radiation (checklist) adequate level for their use of radiation. • Assess what each country has of infrastructure components • Originally 5 countries. • Transfer to database • Create an Action Plan to remedy the deficiencies, • Expanded to about 50 countries with a target compared to the necessary infrastructure • for the Agency to achieve the upgrading by the year 2000. • for the Country • Carry out the Action Plan • Verify that the country then has all the necessary JAERI-Conf components 97-008 Model Project Practice Groups Model Project Practice Groups Group 2: Low to high exposures; risk of serious Group 1: Low exposures; low risk of contamination; off site accident potential contamination and accidents includes: includes: • industrial radiography * medical diagnosis • medical radiotherapy • x-ray fluorescence • x-ray generators ♦ level gauges • industrial irradiators • small amounts of unsealed sources • uranium mining (e.g. radioimmunoassay) • non-power reactors Model Project Practice Groups Group 3: Multiple exposure pathways; risk of significant contamination; potential for serious off site accident consequences includes: • power reactors • nuclear fuel cycle facilities JAERI — Conf CT> tso 97 — 008 Country Safety Officer Country Safety Officers are expected to: A member of staff from the Safety Department • Collect and maintain up-to-date information responsible for a small group of countries for: on radiation and nuclear safety - assembling information on the safety status • Provide inputs for Country Profile database - creating an Action Plan • Prepare, in co-operation with TO's and NSRW TC Co-ordinator, action plans - co-ordinating implementation either through the Model Project or National Projects • Monitor progress on implementation of action plans. Action Plans Action Plans • From analysis of completed questionnaires and other The action plan should be based on Member data against requirements for an adequate infrastructure States' needs for developing or strengthening • CO, NSRW-TC Co-ordinator, TO's together with relevant people in TC prepare an Action Plan that, when fully targeted infrastructure elements within a implemented, will achieve adequate infrastructure reasonable period of time. • Action Plan includes - Agency actions - Member State actions • Needs to be agreed JAERI-Conf 97-008 Technical Officers ______Regional Cooperative Agreement • Appraise project requests • Prepare workplans and schedules • Prepare job descriptions and proposals for experts • Prepare procurement requests • Evaluate fellowships • Prepare training courses and workshops JAERI-Conf 97-008 RCA by methods RCA sources from accidents Radiation 2/3 IAEA determined services ______sealed RAS/9/006 (1993-1997) after measurement of the - ures Response about t of tasks is Preparedness from period aspect and Strengthen control consultant and 1/3 rastruc to f on In Coordinators assessment and designation funding and areas Emergency project Project Dose Regulatory Planning Year Expert Intercomparison Training Workshops - - - otection Project National Current donors 5 Agency • • • • Off-Site • • • • RCA Pr Man in: Asian s Safety Zealand e Nam Lanka States Philippines Pakistan Sri Viet Myanmar New Singapore Thailand ramm Basic Response Reference the prog of s of Member Education ude RCA Dosimetry incl and Korea Emergency 6 17 of 00 are Mongolia Rep. Malaysia Bangladesh China India Indonesia Japan Australia S/9/ Implementation Standards Characterization Personal Off-Site A Training • • • • • R There 164 X N tn m a®s j l x s h cn Personal • • • Imple • • • • • - - - - Expert Workshops Personal - - - Training National Expert National Expert 2, Measurement 1. Radon Contamination Calibration External 37 14 men RCA RCA Advisory measurement meeting Dosi missions dosimetry tati dose on seminars networks dosimetry Member techniques on metry implementation protocols assessment monitoring of Groups to the states ____ intercomparisons draft services BSS ______ ("90-'92) guidelines ( ’ _____ 95-'97) Training • • • Po Characterization - - - - - Regional C h a r a c t e r i s t i c s : RCA Course Development Over Significant Use Physiological Anatomical Physical Implementation Biological Dosimetry Emergency Metabolic pulation Countries 1,000,000 of and and phantoms difference seminar f or workshops dosimetry people - - education > response Internal of Rad 50% distance of of from to iation -111 sampled of the in the promote the radiation and Western Reference world BSS Protection learning external population Reference T protection & A Asian E Zf materials P" v ■ Man ;; • cZ" ICHF fUl. U»<1 - 97-008 JAERl-Conf JAERI-Conf 97-008 Intercomparison Exercise for Surface Contamination Monitoring NEXT PAOE(S) left BLANK JAERI —Conf 97-008 In this workshop, special emphasis was given to carrying out the intercomparison exercise with the contamination monitoring instruments that are in daily use at the participants facilities. This chapter gives results of the intercomparison exercise for surface contamination monitoring performed in this training workshop. Every participant had been requested to bring their routinely-used contamination survey instruments with them. Ten participants brought their contamination survey instruments and participated into the intercomparison exercise. The other participants, who could not bring their instruments, participated into the exercise with JAERI survey meters. Table 1 shows the specifications of surface contamination monitoring instruments employed by the participants. JAERI prepared two types of quasi-surface- contamination samples for the intercomparison measurement. They are flat sample sources, but the sizes of the radioactive area are different each other; one is 100 mm by 150 mm and the other is 25 mm <}>. The former is larger than window area of common instruments and the latter is the reverse. This exercise could help the participants to understand that the difference of source size affects the determination of calibration factors for each instrument. The participants evaluated surface contamination of these samples according to the procedures given in the following pages. Table 2 summarizes overall results. The names of countries of the participants are given anonymously, and values in parentheses are the differences between the evaluated and the reference values. It is realized from the table that some evaluations may be seen to have significant deviations. The results of this intercomparison should help to identify problems on the surface contamination monitoring in the member states and to show the necessity of the calibration of instruments and the standardization of evaluation method in order to improve contamination monitoring technique. 169 JAERI-Conf 97-008 Intercomparison Exercise for Surface Contamination Monitoring 1. Purpose To provide participants with the information that can be used to improve the performance of their contamination monitoring method. 2. Instrument and equipment 2.1 Instrument Portable surface contamination monitor (Survey meter) 2.2 Beta-ray source simulating surface contamination Nuclide : l47Pm, 204T1, <'"Sr-9 "Y Dimension : 100 mm X 150 mm, 25mm 3. Procedures 3.1 Measurement procedures 1) Measure the background count rates. 2) Measure the count rates with the beta-ray sources simulating surface contamination. 3) Evaluate the activity per unit area. And indicate the value in attached table. • For accurate measurements, the detector shall be held stationarily during three times the response time. • All measurements shall be performed as the routine way at your facility. Please do not scratch the source surface by the edge of detector. 4) Consider and discuss the main reasons for the difference between the reference value and the evaluated value. - 170- JAERI-Conf 97-008 Data sheet Nation Survey meter type (detector (size), manufacturer) 1) The surface contamination with dimension of 100 mm X 150 mm Table 1 The surface contamination with dimension of 100 mmX 15 0 mm Source Reading value Activity per unit area (Bq/cm 2) 147Pm ( ) ( ) *Sr-Y ______(______L Background value ( 1 Area of detector window ______cm2 Calibration factor ______(______) or Conversion factor ______£______) Indicate the method of evaluation. 2) The surface contamination with dimension of 25 mm Table 2 The surface contamination with dimension of 25 mm # Source Reading value Activity per unit area (Bq/cm 2) 147Pm ( ) 2114-yj ( ) * ’Sr-Y ______£_____ L Indicate the method of evaluation. JAERI —Conf 97 — 008 3) Difference Table 3-1 The difference between the reference value and the evaluated value (The surface contamination with dimension of 100 mm X 150 mm) Source Reference value (Bq/cm 2) Evaluated value (Bq/cm 2) Difference(%) l47Pm 2"4TI "'Sr-Y Table 3-2 The difference between the reference value and the evaluated value ______(The surface contamination with dimension of 25 mm (|)) Source Reference value (Bq/cm 2) Evaluated value (Bq/cm 2) Difference(%) 1J7Pm : *'Sr-Y Evaluated - Re ference Difference (in %) x 100 Re ference 4) Discussions JAERI-Conf 97-008 2 2 2 2 cm cm cm' cm mg mg mg mg 2 30 2.5 thickness Window 4 mesh window window SL open end end type dia. probe probe proportional probe dia. in 2 scintillator PGM of in phosphor probe cm cm cm instruments 160 counter 2.5 GM Beta BP714A Pancake DP2R GM Pancake BP6R14A Bicro Plastic 5 Dual Type 2 cm calibration Bq monitoring in provide to nuclides features 25 Facility for Other Audio contamination , change 3 cpm range 0-3k, surface 10 X of auto , range 2 0-lk, 0-1000 10 XIOOcps auto scale scale scale scale X1,X10, X readout 0-300, 0-30k,0-100k kcpm cps 10', 0-100, 4 kcps kcps kcps kcps 100 10 X1,X10, Digital Analogue change X Quasi-log 0-5 Quasi-log Quasi-log 0-6X0.1, 0-2 0-5 Quasi-log 0-5 0-100, 0-10k, Range X 0-1X 0-10, Specifications 900 3 1 Monitor 122 |3 lebe!) LB RM-20 Model Ratemeter ratemeter meter BACC Table TGS-123C range instruments of RM6 CM5 PCMS) PCM5 Berthold FAG(low NE NE Ludlum survey Mini NE Aloka Eberline RSSD Wide Chinese (cf. Type LANKA THAILAND PHILIPPINES SRI MONGOLIA PAKISTAN KOREA MALAYSIA INDIA INDONESIA JAPAN Nation CHINA 173 Table 2 Results of intercomparison exercise for surface contamination monitoring Participant Surface contamination 100 mmX 150 mm Surface contamination 25 mm (G) 0.33 0.11 0.09 353.0 122.2 14.9 (-97%) (-97%) (-99%) (-68%) (-72%) (-87%) 97-008 (H) 6.0 2.3 No sensitive 22.6 8.3 No sensitive (-54%) (-43%) (-98%) (-98%) (I) 30.1 7.1 2.1 459 114 6 (+132%) (+78%) (-65%) (-58%) (-73%) (-95%) (J) 14.5 3.8 3.2 776 298 29 (+12%) (-5%) (-47%) (-29%) (-31%) (-75%) (K) 25.0 8.33 0.67 231.67 65.0 1.67 (+92%) (+108%) (-89%) (-79%) (-85%) (-99%) (L) 14.6 4.3 3.4 168.1 37 7.1 (+12%) (+8%) (-43%) (-85%) (-91%) (-94%) (M) 4.3 9.13 5.4 166.5 49.5 80 (-67%) (+128%) (-10%) (-85%) (-88%) (-31%) (N) 11.7 4.9 8.7 32.8 15.7 5.9 (-109%) (+23%) (+45%) (-97%) (-96%) (-95%) (O) 14.2 3.2 1.0 856 328 38.4 (+9%) (-20%) (-83%) (-22%) (-24%) (-67%) JAERI-Conf 97-008 Appendix Workshop Agenda and Participants ’ List NEXT PA@E(S) left BLANK JAERI-Conf 97-008 IAEA/RCA Training Workshop on Contamination Monitoring Tokai, Japan October 21 - 25, 1996 AGENDA MONDAY October 21 [Conference Room No. 7] 11:30 - 12:00 OPENING SESSION Chairperson: K. Fujimoto ( NIRS: RCA National Coordinator of Japan ) 1. Opening Address R.V. Griffith (IAEA) 2. Welcome Address by Japanese Government 1) S. Nagayoshi (MOFA) 2) Y.Izumi (STA) 3. Introducing of Workshop Director Workshop Director K. Bingo (JAERI) Deputy Director H. Ishiguro (PNC) 4. Welcome Address by Workshop Director K. Bingo 5. Introducing of Participants 6. Information from Secretariat H. Murakami 12:00 - 13:20 LUNCH 13:20 - 17:35 SPECIAL LECTURE SESSION 13:20 - 14:05 Chairperson: K.Bingo Radioactivity Standards and Calibration Methods for Contamination Monitoring Instruments Makoto YOSHIDA (JAERI; Japan) 14:05 - 14:50 Chairperson: H.Ishiguro Monitoring and Evaluation Techniques for Airborne Contamination XIA Yihua (CIAE; China) (14:50 - 15:20) COFFEE BREAK 15:20- 16:05 Chairperson: K. Fujimoto Monitoring and Evaluation Techniques for Surface Contamination David WOODS (ANSTO; Australia) 16:05 - 16:50 Chairperson: Xia Yihua Features and Properties of Instruments for Contamination Monitoring Richard V. GRIFFITH (IAEA) - 177- JAERI-Conf 97-008 16:50 - 17:35 Chairperson: David Woods Basic Techniques of Decontamination Naoki YOKOSAWA (PNC; Japan) TUESDAY October 22 [Conference Room No. 7] 9:10 - 9:40 SPECIAL STATEMENT Review of the IAEA Program in Radiation Protection R.V. GRIFFITH 9:40 - 12:00 COUNTRY REPORT SESSION I Chairperson: Arlean L. ALA MARES 9:40 - 10:00 BANGLADESH 10:00 - 10:20 CHINA 10:20 - 10:40 INDONESIA (10:40 - 11:00) COFFEE BREAK Chairperson: Hoang Van Nguyen 11:00- 11:20 INDIA (1) 11:20 - 11:40 INDIA (2) 11:40 - 12:00 KOREA 12:00- 13:30 LUNCH 13:30 - 17:20 COUNTRY REPORT SESSION II Chairperson: Sarwar NAQVI 13:30 - 13:50 MALAYSIA 13:50- 14:10 MONGOLIA 14:10- 14:30 MYANMAR Chairperson: Fookiat SINAKHOM 14:35 - 14:55 PAKISTAN (1) 14:55 - 15:15 PAKISTAN (2) 15:15 - 15:35 PHILIPPINES (15:35 - 15:55) COFFEE Chairperson: K.S.Pradeep KUMAR 16:00 - 16:20 SRI LANKA 16:20- 16:40 THAILAND 17:40 - 17:00 VIETNAM WEDNESDAY October 23 EXERCISE SESSION I : [Meeting Room; Facility of Radiation Standard] 9:15 - 11:20 Intercomparison Exercise for Surface Contamination Monitoring 178- JAERI-Conf 97-008 (10:20 - 10:40) COFFEE BREAK 11:20 - 12:00 Technical Demonstration : R.V.Griffith 12:00- 13:20 LUNCH EXERCISE SESSION II : [Conference Room No. 7 & JRR-1 Facility] 13:20 - 15:00 Radiation Safety of Workers at a Contaminated Area (15:00 - 15:20) COFFEE BREAK 15:20 - 17:20 Technical Tour: Waste Management Facilities in JAERI THURSDAY October 24 EXERCISE SESSION III : [PNC; Engineering Demonstration Facility] 9:15- 9:30 Orientation 9:30 - 11:00 Practice on Calibration of Instruments 11:00- 12:00 Practice on Air Dust Sampling and Evaluation 12:00- 13:30 LUNCH 13:30 - 15:00 Practice on Air Dust Sampling and Evaluation ( continued ) (15:00-15:15) COFFEE BREAK 15:00 - 16:45 Practice on Smear Sampling and Evaluation 16:45 - 17:15 Practice on Air Dust Sampling and Evaluation ( continued ) 17:15 - 17:35 Discussion FRIDAY October 25 9:15 - 10:30 DISCUSSION SESSION: [Conference Room No.8] Chairperson: R.V.Griffith Comments: l)Xia Yihua 2) David Woods 3) Participants Discussion (10:30- 10:45) COFFEE BREAK 10:45 - 11:15 CLOSING SESSION: [Conference Room No. 8] Chairperson: R.V.Griffith 1. Certificate Awarding: K.Bingo (JAERI) 2. Closing Remarks: H.Ishiguro (PNC) 3. Closing : R.V.Griffith (IAEA) 11:30 ADJOURN - 179- JAERI-Conf 97-008 IAEA/RCA Training Workshop on Contamination Monitoring Participants List Ms. Aleya BEGUM Institute of Nuclear Science and Technology, Atomic Energy Research Establishment P.O.Box 3787 Dhaka -1000 BANGLADESH Mr. GOU Quanlu China Institute for Radiation Protection P.O.Box 120 Taiyuan, Shanxi 030006 CHINA Mr. K.S. Pradeep KUMAR Radiation Safety Systems Division, Bhabha Atomic Research Centre Mumbai - 400 085 INDIA Mr. G. NATARAJAN Health and Safety Division, Atomic Energy Regulatory Board V.S. Bhavan Anushaktinagar Mumbai - 400 094 INDIA Mr. Gatot SUHARIYONO Center for Standardization and Radiation Safety Research (PSPKR-BATAN), National Atomic Energy Agency P.O.Box 7043 JKSKL Jakarta - 12440 INDONESIA 180 JAERI-Conf 97-008 Mr. Jong-11 LEE Health Physics Department, Korea Atomic Energy Research Institute P.O.Box 7, Daeduk-Danji Taejon KOREA Mr. Bustami bin ABU Malaysian Institute for Nuclear Technology Research (MINT) Bangi, 43000 Kajang MALAYSIA Ms. Navaangalsangiin OYUNTULKHUUR Central Radiological Laboratory, National Centre for Hygiene, Epidemiology and Microbiology Ulaanbaatar - 211049 MONGOLIA Ms. Kay Thi THIN Physics Department, Yangon University Yangon MYANMAR Mr. Sarwar NAQVI Health Physics Division, Karachi Nuclear Power Complex P.O.Box 3183, Karachi PAKISTAN Mr. Naseer MOHAMMAD Pakistan Atomic Energy Commission P.O.Box 27 DERA GHAZI KHAN PAKISTAN 181 JAERI-Conf 97-008 Ms. Arlean L. ALAMARES Philippine Nuclear Research Institute Commonwealth Avenue Diliman, Quezon City PHILIPPINES Mr. Sunil Shantha THENUWARA HANNADIGE Atomic Energy Authority 696, 1/1 Galle Road, Ceramics Building Colombo 3 SRI LANKA Ms. Fookiat SINAKHOM Waste Management Division, Office of Atomic Energy for Peace Vibhavadi Rangsit Road Chatuchak Bangkok 10900 THAILAND Mr. Hoang Van NGUYEN Nuclear Research Institute 13, Dinh Tien Hoang Dalat VIETNAM - 182- JAERI-Conf 97-008 Lecturers: Mr. David WOODS Radiation Protection Services, Australian Nuclear Science and Technology Organization PMB1, MENAI, NSW 2234 AUSTRALIA Mr. Richard GRIFFITH Division of Radiation and Waste Safety, Department of Nuclear Safety, Wagramerstrasse 5, P.O.Box 100 A-1400 Vienna AUSTRIA Mr. XIA Yihua China Institute of Atomic Energy P.O.Box 275-84 Beijing 102413 CHINA - 183- JAERI-Conf 97-008 Participants list (Japanese): Mr. Shoichi NAGAYOSHI Ministry of Foreign Affairs Mr. Yoshinori IZUMI Science and Technology Agency Mr. Kenzo FUJIMOTO National Institute of Radiological Sciences Mr. Kazuyoshi BINGO Japan Atomic Energy Research Institute (JAERI) Mr. Katsumune YAMAMOTO (JAERI) Mr. Mikio FUJII (JAERI) Ms. Kirnie OKUMURA (JAERI) Mr. Hiroyuki MURAKAMI (JAERI) Mr. Hsuimu Y ABUT A (JAERI) Mr. Masahiro NISHIZA (JAERI) Mr. Michio YOSHIZAWA (JAERI) Mr. Akira ENDO (JAERI) Mr. Fumiaki TAKAHASHI (JAERI) Mr. Tetsuya OISHI (JAERI) Mr. Keiyi SfflMOOKA (JAERI) Mr. Hideharu ISHIGURO Power Reactor and Nuclear Fuel Development Corporation (PNC) Mr. Kazushige NINOMIYA (PNC) Mr. Kunihiko SHINOHARA (PNC) Mr. Tomohiro ASANO (PNC) Mr. Kimio NODA (PNC) Mr. Kiyoshi FUJII (PNC) Mr. Satoshi MIKAMI (PNC) Mr. Hiroshi TANAKA (PNC) Mr. Yasuhisa ITO (PNC) Mr. Shiiyi HORIUCHI (PNC) Mr. Mitsunori SUZUKI (PNC) Lectures: Mr. Naoki YOKOSAWA (PNC) Mr. Makoto YOSHIDA (JAERI) - 184- (si) 6 *1 jic/mwa *2 si6#m$m,6#R *5 SlfgffltS m g to Id ^ g to Id to ftsSIS Id H S 3 y — h vu m ». »5. 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(cm2/s) 6.89476 x 10" 7.03070 x 10" 6.80460 x 10" 51.7149 1 J( = 10’erg) kgf-m kW* h caKItlS) Btu ft • Ibf eV 1 cal = 4.18605 JUtMffi) 1 0.101972 2.77778 x 10" 0.238889 9.47813 x 10" 0.737562 6.24150 x 10" = 4.184 J OMt?) 9.80665 1 2.72407 x 10" 2.34270 9.29487 x 10" 7.23301 6.12082 x 10" = 4.1855 J (15 "Cl 3.6 x 10‘ 3.67098 x 10s 1 8.59999 x 105 3412.13 2.65522 x 10* 2.24694 x 10“ = 4.1868J(BBsKa*) 4.18605 0.426858 1.16279 x 10" 1 3.96759 x 10" 3.08747 2.61272 x 10" EW'F 1 PS < (AM, /; ■ 1055.06 107.586 2.93072 x 10" 252.042 1 778.172 6.58515 x 10“ = 75 kgf-m/s 1.35582 0.138255 3.76616 x 10" 0.323890 1.28506 x 10" 1 8.46233 x 10" = 735.499 W 1.60218 x 10"19 1.63377 x 10"' 4.45050 x 10"26 3.82743 x 10"“ 1.51857x 10"' 1.18171 x 10"" 1 A Bq Ci »S Gy rad S3 c/kg R 1 Sv rem W i 2.70270 x 10"' 1 100 1 3876 a 1 100 m * « B 3.7 x 10'“ 1 0.01 1 2.58 x 10" 1 0,01 1 (86 ^ 12 fl 26 [lift)