AHomis Research Center
M.A. GOMAA R.M.K. EL-SHINAWY I A.T. ABDEL-FATTAH
(7th October 1991) CONTENTS
Page
1. Tentative Agenda. 1 2. Paper No 1: Fulure Sceintific Trends in Radiation Protection and Cevil Defence M.A. Gomaa. 3. Paper No. 2: On the Implementation of 1990 ICRP Recomendations M.A. Gomaa and M. El Naggar. ^3 4. Paper No. 3: Thermoluminescence Properties of Bone Equivalent Calcium Phasphate Ceramics, B.A. Henaish, A.M. El-Agramy and W.I. Abdel Fattah 49 5. Paper No. 4: Potassium body burdens in occupational users of Egyptian nuclear research Centre M.S. Abdel - Wahab, S.K. Yossef, A.M. Aly and A.T. Magda. 61 6. Paper No. 5: Transport of Radionuclides in Fresh Water Stream. W.E. Abdel Malek, A.T. Ibrahim, in R.M.K El-Shawy and N. Kamal. 76 7. Paper No. 6: Water treatment Process for Nuclear Reactor, M.A Marwan, N.S. Khattab A.N. Hanna. 77 8. Paper No. 7: Research Activities Related to Internal Contamination and Bioassay and Experience Gained in this Field, A. El-Sharnouby, M. Aziz, A shaaban and A. Atwa. 78 Page
9. Paper No. 8: Environmental Radiation Monitoring
in Inshass, A.T. Abdel Fattah, M.W. Essa, and M.A.
Gomaa. 81
10. Paper No. 9: Safe Transport of Radioactive
Materials and Other Activities, R.M.K. El - Shinawy 105 : THE SECOND RADIATION PROTECTION AND :
9.00 REGISTRATION
9.30 OPENING
10.00 FUTURE SCIENTIFIC TRENDS IN RADIATION PROTECTION AND
CIVIL DEFENCE, M.A. GOMAA
M ::::::BREAK: SCIENTIFIC AND TECHNICAL POSTERSm 11.00 ON THE IMPLEMENTATION OF 1990 ICRP RECOMMENDATIONS
M.A. GOMAA AND A. EL-NAGGAR.
11.30 THERMOLUMINESCENCE PROPERTIES OF BONE EQUIVALENT
CALCIUM PHOSPHATE CERAMICS B. HENAISH A. EL-AGRAMY
AND W.I. ABDEL-FATTAH,
11.45 POTASSIUM BODY BURDEN IN OCCUPATIONAL USERS OF
EGYPTIAN NUCLEAR RESEARCH CENTERS, M.S. ABDEL-WAHAB,
S.K. YOUSSEF, AM.ALY AND MAGDA T. ABBAS
12.00 TRANSPORT OF RADIONUCLIDES IN FRESH WATER STREAM W,
E, ABDEL MALIK, A.T. IBRAHIM AND R.M.K EL-SHINAWY AND
N. KAMAL
12.20 WATER TREATMENT PROCESSES FOR NUCLEAR REACTORS.
M.A. MARWAN M.S. KHATAB AND A. HANNA 12.40 EXPERIENCE GAINED IN BIOASSAY, INTERNAL DECONTAMINATION AND LOW LEVEL WASTE TREATMENT A. EL-SHARNOUBY, M. AZIZ., A. SHABAN AND A.N.ATWA 13.00 ENVIRONMENTAL RADIATION MONITORING IN INSHASS AREA A.T. ABDEL FATT AH, M.W. ESSA AND M.A. GOMAA
BREAK.:EXHIBITION•:•:•:•:•:•: :(7rio.o.CT.:i99i).. •:•
14.00 SAFE TRANSPORT OF RADIOACTIVE MATERIALS AND OTHER ACTIVITIES R.M.K. EL-SHINAWY 14.30 RECOMMENDATIONS 15.00 ADJOURN 2nd Rad. Prot. Conf. (1991), Paper No. 1
•FUTURE SCIENTIFIC TRENDS IN RADIATION PROTECTION AND CIVIL DEFENCE' M.A. GOMAA
Radiation Protection and Civil Defence Dept. Nuclear Research
Center, Atomic Energy Authority, Cairo, Egypt
ABSTRACT:
The aim of the present study is to review current and future scientific trends in radiation protection and civil defence of the nuclear research center of the Atomic Energy Authority of Egypt.
The main areas of interest are summarized as follows :
1. ORGANIZATION AND TRAINING:
Up dating radiation protection legislation,
Information transfer via translation to Arabic of latest I AEA &
ICRP Publication,
Establishing data base for rad port services in Egypt.
Updating radiation protection training programmes, and
Application of decision - aiding techniques in radiation
protection (justification and optimization).
2. OCCUPATIONAL RADIATION CONTROL:
Mixed field external personnel dosimetry/detection,
Internal dosimetry / Biophysis/W B counter / Bioassay,
Area monitoring air and surface, and
Updating radiation format. -4 - GOMAA
3. PUBLIC RADIATION CONTROL:
Environmental programme around Inshass reactor,
Transport of radioactive materials,
Transfer of radioactivity in aquatic environments,
Storage of radioactive wasts, and
Annual exposure to members of the public - natural radiation,
medical and others.
4. EMERGENCY:
First aids decontamination,
Decontamination of personnel and equipment,
Dose calculation after accidents, and
Handling radiation accidents / emergency planning.
5. QUALITY ASSURACE:
Calibration of radiation sources and measuring devices, and
Intercomparison personnel dosimetry programme.
The proposed plan can be completed within 1992 - 1997 national plan with the support of IAEA technical assistance, research Contract and proper funding from the Atomic Energy
Authority of Egypt.
INTRODUCTION :
Research activities in the field of radiation protection and civil defence were summarized in the book entitled "Research activities of Radiation Protection Deprartment - 1961 - 1990". 2nd RFC 11991) "5"
The Frist conference of radiation protection and civil defence Dept. was held at the Rad. Prot. Dept. NRC AEE on 27/11/1989. In which technical reports as well as research work in progress were presented. During the Seond Conference of the Radiation Protection and Civil Defence Dept. technical reports, research papers as well as work in progress shall be presented. The seond conference shall be held at K. A. Mahmoud Hall on 7/10/1991. The main aim of the present study is to review current and furture scientific research trends in radiation protection and civil defence. In the present study, the author followed IAEA publication No. 101 entitled "Operational Radiation Protection : A Guide to 2 Optimization" . In which the major elements of an effective operational radiation programme are the following : 1. Orgaiztion and management. 2. Personnel selection and training. 3. Occupational radiation control. 4. Public radiation control. 5. Quality assurance. 6. Emergency operations. Members of the scientific staff, specialists and technicians of the radiation protection dept. are engaged in the last thirty years in these areas. - 6- COMAA
Future scientific tends in radiation protection are needed to be known for two main reasons. 1. The third national plan 1992-1997 is understudy, in which scientific programmes should be included. 2. The executive regulation of the atomic energy authority issued in 19913 shall be implemented completely soon. Scientific trends shall also be included in the reorganization of the technical and scientific activities within the various centers of the AEA in general and in the NRC and the National Center for Nuclear safety and radiological protection in particular. 1.0 ORGANIZATION AND TRAINING: In this section scientific trends may be classified into : 1. Legislation and Organization. 2. Information transfer, and 3. Training. Future scientific trends are aimed into : 1.1 UP DATING RADIATION PROTECTION LEGISLATIONS: The latest published legislation in Egypt is the ministry of Health order No. 265 (1989)4 which deals with the Industrial Radiography Instructions. In 1990 - Rad Prot. rules and procedures were distributed among NRC members. The AEA formed a committee by AEA order No 14 (1991)6 in order to evaluate radiation protection and radiological control 2nd RPC (1991) - 7 -
matters within the AEA. New rules and procedures are
reformulated according to latest IAEA and ICRP
recommendations, committee work is in progress. The Ionizing
radiation law of Egypt (I960)7 needs to be updated.
1.2. ORGANIZATION:
A study was completed by the author (1991)8 aimed to establish
the General Directorate for Radiation Protection and industrial
safety and the scientific Rad. Prot. department of the nuclear
reserach center. In this study, the directorate is composed of three
main directorates :
a. Physical surveillance directorate,
b. Radiation services directorate, and
c. Industrial saftey directorate,
THE SCIENTIFIC DEPT. IS COMPOSED FROM SEVERAL UNITS:
a. Personnel dosimetry unit,
b. Environmental monitoring unit, and
c. Public radiation control unit,
Rad. Port, infrastructure research work is needed in order to :
1. Minimize repeated research activities in progress in the centers
of the AEA.
2. Proper selection of the AEA divisions in which radiation
protection departments represet an integral parts or its
reunification in a single AEA division. - 8 - COMAA
Furthermore, research work is needed in the field of "Application of decision aiding techniques in radiation protection". In 1990 - 19919", litte experience was gained by the author in this field, where the cost effectiveness and cost-benefit analysis was used in radiation shielding studies for Am-Be neutrons. ALARA group formed by AEA order no (1990) should complete its work and ALARA center should be formed. The atomic energy authority formed several committees to deal with organizational aspects of handling and dispose of radioactive waste11, transport of radioactive materials and 12 calibration (National Calibration Institute) . Research works are needed in these areas. 1.3. INFORMATION TRANSFER: Research activities are needed in the fields of : a. Translation of IAEA and ICRP recommendations to Arabic. b. Establishement of data base for radiation protection purposes. 1.3.1. TRANSLATION: Three important documents were translated into Arabic recently by Salama and revised by the author : 1. IAEA, Safety Series No. 104, extension of the principles of radiation protection to sources of potential exposure, 1990. Revised by Gomaa and Hammad. 2nd RPC (1991)
2. IAEA, Safety Series No. 101-operational radiation protection : a
Guide to Optimization, 1990 revised by Gomaa.
3. ICRP summary of 1990 recommendations of the international
commission on radiological protection, 1991. Revised by Gomaa
and El - Naggar.
Future work is aimed to translate the main 1990
recommendations of ICRP into Arabic as well as latest IAEA
publications.
1.4. DATA BASE FOR RADIATION PROTCTION PURPOSES:
IAEA through its radiation safety section and IAEA Advisory
group proposed in 199016 the following :
1. Creation of a data base of date bases project by IAEA staff,
2. Initiation of coordinated research programmes to transfer
appropriate data bases and programmes to IAEA, and initiation
of coordinated research programmes to fill the important gaps
in the data bases programme.
The subject arese for date bases are : 1. Sources. 9. Results of exposure. 2. Transport. 10. Optimization in ALARA. 3. Dispersion. 11. Rad. port practice. 4. Shielding. 12. Regulation and standards. 5. External Dosimetry. 13. Health effects. 6. Internal dosimetry. 14. Medical countermeasures. 7. Assessment of exposure. 15. International Expert Directory. 8. Measurement of exposure. 16. Training and expert. -10- GOMAA
Henuish and Gomaa are working in these areas :
1.5 RADIATION PROTECTION TRAINING PROGRAMES:
1.5.1 BBASIC RAD. PROT. TRAINING PROG. (BRPTP): Members of the staff of the Rad Prot. Dept are engaged in teaching dosimetry as well as radiation protection aspects in the basic Rad. Prot. Training courses at: 1. Nucl. Safety and Rad. control center courses . 2. Middle East Regional Center for Radioisotopes course. 3. Faculty of medicine/Alexandria University course. Efforts are made to up date the course syllabus according to IAEA training courses , and IAEA and JCRP publications. 1.5.2 ADVANCED RAD. PROT TRAIN PROG (ARPTP): Advanced rad Prot. train, programme were designed by the staff members of the rad Prot. Dept. in 1991 . It is designed in order to : a) Up grade health physicists who attended BRPTP to the rad Prot. expert level. b) Up grade specialists and assistant lecturers for pre-qualification to senior jobs at the rad port directorate or departments.
1.53 MEDICAL PHYSICS DIPLOMA: The atomic energy authority and the faculty of medicine, Alex. University discussed the need to establish diploma in the field of 2nd RPC (1991) ~U~
Medical physics. During the course of study students shall attend on job training in Rad. prot. matters at AEA. The letter of agreement was signed in 1990 and the administrative matters are in progress.
1.5.3 HANDLING RADIOLOGICAL AND NUCLEAR ACCIDENT
TRAINING PROG.: The first training course on handling radiological and nuclear accident was completed last Oct. 1990. The course was given by AEA members to military physicians and nurses. Training facilities and equipment are needed in order to up date BRJTP, implement the ARPTP, the Rad. Prot aspects of the Medical physics Diploma and the Radiological and Nuclear Accident Programme.
1.6 THE ATOMIC ENERGY TECHNICAL INSTITUTE (AETI): A committee was formed by AEA order 1991 in order to study and estiblish the Atomic Energy Technical Institute (AETI). The aim of the AETI is to train graduates of secondary schools (general and Technical) to assistant users of ionizing radiation. The graduates of AETI shall be trained in one of the following areas as: 1. Industrial radiography. 2. Health physics/technician. 3. Operator of large radiation facilities. -12 - GOMAA
4. Maintenance technician of nuclear-electronic equipment. 5. Radioisotope laboratory technician. AEA contacts with the Ministry of Higher Education led to aprove the AEA proposal and formation of joint committee. The joint committee agreed upon the general lay out of the institute plan. The detailed syllabus of the two year course was designed by the AEA subcommittee. And the course plan was forward to the AEA chairman in June 1991, for Approval. The AETI matters were forward to the AEA vice chairman for training and international co-operation for further study. The vice chairman suggested that the first semester of the course to be taught as A.E.A. programme to AEA trainee. The proposed date of the programme is 15 Nov. 1991 for 3 month. Successful candidates are promoted to the higher Atomic Energy training programmes. The estimated 5 years budget of the AETI is LE 3.000.000 and it is proposed to be included in the next five years plan of Egypt.
2.0 OCCUPATIONAL RADIATION CONTROL:
In the field of occupational radiation control attention is paid to the individual and the work place.
2.1. FOR INDIVIDUAL: Research is in progress in the fields of: 1. Radiation detection physics of radiation effects. 2nd RPC (1991) -13"
2. Mixed field external personnel dosimetry : a) Development of personnel dosimeters. b) Neutron and radon dosimetry. c) Implementation of ICRU and ICRP recommendations. 3. Inernal personnel dosimetry. a) Development of techniques for quick method of determination of radionuclides in Body (W.B.C and Bioassay): b) Internal dosimetry and biophysical modelling. c) Implementation of ICRP recommendations.
2.2. FOR THE WORK PLACE: Research and development in the field of: a) Air contamination and radon in work place. b) Surface contamination. c) Air kerma measurements. d) Implementation of ICRP recommendation.
3.0. PUBLIC RADIATION CONTROL: In the field of radiation control attention is paid to the non occupational worker within the AEA and to members of the public. - 14 - GOMAA
3.1 FOR NON RADIATION WORKER IN PROGRESS IN THE AREA OF RESEARCH: a. Environmental radiation monitoring around Inshass reactor. b. Storage of radioactive waste (See Rad. Prot. Dept. report 1991 )22. c. Transport of radioactive materials from work place to another and to SRWF facilities. d. Air kerma measurements and personnel dosimatry during nominal working conditions. 3.2 FOP- MEMBER OF THE PUBLIC 3.2.1 TRANSPORT OF RADIOACTIVE MATERIALS IN AQUATIC ENVIRONMENT: a) Passage of shipment is Suez canal*. b) Radionuclides transport in Ismailia canal (Fresh water stream). c) Radionuclids transport in northern coast. d) Radionuclides transport in river nile. e) Dosimetric modelling.
ANNUAL EXPOSURE TO MEMBER OF THE PUBLIC-NORMAL: 1. Natural background radiation. a) Cosmic rays. b) 5-rays from the ground. 2nd RFC (1991) '15'
c) Air, water, soil, food, radon.
3.2.3 MEDICAL EXPOSURE:
1. External sources.
2. Nuclear medicine.
3.2.3 OCCUPATIONAL EXPOSURE:
3.2.4 TECHNICALLY ENHANCED EXPOSURE:
a) Consumer products.
b) Cosmic rays during flight.
c) radiation control in food.
33 RADIOLOGICAL CONTROL:
1. Imported food.
2. Consumer products.
3. Other imported items.
4.0 EMERGENCY:
1.4. ON SITE:
4.1.1. INDIVIDUAL - EMERGENCY PLANING:
1. First aid decontamination.
2. Decontamination of personnel.
a) External.
b) Internal.
3. Handling radiation accident.
4. Dose calculation during and after accident. -16 - GOMAA
4.1.2 Area Monitoring (Non - Radiation - Worker) :
1. Rad. Monitoring Network.
2. Air Monitoring.
3. Radioactivity in soil.
4. Radioactivity in water.
5. Source term calculation.
6. Handling radiation accident (control of entry).
7. Dose calculation.
4.2. OFF SITE* 23 1. Development of National incident involving , ionizing
radiation Research and development in Emergency plaining.
a) Sheltering.
b) Radioprotectors.
c) Evacuation.
d) Control of access.'
e) Personnel monitoring.
f) Medical care.
g) Control on food supply,
h) Control on water supply.
4.3 INTERVENTION:
a) Intervention at workplace.
b) Derived inteivention levels for food stuff.
c) Exemption rules. 2nd RPC (1991) -17'
5. QUALITY ASSURANCE: In order to be sure that radiation protection programmes are effective, the main elements to be included in the future trends of the progamme are the followings : 1. Calibration of sources and calibration of measuring devices. 2. Intercomparison personnel dosimetry prog.. 3. Up grading radiation protection forms. 5.1 CALIBRATION: Joint committee for co-operation with the national institute of calibration were formed by AEA order No 47 (1991). Its aim is to study all aspects of co-operation co-ordination, research, and studies and joint experience between the authority and NIC. Studies aiming to M.Sc. on Ph.D. are going on through channels of co-operation between AEA-NIC. in the fields of : 1. Construction of radon chamber. 2. Quality control of radioactivity in foodstuff. 3. Dosimetric estimation for pre-dosed sample. Furthermore, the dosimetric calibration laboratory of the Dept. needs updating (surces and measuring devices). 5.1.1 SOURCES: Reserarch and development are needed in order to: 1. Collect major sources in use of NRC in a large area sources hall (neutrons, 5-rays, and x-rays). 2. Install sources, so that working with them do not lead to -18 - GOMAA
expose the user to unnecessary exposure from the other sources. 3. Calibration of the sources.
5.1.2. MEASURING DEVICES:
5.1.2.1 PERSONNAL MEASURING DEVICES: The following items need periodic calibration : 1. Pocket desimeters and pocket alarm dosimeter. 2. Film badges and TLD. 3. Track detectors and others.
5.1.2.2. AREA MEASURING DEVICES: The following items needs periods calibration : 1. Neutron rem - meters. 2. 5-ray survey meters. 3. Contamination meters. 5.2. INTERCOMPARISON: Future work is aimed to participate in the intercomparison programmes for personnel dosimetry on the national and international levels. 5.3. RAD. PROT-FORMS: Research and development is needed in the following items : 1. Up grading personnel radiation record forms. 2. Radiation monitoring at work place. 3. Radiation monitoring out side work place. 4. Calibration forms. 2nd RPC (1991) -19 -
Annex -1 Proposed 1991 -1977 Five Years Plan Radiation Protection Activities
The main elements of the 1992-1997 Radiation Protection Activities and Atomic Energy Technical Institute Plans are as follows : 1. Centralization of the radiation protection research and services project Budget: 1300000 LE. 2. Protection services for security, safety and physical protection project. Budget: 1100000 LE. 3. Envrionmental radiation monitoring at Inshass Area project. Budget;! 300000 LE. 4. Updating dosimetric calibration laboratory project. Budget: 450000 LE. 5. Decontamination of surfaces and equipment central labroatory project. Budget: 750000 LE. 6. Atomic Energy Technical Institute project. Budget: 3000000 LE.
IAEA SUPPORT: 1. IAEA Technical Assistance Project No. EGY/09/26. - 20 - GOMAA
Personnel Dosimetry - 1991 - 1992 (in progress). 2. IAEA proposed Research Contract - submitted last April 1991. Radiation Medical Exposure in Egypt. 3. IAFA Technical Assistance Project-submitted to AEA, hence IAEA, Sept. 1991. Phase II of Environmental Radiation Survey Around Inshass Site (Phase I-IAEA Technical Assistance Project No EGY 9/017). NATIONAL SUPPORT: 1. Scientific Research Accademy Project, environmental radiation survey in Egypt. (In progress -1990 -1992). LOCAL SUPPORT: 1. Radiation protection advisory and services, Scientific research supporting council of AEA 1989 - 1992. REFERENCES: 1. W. Abdel-Malik, Research activities of Rad. Prot. Dept., (1961-1990), Rad. Prot. Dept, NRC, AEA, Egypt., 1991. 2. IAEA, Operational Radiation Protection : A Guide to Optimization, Safety Series, 101, Vienna, 1990. 3. The Executive Regulation of the Atomic Energy Authority, Republican order No 47, Egypte, 1991. 4. Radiation protection procedures for industrial radiography Ministry of Health order No. 265, 1989 El - Waqah El - Masrial 1991. 2nd RPC (1991) -21-
5. M.A. Gomaa and R.K. El-Shinawy, Radiation protection rules and procedures - short Note, rad. Prot. Dept, NRC, AEA, 1990. 6. Atomic Energy Authority Order No. 15-on the formation of Radiation Protection and Control Committee, 1991. 7. The Ionizing Radiation Law, No. 59, I960, Egypt. 8. M.A. Gomaa, A study for the establishment of NRC Directorate for radiation protection and industrial safety and the scientific department of rad. prot. April 1991, Rad. Prot. Dept., NRC, AEA, Egypt, (In Arabic). 9. M.A. Gomaa, on the use of decision - aiding, techniques in radiation protection, Nov. 1990, IAEA-Tech. Comm-Meeting, IAEA, Vienna, 1990. 10. Atomic Energy Authority order No. 59, - for study NRPB book, ALARA between theory and practice, 1990. 11. Atomic Energy Athority order No. 14, for Handling and dispose of radioactive waste, 1991. 12. Atomic Energy Authority Order No. 47, for co-operation with National Institute for calibration, 1991. 13. H.N. Salama, Translation of (IAEA - SS - 104 (1990): Extension of the principlal of Rat. Prot. for Potential Sources) - Revised by Gomaa and Hammad - In Arabic - Rad. Prot. Dept., NRC, AEA, 1991. 14. H.N. Salama, Translation of (IAEA-SS-101 (1990). Operational - 22 - GOMAA
Radiation Protection). Revised by Gomaa - In Arabic - Rad. Prot. Dept, NRC, AEA, 1991. 15. H.N. Salama, Translation of (summary of 1990 Recommendations of the ICRP). Revised by Gomaa and EL - Naggar - In Arabic Rad. Prot.. Dept., NRC, AEA, 1991. 16. IAEA Recommendation of the IAEA Advisory group meeting on Rad. Prot. data base services for member states, IAEA, Vienna, 26-28 Nov. 1990. 17. B. Henaish and M.A. Gomaa, work in progress in Rad, Prot data base services in Egypt. Rad. Prot. Dept., NRC, AEA, Sept. 1991. 18 AEA-Secretory General order No. 22, 1990 for the supervision of training in radiation safety, Egypt. 19. IAEA Technical Report Series, Radiation Protection Training, No. 280,1990. 20. Advanced Rad. Prot. Training programme, letter to the chairman of AEA, April 1990. 21. AEA order No. 41 study and establish of Atomic Energy, Technical Institute, 1990. 22. W. Abdel Maliu et al., SRWSF, Rad. Prot. Dept. report, NRC, AEA, 1991. 23. AEA order No. 53 (1990): to study NAIR - NRPB booklet. 2nd Rad. Prot. Conf. (1991), Paper No. 2
ON THE IMPLEMENTATION OF 1990 ICRP RECOMMENDATIONS Muhammad A. Gomaa, and Anas M. El - Naggar Radiation Protection Department, Nuclear Research Center, AEA Cairo, Egypt
ABSTRACT: The International Commission on Radiological Protection (ICRP) adopted new recommendations on radiation protection in November 1990. These, recommendations are based on the latest radiation risk estimates derived from the reanalysis of the atomic bomb survivor data and other epidemiological studies. These recommendations are published in 1991 as ICRP Publication 60. Based on biological aspects, the Commission adopted a multidimensional concept of "Detriment", and assessed the distribution of detriment in organs and tissues. This introduced new terminologies and some changes in tissue weighting factors. The important feature of the new ICRP recommendations is the introduction of new dose limits for the control of Occupational Exposure, and control of Public Exposure. This treatise presents a concise review of the fundamental features of the new ICRP recommendations, the biological criteria, - 24 - GOMAA & NAGGAR
and the developed conceptual framework of "Individual Dose, and Risk limits" to replace the previous concept of System of Dose Limitation in the ALARA triad. Speculative proposals of amendments to the Egyptian Regulatory Authorities based on the new ICRP publication 60 recommendations are discussed. An appraisal to implement the new ICRP recommendations in Egypt, several implications arise as consequence of the new Limits on Effective Dose. These implications require consideration by the Regulatary Authorities aiming at modification of Radiation Protection framework. Major modifications include recategorization of occupational workers, reclassification of working aress, and the establishment of source-related constraints by computing the necessary shielding required for the currently operating and new radiation facilities. Further relevant aspects are also considered.
1. INTRODUCTION:
National Radiation Protection Authorities are continually introducing appropriate -modifications in the radiation protection framework regulations. These modifications are adopted according to the recommendations of International bodies concerned with Radiation Protection, and which are consequent upon rigorous analysis of relevant data presented by scientific institutions on the subject of the biological effects of ionizing radiations. The most 2nd RPC (1991) "25"
leading authority on Radiation Protection is the International Commission on Radiation Protection (ICRP). In Egypt, the Ionizing Radiation Law is the Republican order No. 59, for the year 1960. It deals with Regulatory and Organizational Aaspects of Radiation Protection from Inoizing Radiations. The executive Legislation of the Law was issued by Ministery of Health order No. 630 (1962). The Atomic Energy Authority (AEA) operated the Radiation Protection measures according to the articles of the law. However, later in (1986), the Atomic Energy Authority (AEA) adopted the International Atomic Energy Agency (IAEA) Saftey Series No. 9 (1982), entilted "Basic Safety Standards for Radiation Protection". This same publication was also adopted by the Ministry of Health for Radiation Protection Procedures in Industrial Radiography, according to ministerial order 265, (1989).
The IAEA Safety Series No. 9 (1982) "Basic Safety Standards for Radiation Protection" adopts the Recommendations of the ICRP appearing in publication" 26 (1977). It is expected that scientific institutions as the IAEA, WHO, ILO, and similar national bodies will develop new standards of Radiation Protection to conform with the new ICRP publication 60 (1991). It is also expected that the Radiation Protection Reulatory - 26 - GOMAA & NAGGAR
Authorities in Egypt will recognize the recommendations of the the new ICRP publication, and will consider the potential impacts of their implementation. 2. REVIEW OF ICRP PUBLICATION 60: The ICRP publication 60, (1991) is intended to assist the regulaory, advisory and management bodies in matters related to the protection of man from the hazards of ionizing radiations. It emphasises that Radiological Protection cannot be conducted on scientific considerations alone, and the value judgements have to be made about balancing of risks and benefits. Radiological protection aims at avoiding "Deterministic Effects" by setting dose limits below their thresholds. However, the problematic enigma underlying the occurrence of "Stochastic Effects" with low frequency even at lowest doses, necessitate that this has to be taken into account in setting "Dose limits". 2.1 Quantities in Radiological Protection:
ICRP publication 60 introduces new "Quantities" in radiological protection based on macroscopic dosiznetrric considerations. The Mean Absorbed Dose in a Tissue or Organ
(DT), is the energy absorbed per unit mass. The Equivalent Drw»>
in a Tissue or Organ (Hj.), is formed be weighting Factor (WR);
(HjsDT. WR). The Effective dose (E) is formed by weighting the
Eqivalent Dose by the Tissue Weighting Factor (WT), and 2nd RPC (1991)
summing over the tissues (E= Hj Wj). The Effective Dose is the sum of the Weighted Equivalent Doses in all the Tissus and organs of body given by the expression, where (H-r.) is the Equivalent dose in tissue or organ (T), and (Wj) is the weighting factor for tissue (T). It is a quantity expressed as the sum of doubly weighted absorbed dose in all tissues and organs of the body. The Committed Effective Dose (E) is the time intergral of the effective dose rate following an intake of radionuclide (where t) is the integration time in years) following the intake. The Collective
Effective Dose (E/-Oii) is the product of mean Effective Dose in a group and the number of individuals in that group. It can be thought of as representing the total consequences of an exposure of a population or a group. The Unit of the absorbed dose is the Gray (GY), and the Unit of the equivalent and effective doses is the Sievert (Sv). The values of Radiation Weighting Factors (Wn), and of Tissue Weighting Factors (Wy) are given in Appendix 1. 2.2 Biological Considerations:
The recommendations of the ICRP included in Publication 60 are based essentially on the biological aspects of the effects of ionizing radiations. "Deterministic Effects" result from cell depletion, or sufficient impairment of tissue function, after large doses. The probability of such harm is zero at low doses. - 28 - GOMAA & NAGGAR
However, above some level of dose (the threshold for clinical effect) the probability and the severity of effect increase with dose. The threshold for these effects is about 2 Gy. Deterministic effects in children exposed in utero should be considered to prevent the occurrence of mental retardation and other effects. "Stochastic Effects" may develop resulting from radiation induced modification of a somatic cell. This may subsequently, after a latent period, progress into malignant transformation of the cell. The probability of cancer induction increases with increments of dose probably with no threshold. "Hereditary Effects" occur when the damage occurs in a genetic cell whose function is to transmit genetic information to later generations. These effects appear in the progeny of the exposed individual.
2.2.2 Fatal Cancer Risk Probabilities: The UNSCEAR and BEIR estimated the lifetime Cancer risk by considering the accumulated data to 1985 of Japanese atomic bomb survivors, the new dosimetric techniques (DS 86), and lifetime projections by multiplicative and modified multiplicative models for high dose, and high dose rate exposures. These indicate that the form of the dose-response relationship is linear quadratic for Low LET radiations. The linear coefficient at low doses or low dose rates is obtained from the high dose, high dose rate estimates of risk by dividing by a Dose and Dose Rate Effectiveness Factor (DDREF) of 2. 2nd RPC (1991) -29-
Based on this data, the ICRP estimated the Nominal Fatal Cancer probabilities for a working population, and for a general population, which differ because of the greater sensitivity of young individuals. This is given in table 1.
Table (1): Nominal Probability Coefficients for Stochastic Effects.
Exposed Population Adult Workers Whole Population
Radiation Effect Detriment (W2 Sv'1)
Fatal Cancer* 4.0 5.0 Non Fatal Cancer 0.8 1.0 Severe hereditary effect 0.8 1.3
Total 5.6 7.3
* For fatal cancer, the detriment is equal to the Probability Coefficient Estimates of the distribution of fatal cancer risk among organs, and the length of life lost for cancer in each of these organs. These are calculated by the Commission after further analysis of Atomic Bomb survivors data. 2.2.2 Hereditary Effects Probability Coefficients: These are also based on the assessments of the UNSCEAR and BEIR experimental data on genetic effects in animals, which are comparable to the estimates of the corresponding effects in man. The Nominal Probability Coefficient of severe Hereditary effects in all generations after low dose and low dose rate are given in table 1, for the working groups and for the general population. - 30 - GOMAA & NAGGAR
2.2.3 Detriment: The term Detriment is used to represent the combination of the probability of occurrence of harmful health effect, and a judgement of the severity of that effect. In this respect, a multi-dimensional concept is adopted, whose main components are: (1) The probability of attributable fatal cancer. (2) The weighted probability of attributable non-fatal cancer. (3) The weighted probability of severe hereditary effects. (4) The length of life lost if the harm occurs. All those components are stochastic quantities. The value of this Aggregated Detriment at low dose is in rounded values 10"16 Sv , shown in table 1. The distribution of detriment in organs and tissues is assessed by considering a) Probability of fatal cancer in each tissue or organ, b) Multiplying by an appropriate factor for non-fatal cancer, which is determined by the lethality factor for that cancer, c) Adding the probability of severe hereditary efects. d) Adjusting for the relative length of life lost. The distribution of aggregate detriment among organs is represented, after appropriate rounding, by the Tissue Weighting factor (Wy). The Commission adopts the term Detriment as a complex concept combining the probability, severity, and time expression of harm. 2nd RPC (1991) "31"
2.3 Framework of Radiological Protection:
Radiation "Practices" are human activities that increase the
overall exposure to radiation. "Intervention" are human
activities that decrease the overall exposure by influencing the
existing causes of exposure.
Four Types of Exposures are Recognised:
a) Occupational Exposure, which is the exposure incurred at work
and essentially as a result of work.
b) Medical Exposure, which is the exposure of individuals during
diagnostic'or therapeutic procedures.
c) Public Exposure, which comprises all other exposures.
d) Potential Exposures, are those that have a potential to occur in
Practices and in Intervention, but with no certainty that they
will occur.
23.1 System of Protection in "Practices":
1. The Justification of Practice.
2. The Optimization of Protection.
3. The Individial Dose Limit, and Risk Limit.
The exposure of individuals resulting from the combination
of all practices should be subject to dose limits, or to some control
of risk in case of potential exposure.
2.3.2 System of Protection in "Intervention":
This is based on the following criteria: - 32 - GOMAA & NAGGAR
The proposed intervention sould aim at the reduction of detriment resulting from process of Dose reduction. This should justify the harm and cost of intervention. The form, scale and duration of intervention should be optimized to give maximal reduction of detriment. Dose Limits do not apply in case of intervention. Intervention is justified to avoid the occurrence of Deterministic Effects. 2.4 The Control of Occuptional Exposure:
2.4.1 Dose Constraints: This constitutes an important feature of Optimization. This depends on the source-related values of individual dose for deciding the form of Optimization. The type of occupation dictates the levels of individual doses likely to be incurred in a well managed operation. This information is used to establish dose constraint for that type of occupation. 2.42 Dose Limits: The ICRP aims to establish a "Dose limit" for a defined set of practices, for regular and continued exposure, above which the consequences for the individual would widely be regarded as unacceptable. Previously, the commission used the attributable probability of death or severe hereditary disorders as the criteria for judging the consequences of exposure. Although, this quantity remains a major factor, however, it is not regarded by the Commission as 2nd RPC (1991) -33-
sufficient to describe the detriment.
The Commission recommends a limit on Effective Dose of
20m Sv per year, averaged over 5 years (100 mSv in 5 years); with
further provision that. the Effective dose should not exceed
60mSv in any single year. The dose limits for occupational
exposures are shown in table (2).
Table (2): ICRP Publication 60 RECOMMENDED DOSE LIMITS W
Dose Limits Application Workers Public Occupational Exposure
Effective Dose 20 mSv per year 1 mSv per averaged over 5 years year(3) defined period (2)
Annual equivalent Dose: • Lens of the Eye. 150 mSv 15 mSv • Skin (4). 500 mSv (5) 50 mSv • Hands and Feet. 500 mSv none
1) These limits apply to doses from External Exposures during the
specific period; and the 50-year committed dose from intakes
during same period (to age of 70 years for children).
2) The effective dose should not exceed 50 mSv in any single year.
Additional restrictions apply to occupational exposure of
pregnant women.
3) In special circumstances a higher value of effective dose could
be allowed in single year, provided that the average over 5 - 34 - GOMAA & NAGGAR
years does not exceed 1m Sv per year. 4) The limitation on Effective Dose provides protection for the skin against Stochastic Effects. However, an additional limit is required for Localised Exposures in order to prevent Deterministic Effects. 5) This is averaged over 1 cm2 regardless of the area exposed. * For Internal Exposure, the Annual Limits of Intake will be based on a Committed Effective Dose of 20mSv. The estimated intakes may be averaged over 5 years. The occupational limits for Radon are under review, meanwhile, the values given in ICRP publication 47 (1986) (remain valid. 2.4.3 Occupational Exposure of Women: Once pregnancy is declared, the Equivalent Dose Limit to surface of Abdomen is 2 mSv for the remainder of pregnancy. Also limiting intakes of radionuclides to 1 /20 (0.05) of the Annual Limit on Intake. The use of source-related dose constraints will provide adequate complicance with this limit. Employment of women should be of a type that does not carry a high probability of high accidental doses and intakes. 2.5 Control of Medical Exposure: Simple routine diagnostic procedures based on common indication do not require case by case justification. However, this 2nd RPC (1991) "35'
case by case justification may be required for complex investigations, and some therapeutic procedures. Optimization of protection in diagnostic radiology offers a scope for dose reduction. Dose Constraints should be considered when medical procedures are not of direct value to exposed individual as in scientific studies involving volunteers. The ICRP recommends that Dose Limits should not be applied to medical exposures, on the grounds that medical exposures are intended to provide direct benefit to individuals. However, the dose to the patients should be as low as is compatible with medical purposes. Also, the practice, should be justified, protection optimized, and both machines and procedures should be of high quality performance to avoid unnecessay exposures.
Diagnostic and therapeutic procedures causing exposures of abdomen of women (likely to be pregnant) should be avoided unless it is strongly indicated. Information on possible pregnancy is obtained from patient herself. If the most recent expected menstruation has been missed, and there is no other relevant information, the woman is assumed pregnant. 2.6 Control of Public Exposure:
In normal situations, public exposure is exercised by the application of controls at the source rather than in the environment. These are done by procedures of constrained - 36 - GOMAA & NAGGAR
optimization and the use of prescriptive limits. The "critical group" are individuals subject to similar exposures from a single source, and who are typical of the most highly exposed by that source.
2.6.1 Dose Limits: The scope of Dose Limits for Public Exposure is restricted to doses incurred as the result of practices. Doses incurred in situationa where "Intervention" is the only available protective action, are excluded. Radon in dwellings and open air, natural or artificial radioactive material in environment and other natural sources, are examples of situations influenced by Intervention. Doses from these sources are outside the scope of dose limits for public exposure. The commission recommends a Limit on Effective Dose of 1 mSv per year for Public Exposure. In special circumstances, a higher value could be allowed in a single year, provided that the average over 5 years does not exceed 1 mSv per year. The commission recommends annual limits of 15 mSv for the lens of eye; and 50 mSv per year for the skin averaged over any 1 Cm2 regardless of the area exposed. These limits aim to prevent deterministic effects to these organs, which will not necessarily be avoided by the limit on the effective dose. The dose limits for Public Exposure are given in table (2). 2nd RPC (1991)
2.7 Other Aspects of ICRP Publication 60: The main text of the publication deals with further issues regarding the following aspects: 1. Potential Exposures. 2. The System of Protection in Intervention. 3. Radon in Dwellings. 4. Intervention after accidents. 5. Practical implementations of Commission's Recommendations. Importance for operational level of radiation protection, the designation of controlled and supervised areas to be decided by operating management. The classification of working conditions based upon expected dose is no longer recommended. Dose measurement, medical surveillance and emergancy planning are also considered. 6. Bases for judging significance of the effects of radiations. 7. Quantities used in radiological protection.
3. APPRAISAL ON IMPLEMENTATION: ICRP publication 60 is a substantial document because it contains great deal of explanatory information. It is strongly recommended that those concerned with Radiation Protection should become familiar with the main features of this document. The basic recommendations are straight forward and continue to emphasize that Dose Limits are not to be viewed as - 38 - GOMAA & NAGGAR
the primary consideration in Radiation Protection, but rather, to provide a "clearly defined boundary" for the process of ALARA (ie. keeping doses as low as reasonably achievable); with social, and economic factors taken into account. It is the ALARA process that should determine the doses to which both workers and public are exposed. The Committee recommends an Occupational Limit of an average of 20 mSv Effective Dose per year over five years (100 mSv in five years) to an individual, with the further provision that the dose should not exceed 50 mSv Effective dose in any one year, with the five year period defined by the regulatory authority. The Committee recommends a Dose Limit for Exposure of Public of 1 mSv Effective Dose per year, which may be averaged over five years only in special circumstances. The Committee recommends a Dose Limit for Pregnant Women workers with Atomic Radiations of 2 mSv to the surface of the Abdomen, and 0-05 (1/20) Annual Limit on Intake for radionuclides, during the remainder of pregnancy once the pregnancy has been declared. This is assumed to give lmSv to faetus. The adoption of the ICRP recommendation of 1 mSv per year for the Public exposure, will designate new definition for the Atomic Radiation Worker, who will receive more than 1 mSv per 2nd RPC (1991) "39"
year. The Regulatory Authority will classify such workers to be subject to Radiation Protection measures, personal monitoring, medical surveillance, and the Occupational Effective Dose Limits. There is however, a pressing need to revise the definition of an Atomic Radiation Worker in the light of the new ICRP recommendations.
The second implication consequent on the new ICRP recommendations is the operating levels for emissions from nuclear facilities. These have to be reassessed by applying the ALARA principle using the detriment associated with the new risk factors. The devolopment of radiation protection programme, entails that the ALARA principle must be applied. All occuptional doses above 20 mSv per year must be brought below that limit regardless of cost. Once this is achieved, the ALARA process must be applied to reduce the dose further. Since the Risk per Sievert is now considered to be greater that its former value, the Detriment against which expenditure on protection is judged in the application of ALARA will likwise increase. All previous ALARA analysis must be reviewed to take these changes into account.
The ICRP based its recommendations for the public dose limit on a comparison with natural background rates which are - 40 - GOMAA & NAGGAR
generally about 1-2 mSv per year including doses from naturally occurring radon daughters. A projection of lifetime risk of fatal cancer was also considered which is about 1:10000 for 1 mSv per year for 70 years. The ratio of new limit to old is 1/5 for public compared or 2/5 for workers. All radiation facilities must revise their ALARA process for occupational exposures, and implement adjustments to on-going operations. New facilities must be designed to comply with 20 mSv per year occupational dose limit. 3.1 Classification of Workplaces and Working Conditions: Control of radiation sources requires formal designation of workplaces containing these sources. The Commission uses the following criteria: Controlled Area is one in which normal working conditions require workers to follow well-established procedures and practices aimed specifically at controlling radiation exposures. The occupational dose limit of 20 mSv per year is implemented, and the ALARA practices are introduced to reduce the dose further. Supervised Area is one in which working conditions are kept under review, but where special procedures are not normally needed. Doses to workers in the supervised areas, should be confidently less that 3/10 of the occupational dose limit. 2nd RPC (1991) - 41 -
Designated Area is defind in terms of dose rates at the boundary. The aim should be to ensure anyone outside the designated area will not be regarded as occupationally exposed. The actual doses received outside the designated areas are kept below the dose limits for public exposure. The designation of controlled and supervised areas should be decided either at the design stage, or locally by the operating management, on the basis of operational experience and judgement. The classification of working conditions based upon expected dose exposures is no longer recommended. 3.2 Source-Related Constraints: The is an important feature in the system of radiation protection. Constraints should be applied at source in form of suitable shielding to comply with the prescribed dose limits. This will necessitate the computation of added shielding thickness for the currently operated radiation facilities, and primary shielding for new facilities. These shielding computations are performed to meet with the boundary conditions prescribed by the Commission for workers in controlled and supervised areas.
3.2.2 Dose Reduction Factors for Existing Facilities: For design puposes, the existing radiation facilities were designed so that the primary X-ray beam, gamma ray beam or neutron shielding for boundary condition was limited at 50 mSv - 42 - GOMAA & NAGGAR
per year, according to ICRP (publication 26). According to the 1990 ICRP (publication 60) recommendations, the design objective is 20 mSv per year. In order to implement the new recommendations, the existing radiation facilities require additional shielding. The calculated Dose Reduction factors are calculated accordingly. DRF-1= recommended annual dose limit (1990) annual dose limit (1976). = 20 mSv V/50 m Sv. = 0.4 for workers (1) A second Dose Reduction Factor (DRF-2) for protection of the members of the public is- given by: DRF-1= annual dose limit (1990) annual dose limit (1976) = 1 mSv/5mSv. = 0.2 for public (2) Relations (1) & (2) can be rewritten in terms of half value thickness and the additional shield thickness X-,, X~ are related with Half value thickness by the following relations:
X1 (DRF-1) = 1.32 HVT.
X2 (DRF-2) =2.32 HVT. where HVT = one half value thickness. In Table 3 the calculated additional shield thickness for X-ray sources are tabulated. The numerical values of HVT for lead (Pb), and concrete were taken from ICRP publication-33. 2nd RFC (1991) "43"
In Table 4, the added lead and concrete thickness for gamma radiation facilities to meet boundary condition of 20 mSv per year is presented. The half value thicknesses for Ir-192, Cs-137, Co-60 and Ra-226 gamma rays were taken from ICRP publication-33. For (Americium-Beryllium) fast neutrons, the required thicknesses of water, paraffin, iron, concrete and graphite to meet the new ICRP boundary conditions of 20 mSv per year were calculated. Table 5 gives additional shield thicknesses required. The half value thickness values for different shielding materials were taken from Turner (1986). 3.2.2 Primary Radiation Shielding for New Radiation Facilities:
The ICRP-piblication No 33, the primary radiation shielding to meet boundary condition of 50 mSv per year are given. In the Egyptian regulation, ministry of health order No. 630, 1962, primary radiation shielding for boundition of 50 mSv per year are also included. The results of calculation of thickness required to reduce exposure to workers to 20 mSv per year are giving in Table 6 for X-ray sources (50 kV, 100 kV and 150 kV). Table 7 gives the calculation of thicknesses for Cs-137 gamma rays. Table 8 gives the calculation of thicknesses for Co-60 gamma rays.
Future work is aimed to extend these calculations to other X-rays sources, gamma ray sources and neutron sources, in order - 44 - GOMAA & NACGAR
to upgrade the Egyptian radiaction facilities according to ICRP publication-60.
REFERENCES:
1. Republican Order No. 59,1960, Egypt.
2. Ministry of Health Order No. 630,1962, Egypt.
3. ICRP publication-26, 1977.
4. IAEA, Safety Series No. 9,1982.
5. Atomic Energy Authority Order No. 413,1986, Egypt.
6. Ministry of Health Order No. 265,1989, Egypt.
7. ICRP publication 33,1982.
8. ICRP publication 60,1991.
9. J.E. Turner, Atoms, Radiation and Radiation Protection,
Pergamon Press, 1986. 2nd RPC (1991) -45-
Table (3): Added Shield Thickness for X-ray facilities for Boundary Condition of 20 mSv per year.
Additional Shield Thickness in Cm
X-ray Xt (DRF-1) X2 (DRF-2) Source Fb Cone. Fb Cone.
50 KV 0.007 0.53 0.012 0.93 100 0.033 2.11 0.058 3.71 150 0.038 2.90 0.067 5.10 200 0.055 3.43 0.097 6.03 250 0.11 3.70 0.20 6.50 300 0.22 4.00 0.35 6.96 400 0.33 4.00 0.58 6.96 1 M.V 1.00 6.07 1.76 10.67 2 1.52 ' 8.05 2.67 14.15 3 9.11 16.00 4 1.95 11.1 3.43 19.5 6 2.03 13.5 3.66 23.66 10 2.23 15.44 3.92 27.14
Table (4): Added Shield Thickness for Gamma Ray facilities in Use for Boundary Condition of 20 mSv per year.
Additional Shield Thickness in Cm
Application Xj (DRF-1) X2 (DRF-2) Fb . Cone. Fb Cone.
Ir-192 0.792 5.41 1.39 9.51 Cs-137 0.924 6.47 1.62 11.37 Co-60 1.58 8.08 2.78 14.15 Ra-226 1.71 9.24 3.01 16.24 -46- GOMAA & NAGGAR
Table (5): Added Shield Thickness for Fast Neutron Sources in use to meet Boundary Condition of 20 mSv per year.
Additional Shield Thickness in Cm Application Xt (DRF-1) X2 (DRF-2)
Water 8.9 15.6 Paraffin 8.6 15.2 Iron 5.8 10.2 Concrete 10.3 18.1 Graphite 11.7 20.5
Table (6): Primary radiation Shielding to meet Boundary Condition of 20 mSv per year.
X-ray Work Load Lead thickness in cm Lead thickness in tcm Source mA. minlw lm 2m 4m 8m lm 2m 4m 8m
50 KV 1000 0.033 0.033 0.025 0.016 4.16 3.3 2.5 1.6 250 0.033 0.025 0.016 0.008 3.3 2.5 1.6 0.8 60 0.025 0.016 0.008 — 2.5 1.6 0.8 — 15 0.016 0.008 — — 1.6 0.8 — — 4 0.008 — — — 0.8 — — —
100 KV 4000 0.330 0.267 0.212 0.161 23 18 16 13 1000 0.267 0.212 0.161 0.110 18 16 13 8.5 250 0.212 0.161 0.110 0.068 16 13 8.5 6.0 60 0.161 0.110 0.068 0.042 13 8.5 6.0 2.5 15 0.110 • 0.068 0.042 — 8.5 6.0 2.5 1.7
150 KV 4000 4.0 3.5 2.75 2.29 38 33 28 23 1000 3.5 2.75 2.29 1.61 33 28 23 17.5 250 2.75 2.29 1.61 1.14 28 23 17.5 12.5 60 2.29 1.61 1.14 0.68 23 17.5 12.5 7.5 15 1.61 1.14 0.068 0.40 17.5 12,5 7.5 3 2nd RPC (1991) -47-
Table (7): Primary Radiation Shield Thickness for Gamma Ray Facilities in use to Meet Boundary Condition of 20 mSv Per year.
Work Load Concrete Thickness in Cm Application Gylweek atlm lm 2w 4m 8m
Cs-137 240 100 90 81 68 120 94 84 73 63 60 90 81 68 60 30 84 73 63 55 15 81 68 60 50 7.5 73 63 55 45
Table (8): Primary Radiation Shield Thickness for Gamma Ray Facilities to Meet Boundary Condition of 20 mSv Per yaer.
Work Load Concrete Thickness in cm Application Gylweek atlm lm 2m 4m 8m
CO-60 240 125 112 102 88 120 118 107 94 82 60 112 102 88 76 30 107 94 82 70 15 102 88 76 64 7.5 94 82 76 58 - 48 - GOMAA & NAGGAR
Appendix. 1 TISSUE WEIGHTING FACTORS* * This represents the distribution of Aggregate Detriment among tissues.
Organ or Tissue Weighting Factor (W
Gonads 0.20 Lungs 0.12 Red bone marrow 0.12 Stomach 0.12 Colon 0.12 Skin 0.01 Bone surfaces 0.01 Thyroid 0.05 Liver 0.05 Esophagus 0.05 Breast 0.05 Bladder 0.05 Remainder*
RADIATION WEIGHTING FACTORS*
Type and Energy range** Radiation Weighting Factor (Wfc) Photons, all energies - \ Electrons and muons, all energies*** 1 NEUTRONS, ENERGY < 10 KeV 5 10 KeV TO 100 KeV 10 > 100 KeV TO 2 MeV 20 > 2 MeV TO 20 MeV io > 20 MeV 5 Protons, other than recoil protons, energy 5 >2MeV Alpha particles, fission fragments, heavy nuclei 20
* All values relate to the radiation incident on the body or, for internal sources, emitted from the source. ** The choice of values for other radiations is discussed in Annex A of ICRP 60. *** Excluding Auger electrons emitted from nuclei bound to DNA. 2nd Rad. Prot. Conf. (1991), Paper No. 3
THERMOLUMINESCENCE PROPERTIES OF BONE EQUIVALENT CALCIUM PHOSPHATE CERAMICS B. A. Henaish#, A. M. El-Agramy# and W. I. Abdel-Fattah* #) Atomic Energy Authority, Cairo, Egypt *) national Research Center, Dokky, Cairo, Egypt.
ABSTRACT: Characteristic TL - glow curves of both natural and synthetic
Ca,(PO4>2 powder have been investigated. Synthetic TL-samples, particularly those which have Ca/P ratio = 2.0, indicated improved glow curves with multi-peak spectra (at 75"C, 175*C and 300"CD and higher TL-sensitivity; about 73 nC/Gy for 3 mg powder sample.
All investigated Cao(P04)2 samples showed nearly regular systematic dependence in their Y-induced Tl-response throughout a useful dose range, from about 1 Gy to about 10 KGy, which is proper for therapeutic And processing levels.
The proper heat annealing course for regeneration of irradiated TL-samples was found to be about 350 *C for 20 minutes. INTRODUCTION: The general requirements for phosphors useful for thermoluminescence dosimetry (TLD) have been summarized by Shulman (1). These are (1) sensitivity, (2) long term information - 50 - Henaish et al.,
storage, (3) simple glow curves, (4) thermoluminescence spectrum of short wavelength, (5) stability of traps, (6) activators and host lattice. Previous study on Egyptian cement as naturally occurring material sowed a common simple glow peak from 180°C to 188*C for the different types of investigated cement (2). These glow peaks was attributed to the release of electrons from Mn and Al-centers as foreign atoms inside the TL-powder. The natural TL-materials such as ieldspar, barite, quartz, fluorite and sand have been examined by a number o searchers (3-8).
This study is aimed at developing a Ca-CPO^ky-ray sensitive TL-dosimeter for y-ray measurements in reactor core/y-cells, radiation sterelization, food processing plants, etc. The material chosen for this study is Ca-SPO^^ which is abundant in nature and is available at low cost. This TL-compound, in addition to its use as TL-dosimeter, can also be used in depth dose measurements since it is nearly bone equivalent for interaction with ionizing radiation.
TL-properties of natural and synthetic Ca3(PO4)2 prepared with different values of Ca/P ratio are examined, a) Sample Preparation: Four tricalcium phosphate ceramic batches were prepared - 51 - Henaish et al.,
from chemical compounds dicalcium phosphate dihydrate as well as from natural bone ash as a source of calcium and phosphorous. Chemical compounds were used to prepare the synthetic samples while the bone ash fore* natural. Various ca/P ratios.were achieved 1.6,1.8, and 2.0 to simulate those in different body organs, mature ceramic bodies were produced at 1250°c -1300"C. b) Apparatus: The TL-powder samples, after proper annealing, were distributed into labeled polyethylene capsules and irradiated by y-radiation from Co-y-cell of dose rate 150 Gy/h. The irradiated samples were handled carefully and protected from light during all stages and then stored in dark at room temperature. The Tl-intensities of the irradiated samples, after 24 hour post-irradiation storage, were evaluated by a Harshaw-2080 TL-picoprocessor using 3 mg of each powder, each experimental point given in this work is the mean value of 5 TL-readings.
RESULTS AND DISCUSSIONS Tricalcium phosphate ceramics did not give detectable TL-response without irradiation. This may indicate that the electron traps which give rise to TL-light do not preexist but are formed during irradiation (9). - 52 - Henaish et al.,
Glow curves of natural and synthetic y-irradiated C^^ can be shown in Figs. 1 & 2. From Fig. 1, it could be observed that the natural CaJPCL^ has two glow peaks at about 75*C (P^) and about 300 °C (P2). It is clear from thfe figure that P2 exhibits a relatively higher intensity and is the main glow peak for this natural compound (sample B). On the other hand, glow curves of synthetic Ca-^PO^, similar to natural (Ca/P = 60), are also sown in Fig. 1. It reflects three glow peaks at about 70, 150 and 300*C The last peak (P^) is seemed, for the first instance, to be higher than P- and r2- This peak (at 300°C), like that of natural, still dominates the other. In addition. Fig. 1 reflects same results for synthetic Ca3(PO4>2 with excess Ca; Ca/P = 80. It is clear from that curve that introduction of more Ca into the synthetic calcium phosphate, to increase the ratio Ca/P from 60 to 80, resulted in a slight increase in the TL-intensity of the main glow peak accompanied by marked decrease in p_ and P~ without any shift in peak position. The formation of the glow peaks is related to electron and hole trapping levels which rrrrained non assigned to any actual model. The latter is only attributed to electron and hole centers (7). Fig. 2 shows the characteristic glow curves of synthetic Ca,
)2 with Ca/P ratio = 20. According to this figure, a - 53 - Henaish et at,
remarkable enhancement in the TL-intensity of the three peaks is observed with pronounced peak shift towards higher temperature of about + 25 °C. It is also clear that P3 (at 300*C) is no longer the main glow peak as before and P2 (at 175*C) dominates and takes this turn. The change of Ca/P ratio from 80 to 20 may indicate that introduction of CaC>2 leads to creation of additional hole centers in the structure of the synthetic Ca Po^ which acts as TL-quencher. The relation between y- induced Tl-signal and absorbed irradiation dose is represented in Fig. 3. It is clear from this figure that both natural and synthetic CagfPO,^ exhibits regular dose response curve over a dose range from about 8C Gy to about 4 kGy. On the other hand, synthetic samples with less Ca/P ratio than natural showed enhanced sensitivity in TL-signal. The experimental TL-intensity shown in Fig. 3 could be theoretically approximated by a power dose-dependence function as follows:
1) TL (nC) = 7.04592 (DO820172) for sample B 2) TL (nC) = 40.4423.(D°51959) for sample 80 3) TL (nC) = 6.28576(D°975788) for sample 60 4) TL (nC) = 167.214 (D°84656)for sample20 Where D is the y-absorbed dose in Gy. - 54 - Henaish et al..
Al last, an experiment for heat annealing of irradiated calcium phosphate, sample-20, was carried out to examine and recognize the proper course for regeneration. The chosen annealing course depends on two variable parameters; heating temperature (200*C, 250 "C, 300°C and 350"C) and heating time (5, 10, 15, 20, 25 minutes). The results of this experiment are shown in Fig. 4 as relative TL-intensity against heating time. It is clear from this figure that the proper annealing course for synthetic caicium pnosph^.e-ly ;.- .2^°^ ' - ?n minutes. According to the above results, it is clear mat CaglPOJ^ is a paramount TL-compound. In comparison, they possesses a multi-peak glow curves of deep TL-trap centers and nave relative higher sensitivities which may be extended down to about 1 Gy for sample-20.
CONCLUSION: From the present results, it could be concluded that both natural and synthetic Ca^CPO^ powder possesses useful Tl-properties which may be utilized in radiation dosimetry. The synthetic compound, unlike natural, is characterized by multi-peak glow curves which greatly enhanced TL-intensity which may increase the possibility of using this compound {Ca,
(PO4)2) as TL-detector. Irradiation to various y-doses from Co showed regular - 55 - Henaish et al..
y-induced TL-response throughout a useful dose range proper for processing level. However, this dose range may be extended down to about 1 Gy sensitivity for sample 20, and hence, broaden the scope of utilization of this compound.
The proper heat annealing course for sample regeneration was found to be 350 °C for 20 minutes.
REFERENCES:
1. J. H. Shulman, Luminescence Dosimetry (Edited by F. H. Attix),
part 1, 3, USAEC (1967).
2. A. M. El-Agramy, M. A. El-Kolaly and H. S. hafez, Proc. 3rd
Conf. on Optical Spectroscopy and Laser, National
Research Center, Cairo, October 1990.
3. D. J. Me Dougall, Thermoluminescence of geological materials,
London: Academic Press, 1968, PP. 527.
4. Mejdah, Dosimetry Techniques in Thermoluminescence
Dating, Riso Reports 261 (1972).
5. A. M. Eid, M. I. El-Gohary, S.M. Kamal, Radiat. Physic. Chem.
26 (1985) 663.
6. S. K. Youssef, B. A. Henaish and M. A. Adawy, Proc. of 1st
Egyptian British Conf. on Biopys., Cairo (1987) 265.
7. B. A. Henaish, S. K. Youssef and M. A. El-Kolaly, Proc. 1st
Egyptian British Conf. On Biophys., Cairo (1987)
303. - 56 - Henaish et al..
8. S. K. Youssef, B. A. Henaish and A.M. Adawy, Proc. 1st Egyptian
British Conf. on Biophys., Cairo (1987) 275.
9. C. Bettinali, V. Goltardi, G. Ferraresso, Interaction of Radiation
with Solids, Plenum Press, (1976) 281. -57- Henaish et al..
30
q_o opp Synthetic ((50) ooooo Synthetic (00) U_LJ_+ Natural (D)
100 200 300 400 500 Tray Tempera lure ( C)
Fig. (1): Cfiarac t eris lie TL -g lovu curves of natural and synthetic calcium phosphate (samples —B , 60, 80) after irradiation to 100 Gy gamma-dose. -58- Henaish et al.,
800
0 0 100 400 500 Tray Temperature ( C)
Fig.(2): Characieristic TL—glotv curve of synthetic catciurn phosphate (samptes —20) after irradiation to 100 Gy gamma. -59- Henaish et at..
D ^ D - ooooo Caf,(PO4)2(B-natural) DDDDD 20-Synthetic) 10 5 - 60-synthetic) ^ 8- ***** 80-svnthetic) ^ 6 -
4 - o 2-
/ O
8:
D S 4 0)
2.
10 3 T 8- 6- -^^^€T^ O *-/
2-
to i 1 i i i i i 8 2 4 B 8 2 4 S 8 10 10 Absorbed Dose (Gy)
Fig.(3) Relation between garnrna— induced TL—intensity of calcium phosphate arid absorbed dose. -60- Henaish et ah,
ooooo 200°C ODP_D 2f>0°C 300°C 350°C
G
30 Healing Time (min)
Fig. (4): Relative TL-intensity of sample —20 as a function of annealing time and temperature (test gamma-dose 100 Gy. 2nd Rad. Prot. Conf. (1991), Paper No. 4
POTASSIUM BODY BURDENS IN OCCUPATIONAL USERS OF EGYPTIAN NUCLEAR RESEARCH CENTER
M.S. Abdel-Wahab
Physics Department, Faculty of Science, Ain Shams
University, Cairo, Egypt.
And
S.K. Youssef, E.M. Aly and magda T.A.
Radiation Protection department, Nuclear Research Center,
Atomic Energy Est., Cairo, Egypt.
ABSTRACR:
A simple calibration procedure for Inshas whole body counter for evaluating total body potassium have been adopted.
More than 120 Egyptian employee in the unclear research center
(N.R.C.) were studied for their total body potassium (TBK). The potassium values were found to have an average of 2.85 rug K kg body weight for males and 2.62 mg K kg" for females, which is higher than the recommended value given for reference man by ICRP. The TBK varied directly with body build index and is sligtly sex dependent. - 62 - Abdel-Wahab et al.,
1. Introduction:
Natural K is a mixture of three unclides K, K and K in the percentages of 93.08, 0.0118 and 6.91, respectively. It is of wide spread presentation in the environment and the intercellular
composition of food chain as well as in body burden of
humanbeing (Mitra 1989). The body cell mass has been recognised
as the core metabolicaliy active portion of the body (Moore 1963)
and is that portion that consists of all hydrated cells. Since
potassium is primarily intercellular, its determination would give
a quantitative estimate of the body cell mass, provided its
intercellular concentration is known. A typical 70 kg man 40 contains about 140 gm K with approximately 3710 Bq of K
(Cheo-Yeh Lan 1989). As ONLY 11% of the total disintegration
result in the emission of 1.46 MeV y-photons, a typical adult emits
about 2.4 X 10 Y-photons per minute. Since potassium is
primarily intracellular, the measurement of these y-photons by An external counting provide, a method of total body K and K
estimation. The difficulties inherent in accurately measuring the
potassium content of the human body have been reported earlies
(Miller 1963) and are generally related to the whole body counting
system calibration factors. This factors are mainly depend on the
internal absorption and scattering of gamma ryas in subjects of
different body size and shape as compared to calibration standards. 2»dRPC(1991) "63"
It depend also on the geometrical dimensions of the subject for which whole-body counting is performed. However, many authors (Mitra 1989, Boddy 1971, hawkins 1976 and Hughes 1967) has been assessed the sensitivity of whole body measuring system for TBK using a commited reference man calibration phantom regardless the variance of the different subjects geometrical sizes from that of the fitted standard man phantom. Metra (1989) presented a detailed study on the calibration of a shallow shield whole body counter using two large detectors (up and down) with mobile scanning bed for potassium. In the present article we followed a quite different technique for estimatings K in subjects. Predection of potassium calibration sensetivity using stationary scanning chair geometry is concerned. The calibration procedure is also covered the deviation of the body size geometr from the committed reference man. Further, measurements of the TBK level in some selected example of Nuclear Research Center (N.R.C.) occupational users and nonusers will be discussed. 2. Experimental Procedure:
The whole-body counter used for measurements consists of nal (TL) crystal (20 cm diameter X 10 cm thick). The wholebody scanning system usd is the stationary chair geometry. The equipment was designed at the protection department of nuclear research Center (N.R.C.) at Inshassm Cairo, Egypt. The detector is - 64 - Abdel-Wahab et al,
ixed along the interior roof of a pre-1945 steel room of 240 an long X 240 cm width X 195 an high. The inside wall of the steel room is also bounded with 0.1 mm thick special sheets of low activity Pb shield in order to reduce the side effects of environmental natural background on the detector. Figure 1 shows the general arrangement of the electronics. The four photomultipliers are supplied from a single igh voltage power unit and the output from each is fed to two dual spectroscopic amplifiers. The output of these pulses is further mixed using a four channel rotary mixture. The single output pulse is feded through a gate amplifier to 8 K channel pulse hight analyzer of full option computer system. The net count are used for the determination of total body potassium (TBK). To reduce the variable 40K background, a similar human being phantom filled with distilled water was used before and after subject measurement for one hour. The daily average counting of the bacjground level in the interested region of potassium is found, to be about 3000 ± 55 count hr" . The electronic stability of the counting system were daily checked by using reference sources of K, Co and Cs, and was found to be less than 3.4% for potassium and 1-2% for the other sources.
The whole body counting system was calibrated using aqueous solution contained different concentrations of KC1. 2nd RPC (1991) •65"
Several polyethylene bottles of identical dimensions were used for this purpose and were stacked close to each other to form a configuration shape of a commited reference man assembly phantom. A correction is also made for the changing of the phantom size. The calibration sensitivity of the whole body counter can be predicted using the following simple formula X = AB Where: A represent the potassium activity in count/gm for the commited reference man assembly phanto and B is a correctio factor for the deviation of the body size from the fitted standard reference man size. 3. Resits and Discussion: Figure 2 presents a linear plot of the measured counting activity of K against concentration for commited reference man phantom. The fitted line is a least square fit to the data using the formula. y = 31.799 X-0.1307 Chr"1 The claculated average count per gram (A) is equal to 31.677 ± 2.5 C gm"1. The calibration scans were performed in the same way as those adopted for subjects regarding the body size. Furthermore, the style of the Egyptian subjects have the following specifications: - 66 - Abdel-Wahab et al.,
width 922-40 cm), anterior posterior thickness 99-25 cm) and hight (145-190 cm). The present whole body measurements using the chair geometry is only scanned the most condensed part of the human being (cover a volume extended from the neck up to knees), hence forth, a further correction factor has been made for the deviation of the body size volume from the commited by using a recommended index volume procedure for the most condensed part. Figure 3 shows the variation of the relative correction factor with body volume ;:;u=r.-- I1 can be seen that the observed relative correction factor have an exponential form which can be predicted using the formula B = 1.8377 e-016299X as indicated by the solid exponential line. The minimum detected limit of the whole body counting system for K was found to be in the range of 225 count for one hour wich is eqivalent to 10 gm of K = 265 Bq for one hour, beyond the measured values of background. Figure 4 shows the estimated overall experimental statistical error Q in the calibration factor X as a function of K activity in count hr" . An interlaboratory comparison of total body potassium has been made in about 120 occupational users and nonusers of the 2nd RPC (1991) • ~67'
Egyptian N.R.C. for both sexes. The potassium values for the subjects (user and nonuser) of both sexes as a function of body weight are sown in both figure 5 and table 1. The results of TBK estimation showed an average K body burden values of 2.85 gm K kg" for males and 2.62 gm K kg" for female wile for nonuser subjects was found to be 2.69 gm k kg" for male and 2.38 gm K kg" for female. This is significantly higher than the value of 2.0 gm given for reference man (ICRP 1974). The solid line represent the average fitted potassium content in male and female subjects which can be predicted using the emperical formula y = 1.748 X +67.4 gm In figure 6, the estimated values of total body potassium are plotted against (W/H)1/2, where W is the weight (kg) of the subject and H is the hight (m). The parameter (W/H)1/2 was chosen because it represents the average body thickness and thus has bee.; accepted as a body-build index (Gupta 1976). The graph indicates that the potassium content increases with body-build index, but is slightly sex dependent. The solid lines of the fitted anomalous of the potassium content can be deduced from the following regression formula y = 11.032 X + 120 gm for male, and y = 8.504 X +128.4 gm for female Mostly, data of figures 5 and 6 showed a scattered and - 68 - Abdel-Wahab et al.,
statistical distribution of the potassium content, this may be ascribed to the nonuniformity proportion of subject hight with their correspondence of the fitted weight. However, whole body counting system using chair geometry is mainly scanning the most condensed region of human being (80% total body weights). For this reason it is reasonable to predict the relation between the fitted volume of the subject (from neck to kness) with the total body potassium activity. The results of figure 7 reflects that relation. The data sows less statistically scattered distribution in the experimentally measured TBK with the change of body volume index. The average body burden potassium can be calculated from the fitted results using the formula: y = 11.699 X10-4 + 131.7 gm/cc for male and y = 7.798 X10-4 + 141.71 gm/cc for female The experimentally deduced activity of TBK according to figures 5,6,7 is relatively higher in comparison with the recommended value by ICRP, i.e. in our experimental finding is found to be averaged 2.65 g K kg" while the ICRP recommended value is about 2.0 K kg" . This may be attributed to the style of the feeded food by Egyptian, which may containes different orders of potassium activity. Prediction of the potassium in human being due to different food varieties are in our scope of study and will be published later. 2"d RPC (1991) -69-
Acknowledgement: The authors are greatly indepted to Prof. Dr. M. A. Gomaa, Head of Radiation Protection Department (N.R.C.) of Egyptian Atomic Energy Establisment for his valuable discussion. REFERENCES: Boddy K., King P.C., Thothill P. and Strong J.A. (1971), Measurement of total body potassium with a shadow shield whole-body counter: Calibration and errors, Phys. Med. Biol. 16, 275-282. Chao-Yek Lan (1989), Bo.dy K and 40K in Chinese subjects measured with a whole-body counter, Health Physics Vol. 57 No. 5, 743-746. Gupta M.M., Bhola G.C., and Gupta N.K. (1976), Body potassium measurement by 40K gamma ray spectrometry, Med. J. Armed Forces India 32,557-566. Hawkins T. and Goode A.W. (1976): The determination of totalbody potassium using a whole-body monitor, Phys. Med. Biol., 21,293-297. Hughes D. and Williams R.E. (1967): The calibration of a whole-body radioactivity counter for the measurement of body potassium content in clinical studies, din. Sci., 32,495-502. International Commission on Radiological Protection (1974): - 70 - Abdet-Wahab et ah,
Reference Man. Oxford, Pergamon Press, ICRP Publication 23. Miller C.E. and Remenchik A.P. (1963): Priblems involved in accurately measuring the K content of the human body, Ann. NY Acad. Sci. 110,175-88. Mitra S., Sutcliffe J.F. and Hill G.L. (1989): A simple calibration of a shadow shield counter for the measurement of total body potassium in critically ill patients, Phys. Med. Biol. Vol. 34, No. 1, 61-68. Moore F.D. and Boyden CM. (1963): Body cell mass and limits of hydration of the fat-free body: their relation to estimated skeletal weight, Ann. NY Acad Sci. 110, 62-71. 2nd xpc (2991) -71-
Table (1): Examples of measured body counting radioassay for composite sex-specific.
Employment Six Total KBody Burden Status Type Number gmKKg-1 gm KJ1000 c**
Male 2.10 - 2.82 4.69 - 5.10 >ly* 20 Female 1.89-2.69 3.96 - 4.70
Axale 2.35 - 2.99 4.86 - 4.84 l-5y 10 Female 2.12 - 2.67 4.25 - 5.06
Male 2.17-2.81 4.46 - 5.31 5-10y 8 Female 2.40 - 2.87 4.65-4.97
Male 2.41 - 2.94 4.72-5.62 10-15y 20 Female 2.29 - 2.86 4.34-4.82
Male 2.20 - 2.57 4.46-4.93 15 - 20 y 25 Female 1.98-2.68 4.37-4.72
Male 2.5 - 3.21 4.75-5.36 20 - 25 y 25 Female 2.10 - 2.60 4.57-5.12
* Non users. ** Index volume of chest abdomen and jointed part of upper legs. - 72 - Abdel-Wahab et ah,
Figure Captions
Figure 1: Schematic diagram of whole body counting system. Figure 2: K-concentration for chair geometry as a function of total whole body counts. Figure 3: Correction factor of body size. Figure 4: Rate of change of K activity as a function of statistical errors. Figure 5: Variation of TBK with body weight. Figure 6: Variation of TBK with body-build index (W/H)1/2. Figure 7: Variation of TBK with volume index of neck of neck up to kness. 2nd RFC (1991) -73-
b 0 s! 1 a s
2000
10 20 30 50 60 Concentration (gm)
rig. 121 -74- Abdel-Wahab et al,
0.5
Volume Index 1x10 )
Flq. U)
iOO 800 1200 1600 Potassium Activity (C tic I
Fig. HI 2 d " RFC (1991) -75-
k Noil - ii^tT ri 210r x M.1I0
- I-1 vuia 1 u
HI 1 1S0 3 -
170-
150 1 1 1 50 60 70 00 90 Oody Wclqllt fig. I5 210
4 Non -uscra • ^
• •rn.ilf) * S 190 - • 3
i 170 1 1 1 5 6 7 [kxly-Uullil Index Ffg. (6)
Nui -UUCT3 ^^ Male . ~^ * louule
^At^^
25 35
Kiq. (7) 2nd Rad. Prot. Conf. (1991), Paper No. 5
TRANSPORT OF RADIONUCLIDES IN FRESH WATER STREAM
W.E.Y. Abdel Malik. A.S. Ibrahim R.M.K. El Shinawy and N.H. Kamel Radiation Protection Department, NRC, AEE, Cairo, Egypt.
ABSTRACT: The physico chemical and mineralogical properties of Ismailis canal bottom sediments (ICBS) showed that it is mainly silica (major) with some montmorillonite and kaolinite (minor). Sorption of the radionuclides by the ICBS was found to follow a three phase process rapid moderate and slow phase reaction. The reaction half life of each phase was calculated for the different isotopes studied. The sorption capacity for different size fractions has been calculated and the cumulative sorption capacit of all sizes have been compared with that of the natural bottom sediment. The sorotion capacity varied with pH and increased linearly with earrier concentration according to the Freundlich type of isotherm. Measurements of the natural radioactivity level showed the presence of trace amounts of natural radionuclides in the ICBS environment. 2nd Rad. Prot. Conf. (1991), Paper No. 6
WATER TREATMENT PROCESS FOR NUCLEAR REACTORS
M. A. Marrwan M.S. Khattab A.N. Hanna AEA, Nuclear Research Center, Inshas
ABSTRACT: Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor colling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerants to obtain the optimum quantity of pure water which reached to 15 cubic metar instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30%. Output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter afficiency are also determined. 2nd Rad. Prot. Conf. (1991), Paper No. 7
RESEARCH ACTIVITIES RELATED TO INTERNAL CONTAMINATION AND BIOASSAY AND EXPERIENCE GAINED IN THIS HELD
A. El-Sharnouby, M.Aziz, A. Shaaban, and Atwa Internal Contamination and Biossay Unit, radiation Protection Department, Nuclear Research Centre
ABSTRACT: Recent nuclear technology and the wide spread use of radioisotopes mainly in the form of chemical and biological tracers lead to increasing the risk of internal contamination. It is for this reason the essessment of internal contaminotion and the chemical treatment should be stressed. There are two concepts of prime importance-Knowledge of the level of risk and Knowledge of the urgercy of treatment. The present paper deals with these points and the steps taken for adminstration of first aid to internally contaminated persons. The research activities of the unit are devoted also to serve this goal and mainly includes: • Investigation of the complexes formed by chelation of transition, lanthanide and actinide elements with separate or mixed ligands. Such in vetro studies could be applied to accelerate the excretion of these elements from the body. 2nd Rpc (wgl) - 79 -
• Application of some recent separation techniques as foam separation, gel-liquid extraction and polyethane foam separation to reconcentrate the trace elements from their solutions before determination in different media or for treatment of low level wastes. • Studing the recent trends of incorporating low fi-activity wastes in polymers. The method was applied for incorporating tritiuim wastes in epoxy resins. The leaching rate of H-3 from the epoxy-resin product was found to be between 1.2 X 10" to 1.3 X 10 7 d 1 under different experimental conditions. • Radiochemical studies involving the sorption and complex formation of different radioelements with some soils ingredient. The work aimed to investigate the possibility of complex formation of certain radionuclides (13 7Cs , 60Co , 99Tc) with humic acid. These radionuclides reach the envirnoment accedentaly or through normal releases from nuclear installations. • Bioassay Laboratory. The radiochemical laboratory of bioassay was supplied last year with a liquid scintillation counter for low beta energy measurments and steps are taken now to supply the laboratory with a fluorometer for natural uranium determination in urine. Prelimenary expermenets are done on urine samples to examin the efficiency of determination of -80-
gross beta activity and natural uranium. Oiher stndies involving: - Particle Track Etch Techniue in the blood of workers. - Medical aspects of radiation accidents. - Erythrocyte life span determination. - Biochemical & haematological changes in radiation accidents. 3 - Autoradiography of LNA, with H labeling. 2nd RPC (1991) 2nd Rad. Prot. Coflf.-(1991), Paper No. 8
ENVIRONMENTAL MONITORING AT INSHASS AREA BY
A. T. Abdel fattah, M.W. Essa & M.A. Gomaa
ABSTRACT: Environmental monitoring of radiation level in some food-chain samples, air-particulates and other items carried out in Inshass after 1985 indicates that at the period of Chernobyl accident an increase in the level had ben noticed exactly on the 10th day after the accident. This has been detected easily in air filters than in other samples. Air-particulate sampling outside the research reactor in Inshass together with other gamma spectroscopic analysis of soil, water and plant samples indicate that the vicinity of the reactor is withen the natural background level. Inside the reactor, a noticeable level from radon and/or thoron decay products was found when the ventilation of the reactor hall was switched off. During Arabian Gulf war (Feb. 1991) experimental results indicated no presence of artificial radionudides (such as 1-131 and Cs-134, 137) either in the air or in other environmental samples. Measured radioactivity was mainly due to radon and thoron daughters. Meteorological data at Inshass local station indicates -82- Abdel Fatah et al.,
that the prevailing wind direction was 295* (North-West) which is in agreement with that of Helwan station at high altitudes. It is not expected that this wind direction NE or SE will help in carrying radiation debris or gases to Inshass (i.e. Egypt).
INTRODUCTION
Historical Approach: Environmental monitoring is one of the essential tasks of the Radiation Protection Department since the begining of the Egyptian Atomic Energy in 1956. Environmental Monitoring Unit (EMU) is one of the main divisions of this department concerned with such work since 1960 " . Environmental monitoring in Egypt in 1960 covered six cities: cairo (or Inshass), Alexandria. Marsa Matrouh, Sewa and Aswan (Fig. 1). In the first station the monitoring includes gummed paper, ion-exchange columns, air filter and pots beside 90 Sr separtion and determination in some environmental items (water, vegetotion, meat, milk, soil and human bones). In the last four stations the monitoring includes only gummed paper and ion-exchange columns. In 1962; the monitoring net was increased to include the capital cities of all the governorates of Egypt (Fig. 1) where 2nd RPC (1991) "83"
diffirent food-chain and other environmental samples were collected and sent to the central laboratory at Inshass for radiossay. In 1964; another two stations in Sinia Nekhel and El-Kontella were added (Fig. 1) including automatic continuopus air monitors for measuring a/B/y activity of dust particle collected on movable filter tape, beside seismographs. Continuous automatic air monitors were added to.stations: Alexandria and Port Said beside that in Inshass which contains 2 monitors of different types.
In the time of two wars (1967 and 1973) and due to priority of defence and other critical national demands, the necessary funds for maintenance and renewal of the counting and other equipments were diminished to minimum. Consequently, the activity and organisation of this system were greatly affected and nearly stopped for several years. In 1984 and due to the report of the IAEA on the situation of the Egyptian reactor safety , the IAEA approved the CP project for environmental survery around the reactor (EGY/9/017). In 1986-1989 the following major equipments had been recieved from the IAEA:
• Multichannel analyzer with a HPGe detector. • Low alpha/beta/gamma counting system. • Meteorological tower with sensors and data aquisition system. • Manual air smaplers. -84- Abdel Fatah et al.,
In 1986 the accident of Chernobyl took place. The Environmental Monitoring Unit (EMU) made the necessary measurements on samples taken from Inshass, cairo, Alexandria and Port Said (food-chain, soil, water, sediment and air particulates) . According to orders of Prime Minister; to make two laboratories for radioactive testing of imported food items at the two major ports of Egypt Alexandria and Port said; the EMU supplied these two laboratories by necessary equipments (Single Channel analyzer Nuclear Interprice SR-5 with 3" Nal crystal, standard Cs-137 source, top loading balance) and perssonels. In 1988 a survery was carried out in and around the research reactor at Inshass. In 1991, an emmergency situation arised due to the war in the Arabian Gulf and the expectation of a radioactive release as a result of bombing nuclear installations and threats of using nuclear weapons in the area of the middle east. The EMU carried out many measurements including 24-hours air-particulate monitoring, soil, weter, vegetations and meteorology. In 1991, the EMU made a proposal for an IAEA CP project (1993/1994) as phase II of project (EGY/9/017) to treat the gabs and deficiencies appeared in phase I. The major equipments requisted in this proposal are: 2nd RPC (1991) "85"
• Large furnance. • HPGe detector and lead shield • Liquid scintillation counter. • Alpha detection system. The aim of this report is to summarise the results obtained after 1985 illustrating the possibility of any artificial increase in the background level. The monitored areas and dates included in this study are: • During Chernobyl accident 1986. • Inshass 1988. • Inshass 1989 (Farm of El-Enani). • Inshass during the Arabian Gulf war (1991). • Inshass (1991).
II. Experimental:
1. Sampling: Sample should be chosen to give the maximum amount of useful informations^ ' '.
a) Air Particulate Filter High volume air sampler (Stablex) with Whatman filter paper or active charrcoal filters were used, sucked air was about 0.849m min and working time was 15-30 minutes. The filter were counted after 5 minutes and after 6 hours. The counting systems used were: -86- Abdel Fatah et al,
a) Single channel analyzer (Nuclear Interprise SR5 with a 3" Nal well type crystal. The window was adjusted once for 137Cs (E = 0.66 MeV) and once for 131I (E = 0.36 MeV). b) Low background alpha/beta/gamma phoswich detector (Tennelec LB 1000) and c) Multichannel analyzer with PC and HPGe detector, b) Other Samples: Other environmental samples such as water, plants and soil, were counted either fresh by gamma spectrometer or after ashing at 450°C by the low alpha/beta/gamma counter. 2. Counter Calibration: Gross beta counter (either GM or Phoswich detectors) were calibrated using the natural potossium as KC1 or K-CO,. The natural potassium contains K-40 in the ratio of 0.0118%. Different weights of potassium salt were counted and the resulted efficiencies were drawn against these weights (Fig. 2). The HPGe detector (or the Nal crystal) was calibrated using a mixture of several isotopes: Co-60, Co-57, Cs-134, Ba-134 and Y-88 (Fig. 3). 3. Meterorology: The meteorological tower used (Fig. 4) had been erected with its sensors at the level of 15 meters according to the instruction of ths IAEA visiting expert Dr. Robej. The included sensrs of this 2nd RPC (1991) . "87'
system are of wind speed, wind direction, temperature, relative humidity, a tmospheric, pressure solar radiation and rain. Power supply of the system is through either normal electric current or battery or solar cell. The whole system: measuring, recording, reading and printing of the diffirent data is done automatically by the use of a suitable programme applied in the data aquistition unit of the system. The programme used gives a reading as a mean value for each hour of the day and another mean value for each day.
RESULTS & DISCUSSION 1. Calibration of counting systems: a) The calibration of the Low Beta counter for gross beta activity by the use of KCL and K-CO- is seen in (Fig2>-
The counter efficiency is about 26-29% for K2CO3 and KCL respectively at a weight of 1 gm each in a planchet of 5 cm diameter. b) The (HPGe) detector calibration curve is seen in (Fig 3). The media of the isoteopic mixture are i-filter paper 10.5 cm diameter representing the filter paper of the high volume air sampler, ii-solution, iii-sand. 2. Measurements during Chernobyl accident (1986): Results of the air activity concentrations are seen in figure (5). This figure indicates the presence of high activity -88- Abdel Fatah et al.,
concentration in air of about 18.5 Bqm at the 10th day of the accident (5 May 1986) which then decreases to a value of about 0.5 Bq.m . Rain fall, occurred in Cairo on may 12,1986 and an effect of radioactivity washing in air is observed. For comparison of the air activity concentration in Egypt on May 5, 1986 to those, e.g. in. FRG, it is to be noted that: on April 30, 1986 it was 100 Bq.m in Karlsuhe and 130 Bq.m"3 in Munich(8). Gross gamma activities of different environmental samples (soil, water, vegetable and milk) at three different sampling sites (Cairo, Alexandria and Port Said) are seen in the table (1). The
Table (1): The gross gamma activity in soil, water vegetables and milk in Cairo, Alexandria and Port Said in Bq. Kg . May 1987 May 1988 Station May 1986 (mean volue) (mean value)
Soil Cairo . 5.57 ±1.3 3.3 ±0.9 2.09 ±0.7 Alexandria 3.59± 1.0 Port Said 6A ± 1.0 Water Cairo 3.0 ±1.2 1.12 ±0.03 0.98 ±0.02 Alexandria 2.15 ±1.1 Port Said 2.5 ±1.1 Vegetables Cairo 4.7 ±1.1 4.0 ± 0.12 2.6 ±0.3 Alexandria 4.1 ±1.3 Port Said 3.8 ±1.5 Milk Cairo 2.0 ± 0.5 1.7 ±0.1 1.6 ±0.3 Alexandria 1.4 ±0.5 Port Said 1.28 ±0.9
* Milk in Bq.L"1 * Cairo station also stands for inshas. 2nd RPC (iggi) -89-
values in May (1986) are not significantly higher than those in May 1987 and May 1988. Gamma spectrometric analysis indicates the presence of some trace amounts of Cs-134, Cs-137 and Co-60 in case of soil, water and vegetation with minimum values for water and maximum values for vegetables. 3. Mointoring at Inshass Research Reactor Site (1988): High Volume Air Sapling: All samples taken outside the reactor building, at the reactor entrance and inside the reactor
Table (2): Example of gamma spectrometric analysis of air filter by high volume air sampler, in the reator hall.
Energy (KeV) Bq. m~3 Nuclide
662.5 3.7 -10-3 137 Cs 1174.9 2.4 -10-3 60 Co 1335.3 4.1 -10-3 60 Co
hall, and after the elapse of about 16 hr to allow the natural radioactivity present in the filter paper to decay, showed a very low concentration of artificial radioactivity. On the other hand, samples taken just over the reactor cover and 5 meters away from -90- Abdel Fatah et al.,
Table (3): Example of gamma spectrometric analysis of smear tests.
Energy (KeV) Bq. m'3 Nuclide
662.5 0.008 137Cs
1174.9 3 0.9 X 10" 60Co 1335.3 1.4 X 10-3 60Co
Location: Pump room, at the floor, near teh last pump. Area: 100 cmZ
Energy (KeV) Bq. m~3 Nuclide
133.9 0.173 144Ce
662.7 0.996 137Cs 1115.6 0.049 65Zn 1174.4 0.000 60Co 1334.9 0.088 60Co
1.393 2nd RPC (1991) -91-
Location: Pump room, at the floor, near teh last pump. Area: 100 cm?
Energy (KeV> Bq. I'1 Nuclide
133.9 ' 2.0 X 103 144Ce 145.7 • 8.2X101 161Ce 487.9 5.9 X 101 140La
497.7 1.7X101 103RU
662.7 5.8 X 102 137Cs 697.3 1.6X103 144pr 766.7 9.5 95Nb 1116.6 2.3 X 102 65zn 1174.4 1.95X102 60Co 1335.1 1.95 X 102 60Co Background 1463.5 40K 1 1598.8 4.98 X 10 140La
this position showed noticeable traces of 137Co. Table (2) shows
an example of such measurement.
Smear test evaluation: The results of the smear tests, carried
out inside the reactor building in different locations, showed that
the contamination is much smaller than the limit value which-is 9 (9) 37 bq cm" , . Table 3 shows two examples of such tests.
4. Monitoring at El-Anani Farm (1989):
El-Anani farm lies close to the reactor area about 5 km to its
North-West direction. Soil samples measured were taken from
the fields of plums, grapes, guwava and pears. Table (5) showes
the gamma spectrometric measurements of these samples. Soil of -92- S.l Fatah et al.,
Table (5): Gamma Spectrometric Measurments of the Soil of El-Anani farm (1989)
Energy Nuclide Activity (Bq. Kg-1) Plums Grapes Guwava
239 Pb-212 B.G. 1.69X101 L.49X101 325 Pb-214 B.G. 1.25X101 1.14X101 583 Ti-208 B.G. 1.92X101 1.35X101 609 Bi-214 1.01X101 B.G. 1.3 X 101 662 Cs-137 • 1.03X101 1.55X101 1.18 796 Cs-134 3.1 B.G. B.G. 911 Ac-228 1.26 X 101 2.28 X 101 9.59 969 Ac-228 1.35 X 101 1.86X101 9.62 1120 Bi-214 2.76 X 101 3.36 X 101 2.13 X 101 1171 Co-60 2.1 B.G. B.G. 1238 Bi-214 3.55 X 101 B.G. B.G. 1333 Co-60 2.4 B.G. B.G. 1461 K-40 1.63 X 102 1.01 X102 1.11X102 1765 Bi-214 B.G. 3.6 X 101 B.G
pears field shows background while those of plums, grapes and guwava indicate the presence of trace amounts of some radionuclides, Irrigating water was found to be background.
A soil sample from El-Moniar, which is vellage about 15 km
North-west to Inshass reactor area i.e. in the same direction as
El-Anani Farm, was measured. Table (6) shows the results of this 2nd RPC (1991) -93-
Table (6): Gamma Spectrometric Analysis of El-Moniar Soil (1989).
Energy (KeV) Bq. 1-1 Nuclide
239 1.67X101 Pb - 212 296 9.8 Pb-214 339 1.08X101 Ac-228 352 1.1X101 Pb-214 583 1.5 X101 Tl - 208 610 1.5X101 Bi - 214 662 2.58 Cs-137 796 1.1 Cs-134 912 1.15X101 Ac-228 970 6.25 Ac-228 1120 9.25 Bi - 214 1460 2.7 X 102 K-40 1623 4.03 X 101 Bi - 212 1765 1.85X101 Bi - 214 1850 2.4 X 101 Bi - 214
measurements indicating the presence of nearly the same
radionudides in nearly the same concentrations as those found in
El-Anani farm.
5. Measurements at Inshass Research Reactor (1991):
All measurements of air-particulate using high volume air
sample taken outside the reactor, at the reactor entrance, inside -94- Abdel Fatah et al.,
the reactor hall either the reactor is shut-down or working and in
the pump room were analysed with gamma spectroscopy after the elapse of about 16 hour to allow the natural radioactivity to decay showed background level. This result is nearly identical to that obtained in 1988 (see Table 2).
The results of the smear tests carried out at different locations inside the reactor building e.g the top of the reactor cover, the side rings around the reactor cover are les sthan those obtuined in 1988.
Table (7): Gamma spectrascopy of soil samples from infront of the reactor.
Energy Nuclide Activity (Bq. Kg"1) (KeV) Surface Soil Deep Soil
238 Pb-212 16 13.3 325 Pb-214 8.14 8 583 Tl - 583 17.5 16.5 609 Bi - 214 7.6 7.9 662 cs -137 6.9 7.7 911 Ac-228 9.59 10.4 1461 K-40 171.0 143 1596 La -140 0.98 1.3 2nd KPC (1991) -95-
The plant sample taken from location infront of the reactor showed bacground level.
The soil samples taken from location infront of the reactor as surface soil 0-3 cm or deep soil 3-7cm deep showed nearly equal levels (Table 7). 6. Mointoring During Arabian Gulf war (1991): This monitoring was carried out using the following procedures: a) Air Kerma mneasurement using gamma monitor (Tennelec
Table (8): Gamma ray measurements.
Location nGylh
Reactor Building 63 ±6 Nuclear Physics building 69 ±11 Engineering building 78 ±15 Radiation Protection building 68 ±10 Decontamination building 72 ±11 Rad. Biology building 80 ±24 Agricultural building 75 ±15 Wot Laboratory Centre. 61 ±11 Abu zabal Village 76 ±14 - 96 - Abdel Fatah
EPS-1 to measure air kerma rate around the various facilities of the NRC at Inshass. Table (8) containing the results of these measurements indicates that the air kerma varies from 61-80 nGy/h. The normal backgound gamma-ray air kerma is 35-80 nGy/h. Then the radiation level is withen the normal gamma ray background range. b) Radio-pollutants in Air: Air sampler was -used to collect air-particulates through Whatman filter paper No. 41 every 6 hours at 6 O'clock then 12,18 and 24. Radioactivity were measured using: • Single channel analyser, Nal (Tl) crystel 3" well type for relative
111 -117 gamma measurement of I and Cs at 0.36 MeV and 0.66 MeV photopeaks respectively. • Alpha/beta/gamma low counter Tennelec LB 1000 with phoswich detector. • Ga:ama spectrometric measurement using multichannel analyzer Nucleus with HPGe detector. All the results of these measurements indicate that there is no artificial radioactivity had been detected during this period which continued for about 3 months Jan. Mars 1991. Measured radioactivity was mainly due to radon and thoron daughters.
7. Meteorological Data:
The metearological system consist of tower of 15 meter hight 2nd RPC (1991) ' 97"
direction, temperature, relative humidity, atmospheric pressure. The data inducated that the prevailing wind direction was 295° (i.e. North-West) which is in agreement with that of Helwan station at high altitades. It is not expected that this wind direction NE or SE will help in carrying radioactive air particulates or gases from the Gulf area or the east to the west (Egypt). CONLUSION
From the results obtained in this report we can conclude the following: • During Chernobyl accident (1986) there is an evidence that radioactivity had reached the terretory of Egypt. Exactly in May 5,1986. • Monitoring in and around the research reactor at Inshass indicate very low level of radioactivity during periods of 1988 to 1991. • Mointoring far away from the reactor site 5-15 kilometer down wind direction indicated normal level of natural raduoactivity and low level of deposited Cs-137 of Chernobyl origin. • Monitoring during Gulf war time indicated normal or background level of radioactivity. • Meteorological data mainly wind direction 295* (N - W) will not help in carrying radiation debris or gases from Gulf area to Egypt. - 98 - Abdel Fatah et
REFERENCES: 1. k.A. Mahmoud, M.K.Moloukhia & A.T. Abdel fattah: Si->J Stable Sr and stable Ca in soil, food items, w?fer and human bones. UARSCEAR Vol. 3-4 (1961). 2. K.A. Mahmoud, A.T.Abdel fattah & M.E.M. Abu Zied: latest 1961 nuclear explosion test series on the fallout over UAR during Sept. to Dec. 1961. UARSCEAR Vol. 5-1 (1961). 3. K.A. Mahmoud, A.T. Abdel fattah & R henry: Report on the fallout radioactive content on UAR during 1962. UARSCEAR Vol. 5-1 (1963). 4. K.A. Mahmoud, A.T. Abdel fattah: Experience dev^1 ' in monitoring and control radiation in UAR 4th Conf. on-Peaceful Uses of Atonuc Energy, Geneva Switzerland 1974. 5. L.R. Rogers: nuclear Power safety. Agencu expert report for the project EGY/4/08. IAEA-TA-2261 (1984). 6. A.T.Abdel fattah, M.W. Essa: Prelcminary environmental monetoring following Chernoby assident Isot. & Rad. Res. 23,1,1-81991. 7. E.R. devera, B.P. Simmons, R.D. Stevens and D.L. storm EPA-600/2-80-018, Washington D.C. (1980). 8. B. Kratochril and J.K. taylor, Anal. Chem., 53,924, A (1981). 2nd RPC (1991) - " "
9. Doerfel & H. and Piesch E., Radiation Protection Dosimetry. Vol. 19 (1987). 10. Bundesgesetzblatt Nr. 125, 1976. radiological Protection Ordinance Oct. 13 (1976). -100- Abdel Fatah et al.,
* 32
2'JO Km
(617 Km l'rora Cairo)
28 28-*
J>
I'O 32 I1'!;-:. "1 : UonHnt-lTiR .'.Jtiifcionn t ninoe 1'X'iO t Ca.1 ro,Ale.x.'in(lrJ./itMatruh, Giwn, AHWDII. Hince ]/ib2 i lorfc Said, Domiettn, Ilooetta, 131 Mansura, Kaphr El i3}ieik]if Tuntu, Damnnhur, Zakazeek, Banim, Shebeen 1SL Koine, Giza, El l-'ayom, Baiii Swefa, Al minya, Aayut^ Sohag, Kena, Luxor, liinc.c .l';i,.| • JJckhel, Kl Kontellu. 2nd RPC (1991) - 101 - (U 0.6 0.8 We ight in g. Fig.I 2): K- 40 calibration curve ( Low beta counter ). -102- Abdel Fatah et al., 60 — in Calibration medium : o Filter paper 10.5 cm -m n Cr\ |i | f inn OUILJ IIUl 1 . . Sand n 1.0 2 Ba- 1 10 n n . i \m u c -ift n y 20 \ 2 ^ 00 1 oo LU f-8 8 i 1 - 7^«—«—. • 0 1 1 i i i 200 £00 600 800 1000 1200 U00 1600 1800 Energy in KeV Fig. ( 3)Gamma spectrometry calibration curve (3 HPGe detector) 2nd RPC (1991) -103- Fig. 4 : Meteorological Tower at Inshass. - 104 - Abdel Fatah et al.. u 12 <*> 10 f- 1 CO \ 6 : A — 1 / 2 0 1 1 1 A, I i i I . i . *t**T "7 7*T*\ ! i i ^i^T*^ , 30 5 10., 20 30 9 April May June 1986 Fig. (5 ): Aerosole activity in the time of Cherobyl accident 2nd Rad. Prot. Conf. (1991), Paper No. 9 SAFE TRANSPORT OF RADIOACTIVE MATERIALS AND OTHER ACTIVITIES (*) Rifaat M.K. El Shinawy Radiation Protection Department, NRC, AEE ABSTRACT This report deals with the different activities in the area of safe transport of radioactive materials. Description of the aim of IAEA safe transport of radioactive materials committes are given. The IAEA committee, dealing with transport of radioactive materials are : 1. Standing Adivsory Group for Safe Transport of Radioactive MateriaR (SAGSTRAM) and 2. Revision of the Safe Transport of Radioactive Materials (RESTRAM). The technical and Protective measures taken during transport of radioactive materials through the vital pass way, Suez Canal are mentioned. During the period from January to June 1991, the number of vessels crossing the Suez Canal and carrying on board radioactive materials of different types, were 44 Vessels of different nationalitites. The total activity trnasported was 76. 340 TBq (2.063 KCi). (*) Full Text in Arabic. 2nd RPC (1991) -106" Also the protective measures taken during the transportation of the large radioactive y- irradiator, from Alexandria port to the National Center for Radiation Research and Technology (NCRRT) [Co - 60,366/TBq (98940 KCi)] were mentioned. Other radiation protection activities mentioned in the report are in the fields of : 1. Safe storage and handling of radioactive materials inside and outside Inshas Site, 2. Assessment of radioactivity in imported foodstuff, 3. Emergency response during the Gulf war, 4. Monitor^ tr of radiation levels in the Egyptian environment. 5. Training and supervision in the field of radia;,on protection. >i'>4/ iUli/f ^^^^/^./^-v 11 olLJLLI ^>^ JU-MH yJUJ uLi -V T • jf^iL ijii^Ji iktnf i OJU M1I/I-/1-V LIS^JI (M-i t^- .i -f J-ii t-itU .a -0 U*W c^ij .4.1-Y V>- JU /.JUJI -Y V>L JUI / .JU-JI - JUA JU>I /JUJI-Y Ji* /JU-JI ~\ »JU-JI -V J V^J /J—-JI -0 I1U/I-/1-V j v-JU /jjiTJI ib-VI JU-JI . JU-. JS" ,y M-iJI *kli\ vJLcl ^lyo-l '.. .34 & i-iUUI I! l"t( -UUI J 1 . i.l . JaiU- j^jJI jX . i.l • I11I/I-/1-V JJL" iilkll iu* ^ i-JLJI i-JLJI rLJVI J^IS" ili Z, U/JUJ ^jLUI J-IT JU»W cuoj j^iTJI iU-VI I11I/I-/1-V U*,c*ij .a.f iiLJI Us. ^JUI ^jdl ^lijjlj ^lijl ^ JjVI ^JUJI >cjll ^» jj I* >jl oL JjAy . ^ ^A^ . ^JLCLJI crwjjll .Jjlj u-ijjdl iL> .LacI iiLJI jijil ^x» -Y c»»»j .4.1 . iibkll tf iLU i^lyi cJjb- LjJJI SilUI i—Jl SsJlSjJI g*yi jljU^ I^JII ^J .i.l iJU. * .(^^A^ I11I/I-/1-V <> rfJjli? c J * 0-* »-^L^!l JL>VI oLli-,1 ^x» ^J»y *J Lf . o^jLJIjolJl...i7..^l iU^JI ^j. i. I JuJU SJUI w^ill JLjl ^S. Jijj >cjll ^.l^J ^ ZJIJUJI AIJ-JJ ^Ui^l Ol^l oUU i»> - rLil - -"V obUI -A Mil/I-/I-V ji -^ ol.lidl ,>. ^^UJij iLU*j .yjk; / **ill AI^LI JB -N .4.1 JUJU kij li* j/U, i^^l Ujfl^.j ^JU if^UI ife-t, ioill iilUI iu*J iJLti I11I/I-/1-V j ^JU .i.i Ju-JI U 1>u ^^ JLU UjU ^JJij i>^wt _>ijjji xx. Jur. A. .U J**.-.?.' j^-i ^ i*Lj>l J>U- ^uU^JI |MJ (jJ^w-* Jilj JuUj li* j > ^Aa .^j \ ^vr fu ^ c^yI. LUII J\* pjai 1 1 SLJ .^JAJI v^^i c ^ J^* v*tt Ju»l .a.! JU-JI 'j4 L-ldl »JU ^ IJU f' o' >^J i-^LxiVI oL-LiJI iU-l xji JU>I .i U .j> ^ (.Lj-iJI) ilijutl j-a- ^ ;.,.U1I I11I/I-/1-V \. JL» ^U-iVI ,>. <^>^IIJ oLIjjJI >cjll iUij Jaii L iVl ili^JI - y : ^ycs i-Juill •**) ojlijJI l»auiJ jLiLjJI jwJkil >4>MII ^ r. cujj . i. i JJVI >Ejii j> us" ^ t • i JU*K .i.i ^UJI jijll fj^ifj i-K ^ id 4.1.5T...Ui^JLJI oUWVI ^ • • • .J.I ^ ^ • iiJ -u*UiVI iU^D iJjJI iijJUl OL-VJ olL,]*; JJ»- > ^ • • rLxll Jut • Uj. i - i_f»J>uA\ *Ji*)l Jut . i - ^jiu^ ,yj-—J . J pL-ei ,y (.jj-b^JI N> to JUfi JLAMI.A >U /tJU*Jf - ^iJi.a* JUc. J Cjtij.i.l - J-i-i i_iiU.i - ^jJtll (JLaj. i Vjij Jl»l/t.u.JI - (_jUa> ju— ju9x«.*. i - jbj^ rl-iJI J-* JU»V>.>. A jtJa^jl fliallf iLjiigJI ^mil juLjil juldl /i -n- v \r olliJI Uluj JWJI £»jLdl -^U-i^l ^^Jl t^uiJI ^y AJ-J^1I .^J-M Y £. tjxe. Jill JL£.A - jL«-i AUI J-*.J - Ji>c }\j+.i - tjtjljjj) jjj\.i tJUu.l - rbJtil JUt r M/\./>« - v s Y J Y * r jiia^l fliaJlj ujli»JI ^-uU juiLJI juldl M \ - - > \ - Y I - Y Lulll oLUi) &Si\ cio-Jij JJIJCJI lit -0 \ Y \ - r 0 \ • - \ r Y r 0 0 r r \ .ji* uujLa>l j~£ • \ r CM - JL±\ i- H1I/I-/1-V u re ^^^^/^./v re ^^^^/^./A ^^^^/^./^ ^^^^/^./^r r ^^^^/^./^v r I11I/I-/1-V TECHNICAL AND SCIENTIFIC POSTERS ROLOG1CAL STATION ^• ) JLCJJVI ik>v» N-\ COUNTING SYSTEMS ^L. *VI i.ll l1 t;:l y~\ ^loij^a^i Y • ^^U^lg.^ \-Y 7J | J| I; V —Y u.^WloL, r-Y aw „ ( , I 1 ( / V 1" " • 1*1 ^t A V -- - i . |.| Jl 1/ V \. -Y . JJLJIj oUUdl, Ajbll aiji -.J»J oljWI \ N -Y ^Jb)l '.Ji>j oljUl > Y-Y M-Y . ^^^^^u . ijyI >ljJ-lj VrfUil Jj^j oljWI N o-Y M1I/I-/1-V iuLJI ij WASTE STORAGE FACILITY AT INSHAS \-e IAEA RESEACH CONTRACT Y-o V ^ • ENVIRONMENTAL MONITRING FOLLOWING CHERNOBYL ACCIDENT RADIATON MONITORING AND CONTROL : SYSYTM. . «~>j.a-J*.* »JJ» ^\*Si\ -yt LUjJI oUlj>lj Jiftlji Mil /I -/I -V lkA. JUfi .i/jU-Jlj 1^>>B JUI /.JU-JI uvJiJIj (>-JLa>*iflj i^o-uJI ii-> .LaxI ^*> ^yJI iLul ^ JjU LT . «—iJI (j--Jj . A. I Jij ^^^^ - ^^^. ru *bu. ^ tft-li)lj ^JUAIIJ ^JtfLJll uwjjdlj ^^ijjJI ilJk • Lit I ,_pi4Lj ^tfydl j-a>j oUI J-Ull jS> - 4u^JI O JLc c>«hn4 Xrf.ijb. X^ J^ /.JU-JI 1. Neutron Rem-Meter (Eberline-EPS-1) ol^yJLI iiilSUI lcyL\ JJL** ^Li jLf>. polyethylene moderated sphere He-3 det. 2. Radiation doserate meter (Eberline-ESP-1) U*J_J L.U i B- Detector 3. Radiation doserate surveymeter ( Dosimeter ) UUj Lu< i*iV ii» B- detector 4. Radiation monitoring detector - (Mini Rad-D. UU i*iV ii» dosimeter) Model number 3029). 5. Super Mini Radiation monitor dosimeter (Model ^U; ii»l£ll i**L*iVI i*>I-l number 3500) 6. Contamination monitor (Dosimeter-Model 2000) Pancake Prob. 7. Radiation doserate meter (RDM-1-Nuclear ii»l£if i*cLuiVI ic>J-l Enterprises) 8. Air Sampler (staplex) 9. High volume Air Sampler (nuclear asociates) .**&! |»*J>I ^'» 10. ALARM Dosimeter (1888-B-Dosimeter) 11. Pocket ALARM (low energy dosimeter Model 002.2000 mlR 12. Dosimeter charger Model 909 (Dosimeter) 13. Dosimeter charger (Victoreen) 14. Flim - badge 15. Kodak Personal Monitoring Film type 2. iiilSUl 16. High dosed film and Low dosed film 17. Transmission Densitometer (EDT 1105) I11I/I-/1-V 18. TLD system (Harahaw - 4000) 19. CPS computer. TUM Printer. Star CC-15, UPS ^'>' J***^. i^^i^' **>' 20. TLD Card. r. icLJI ^ii ^ dUij (yjll C-UJIj ijlijJI |*~J J-JJ ***> JU>I jLk»u> . i.i JU-J! LfJI .i.l JUJI oL^ iiLijJI <»MJ ^^JJ .a.I JL-JI jJul - tjjJI ijlju ,_,»_, ^U J^i yjjdl ^LiJIj iU^JI ^ J»LSJ rU Jtjju .jjdl ^J, i.l JU-JI fli ft liar iJU^JI JLcVI jj«ar *U-\ ljxi\ (^J, s.i JU-JI^ *jju}| j.U> ^ M1I/I-/1-V - > Y I11I/I-/1-V TY - \f Njt A v n o t r r i-JLc ij4»lj i-X* ^^v<^ wvv ^^vr i^f ]U^ | f M11/I-/1-V rr -Y- N V u I11I/I-/1-Y r U ^^ n n i Y^ Y r ^'^ YY u ^^ tY •v Y^ - ^ M ON - A \. - Yt \> r. oljLUI V v. tv 0 YY Y \. Y\ YY \\ To n 1\ Y^ - - • t i| , •( J ifua^jl fliaJIg flLj ^^^ No Y"\ NO A i NV Nt N A YY YN ^ ^ Nf V V N^ NY - r \r NV - - NY V - i ^ YY 00 NV - N"\ YV 0 V V r 0 A •v S^llJflfc«rJ - A No r Yo \ N"\ NN j jua^JI fliaJI| iLLi^JI ^uii ^lill juicJI ^^ Mil/I-/I-V rv p.SLJ Yfl ^ 0 i V V - V Y Y Y - - N \ - - > > - Y Y Y \ S Y> Yr 0 Y 0\ 11 \. A LJ -rti j \\\. L- * ON x all Y 0 • = yUjJI OUJJLI g OJI «• = JI r. = TA I1H/I-/1-V Table 1 Collective Effective Dose Equivalent for Radiation Workers in the Nuclear Research Centre of the Atomic Energy Est, Egypt COLLECTIVE DOSE COLLECTIVE NUMBER OF YEAR NUMBER OF IN YEAR DOSE IN PERSONS MAN - Sv PERSONS MAN - Sv 1964 360 1.056 1972 206 0.294 1965 424 1.456 1973 116 0.153 1966 482 1.393 1974 206 0.635 1967 476 1.212 1975 202 1.088 1963 472 0.637 1976 163 0.420 1969 495 1.0.06 1977 157 0.393 1970 345 0.572 1978 195 0.903 1971 306 0.454 1979 323 0.671 Table 2 Cases of Over - Exposure in the Nuclear Research Centre COLLECTIVE DOSE IN PERCENTAGE mSv NUMBER OF DEPT. YEAR PERSONS DEPT. CENTRE 1964 65 65 95 95 4 6.25 1.11 Phys 1965 65 1 1.66 0.23 Plasma 1967 54 54 - 2 1.33 0.42 Ch and Phys 1968 95 1 2.94 0.24 Neut. Ph. 1975 95 110 2 3.45 0.99 Biol. 1978 93 55 - 2 2.27* 1.03 Chem. TOTAL 12 0.25 * Assuming the number of workers in the department of Chemsitry are 88 persons, radiation records show data recorded for 11 workers only. r\ M1I/I-/1-V Collective Dose Distribution Atomic Energy Authority No of Workers Egypt 1960 -1990 11 2 1 0 I 5 50 50 100 100 150 150 200 200 250 250 300 mean - mSv 32% 115.2 Collective Dose 56% 201.6 10% 36.0 2% 7.2 Annual Nat. Background Level = 1 mSv Note: average annual.. Exposure = 0.66 mSv Annual exposure Annual Collective Exposure = 0.463 manSv due to work No of Rad. Workers = 702 man in A.E.A = excess * For 2400 Worker, the exposure due to 102 natural background average exposure = 0.66 x = 0.29 mSv. radiation of 47 days 2400 only. * For 3700 worker (1990) average exposure = 0.66 x 102 =_ 0.12 mSv. 3700 M1I/I-/1-V No of Workers No of Workers 86 29 51 11 0-5 55-50 50-100100-150150-200 m Sv 0-5 5-510 50-100 100-150 m Collective Dose Collective Dose 5 m Sv = ADL member of public 5 m Sv = u-'E-C 50 m Sv = ADL occpational 50 m Sv = "E'C 100 m Sv = 2 ADL occpational 100 m Sv = •j>:;r 150 m Sv = 3 ADL occpational 150 m Sv • t » t * _ 11 200 mSv = 4 ADL occpational No of Workers No of Workers 75 51 34 23 19 1 I I 0-5 5-50 50-10I0 100-150 150-200 200-250 m Sv 0-5 5-50 50-100 m Collective Dose Collective Dose I11I/I-/1-V n No of Workers 34 11 I i 0 0 1 I 0 5 5-50 50-100 100-150 250-300 m Sv Collective Dose No of Workers No of Workers 37 46 i 31 26 I II 0 5 5-50 50-100 m Sv 0 5 5-50 m Sv Collective Dose Collective Dose rr 5 m Sv 32% 56% 2% 150 mSv 100 m Sv 10% 50 m Sv Collective Dose 1960 -1990 ATOMIC ENERGY AUTHORITY EGYPT I11I/I-/1-V rr r. tv As - - - Y"\ YV UN \r. • XI ^^ V^ v^ ,n UL \^^ ^A re v. 6\ Yr rr NA,L N,At \.. M \ov voo c • UY J J VN - • yn \r^6 V.Y YY"\\S wr rtt ^^^ UN YYY tAY I11I/I-/1-V ^^A^ lj\ (TA) r. ij- U-) IV e • LA At£ t rt rv A W rt .- Yj"V (TT) A V^ L\,A (TO Tj) ()() e • tv oljUJI rr 1jt"V« (TY1) rrr rtr A^V ^ SAA o>-- itLJI ,> yjjdl gliillj i-UjJI pjtf ^UJI ^^JUJI J}U Ji* dUij *—ill *j~*v> yj*}\ JUC jjji JLJ" .j.i /.^>jii itii ^^^^/^ -/v ^'LJI ^^JWI yjll j> cbUI jut >U» JU^I . A.I .u-JI ,>- oUS* c- jdi ^tijij *t\sj\ pji JjVi .^juji ^i j> ^UJJI ^.ir ju~. cujj ij. N ^AV ^-J> ^fi thu UJI j2jll aU*;l J*. jj -T J»LLJI oli JIJJ &H\ JJLJI - ^jLill J*tf JUM CUJJ .4.1 JL-JI jcuil JOJJ II* I11I/I-/1-V .#11 Ll* (j*-** CJS hsj^\ ^*j* |.Ljl iJLrL^VI 2^0)1 j^iu - N ^^ • f U JucUiVl ili^JJ iJjjJI *^dJI oUy j-kr - N ^^N) iJjjJI i^JLJIj iijJJI iiUiU i j^UI JJU ^> rUii ii-LVI rv J»LLJIj i^Ui1*! iilSjJf ikUl ^>C o- i^JI /U i-J-JI o Ltl . ^'Jdll ^LijJIj ilijJI ,»-J (^ JL^i-1 -I 2L4JI) iJ^I 2ki;Vli fUJ UJUJ^J ijf.j.Vl jJ>J yi U\ t^JUJI >c>U ILISUI JLcVI ^ -V jJL* ijo> ^i (OJMJI **i jjju^>) iJlijJI oUjilj oljLi.i-VI pjj-1* i-Jlj-» N • • • .4.! J-liUI J^ft>U» JU*.| - v ju^l .gLiaJlj iLAijJI ^ juLJI ,/UrJI ^M)l 111171 -/I -V 2 nd Rad Prot. Conference, (1991) Paper No 9 I J»LUI oli iljJJ ^\ JUI Jl^- ^i .AAill OIUU ^Vl kUo-aifl Jl^w y,* - M1I/I-/1 - V ,JL)I iilkU ZJjJl .^.Vl JJLJI Jliv ^ .^jJJI UlkU iJjjJI iJlTjJl JoljjSL. ji iJjjJI oLkdl jf. L oli iljJU ^1 JiJW i-»lil ZJjjJI Uii. „• LJja oLku O^Jj AoJUl SSIUI iJjjJI Utf JI iiUVli iJ,JJI -iVI LU^IJ iJjJI i^dl ,5* tjjLatl iOi^l oL-»>dl ^jb iJLSL* c (International Commission on Radiological Protection ICRP) C>-JI Js. ^L^iVI ol^sJll i^f ,>- ^1 JJUI oUU JU '•"-" JUI Jl** ^ oL~f j* UjjJI aiOJI ,>* jjuy LJ oitju; ^ 1+*^. U JLJL jjj^J £jXl\ ALUII ^*J \^^"\ fU ^JI jl> ,JA (CRP) ^l>dl i-i»JI g.yi (.Li jj^wi? ijj^ Aip J J^aij JJJIJ RADTRAN - IV Jjudl ^.U^JI r^ jJUIj Jfcl) 5663/R1/RB *>j ^>>w JJUJ • LiltVI JjjJI uu ii^iJL* i—l»y 2f*ijJ |»L* »JJJ-A< Uaul AJJH .JJfcll oLL*. UJI *j>JI ^ J-uJI JU-j f . LxcVl JjjJI HyJL iJL^* i--o-^ ^>U £^"j 0^ L'U^VIj ^Vl JUI JU-. aljlI JB oUL*I ^JULllj uJ^-il il^SU ^ill ZSUdl iJjJI D JJIj ^yjuJI ji> ^ (RESTRAM) ^Ui^l J»UJI ofi »ljJU ^1 JiJI JJ ^1 JJLJI Jtclji i»>l^c it^^Jll »i J>IIJ «"V^j cLVI aLJL» ^UAVI J»LUI oli iljJLJ ^fl JSJI U f .A. .rv ,v iUVU iJjjJI jt^Jill iJii-j iyjjVI LjLiiiVI i«j-jJI L* <>uJ,j L;-ifc>'j I11I/I-/1 -V .UaJ vJ»j *Jj . L .(ICRP) L^LuiVI iili^U iJjjJI *Jad\ c 5663/R1/RB, 4292/RB rUjl &j»J\ *>u <> JI^VI -r RADTRAN - IV& INTERTRAN. r" ysJI jJUdl uj, -GJUJIJ (^^o-JI / U^L-VI /JU iLJ ^c ^eUulV! J»LLJI oli il^JJ ^V! Jidl oLJLe JLt ^li^JI ^1 Jl^iVI - oli aljll JJu oULfi r,..b;T,, i-»UI-l LJLiuJI oLJL*-Jij sjjjii jidi (jaj.ijj JIJU»I p&s liTj. iikidi jaji J;UJ rij^ oli il^JJ ^Vl JiJJ iiJLidl JUtVL UISII i-^LJI *l*UI J-^U JL*1 j^* rv.rv >• >o r/v\ - rr /1-v -,U» ^ SJyUll ^U^VI J»UJI ^-5" j^ji ^UJI Jj UF6 NAA^ Th-230 U3O4 H-3 UO2 ro^ —— Co-60 V « J^ t*-> \) : ii I11I/I-/1 - V ii - rv\\ Ujj V c TYPE PL rroN - Yo.^^rA'^: i_,ui,•*, .i*L,/4ljj>L. Ne- (j (AECB - CDN/2012/BLU): *>)l ,^-y > Uif^ll ol^ (REV. 17) 31/3/1992: » APRIL -1990: (58630A) 21/9/1990: i^yi j (58630 A) L4/9/1990: Vv-JI jLSl iilfi * UJ ^U i^-^^LI oULJU l4.u.ilk«j iJU i*w»jll oULJI i*> JoSV, ^i!l iilkU iJjJI iJlTjJl ^ .jiLJIj ^1 JJUU i-,U.I o M1I/I-/1 - V Radioactive Waste Storage Facilities of Atomic Energy Authority - Egypt - 1991 JLII obUJJ jjU-J' jiikuJ ^1 Ailz^JI JlkuJJ ^M^irJl ^-Jl ^oidl J_,Lw r tl^JI /i LT. CJ>OI J*U»J jJLallj obcULT iikiJI i^jl g^ J> . /r N • - A { l+Lu v^J JJJJI uLaJI ^*c Oli **ill oLUdl S^J ijl. JUI ajfcll Ju_»>- J\ iiUVL . ^I4JI otjicdlj itUUJ ii-J (\t) pij jtyOli iu+JI ^iijljl .^JUw ^j .i.\ JL-JI J^ a^Lllt iJ;JJl M1I/I-/1 - V ilT ^ j,,,uT7tl kiJI oLU ^ iikidl -uclll jJlkJI fAicu-i >U)I Luis- r>JI UT I11I/I-/1 - V . obJI LU ^ ijiLJI ioLjjJI oljl^ZU li-t ^UiVI Ajlall & nJ . iijji-il IJliiJI .a.l o iLaj LJl llk.;.f i^LulVI oLji-ilj (S • ^•ii.„.&!) pJi-l i*jl i«lju JLu »^IT.I.-* ui^loL-UUI ju^\\ fiiaJlf iL I11I/I-/1 - V iVi oii>uii £— gj^t* ^i tfjiiii -r 1 UilJI lit\*Si\ *s.JL\ 1- Regulation for the safe transport of radioactive materials IAEA, 1985 Edition (As Amended 1990). - MS- ioJJI Zilkl) UjjJI XJIOII ^LiuiVI J»LLJI oli iljJJ ,>.Vl .^^^^/'^/^t u!l -o