ABSTRACT
Year : 2023 | Volume
: 46 | Issue : 5 | Page : 440--453
Theme 9. Regulatory framework: System of protection, standards and regulation
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Abstract - 91511: Implementation of ALARA during replacement of heat exchangers at reprocessing plant
J. P. N. Pandey1, Pankaj Kumar1, D. K. Pandey1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: A reprocessing plant designed for reprocessing of PHWR spent nuclear fuel was operated for more than 10 years as per the design intent. Thermosyphon evaporators were used for concentration of product solution and for the liquid waste evaporation.[1] After intended batch of operation two heat exchangers (HXs) of these evaporators got failed due to leakage development in their tube section. Therefore, replacement of these failed HXs were planned after regulatory approval. Replacement work was associated with high potential of external & internal hazard as it requires personal entry to the process cell for cutting, welding, floor decontamination and other associated works including quality assurance of weld joints. These evaporators were located in two different process cells where personal entries are prohibited due to high background radiation levels. Therefore, a scheme was worked out to carryout replacement work by limiting the individual dose within prescribed limit and controlling the collective dose within a dose budget assigned by regulatory authority. Efforts were made to control the individual and collective exposure by bringing down the background radiation level in process cells as per the ALARA principle which is being presented in this paper.
Materials and Methods: Hot spot identification: Radiation mapping of process cells was carried out to identify the hotspots on equipment & transfer lines which were contributing significant radiation dose in working area. Radiation mapping was done from cell top opening by lowering a gamma survey meter (silicon diode-detector) as well as by making physical entries. Chemical decontamination with mixture of caustic & carbonate solution followed by washing with DMW was found very effective to bring down the radiation level of hotspots below targeted levels. Gradual reduction in radiation level of cells with time is shown in [Figure 1]. Contamination Monitoring and control: Inhalation hazards due to airborne contamination generated during cutting and welding jobs were avoided by providing local ventilation near the working spot in addition to existing cell ventilation. This helped to prevent build-up of airborne activity near the working area. For monitoring purpose, one Continuous Air Monitoring Systems (CAM) with extended suction probe and remote display system as well as Personal Air Samplers (PAS) were also used for work place monitoring during each cutting & welding job. Measurement of airborne contamination level helped in deciding appropriate respiratory protective wears used during cutting and welding. Total 232 personal entries consuming 755 man-hours were utilized for replacement of both heat exchangers. Detectable airborne contamination was observed in 13 entries only. Details of personal entries, man-hour consumed and collective exposure are presented in [Figure 2]. Exposure control: Internal exposure was controlled by use of appropriate PPEs supported by indigenously developed fresh airline manifold which gives alarm on stoppage of air supply to respirator. Also, one dedicated crew of skilled persons were deployed for undressing of contaminated PPE and work was successfully completed without any inhalation hazards. External exposure was controlled using Electronic Personal Dosimeters (EPDs) having alarms for dose, dose rate and time. For individual dose control “Distance” factor was used by deploying an automation system (scissor lift) which could carry the radiation workers to desired height for intended works.{Figure 1}{Figure 2}
Results and Discussion: Steam Strippers and Product Storage Tanks were the major hot spots which were identified by radiation mapping. Radiation level near working area were reduced by internal decontamination using mixture of caustic and carbonate followed by flushing with water. Airborne contamination and transferable contamination were controlled below targeted value.
Conclusion: Replacement of Heat Exchangers of Product evaporator and waste cycle evaporator was completed consuming 25% & 70% of authorized dose budget without any incidents of personal contamination with individual exposure well within regulatory limits.
Keywords: ALARA, dose control, heat exchangers, reprocessing plant
Reference
Nam-Suk J, Hee-Seock L, Arim L, Sang Eun H. J Radiol Prot 2021;41:S150-9.
Abstract - 92240: Consideration of the future direction of the Korean radiation regulation against crises
Euna Lee, Chae-Eon Kim, Euidam Kim, Yoonsun Chung
Department of Nuclear Engineering, Hanyang University, Seoul, Republic of Korea
E-mail: [email protected]
COVID-19 has taken over the entire world, and the radiation regulatory system was no exception. In the IAEA (International Atomic Energy Agency) survey, 97 out of 123 countries (78.86%) reported that the regulatory body (RB) had to have limited regulatory activities during the COVID-19 period.[1] The Korean RB, KINS (Korea Institute of Nuclear Safety), tried to implement the regular inspection, but the schedule was inevitably delayed due to the COVID-19 distancing. They tried to reschedule the regular inspection and changed the method depending on the type of facility. For example, regular inspection of medical facilities sensitive to the virus has been postponed or substituted with the documentary examination, which includes a self-inspection report, a radiation source management inspection report, and a facility site video.[2] Several other countries also had difficulty with radiation regulation due to COVID-19. Countries including the USA, Canada, France, Poland, Ukraine, Japan, and Australia that we investigated, renewed their regulatory system against COVID-19 based on remote-inspection methods. Virtual or IT tools including E-mail, audio or video file, or online conference were used to assist with the remote inspection. Even before COVID-19, the inspection priority existed in some countries (USA and Australia). After the COVID-19 breakout, more countries have introduced new inspection priorities which prioritize the needs of essential agencies in the global crisis (USA, Canada, and France). Through these attempts, the RBs of each country tried their best to carry out normal radiation regulations even in this epidemic situation. Meanwhile, the renewed regulatory method against COVID-19 in Korea was implemented in 2021. Thus, we need to establish systematic procedures in advance to deal with possible future emergencies. Indeed, since Canadian RB prepared robust business contingency plans for the nuclear sector after the SARS (severe acute respiratory syndrome) happened in 2003, they could maintain the stability of their radiation regulatory system during COVID-19.[4] Therefore, we specified the process of Korean regular inspection as 'contactless inspection', which combines the video and documentary examination. Under this organized detailed procedure we constructed, the RB and examinee can prepare and implement the regular inspection thoroughly even in these crises. If needed to check the real-time situation in the facility, new technologies such as Augmented Reality tools (AR glasses, AR masks) and web conference platforms (Zoom, etc.) can also be used. Moreover, preparing detailed plans, the Korean RB might need to establish explicit standards related to classifying the riskiness of the radiation facilities more in detail as the USA and Australia do. This standard would allow the RB to allocate regulatory resources efficiently to facilities that need the regulation most urgently. In this study, we synthesized changes in the international regulatory system and considered the future direction of the Korean regulatory system. Now, to maintain the stability of the regulatory system in global crises, it is time that we have to prepare systematic procedures rapidly with this pandemic as a chance to advance.
Keywords: Contactless inspection, COVID-19, regular inspection, regulatory system, systematic procedure
References
International Atomic Energy Agency (IAEA). Impact of COVID-19 Pandemic on the Regulatory Activities for the Safety of Radiation Sources. Ver. 2. Vienna, Austria: International Atomic Energy Agency; 2020.Korea Institute of Nuclear Safety (KINS). The Information on 2021 Regular Inspection (Documentary Examination Organization); 2021. Available from: https://rasis.kins.re.kr, KOREA.Nuclear Regulatory Commission (NRC). NRC COVID-19 Update, NRC COVID-19 Update | NRC.gov, USA; 2022.Canadian Nuclear Safety Commission (CNSC). CNSC's Response to COVID-19, CNSC's Response to COVID-19. Canadian Nuclear Safety Commission, Canada; 2020.The French Nuclear Safety Authority (ASN). ASN REPORT on the state of nuclear safety and Radiation Protection in France in 2020. France; 2021.Nuclear Regulation Authority (NRA). NRA Annual Report. Japan; 2020.National Atomic energy Agency (NAA). Annual Report; Activities of the President of the National Atomic Energy Agency and Assessment of Nuclear Safety and Radiological Protection in Poland in 2020. Poland; 2020.State Nuclear Regulatory Inspectorate of Ukraine (SNRIU). Report On Nuclear and Radiation Safety in Ukraine for 2020. Ukraine; 2020.Australian Radiation Protection and Nuclear Safety Agency (ARPANSA). Annual Report 2020-21. Australia; 2021.
Abstract - 92408: Regulatory framework for radiation protection for nuclear and radiological facilities
Shailesh Deolekar, N. L. Sonar, S. S. Prasad, K. P. Dixit
BARC Safety Council Secretariat, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Radiation and radioactive substances are being used for various societal applications. However, due to the associated risk with radiation; India has accorded high priority for protection of the personnel, the public and the environment from nuclear and radiation hazards without compromising societal benefits. Safety is ensured by an effective legal and governmental framework which is necessary to regulate the facilities and activities that give rise to radiological risk. Government of India has issued Act and Rules such as, Atomic Energy Act, 1962, that provides the basic regulatory framework for all activities pertaining to atomic energy programme in India.[1] Radiation Protection Rules, 2004[2] specifies the functions and responsibilities of the employer, the licensee, the workers and radiological safety officers. Atomic Energy (Factories) Rules, 1996[3] enable statutory authorities to enforce regulatory control over inspectors, workplace hygiene, safe use of machinery, workers, protective equipment, etc. Atomic Energy (Safe disposal of radioactive waste), Rules, 1987, enables the provisions for safe management of Radioactive waste. Regulatory body appointed by Government of India under Atomic Energy Act is entrusted with the responsibility of protecting workers, public and the environment against harmful effects of ionizing radiation. This paper describes the experiences, advancement and challenges faced by the regulatory body during implementation of radiation protection in nuclear and radiological facility.
Regulatory Framework: A Regulatory framework is established for nuclear and radiological facilities in BARC to ensure radiological protection in the activities associated with nuclear and radiation facilities. The regulatory framework adapted for radiation protection of BARC facilities comprises of a three-tier safety review process involving approx. 1000 experts drawn from various fields to support the regulatory safety review processes.
Challenges to Regulators: The regulatory body faces challenges due to the diverse nature of nuclear and radiological facilities in BARC such as, fuel fabrication, research reactors, reprocessing, waste management, accelerators, biomedical, radiochemical, etc. With the advancement of the knowledge, the new technologies are introduced day to day in different fields. One of the major challenge posed to the regulator is to develop regulatory competence on the new technologies introduced. Some of the recent challenges are advancement of new waste management practices for separation of useful radionuclides from the radioactive waste, integration of the nuclear facilities like reprocessing, waste management and fuel fabrication, new designs of high energy accelerators, etc. Regulatory documents are not available for the new technologies/fields and such regulations are required to be evolved based on experience. Preparation of several regulatory documents are already undertaken based on the new topics and recent technological advancements. As most of the facilities are getting older in BARC, life extension of such facilities and their subsequent decommissioning is a major challenge to be faced in near future. Regulators are realizing the challenges and acquiring/ improving their knowledge base to take up the challenges. The domain knowledge of other regulatory bodies and reputed national and international institutes/ agencies such as, AERB, IAEA, ICRP, USNRC, etc., are also referred to for getting the information on experience feedback and improving upon safety.
Radiation Protection: The regulatory control is achieved through the process of licensing. A regulatory consent is required to be obtained from the regulatory body for siting, design & construction, commissioning, operation and decommissioning of a radiation installation. Separate consents are needed for sealed /unsealed sources, radiation generating equipment, safe transport of radioactive material and disposal of radioactive waste/disused sources. After grant of regulatory consent, regulatory body continues to ensure safety/radiation protection by closely monitoring the operation of facility/ activity through the direct and indirect modes of safety verification, such as regulatory inspection, report on safety performance, radiological safety status, dose budgeting, event reporting, etc. The regulatory body has published various documents in the field of radiation protection. The regulatory body is conducting regular training courses on safety and regulatory measures for nuclear and radiological facilities. In addition, Theme meetings are conducted periodically on important safety related issues/ topics.
Conclusion: The regulatory safety framework for radiological protection in BARC is well established and delineated through the safety documents published by the regulatory body. The regulatory body has put a consistent effort for improvement of safety status of the associated facilities/activities and towards promotion of a well-balanced safety culture in the nuclear and radiological facilities.
References
Atomic Energy Act; 1962.Atomic Energy (Radiation Protection) Rules; 2004.Atomic Energy (Factories) Rules; 1996.
Abstract - 92409: Consideration of dose constraints for non-radiation workers at a nuclear site
Preetha Jayaprakash
BARC Safety Council Secretariat, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: As per the regulatory guidelines, 1 mSv/y is set as the dose limit that any individual, beyond the Site boundary of a nuclear facility, can receive, during its normal operation. The dose limits for the occupational exposures (above 18 yrs) is set as 20 mSv/y averaged over sliding scale of 5 years. Further dose limits for students and apprentices is set as 6 mSv/y. This article proposes for regulatory consideration of annual dose constraint required to be set under planned exposure situation for non-radiation workers within a Nuclear Site. At a nuclear Site, there are several groups of supportive workers like the administrative staff, temporary staff in the areas of cosmetic maintenance, gardening staff etc. and scientific staff working in its premises. Presently, there is no consensus on this group of worker who are neither categorized as an occupational worker nor a member of public. Such a group of non-radiation workers within the nuclear facility site boundary needs to be covered under the regulatory framework for safety. References for defining 'member of public' and dose limits/constraints: The statement given in IAEA GSR-Part 3 para 3.78 which is stated as “Employers, registrants and licensees shall ensure that workers exposed to radiation from sources within a practice that are not required by or directly related to their work have the same level of protection against such exposure as members of the public” implies that the non-radiation worker at Site needs to be considered as 'public'. The same is brought out in the AERB definition of 'member of public' which is stated as 'Any individual in the population except for one who is subject to occupational or medical exposure'. As per regulatory requirements, the dose constraints are set for various facilities in order to ensure that the public dose limit is not exceeded under normal conditions of operations. Sometimes it may happen that the maximum dose from releases from a particular facility occurs within the Site boundary. If multiple such facilities contribute to a higher dose within the Site boundary, question arises as to dose constraint that needs to be applied for the non-radiation workers within the Site. This sets a need to evaluate a theoretical prospective dose to these workers considering them as public or defining a separate dose constraint or dose target for these workers. United Kingdom's independent safety regulator ONR (Office for Nuclear Regulation) in its document 'Safety Assessment Principles for Nuclear Facilities' 2014 Edition Rev. 1, published in 2020, has brought out targets and a legal limit for effective dose in a calendar year for any person working on-Site from sources of ionising radiation. The Legal limit is set as 2 mSv/yr for employees other than radiation worker. The limit is kept slightly higher than 1 mSv possibly taking into consideration that there are no infants in the affected group. Dose Constraint set at Siting Stage: One of the basic criteria considered for site evaluation of a nuclear facility in India, is the possible radiological impact on the environment and the public. The dose constraint for a nuclear facility is approved by the regulatory body at the Siting stage ensuring that the public dose limit of 1 mSv/yr for the Site, is not exceeded. It is the dose to the public at Site boundary that is used to arrive at the dose constraint for the facility. Since there is no regulatory requirement to consider the applicability of dose constraints to the non-radiation worker within the Site, the impact on these workers may be overlooked.
Conclusion: Regulatory body apportions dose for multiple facilities at a Site considering the radiological impact due to releases from a facility. Though doses are evaluated at various distances from the facility, it is the maximum dose beyond the Site boundary that is considered for Dose apportionment. At times, the releases from facilities could be from a chimney or a stack whose impact could be lesser or greater at shorter distances, depending on the type of radionuclide. If the dose is higher at shorter distances, then the dose to the non-occupational workers at the nuclear Site needs to be looked into to comply with the IAEA GSR-Part 3 requirement. For this, the regulatory body may define a separate dose constraint for the non-radiation workers or, if they are to be considered as public, which follows a dose limit of 1 mSv/yr, then the prospective dose to the public within the Site also needs to be taken into consideration while apportioning dose to a facility. This would help in bringing the dose to non-radiation worker within the Site, under the regulatory ambit.
Keywords: Dose limits, non-occupational worker, occupational exposure, public
References
AERB. Safety Code for Site Evaluation of Nuclear Facilities. AERB.AERB. Safety Glossary 'Glossary of terms for Nuclear and Radiation Facilities and Associated Activities. AERB; 2022.IAEA. GSG-7, Occupational Radiation Protection. IAEA; 2018.IAEA. GSR-Part 3, Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. IAEA; 2014.Office of Nuclear Regulation, Safety Assessment Principles for Nuclear Facilities. 2014 ed., Rev. 1; 2020.
Abstract - 92418: Legal, regulatory and technical aspects of exemption and clearance of source (radioactive materials or radiation generating equipment) in India
K. Jolly Joseph, Soumen Sinha, P. K. Dash Sharma
Atomic Energy Regulatory Board, Niyamak Bhavan, Mumbai, Maharashtra, India
Email: [email protected]
Introduction: The sources (radioactive materials or radiation generating equipment) used for various applications in industry and medicine are regulated to protect the radiation worker, public and the environment from the harmful effects of radiation. Exemption is the deliberate omission of a source or practice from some or all aspects of regulatory control on the basis that the exposure (including potential exposure) due to the source or practice are too small to warrant the application of those aspects. Clearance is the process of removal of the regulatory control from the radioactive material or radioactive substance within the licensed facilities or activities (AERB Safety Glossary, 2022). Thus the concept of exemption and clearance provide framework for regulatory control of source commensurate with its radiological hazard and thereby enabling the effective utilization of regulatory resources. Clearance level facilitates effective reduction of volume of material disposed as radioactive waste, thereby reducing the cost involved in its management. This paper briefs about the legal, regulatory and technical aspects for the exemption and clearance of source in India. Legal and Regulatory aspects: Legal provisions of exemption for both radioactive substance or material and radiation generating equipment, devices or appliances emitting radiation is stipulated in the Rule 5 of the Atomic Energy (radiation protection) rules(RPR), 2004[1] promulgated under the enabling provisions of the Atomic Energy Act, 1962. As per this rule, the radioactive substance or materials and radiation generating equipment, devices or appliances emitting radiation are exempted from Rule 3 of RPR, 2004 if it is not exceeding the radiation level or the limit notified by the Central Government respectively. Thus the radiation levels for the radioactive substance or material and limit for the radiation generating equipment are to be notified by the central government in order to exempt them from Rule 3. In 02/02/2021, central government issued the notification S.O. 503 (E), authorizing Chairman, Atomic Energy Regulatory Board (AERB) to specify radioactivity levels in substance or material deemed to be a radioactive substance or radioactive material. Central government issued the notification S.O. 504 (E) in 2/02/2021, which authorized Chairman, AERB to determine the limit below which the use of the radiation generating equipment, devices or appliances emitting radiation are to be exempted. In 07/12/2021, AERB issued the safety directive (No.1/2021) specifying the radioactivity levels in substance or material deemed to be a radioactive substance or radioactive material. Exemption levels for moderate and bulk quantities of radioactive materials in terms of activity concentration and total activity are stipulated in this directive as regulatory conditions. Furthermore, the clearance levels for the bulk amounts of radioactive substance or materials in terms of activity concentration are also specified in this directive. AERB issued the safety directive (No.2/2021) in 07/12/2021, specifying the exemption criteria for the radiation generating equipment, devices or appliances emitting radiation. Technical aspects: The exemption and clearance criteria and levels stipulated in the safety directives are based on the International Basic Safety Standards for Protection of the International Atomic Energy Agency (IAEA GSR Part 3, 2014). The exemption and clearance levels recommended by IAEA GSR Part 3 for some of the major radionuclides were validated based on the Indian scenario of exposure situations and conditions before its inclusion in the safety directive following relevant IAEA recommendations and guidelines. Dose criteria for the exemption and clearance are also specified in the safety directives. Owing to the exempted or cleared source, effective dose expected to be incurred by any individual is of the order of 10 μSv or less in a year under all reasonably foreseeable circumstances and for low probability scenarios it does not exceed 1 mSv in a year. The limit for exemption of Radiation generators are such that in normal operating conditions the dose rate shall not exceed 1 μSv/h at a distance of 0.1 m from any accessible surface of the equipment or the maximum energy of the radiation generated is no greater than 5keV. The prescribed limit for exemption of equipment containing radioactive material is that in normal operating conditions the dose rate shall not exceed 1 μSv/h at a distance of 0.1 m from any accessible surface of the equipment.
Conclusion: The exemption and clearance of source in India is based on well-structured legal and regulatory framework and commensurate with international standards and guidelines.
Keywords: Clearance, directive, exemption, notification, source
References
Atomic Energy (Radiation Protection) Rules; 2004.International Atomic Energy Agency. Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. Part 3. Vienna: IAEA GSR; 2012.
Abstract - 92419: Radiological impact assessment for nuclear facilities for protection of the environment
Ritu Raj, S. Sinha1, S. P. Lakshmanan, S. K. Dubey
1Directorate of Radiation Protection and Environment, 2Directorate of Regulatory Affairs and Communication, Atomic Energy Regulatory Board, Mumbai, Maharashtra, India
E-mail: [email protected]
In order to meet the fundamental safety objective, the people and the environment need to be protected under all circumstances from harmful effects of ionizing radiation. This is achieved by complying with the general safety requirements which are established on basis of ten safety principles. However, historically, international community on radiation protection was of the view that standard of environmental control needed to protect people will ensure that other species are not put at risk. Nevertheless, recent international trends in this field shows an increasing awareness of the vulnerability of the environment and a need is felt that environment also should to be considered for optimization of protection and safety. Hence, it is necessary to show that the biota is protected through an explicit dose assessment. The main environmental protection objectives are to prevent or reduce the frequency of deleterious radiation effects in the environment to a level where they would have a negligible impact on the maintenance of biological diversity, the conservation of species, or the health and status of natural habitats, communities, and ecosystems. For assessing and controlling the effects of radiation on flora and fauna, ICRP's approach is based on the concepts of 'reference animals and plants', a 'representative organism' and criteria in the form of 'derived consideration reference levels' can be used. Reference animal and plants are identified from representatives of marine, terrestrial and freshwater ecosystems located in the area where radioactive effluents from operating nuclear and radiation facility are discharged. The dose rate is estimated for the assessment of the impact on flora and fauna. The reference area around the source (typically 100-400 km2) is so selected such that is sufficiently large to allow mixing of the effluents with the environmental media and the number of individuals of the species considered in the assessment is also suitably large. Activity concentrations in flora and fauna are estimated using the estimated source term for normal operation, environmental dispersion, transfer models, information on the times spent by the different species in different habitats (e.g. on or above soil, in the water, in aquatic sediments). Dose rates from internal and external exposures of reference animals and plants relevant for the ecosystems under consideration are estimated. In order to assess the impact on the non-human biota, Derived consideration reference levels (DCRL) are used. These are a set of dose rate bands within which there is either no evidence or only some evidence of damaging effects of ionizing radiation on individuals of the species. For dose rates below the lower bound of the bands, no effects have been observed or no information on effects is available. DCRL do not represent limits; rather, they are points of reference. When the estimated doses to the representative organisms are within the band or close above the upper bound of the band, the radiological situation can still be considered acceptable. However, such a result would likely warrant a closer examination of the possible impacts on the environment, which would need to take account of a number of factors. Typical value of derived consideration reference levels are in the range of 0.1-100 Gy/day for animal and plant. This paper discusses the important aspects on methodology and approach for protection of environment for addressing aspect related to radiological impact assessment for Flora and Fauna.
Keywords: Biota, dose, environment, reference level
References
ICRP. Dose Coefficients for Nonhuman Biota Environmentally Exposed to Radiation. ICRP Publication 136; 2017.ICRP. Environmental Protection – The Concept and Use of Reference Animals and Plants. ICRP Publication 108; 2008.IAEA. Prospective Radiological Environmental Impact Assessment for Facilities and Activities. Vienna: IAEA GSG-10, IAEA; 2018.
Abstract - 92428: Regulatory review of radiation safety aspects of 700 MWe KAPP-3 commissioning for improvement in radiation safety aspects in upcoming 700 MWe NPPs
J. Mandal, N. Khandelwal, S. K. Pawar
Directorate of Radiation Protection and Environment, Atomic Energy Regulatory Board, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Kakrapar Atomic Power Project (KAPP-3), the first indigenous 700 MWe nuclear reactor of India, achieved first criticality on July 22, 2020. Subsequently unit was operated up to 50% FP (Full Power) to carryout phase C commissioning activities. In this paper we discuss, authors experience of Regulatory review of Radiation Safety Aspects of KAPP-3 (700 MWe) commissioning so as to confirm regulatory compliances and use the experience gained for improvements in Radiation Safety at upcoming 700 MWe NPPs in India.
Review Methodology: The Radiological safety aspect of KAPP-3, specifically with respect to initial fuel loading, low power operation, radiological works carried out in the control area, collective dose and individual exposure, radiological conditions, adequacy of the radiation shielding, radioactivity in the nuclear systems and effluent releases were studied and compared with the same of TAPS-3&4. The observations and areas for improvement based on the study are identified for enhancing the radiological safety at KAPS-4 and upcoming 700MWe NPPs.
Observations and Discussions: (a) Collective dose for initial fuel loading in KAPP-3 was 22 p-mSv and in TAPS-3 was 21 p-mSv, this is comparable (b) Dose rates on the DDU (Deeply Depleted Uranium) bundles are 5 to 10 times higher than Nat. U and DU bundles (c) In initial three years of operation, TAPS-3&4 completed 415 FPD (Full Power Day) and KAPP-3 completed 31 FPD. KAPP-3 FPDs were less due to modification work based on the commissioning experience. During this period dose consumption at KAPP-3 was 600 p-mSv (50% dose in modification) and TAPS-3&4 was 1256p-mSv (40% dose in U-4 BSD) (d) Accessible area radiation field at 50% FP has shown a considerable decrease at KAPS-3 as compared with TAPS-3&4 due to enhanced shielding at HFU (Horizontal Flux Unit) (sr. No.7), LZC (Liquid Zonal Control) (Sr. No. 12&13) & D2O lines (sr. No.9). (e) In KAPP-3, the thickness of End Shield has been reduced to 920 mm compared to 1120 mm in TAPS 3&4 to accommodate feeder of higher diameter. The source of radiation field in the FM vault in KAPP-3 is increased due to core radiation and volume source from feeder & header cabinet. At 50% FP, the radiation field in FM vault by RADAS (Radiation Data Acquisition System) was 15-20 mSv/h at TAPS-3&4 and KAPP-3 it is 30-40mSv/h. Thus increase by two times at 50% power level (f) in pump room (115m ele.), additional shielding has been provided for SG, PCP casings as compared to TAPS-3&4. Radiation data at 115 m elevation in pump room at 50% FP was 20-25 mSv/h by RADAS at TAPS-3&4 and radiological survey data at KAPP-3 pump room indicated rad. field in the range of 0.2 to 10 mSv/h with maximum on AR (Adjuster Rod) cooling line (5-10mSv/h) (g) The PPP (Primary Pressurizing Pump) and F/M pump strainers are shielded by lead ball filled shielding box at KAPP-3, the dose rates are as per design values (h) AGMS (Annulus Gas Monitoring system) delay tank has been eliminated by providing delay in tubes in FM vault. In TAPS-3&4 the AGS system field at 50% FP was observed as 0.20-0.30 mSv/h and KAPP-3 it was 5-7 μSv/h at 50%FP (i) Theoretical tritium activity build up in Moderator system and PHT (Primary Heat Transfer) system is expected to be in similar order. The tritium activity build up in the PHT and moderator system is required to be studied further for least up to 500FPD to compare the tritium activity build up rate with TAPS-3&4 (j) Cobalt based and Vanadium based SPNDs were used in TAPS-3&4. The surface dose rate of irradiated cobalt SPNDs (Self-power Neutron Detector) was about hundreds of Sievert per hour resulting in increase of collective dose for handling of these cobalt SPNDs. In KAPP-3&4, the SPNDs are Inconel & Vanadium and the high surface dose rate at SPNDs due to cobalt element could be eliminated.
Keywords: Collective dose, radiation field, radiation shielding, radiological safety
References
TAPS-3&4 Collective Dose Budget Proposal and HP Report (for the year 2005 to 2007).KAPP-3&4 Collective Dose Budget Proposal (for the year 2020 to 2022) and HP Report (for the year 2020 to 2021).KAPP-3&4 Safety Analysis Reports, Section-9 and 10; April, 2018.
Abstract - 92477: Proposal of regulation framework for large-scale particle accelerator and legislation in the Korean Nuclear Safety Act
Hee-Seock Lee, Nam-Suk Jung, San Sun Han1
Pohang Accelerator Laboratory, POSTECH, Pohang, 1Korean Institute of Nuclear Safety, Daejeon, South Korea
E-mail: [email protected]
In the old Korean Nuclear Safety Act, the radiation-related regulation had concerned mostly radioactive isotopes or X-ray generating devices. Even though the number of small-scale particle accelerators like medical accelerators has increased, the main framework was not changed so much. However, the constructions and operations of large-scale particle accelerators have continued since 1994 when the first large-scale particle accelerator, Pohang Light Source, was constructed. Now several large-scale particle accelerators including proton therapy machine are operated or in construction. Therefore, the weak points in existing Nuclear Safety Act or the limit of rule application were disclosed. We have studied to make practical regulation framework based on the scientific results. Some of the results were introduced at IRPA15 in Seoul.[1] New regulation framework of large-scale particle accelerator was proposed to Korean regulation body. First the base concept to classify the accelerators considered the amount of secondary radiation and radioactive materials produced by primary beam, which intensity is also important. The [Figure 1] shows the beam power level of many particle accelerators in world. The beam power is directly to determine the production level of secondary radiations. In 2021, the first article was inserted to Korean Nuclear Safety Act for separating large-scale particle accelerator from the regulation framework of general radiation-generating devices. And it was approved by Korean Congress after moderating the frame a little. It will take effect in next March. At present, the lower statutes are under discussion. It will be completed before next March. This study will be done continuously for regulatory guide and other protection practices. The classification idea of particle accelerators was suggested as shown in [Table 1].[1] The [Figure 2] is the example of domestic accelerators classified by proposing new regulation frame. There is some discrepancy between contents proposed in [Table 1] and one in [Figure 2]. It was discussed seriously. In this paper, we will introduce the scientific results which new framework were based on. The lower statutes will be presented with the important factors which are used to determine new regulation articles. This work was supported by the Nuclear Safety Research Program through the Korea Radiation Safety Foundation funded by Nuclear Safety and Security Commission.{Figure 3}{Table 1}{Figure 4}
Keywords: Large-scale particle accelerator, nuclear safety act, regulation
Reference
Nam-Suk J, Hee-Seock L, Arim L, Sang Eun H. J Radiol Prot 2021;41:S150-9.
Abstract - 92557: Prescribed limit of uranium in drinking water in India – whether to follow WHO guideline value or case of a site-specific scientific research
S. K. Jha1,2, S. K. Sahoo1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Uranium is an emerging geogenic and insidious contaminant in drinking water and a greater concern for the members of the public, policy makers and government. Its occurrence in groundwater is due to natural weathering process of rocks and soils associated with the primordial radionuclide since the evolution of the earth. Crustal average concentration of natural uranium is around 3 mg/kg (ppm) which is major source of natural uranium in groundwater, besides few anthropogenic sources. Variation of uranium concentration over a wide range in groundwater is a function of multiple physico-chemical parameters of both host rock and interacting water media. It is, therefore, a guideline value or prescribed limit is necessary for drinking water to avoid any potential adverse health impact on the members of the public. World Health Organization (WHO) recommended a guideline value of 30 μg/l (ppb) for uranium on the basis of chemical toxicity while 10 Bq/l (800 μg/l) for 238U and 1 Bq/l (12.6 μg/l) for 234U on the basis of radiological toxicity.[1] WHO guideline values on chemical toxicity and radiation dose consideration still a global case of investigation. In addition to this, US EPA recommended a primary limit of 30 ppb for uranium in drinking water while there is a divergent limit for uranium in drinking water across the countries i.e. 2 μg/l (ppb) in Japan to 964 μg/l (ppb) in Czech Republic.[2] Atomic Energy Regulatory Board, India has recommended a limit value of 60 ppb for uranium in drinking water on the basis of radiological dose criteria and appropriate safety factor. Of late, Bureau of Indian Standards has promulgated the acceptable limit value of 30 μg/l (ppb) for uranium in drinking water,[3] adoption of WHO guideline value. Concerns raised about the scientific validity of the stipulated limit and meeting its purpose considering the heterogeneity in physiology, water ingestion rate and other underlying attributes. Moreover, unlike the WHO guideline value and US EPA Limit background document and methodology for derivation of the limit value, no scientific or technical literature about the presumptions and methodology adopted in deriving the limit is available in the public domain. The standard water intake rate of 2 litres per day which is more appropriate to western countries might not be valid in a tropical climate like India. ICMR-National Institute of Nutrition recommended water requirement for various ages and gender on the basis of energy requirement. For adult man and woman with moderate activity, the recommended total water requirement per day is 4.05 litre and 3.2 litre, respectively. For children of age group 7-9 y, the recommended total water requirement per day is 2.5 litre while for boys and girls in the age group 10 – 18 years, the total water requirement per day is greater than 3 litres.[4] Similarly, the variation in body weight, physiology of Indian population, socio-economic parameters and lifestyle from the western countries contribute to uncertainty to the tolerable daily intake, water intake through drinking water pathway and the derived limit value. It is also highly necessary to study synergistic effect of both chemical and radiological toxicity to develop a nationally or globally consensus single guideline/limit value like limit on radiation dose to members of the public irrespective of the variation of regional/local attributes. For protection of members of the public, a robust and conservative limit is needed considering all appropriate and representative input parameter for the derivation of the national limit on uranium in drinking water.
Keywords: Drinking water, geogenic, groundwater, limit, uranium
Acknowledgment
Authors wish to thank Dr. D. K. Aswal, Director, Health Safety and Environmental Group for his guidance. The authors also acknowledge the support received from colleagues of UCIL management.
References
WHO. Guidelines for Drinking Water. 4th ed. WHO: 2017.Sahoo SK, Jha VN, Patra AC, Jha SK, Kulkarni MS. Environ Adv 2020;2:100020.BIS, IS 10500: 2012: Amendment No. 3. Drinking Water – Specification. BIS; 2021ICMR-NIN Expert Group on Nutrient Requirement for Indians. Recommended Dietary Allowances (RDA) and Estimated Average Requirements (EAR); 2020.
Abstract - 96155: Industrial safety management towards zero incident initiative at nuclear facilities
Praveen Dubey, Nitin Choughule, Alok Srivastava
Industrial Hygiene and Safety Section, BARC, Mumbai, Maharashtra, India
Email: [email protected]
Front end and Back end nuclear facilities alongwith radiological hazards are having plenty of non-radiological hazards, which are equally important to address. The nuclear facilities may contain hazardous processes and materials. For example, as Non radiological hazards may include physical hazards (Noise, heat stress, illumination, electricity), chemical hazards and mechanical hazards. Occupational workers normally come across other work place related hazards such as poor housekeeping, fire, transport, environment related issues. Non-radiological accidents have direct impacts on the individual involved. They also negatively impact the image of nuclear facilities and their general acceptance by the public. The importance of Industrial Safety has been recognized and addressed by IAEA. Identification of workplace hazards and its control requires sensitive handling from both Workers and Managers. Knowing about hazards and awareness of work related risks, good safety culture and safety motivation are few of the important key parameters in an accident prevention programme. The effectiveness of the accident prevention programme depends on the way the organization inculcates a sense of safe working among all levels to attain the organization's safety goals i.e., vision zero incident at workplace. In BARC, Industrial Hygiene and Safety Section plays a very crucial role in industrial safety management, which begins with the design, planning, procurement, contracting, workplace monitoring, etc. The involvement and participation of employees is boosted by training and a thorough orientation in the work processes. Supervision and control are also important tools. As a premier institute and suffice with many complex activities, a high degree of commitment to safety by management and rigorous and pro-active measures are essential to achieve zero incidents. Self-regulation through safety management system is a proactive and positive attempt to improve safe records. Work-related accidents on sites can happen for a number of reasons. Such accidents can lead to stress, fatigue, illness, distraction and subsequent onsite injury. By fostering a zero-accident safety culture through continuous safety improvements and safe practices will help employees reduce their stress experience and fewer work-related accidents. Inputs from outcome / root causes of accidents provide suggestions which helps to move closer to Zero-Accident initiatives. An element of such initiatives include: (a) A safety policy which states the commitment to safety and health at work. (b) A structure to ensure implementation of the policy. (c) Training to personnel with knowledge of safety. (d) In-house safety rules to provide instruction for achieving safety management objectives. (e) Safety inspections at regular intervals to identify hazardous conditions and for the rectification. (f) A programme to identify hazardous exposure or the risk of such exposure to the workers and to provide suitable personal protective equipment. (g) Investigation of accidents or incidents to find out the cause of any accident or incident and to develop prompt arrangements to prevent recurrence. (h) Emergency preparedness to develop, communicate and execute plans prescribing the effective management of emergency situations. (i) Safety committees to provide a forum for all those involved to discuss issues and pass on information. (j) Promotion, development and maintenance of safety and health awareness in a workplace. Workplace specific risk assessment and risk management are crucial in achieving the goal of accident-free workplace. Improving safety is a never-ending task: even when performed well, it is never finished. In nutshell, we have discussed how BARC is committed to ensure the highest standard of OHS in all activities undertaken in order to achieve zero-accident working period and vision zero incident concept.
Keywords: Accident prevention programme, non-radiological hazards, safety culture
References
Nuclear Series (NP-T-3.3 “Industrial Safety Guidelines for Nuclear Facilities“).Organizational behaviour and safety management in the construction industry. Constr Manage Econ 1989;7:303-19.Pidgeon NF. Risk assessment, risk values and the social science programme: why we do need risk perception research. Reliab Eng Sys Saf 59:5-15.
Abstract - 96224: Evaluation of whole body vibration and assessment of degree of comfort
Garima Singh, M. D. Patel, G. L. N. Padmavathi, Nishith Ghosh, G. Nagaraju, Alok Srivastava
Industrial Hygiene and Safety Section, Health Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: In road transport, occupational heavy motor vehicle drivers are affected by vibrations during their working shifts. The intensity of WBV depends on the bus type, the quality and the type of the road surface and the bus speed.
The indicator of exposure to vibrations is frequency-weighted root-mean-square acceleration value for 8 hours daily exposure, A (8). As per Model Factories Rules 120 (MFR 120) under Section 87 of the Factories Act, Schedule-XXVII–Operations involving High Noise and Vibrations Levels -the daily exposure A(8) limit value is 1.15 m/s2 for whole body vibration. Automotive Industry Standard AIS 153: Additional Requirements for Bus Construction specifies use of tri-axial seat pad accelerometer for measuring the vibration level at driver and passenger seat locations. Accelerometer should capture at least frequency range 0.4 Hz to 80 Hz for human health, comfort and perception. Further it is stated to refer to IS 13276-1: 2000 Mechanical Vibration and Shock - Evaluation of Human Exposure to Whole Body Vibration: General Requirements (adoption of ISO 2631-1:1997). ISO 2631-1 standard defines a frequency range between 0.1 and 80 Hz for the WBV evaluation and given below are the approximate indications of the relation between acceleration level and the degree of comfort [Table 1]. According to the ISO, the duration of measurement of human exposure to vibration should be sufficient to ensure statistical precision and to ensure that the vibration is typical of the exposures which are being assessed. In general, a minimum of 30 minutes should be recorded and analysed according to the ISO procedure.
Results and Discussions: This paper summarises the findings of the study that was conducted to check that the values are within the prescribed safe limits. Evaluation was carried out according to the procedure and criteria prescribed by International standard ISO 2631-1 (1997). All the buses were of same make and were moved on a fixed route at average speed of 35 km/h, vibration was monitored for minimum duration of 30 minutes. The effect of vibration on the comfort level for the same road conditions was analysed. It was observed that overall, the driver was found to be safe as per ISO 2631-1 but the comfort levels were often exceeded; most of the values have the level of comfort in 'uncomfortable' range. This is mainly due to roads having speed bumps, though safe with respect to health, this may decrease the driver's performance and ability to control the vehicle. So the necessary action should be taken through interventions such as using lumbar support, armrests, suspension seats, shock absorbers, maintenance-lubrication of moving components, repairing roads etc. can help reduce whole-body vibration.{Table 2}
Keywords: Back pain, heavy motor vehicles, ISO 2631, whole body vibration
References
Palmer KT, Bovenzi M. Rheumatic effects of vibration at work. Best Pract Res Clin Rheumatol 2015;29:424-39.The American Conference of Governmental Industrial Hygienists (ACGIH). Threshold Limit Values (TLVs) and Biological Exposure Indices (BEIs). Cincinnati: ACGIH; 2019. p. 206-12.Palmer KT, Griffin MJ, Bendall H, et al. Prevalence and pattern of occupational exposure to whole body vibration in Great Britain: Findings from a national survey. Occup Environ Med 2000;57:229-36.
Abstract - 96225: Methods and practices for industrial hazards prevention in radiation facilities
G. L. N. Padmavathi, Garima Singh, G. Nagaraju, Kailash Gharat, M. D. Patel, Alok Srivastava
Industrial Hygiene and Safety Section, Health Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: A comprehensive plan for the industrial accident prevention at BARC - Accident Prevention Program was started in 1962. Thereafter, BARC Safety Council was created in 2000 through a presidential order as per Section 23 (Administration of Factories Act, 1948) of the Atomic Energy Act, 1962 to enforce the provisions of the Factories Act, 1948 and the Atomic Energy (Factories) Rules, 1996 in BARC.
Results and Discussions: BARC Safety Council, the apex regulatory body advises on safety policies, issues directives concerned with safety, safety matters, ensures compliance of safety directives, and promotes good safety culture through feedback from Industrial Hygiene and Safety Section of HS&EG. A positive safety culture can be described as a preventative health and safety culture – where the principle of prevention is accorded the highest priority. Hazard Control methods can be Engineering, Administrative or Personal Protective Equipment (PPE). Engineering controls consist of a variety of methods for minimizing hazards that include elimination, substitution, process control, enclosure and isolation, and ventilation. Elimination is the most preferred method of controlling hazard. When, elimination is not feasible, substitution is the best approach to hazard mitigation. Whenever possible, less hazardous agents are substituted in place of their more hazardous counterparts. This also applies to conditions and activities e.g. substituting toluene for benzene, non-lead-based paints for lead-based ones. Process controls involve changing the way a job/activity is performed in order to reduce risk. Examples of this include using wet methods when drilling or grinding or using temperature controls to minimize vapour generation. Enclosure and isolation aim to keep the occupier away from hazard. Glove boxes are a good example of enclosure and isolation. Interlock systems for lasers and machinery are other good examples of isolating processes. The most common method for ventilation is exhaust systems like fume hoods. Administrative controls alter the way in which work is performed. They may consist of policies, training, standard operating procedures/guidelines, personal hygiene practices, work scheduling, etc. These controls are meant to minimize the exposure to the hazard and should only be used when the exposure cannot be completely mitigated through elimination/substitution or engineering controls. PPE should always be used as a last line of defence and is an acceptable control method when engineering or administrative controls cannot provide sufficient protection. Workers' safety and health is to be ensured by not only prescribing mandatory safety standards that are to be observed but they also require facility authorities to positively take proactive measures for ensuring workers' safety and health.
Some proactive industrial safety aspects are-
Periodic noise level measurementPeriodic illumination level measurementPeriodic testing of material handling equipmentsPeriodic testing of fire extinguishers & fire detectorsRoutine functional checking of eye wash fountainCompliance of AEFR 1996Monitoring of temperature, pressure, flow etc.Adherence to safety guidelines
Other proactive/reactive steps in the field of Industrial hazards prevention are Safety Framework, Safety Performance Index, Hazard identification measures, dissemination of Safety Information, Recording of Industrial Statistics including Injury On Duty (IOD) cases, Accident Investigation and Analysis, Corrective Measures, imparting Safety and Health education to workers and refresher training programmes along with training when new processes are introduced.
Keywords: Accident prevention programme, hazard investigation, safety review
References
Alli BO. Fundamental Principles of Occupational Health and Safety. 2nd ed. Geneva: International Labour Office; 2008.IAEA. Industrial Safety Guidelines for Nuclear Facilities, International Atomic Energy Agency (IAEA) Technical Document, IAEA Nuclear Energy Series No. NP-T-3.3. Vienna: IAEA; 2018.
Abstract - 96231: Industrial hygiene surveillance practices for beryllium workplaces and role of re-suspension factor
Mahesh K. Kamble1, Munish Kumar1,2, Ankur Chauhan1 and Alok Srivastava1
1Industrial Hygiene and Safety Section, Health, Safety and Environment Group, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Due to its remarkable physical, mechanical and nuclear properties, Beryllium (Be) finds wider applications in nuclear, space, defense and various items of day to day use. However, being regarded as one of the most toxic element of the periodic table, its handling without proper safety, training and care can lead to serious health hazards. Inhalation is the major route of exposure for Be and its compounds as dust particles associated with various processes involved during manufacturing/processing are of the range from (<1-100)μ or so, which can be deposited in the lungs and may lead to chronic beryllium disease (CBD) and acute beryllium diseases. Further in case of skin exposure, irritations/allergic reaction to skin has also been observed. Also, IARC has classified Be as confirmed carcinogen to humans. Hence, industrial hygiene surveillance (which includes air concentration & surface contamination monitoring etc.) plays an important role to ensure that the levels at workplaces are maintained within the prescribed values. To prevent Be inhalation, proper ventilation (15 air changes/hour for a medium level facility) & maintaining suitable negative pressure is must in addition to process automation, appropriate containment, proper housekeeping, education & training of workers. Further use of PPEs also plays an important role to eliminate potential health hazards. In addition, dedicated medical program for occupational workers includes annual medical examination of vital body parameters, chest X-ray & pulmonary function test. Also, some of the facilities world-wide perform Be-LPT & monitoring of blood, urine etc. for Be workers. It is noteworthy that for Be, there is no direct monitoring instrumental method practiced although laser induced breakdown spectroscopy (LIBS), (γ, n) reaction based detector & mass spectrometers have been attempted. Hence, fluorometry technique is quite popular for routine monitoring although ICP or GF-AAS based methods are also available (Kumar et al., 2021). Present paper deals with various industrial hygiene practices & reports results on re-suspension factor, K and its role in facilities where Be dust/particles may be present. This is required as airborne concentrations are often calculated by assuming that the contributions due to re-suspension caused during local disturbances from contaminated equipment's are negligible. Further comparison with re-suspension factor values reported in literature for heavy elements like U, Pu etc. is also given. Ideally, ventilation system carries away the Be dust particles but is an ideal situation as local disturbances lead to increased concentration in certain areas especially during maintenance, refurbishment or related activities. For the re-suspension study of Be, areas where possibly of situations leading to air borne concentration were identified and air samples were collected for a given work shift using 0.8μ mixed cellulose ester filter paper along with dust samples by placing 11 cm dia, filter paper at a height representing breathing zone. Measurements were also performed at ground level for surface samples to estimate contribution towards ground level contamination. Samples were leached in 0.1N H2SO4 for 2 hours & after addition of suitable chemicals viz. NaOH to maintain ph (11-12), Na4P2O7 as masking agent & Morin dye, samples were analysed using fluorometry technique.[1] K factor (m-1) was estimated as a ratio of air concentration (μg/m3) to surface contamination (ng/cm2) levels. In the present study, the mean & median values for - i) air concentration were (0.0132 & 0.0036)μg/m3 with range of (0.002-0.0483)μg/m3 & ii) surface contamination levels were (2.79 & 1.79)ng/cm2 with range of (0.52-9.74)ng/cm2. The typical K factor was found to be in the range (10-3-10-5)/m with a mean value of 10-4/m and is in agreement with the values reported in literature albeit with some deviations.[2] From literature, it is also observed that the typical value of re-suspension factor for Be workplace is at least an order of magnitude greater than those i.e. (10-5-10-6)/m reported for radiological facilities handling U, Pu etc.[3] The study indicates importance of local exhaust, proper housekeeping in addition to use of PPEs to minimize the Be exposures and associated risks. It was also observed that wet mopping reduces contamination level by a factor of 5-10 which helps in reducing air concentration.The study shows that the re-suspension factor, K plays has an important role in determining the Be exposures & practices like repeated (wet) cleaning of surfaces, equipment, floor, efficient ventilation & local exhaust may help in minimizing the Be hazards & achieving the revised PEL of 0.2μg/m3 as given by OSHA.
Keywords: Air sampling, estimation of concentration, inhalation and re-suspension factor, toxicity
References
Kumar M, et al. Procedure 2nd DAE Symposia. CTAC; 2021. p. 58.Sansone E. Treatise on Clean Surface Technology. Boston: Springer; 1987. p. 261-90.Brodsky A. Health Phys 1980;39:992-1000.
Abstract - 96521: Overview of ventilation as an engineering control measure in radiation facilities
Aparna R. Sawatkar, J. D. Sharma, Nishith Ghosh, V. K. Pallavee, Alok Srivastava
Industrial Hygiene and Safety Section, HS&EG, BARC, Mumbai, Maharashtra, India
Email: [email protected]
Introduction: In BARC, radiation safety control measures are implemented in a cradle to grave manner i.e. from processing of the uranium ores to reprocessing of the spent fuel and radioactive waste management. Some of the chemicals are used for treatment of radioactive material are hazardous. Ventilation is the most crucial measure to keep the airborne concentration of radiological as well as chemical contaminants within the specified occupational exposure limits. It was observed through Industrial Hygiene Surveillance & Regulatory Inspections conducted by various Committees, that the ventilation system installed needs some alterations and/or are due for periodic maintenance. This paper gives an overview of critical aspects for effective design/ planning of this Engineering Control Measure.
Nature of Hazards: In a workplace, ventilation is used to control possible over exposure to air borne contaminants, possible risk of fire or explosion from flammable gas or vapour levels at or near the lower explosive limit (LEL), or degraded indoor air quality, & also it is crucial to keep Heat Stress of the personnel within recommended limits, Therefore, efficacy of the ventilation system needs to be reviewed periodically.
Observations and Recommendations: Effectiveness of the ventilation system depends on location of the supply and exhaust ports. Ideally the exhaust ports and supply ports are installed opposite to each other. But, special consideration is required while mounting/ installing location of the exhaust ports, if chemical handled/ stored in the area is heavier than air. Exhaust mounted at higher level may not reduce the airborne chemical concentration, effectively. In such case, air needs to be exhausted by mounting the exhaust near ground level. In Radioactive Facilities, once through ventilation is provided. The air is moved from white area to amber area and then after giving proper treatment exhausted through the stack. The flow of supply air and exhaust air is adjusted accordingly to maintain positive or negative pressure required in the specific areas. But, if a Supply port with high velocity is placed in the vicinity of the entrance inside a radioactive laboratory, the air pattern may reverse. The direction of air flow pattern is checked by performing smoke test and recommendations are provided for proper flow balancing. Local exhaust system is used to control air contaminants by trapping them at or near the source area before dilution into the workplace ambient air. The size and shapes of hoods are designed for specific tasks or situations. The air speed (velocity) at the hood opening and inside the hood must be enough to capture and carry the air contaminants. To achieve highest effectiveness, the hood should surround or enclose the source of contaminant or be placed as close to the source as possible. This system should be isolated from general ventilation system. For Local Exhaust hood minimum velocity of 0.75m/s is required which is difficult to achieve when the duct is incorporated in the General Exhaust System. Also, for Fume hood, the Face Velocity needs to be adjusted in the specific range of 0.5 – 0.75 m/s, as higher face velocity may cause turbulence due to formation of eddies which may result into escape of air into the work place from the fume hood. This goes without saying that, achieving the required number of Air Changes only on the basis of Fume hood Face Velocity may not be justifiable. Ventilation balancing becomes very critical and difficult for the Facility, when the Fumehood Exhausts are connected to the General Exhaust System. It is desirable to keep the Fume hood Exhaust System, to be separated from General Exhaust System. Clogged ports, filtering equipment and ducts adversely affect the overall performance of the ventilation system, so periodic maintenance is vital. Also, presence of high-speed rotating fans and high resistance created by fluid movement through ducts, leads to exposure to high noise due to vibration attributed to improper installation or inadequate maintenance of the system.
Conclusion: Effectiveness of engineering control measures depend upon the proper installation and regular maintenance of every devices utilized in the system. Lack of which may lead to overexposure of the personnel and deterioration of the health conditions. While pursuing the best practices available, awareness towards safety conditions will help to develop prudent working conditions.
Keywords: Air Changes, airborne contaminants, control measures, industrial hygiene, ventilation
References
ACGIH. Industrial Ventilation: A Manual of Recommended Practice for Design. 30th ed. ACGIH; 2019.Industrial Ventilation. Canadian Centre for Occupational Health & Safety; 2022.
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