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Year : 2023 | Volume
: 46 | Issue : 5 | Page : 37--119
Theme 2. Radiation Safety and Protection in Nuclear Facilities
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Abstract - 21235: A decade of radiological monitoring at beach sand mineral separation plant Chavara
Jaison T. John1,2,3, R. Sujata1,2,3, S. K. Jha3,4, M. S. Kulkarni3,4
1Health Physics Unit, IREL Udyogamandal, Kochi, Kerala, 2HS&E Group, BARC, 3HS&E Group BARC, 4HBNI, Deemed University, Mumbai, Maharashtra, India
E-mail: [email protected]
The predominant industrial process which require handling of Naturally Occurring Radioactive Materials (NORM) are processing of Beach Sand Minerals (BSM).The Chavara Mineral Division of IREL (India) Ltd is situated in the south west coast of Kerala (8.98210 N & 76.52600 E) within the Natural High Background Area of Kerala, India. Mining of beach sand minerals and processing the raw material to produce Ilmenite, Rutile, Zircon and its products, Sillimanite and Monazite has been continued in this Mineral Separation Plant. The Monazite Upgradation Project (MUP) was also operational and continued to produce upgraded monazite of +96% purity. The radiologically significant component in BSM mining & milling facilities is Monazite - the principal ore of Rare Earths and Thorium. Monazite contains around 0.35% uranium as U3O8 and 9% Thorium as ThO2 with Th-232 series activity concentration in the range 40–600 Bq/g. Zircon also contains uranium and thorium in the crystal lattice. Monazite contains thorium and uranium chain radionuclides in secular equilibrium and do not exhibit any appreciable change in activity with respect to time [AERB, 2013]. Regulatory mandate to perform continuous radiological monitoring of occupational workers in such a front end nuclear fuel cycle facility is the responsibility of Health Physics Unit under HPD, BARC. The hazards in BSM facilities can be both external and internal. External hazards are due to high energy beta-gamma rays of thorium series and internal hazards are mainly due to alpha emitting radionuclides deposited inside the body. External component of occupational radiation dose is monitored through TLD-Chest cards to each radiation workers – issued, controlled, measured and accounted on quarterly basis. This is carried out in co-ordination with TLD Unit under BARC at NFC Hyderabad. Occupational radiation workers varied from 60 to 92 during 2012-2021. The committed effective dose to each radiation workers, from the intake of radionuclides through inhalation pathway, in each year is estimated using measured short-lived Thoron (220Rn) progenies in milli Working Levels (20 mSv =1000 mWL) and long-lived 232Th (DAC= 20 mSv= 220 mBq.m-3) in mBq.m-3 from the airborne particulate samples while applying individual occupancy period. The estimations are carried out as per the standard protocol [based on IAEA Safety Report Series 68, 2011] from annual average levels in air particulate samples of about 250 numbers collected from different plant locations. Analysis of occupational dose data of IREL Chavara during 2012- 2021 [Figure 1] shows that the institutional dose in Person Sv, including external and internal components of regular and contract workers, varied from 0.018 (in 2021) to 0.046 (in 2014) with an average of 0.031±0.010. The average annual individual dose has varied from 0.20 mSv (in 2020) to 0.60 mSv (in 2014), with average of 0.39±0.12 mSv, which is <2% of the limit of 20 mSv. The maximum individual dose recorded in a year varied from 1.37 mSv (in 2012) to 4.77 mSv (in 2014), which is <25% of the annual dose limit. IAEA-SRS.49, 2006 says that the available data from BSM/Monazite based industry suggest that the average dose received by a worker is in the range of 1–8 mSv/annum, with higher exposures being possible where controls are inadequate. Contribution of internal dose in total individual dose, due to short-lived Thoron progenies and long-lived 232Th, is found to be varying from 17% to 53% with an average of 34±10%. The average internal dose varied from 0.086 to 0.24 mSv with an average of 0.13±0.05 mSv per annum. The occupational dose records in the period of 2012-2021 implies, IREL Chavara is consistent to implement ideal levels of occupational dose control practices, engineering safety measures and effective worker-management co-ordinations to maintain the occupational dose below the industry average.{Figure 1}
Keywords: Beach sand minerals, monazite, naturally occurring radioactive materials, occupational dose
References
AERB Safety Guidelines NO. AERB/FE-FCF/SG-5. Radiological Safety in BSM and NORM. Mumbai: AERB Safety Guidelines; 2013.IAEA. Safety Reports Series No. 68. Vienna: IAEA; 2011.IAEA. Safety Reports Series No. 49. Vienna: IAEA; 2006.
Abstract - 21236: Study on radiological aspects in inland mining and refilling by BSM Industry Unit, Chavara
Jaison T. John1,2, R. Sujata1,2, Abinash Sahu2,3, Ajesh Kumar1,2, S. K. Jha2,4, M. S. Kulkarni2,4
1Health Physics Unit, IREL, Udyogamandal, Kerala, 2HS&E Group, BARC, 4HBNI, Deemed University, Mumbai, Maharashtra, 3Health Physics Unit, OSCOM, Berhampur, Odisha, India
E-mail: [email protected]
Sand mining is a practice that is used to extract sand from various environments such as beaches, dunes, inland, dredged from ocean and river beds of interface regions. This raw sand along the coasts of India contains large reserves of heavy minerals - ilmenite, rutile, zircon, sillimanite, monazite, etc. The industrial process of beneficiation for these Beach Sand Minerals (BSM) aka Heavy Minerals (HM) requires raw sand as the input material. The Chavara Unit of IREL (India) Ltd situated in the south west coast of Kerala (8.98210 N & 76.52600 E) is a DAE unit involved in HM beneficiation. Mining of inland area having raw sand rich with HM is also practiced by this unit. The radiologically significant component in BSM mining area is largely Monazite, a Naturally Occurring Radioactive Material (NORM) - the principal ore of Rare Earths and Thorium. Monazite contains around 0.35% uranium as U3O8 and 9% Thorium as ThO2 with 232Th activity concentration in the range 40–600 Bq.g-1. Zircon is also a NORM material that contains uranium and thorium in the crystal lattice with 0.5-1.0 Bq.g-1 of 232Th and 1-5 Bq.g-1 of 238U. The typical raw sand composition used by BSM units like IREL Chavara is 10-40% Heavy Minerals and 55-80% quartz and rest is shells & organic waste material. The area used for inland mining by the beneficiation units like IREL Chavara is later restored with the tailings of the unit, which is >90% of quartz/silica, to the prevailing topography of the area. The area is then refurbished with locally adaptive plantation and later released to the eligible local community for their living. It is important to assess the radiological aspects of pre-mining and post mining scenario as the geology is altered due to extraction of NORM materials like monazite, compared to the original land geology. Surface soil samples (03 each) from inland mining areas – both pre-mining and post-mining at refilled and refurbished conditions were collected of five locations and analysed for NORM radioactivity of 232Th, 226Ra & 40K. Gamma radiation field was also measured with a calibrated NaI(Tl) based low range survey meter during each sampling. The samples were processed and analysed on secular equilibrium as per standard protocol in HPGe based gamma spectrometry system. The 40% RE p-type coaxial HPGe detector connected to 8K MCA system is calibrated using RG-Th and RG-U standards in defined bottle geometry for quantification. 232Th was estimated using gamma lines 239 keV of 212Pb, 727 keV of 212Bi, 911 keV of 228Ac & 2614 of 208Tl and 226Ra/238U was by 186 keV of 226Ra, 352 keV of 214Pb & 1764 keV of 214Bi while 40K by 1461 keV single gamma line. The determined 232Th, 226Ra and 40K concentrations in Bq.g-1 as well as measured and calculated gamma dose rate (using the UNSCER 2000 formula for absorbed dose rate) in μGy.hr-1 is tabulated. The observations clearly show significant reduction in radioactivity levels and gamma dose rate levels at post/refurbished scenario from pre-mining levels, due to extraction of monazite from raw sand. The activity levels of 232Th in the refilled areas were 3% (Karithura-1) to 21% (Karithura-2) of original levels while 226Ra were 3% (Karithura-1) to 26% (Vellanathuruthu) of original levels. The estimated (from respective levels of 232Th, 226Ra & 40K – using UNSCEAR formula) as well as measured gamma dose rate also shows considerable reduction in the same line. On an average the residual gamma dose rate (calculated) is 11 ± 9% of original levels and measured dose rate is 16 ± 11% of original levels. The difference seen in the estimated and measured dose rate is due to cosmic component and measurement errors. The annual external gamma dose with occupancy factor 0.2 were estimated and found to be 0.75 to 11.10 mSv.y-1 in pre-mining scenario while it is only 0.21 to 0.58 mSv.y-1 in refilled scenario which falls in the world wide average of 0.3-1.0 mSv.y-1[2] and is a significant change in NHBRA.{Table 1}
Keywords: Dose, monazite, NORM, pre and post mining
References
UNSCEAR. Sources, Effects and Risks of Ionizing Radiation. New York: UNSCEAR; 2000.UNSCEAR. Exposures from Natural Sources. Vol. I. New York: UNSCEAR; 2008.
Abstract - 21296: Evolution of radiological monitoring systems and methodologies in Indian uranium mining industry: A historical perspective
Amir Hasan Khan
Ex-Raja Ramanna Fellow, Health, Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Soon after discovery of uranium ore at Jaduguda in Singhbhum Thrust Belt, uranium mining was initiated in mid-1950s. After exploratory mining regular underground mining operations were undertaken, with simultaneous installation of an ore processing mill near the mining facility at Jaduguda. A dedicated, independent Health Physics Unit was established in 1965 to monitor gamma radiation, radon, its daughter products and ore dust in the mine and mill. Monitoring for uranium, radium-226 and chemical pollutants in effluents from the mine, mill and the recipient aquatic systems up to several kilometres were taken up. Initially many of the monitoring equipment like scintillation cells or modified Lucas cell wit counting systems were developed in-house. A 100 ml capacity Corning glass flasks internally coated with ZnS(Ag) with tight cork provided with stopcocks capable of retaining high vacuum were used to collect air samples for radon-222 from the mine workings for counting alpha activity in laboratory.[2] This was essential but delicate item for use in mines. It was subsequently improved to use cylindrical aluminium cells with glass window and provided with Swedgelok connector to retain vacuum. An instant and continuous radon daughter working level meter was also developed. Ambient dosimetry using radon, equilibrium factor and occupancy period was used to estimate the dose to workers. Personal dosimeter using SSNDT film and TLD chip with a membrane filter was developed with a calibration system for personal dosimetry. The dosimeter can be used with the miners' belt. Ambient and personal dosimetry results are in good agreement. Radon emanation studies were carried out on the backfilled uranium tailings in the mine stopes and on the surface of the large tailing containment facility using inverted accumulation chambers from which air sample can be drawn in a scintillation cell after a suitable accumulation period.[1] A radon bubbler made of glass was developed to estimate low concentrations of radium-226 and/or dissolved radon in environmental samples. Additionally, low level radon detection system (LLRDS) was developed for measurement of environmental radon. It is also used to measure radon-222 in exhaled breath of workers.[4] The radon monitoring data are useful in periodic planning and upgradation of mine ventilation system too.[1] For example, after a few decades of operations when the mine reached deeper levels, the ventilation system of Jaduguda mine was revamped. By addition of air supply of 20 m3. s-1 to the existing supply of about 90 m3.s-1 resulted in reduction of the average dose by about 30%. There are other monitoring systems and methodologies developed over the years. As several new underground mines have been developed in Sighbhum-East district and one at Tumalalpalle in Andhra Pradesh, the systems developed at Jaduguda are used at all the mines and the ore processing mills in the country. The paper summarises many of the systems and methodologies developed and being used in all the uranium mining facilities and related laboratories in the country.
Keywords: Dose, Monitoring, radon, uranium mining, ventilation
References
Raghavayya M, Khan AH. Radon and Uranium Mine Ventilation, 16th Annual Conference of IARP, Bombay, January 1989; 1989.Khan AH, Raghavayya M. Recent Trends in Monitoring Radiation and Radon Daughter Products in Indian Uranium Mines. Vol. 1. Proceedings of the IV Congress of the International Radiation Protection Association; 1977.Raghavayya M, Khan AH. Radon Emanation from Uranium Mill Tailings Used as Backfill in Mines, Noble Gases Symposium, Las Vegas, USA; 1973.Srivastava GK, Raghavayya M, Khan AH, Kotrappa P. A low level Radon detection system. Health Phys 1984;46:225-8.
Abstract - 21314: A comparative study between gamma spectrometry and colorimetric method for estimating uranium ore grade
B. K. Rana1,2, Ranjit Kumar1, Gopal Verma1, Samim Molla1, S. K. Jha1,2, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Anushaktinagar, Mumbai, Maharashtra, India
E-mail: [email protected]
Accurate estimation of uranium in geological rock samples is very essential for the uranium exploration programme and to recover the uranium economically from the low grade ore for sustainable development.[1] Uranium content in rock samples can be estimated by various non-destructible and destructive techniques. Gamma-ray spectrometry and TBP-Kerosene uranium extraction followed by calorimetric analysis are widely used for the estimation of uranium in the ore as a non-destructive and destructive technique, respectively. NaI(Tl) gamma spectrometry technique can be used for estimation of uranium in ore having radioactive equilibrium between 238U and its daughter products. 10 nos of ore samples from Jaduguda and Banduhurang mines were collected and processed as per the standard procedure prior to gamma spectrometric analysis. About 400 g of powdered ore samples are packed airtight in a cylindrical polyethylene terephthalate (PET) container of 6.5 cm in diameter and 7.5 cm in height, and packed samples are kept for a month to attain equilibrium between 226Ra and its daughter products, and finally counted in 5“×4” NaI(Tl) gamma spectroscopy for 10,000s. 226Ra activity was estimated using 1764 keV gamma rays of 214Bi.[2] By assuming the secular equilibrium between 238U and 226Ra, the ore grade in the ore is estimated. However, for calorimetric estimation of uranium, a known weight (5 g) of powdered ore samples is wet digested and subjected to 10% TBP in kerosene extraction of uranium in nitric acid medium, followed by stripping of uranium from the organic phase to the aqueous phase by saturated sodium sulphate solution. Finally, 50% NaOH (5 ml) and 20% H2O2 (1 ml) are added to the solution and volume is adjusted to 50 ml prior to colorimetric analysis where the absorbance of the solution is recorded at 380 nm using a double beam UV-visible spectrophotometer. Ore grade estimated by these two techniques showed good agreement with each other [Figure 1], with a variation of ± 4 to ± 15 %. The Pearson correlation coefficient (r) was estimated to be 0.993. The mean ore grade was estimated by NaI(Tl) gamma spectrometry and colorimetric analysis technique was found to be 0.068 ± 0.042 and 0.064 ± 0.038 as % eU3O8, respectively, with an overall variation of ± 5%. A paired t test performed on the two set data showed that at the 0.05 level, the difference between the two estimated means was not significantly different from each other. NaI(Tl) gamma spectrometry analysis results can be more representative of the bulk of ore processed in the process plant for large sample size (~400 g) than analysis by the colorimetric technique (5 g). A comparison between the two techniques is given in [Table 1]. NaI(Tl) gamma spectrometry techniques can be used for screening and rapid estimation of uranium in the ore or feed to the process plant within a day when there exists a secular equilibrium between 238U with its daughter radionuclides.{Figure 2}{Table 2}
Keywords: Colorimetric technique, gamma spectroscopy, nondestructive technique
References
Sarangi AK. Vol. 99. Published in the Transactions of the Mining, Geological and Metallurgical Institute of India (MGMI); 2002-2003.Sahoo SK, et al. Radiat Prot Dosimetry 2010;140:281-6.
Abstract - 21333: Attributes and appraisal for radiological safety in uranium mill at Jaduguda
V. N. Jha1, S. K. Jha1,2, Rajesh Kumar3, N. K. Sethy1, S. K. sahoo2, M. S. Kulkarni1,2
1Health Physics Unit, HPD, BARC, 2HBNI, Mumbai, Maharashtra, India
E-mail: [email protected]
Hydrometallurgical milling operation of low grade uranium ore at Jaduguda involve steps like crushing, grinding, size optimization, leaching under oxidant stress (pyrolusite), ion exchange separation and product recovery as Magnesium di-uranate. Depending on the nature of the associated operation the radiological concerns, monitoring perspective and mitigation strategies can vary to a great extent. Although the state of art technology used at Jaduguda uranium mill ensures optimum radiological protection of occupation workers appraisal for the stipulations of regulatory compliance requires a comparison against the existing guidelines / derived limits. Present paper briefly provides the radiological concerns, mitigation measures, existing scenario and comparative level against the stipulated Derived Limits.
Materials and Methods: Radiological monitoring of Jaduguda mill is ensured through the monitoring of the ambient conditions at specified frequency and individual monitoring for different categories. Long lived alpha activity is estimated by collecting air samples using respirable dust samplers and counting after the decay of short-lived daughters. Samples are collected on monthly basis from eleven locations of different sections of the mill. From each representative location six hours sampling is carried out for activity estimation. Ambient gamma survey is carried out in different section of the mill for the evaluation of external exposure. For crushing, screening and grinding section Radium body burden is estimated via collecting the exhaled breath and counting the positively charged daughter (210Po) of radon after specified delay. Regular mill workers are subjected to radon in breath examination once in three years. 222Rn measurement in uranium mill is carried out using the Alpha Guard (PQ 2000 Pro) in diffusion mode. Uranium content in the urine of regular workers from finished product section is estimated at the laboratory following chemical separation using solvent extraction-solid fusion and fluorimetric measurement technique.
Results and Discussion: The physical process involving crushing, grinding and screening the radiological concern is mainly due to the external gamma radiation. The long lived alpha activity (U, 226Ra, 230Th and 210Po) associated with the respirable size dust particulate is substantially low and the resulting inhalation exposure is of the order of 100 to 150 μSv y-1 , which is ~ 2-3 % of the annual limit. Though the radiological concern of the particulate is low the free silica content pose additional risk and the limits are prescribed based on the free silica content of the respirable dust. The area is equipped with dust extraction systems and scrubbers. Chemical treatment of ore leads to modification in radiological solubility class (S to M) of the material and increase in the concentration of uranium; thereby the limits are based on soluble uranium compound consideration. While comparing the chemical characteristics, most soluble radiological “F” class and the moderately soluble “M” class, more restrictive “M” class is considered for deriving the limits. In the product precipitation area, the product is mostly free from radioactive impurities except decay products of 238U for long term hold up of the concentrate. Results of radiological monitoring with typical range are provided in [Table 1]. Operations pertaining to the product recovery are carried out in glass enclosures and the particulate is released through pre and HEPA filter. Provision for spillage collection system, surface smoothing and acidified water washing is made for effective decontamination. In some areas of concerns the required radiological protection is ensured through entry restriction and optimizing of the exposure time. Average occupational exposure of mill worker is worked out to be of the order of 3 mSv y-1, which is substantially lower than the annual limit of 20 mSv y-1 averaged for five years. Conclusion: The radiological monitoring for Jaduguda mill substantiates effective appraisal for the regulatory stipulation. The worked out occupational exposure is of the order of 3 mSv y-1, substantially lower than the recommended annual limit of 20 mSv y-1.{Table 3}
Keywords: Control measures, mill, radiological condition
Reference
AERB. Radiological Safety in Uranium Mining and Milling. AERB Safety Guidelines NO. AERB/FE-FCF/SG-2; 2007.
Abstract - 21436: Hyperaccumulator species for rare earth elements: Phytomining for circular economy
A. C. Patra1, S. Mehetre2, V. K. Thakur1, S. K. Jha1,3, M. S. Kulkarni1,3
1Health Physics Division, Bhabha Atomic Research Centre, 2Nuclear Agricultureand Biotechnology Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Rare earth elements (REEs) are a group of 15 elements from lanthanum to lutetium. Their demand has increased significantly over the last decade due to their increased industrial uses. REEs are mined from a variety of mineral deposits located throughout the world. Recent research has suggested the application of phytomining followed by biomass generation and chemical separation for the generation of mixed REE from milltailings. Our recent findings have shown that the U mill tailings generated form the mining and ore processing facility at Singhbhum can act as a potential source of REEs.[1] Although extensive studies have been carried out to find hyperaccumulator species for REEs globally, there is dearth of Indian literature in this domain. Hence, in this study authors have made an attempt to study REE uptake in plants, in the search of REE hyperaccumulator Indian plant species, which can thereafter be applied forphytomining of REEs from uranium mill tailings as part of circular economy. Indian maize and Chickpea were the two plants investigated for REE uptake by pot-experiments. These two plants were selected since they are monocotyledons and dicotyledons. The plants were grown in plant-growth chamber providing natural environment, conducive for plant growth.Seeds of the plants were sown in soils in which mixture of Rare-earth chloride was added in a predefined ratio and homogeneously mixed. Control growth experiments were also carried out. The crops were harvested on day 1, day 10 and day 20. It was observed that in the mid-term of the experiment, the plants became stunted, but later regained their growth. Segregation into root and shoot parts was carried out and fresh weight was obtained. The plant roots and shoots were dried and digested with electronic grade acids along with reagent blank and standard. Thereafter the samples were filtered and analysed by ICPOES technique. The concentrations of Ce, La, Nd, Pr, Sm and Y were observed to range from 0.02-3.04, 0.03-2.61, <0.01-1.69, 0.04-0.19, 0.04-0.19, 0.18-0.85 and <0.01 μg/ml in the maize samples and from 0.06-2.06, 0.05-1.46, <0.01-1.10, 0.05-0.13, 0.04-0.08 and <0.01 μg/ml in the chickpea samples, respectively.Higher REE uptake was observed in the plant root compared to shoot for both plants. Among the REEs, maximum uptake was observed for Ce>La>Nd>Sm>Pr>Y. Highest concentration of 175ppm was observed for Ce in chickpea root as seen in [Figure 1]. Dicranopterislinearis is a known hyperaccumulator for La and Ce with highest reported concentration of >1000 ppm and 0.7%.[2] To understand efficiency of phytoextraction Bio-concentration Factor (BCF) and Transfer Factor (TF) were determined for all REE. TF values have been shown in [Table 1]. The BCF were in the order of 10-2 to 10-7for the REEs. From the results it can be concluded that these plants show growth in the presence of REEs and also show measurable uptake of the REEs like Ce and La.{Figure 3}{Table 4}
Acknowledgement
Authors express their gratitude to Dr. D. K. Aswal, Director, HS&E Group for his kind support.
Keywords: Phytomining, plant species, rare earth elements, U mill tailings
References
Patra, et al. Potential of Uranium Mill Tailings for REE Harvesting. Proceedings of CTAC-2021; 2021.Shan X, et al. Accumulationand uptake of light rare earth elements in a hyperaccumulator Dicranopteris dichotoma. Plant Sci 2003;165:1343-53.
Abstract - 21528: Study on radioactivity content of soil samples in the coastal region near Brahmagiri, Puri
Abinash Sahu1, P. Prusty1, A. Rout1, S. K. Jha1,2, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Heavy mineral sand deposits occur underwater or may form part of sea beaches or coastal dunes created by wind action over long period of time. They may also occur inland in coastal strip upto few tens of kilometres wide. The highest heavy mineral concentration are found in beach deposits. Deposits of heavy mineral sand may occur in populated areas, leading to high levels of natural background exposure due to the relative high thorium content of the monazite component. Recently AMD has reported that a large amount of heavy mineral deposit is available in coastal region of Brahmagiri, Puri. Monazite is a high value mineral as it contains the rere earth elements, uranium and thorium which are very important in present global scenario. To study the environmental effects of mining these high valued heavy minerals it is highly essential to generate background radiological data. Our baseline study represents radiological condition of soil properties of the pre-mining area stretched over 26 km2 in the coastal area near Brahmagiri, Puri, Odisha, India. The samples were collected from 12 different locations of the study area which is represented by the area demarked in the black line of [Figure 1]. The samples were collected as per the guidelines mentioned in IAEA TRS – 295.[1] The soil samples were dried & grinded to below 40 μm size. The samples were then hermitically sealed for 30 days so that the daughter radionuclides to attain secular equilibrium. The samples were then analysed in a p-type high purity Ge detector system with 40% relative efficiency with respect to the 3” × 3” NaI (Tl) detector system. The HPGe detector system was calibrated using IAEA RGU & IAEA RGTh standards for estimation of the naturally occurring radionuclides of U & Th. The results of the analysis are represented in the [Table 1]. The gross alpha and gross beta. The radioactivity level from Ra228 is linked to Th232 which was then traced back to the monazite content in those samples. 96% pure monazite sample was used as standard for the estimation of the total monazite content in the soil samples. The results presented in [Table 1] shows that there is a considerable amount of monazite content in the soil and beach sand. Monazite being a prime source for Th, rare earth and U, extraction of these high value heavy minerals has high impact on the economic aspect of society.{Figure 4}{Table 5}
Keywords: Gross alpha, gross beta, monazite, radioactivity, radium
Reference
IAEA. A Guide Book for Measurement of Radionuclides in Food and the Environment. Technical Reports Series No. 295. Vienna: IAEA; 1989.
Abstract - 21537: Radioactivity monitoring in solid wastes using gamma spectrometry
R. P. Patra1, A. Sahu1, A. Rout1, P. Prusty1, S. K. Jha1,2, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Department of Atomic Energy, Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Safe and economic management of radioactive waste is necessary for the successful implementation of nuclear power programme. Responsible radioactive waste management requires the implementation of measures that will afford protection of human health and the environment. Since improperly managed radioactive waste could result in adverse effects to human health or the environment now and in the future. In India majority of the Rare Earth (RE) elements are obtained from processing of Monazite, which contains 45 – 60 % of RE along with the radioactive element 232Th, natural U and their decay products. Thus the chemical separation of RE compounds from monazite requires to take care of the radioactive element. IREL(India) Limited has set up a Rare Earth Extraction Plant (REEP) at OSCOM, Odisha to chemically process the Monazite and retrieve the essential elements viz. rare earth elements, Thorium and Uranium etc from it. Measurement of Solid Wastes generated due to processing of Monazite was carried out by using Baltic Scientific instruments made High Purity Germanium gamma spectrometer detector with resolution 2.4 Kev at 1.33 Mev gamma ray from a 60Co source (manual HPGe detector GCDX-40190). The gamma-ray peaks obtained with NaI detectors are very broad by comparison, so that two peaks close to each other cannot be resolved and low- energy peaks may not be easily observed. But with HPGe detector it has high resolution; it can detect close energy peaks. Semiconductor detectors made of germanium or silicon compensated with lithium provides significantly better energy resolution. In IREL Rare Earth Extraction Plant is established to process monazite. During the process of monazite four types of solid wastes were generated. These waste were collected on monthly basis and analysed in alpha beta counting system for gross alpha and gross beta activity and High Purity Germanium gamma spectrometer detector for determination of 228Ra,208Tl,226Ra and 234Th activity [Figure 1]. The activity concentration of 228Ra nuclides was found to be in the range of 60.11 Bq g-1 to 3246.72 Bq g-1. For 208Tl nuclides, the activity concentration ranges between 212.54 Bq g-1 to 1028.23 Bq g-1. For 226Ra nuclides, it ranges from 29.86 Bq g-1 to 2132.16 Bq g-1. For 234Th nuclides, it ranges from 61.56 Bq g-1 to 1662.14 Bq g-1. The activity of collected waste samples were found to lower than the permitted limit. Hence, no radiological hazards were observed. The radioactive waste management is to deal with radioactive waste in a manner that protects human health and the environment now and in the future without imposing undue burdens on future generations. The volume of this waste stream is not large and it is feasible to store it on-site in an engineered facility.[3] The solid waste stored in the solid waste trenches will be processed in future. The study carried out in this work would be helpful to effective management of radioactive waste and provide valuable / useful data for the impact assessment due to storage of solid waste in solid waste trenches (BARC Internal Report 2018) due to operation of Monazite rich beach sand processing of IREL, OSCOM.{Figure 5}
Keywords: Monazite, radioactive waste management, spectrometry, waste storage and disposal
References
Baltic Scientific Instruments. (HPGe) Manual HPGe Detector GCDX-40190.BARC Internal. Radiation Protection and Environmental Surveillance in and Around Orissa Sand Complex (OSCOM). Chatrapur, Odisha: BARC Internal Report; 2018.Gupta MP, et al. Indian Experience in Near Surface Disposal of Low Level Radioactive Solid Waste. Vienna: Proceedings of the Symposium on Experience in the Planning and Operation of Low Level Waste Disposal Facilities, International Atomic Energy Agency; 1996.
Abstract - 21550: Dose assessment methodology for the uranium mill workers at different geo chemical environment
Gopal P. Verma1, S. K. Jha1,2, S. K. Sahoo1, Abhigyan1, M. S. Kulkarni1,2
Radiation Protection Section (Nuclear Fuels)
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Estimation of dose to the radiation worker is being carried out using area monitor, data on radiation level, radon thoron & progeny, airborne long-lived activity and TLD record as applicable in different unit. As per IAEA [Basic safety Series No. 100] large variations exist in calculation of dose to the uranium mill worker in different countries. As per the recent AERB guidelines (AERB/NF/SM/RP-1) actual time spent by worker is to be taken for dose calculation. The radon concentration and ambient Gamma dose rate is measured regularly in the uranium mill atmosphere. The radon progeny exposure is worked out as Working Level Month (WLM) and is converted to dose using conversion parameters [ICRP-65]. The external individual dose is worked out using the ambient gamma dose rate. The information so obtained is converted into individual dose of the radiation worker using occupancy period and the attendance.
Dose Computation Methodology for the Mill Workers: The dose assessment of the mill workers is done using the following simplified approach-
The Radon concentration in the ambient air is measured in regular interval using the AlphaGUARD® portable radon monitor. The Ambient Gamma Dose rate is measured using the radiation survey meter. The occupancy and attendance related information is provided by the facility (Mine). The 8-hour shift period is divided into two representative time fractions according to the time spent at (1.) Plant area and (2) Non-plant area. the time spent have been typically assumed as T1, and T2, as per the job involvedThe dose assessment is carried out in quarterly basis by evaluating the exposure rate and individual occupancy data at the working area.Following methodology is used for the evaluation internal dose:
I. Internal Dose Calculation at the Mill Area:
Equivalent Radon Concentration (EERn)
EERn = (CRni x F) Bqm-3
Where, CRni = Radon Concentration in Bqm-3 at location i and F = Equilibrium Factor = 0.4 (for Mills)
[INLINE:1]
Where, OFi = Occupancy Factor at location i (hours)
170 = No. of working hours per month
1 WL = 3700 Bqm-3
Internal Dose (mSv)i = No. of WLM x 5 mSv WLM-1
Since, 1 WLM = 5 mSv (Ref: ICRP-65)
II. External Dose Calculation at the Mill Area:
The external dose evaluation is carried out using the Ambient Gamma Dose rate.
External Exposure (mSv) = Gi X OFi
Where, Gi = Ambient Gamma Dose rate at location i
OFi = Occupancy Factor at location i (hours)
III. Reporting of the Quarterly Individual Doses:
The total dose is calculated by adding external and internal doses to the individual worker in quarterly basis. Quarterly Dose Received by Individual Mill Worker is given in [Table 1].{Table 6}
Total Dose (mSv) = Ext. Dose (mSv) + Int. Dose (mSv)
The present work established the uniform dose computation methodology to estimate the collective dose of an individual worker of uranium mill present in different geo-chemical environment. The methodology is effective enough in minimizing discrepancy the estimation of collective exposure of the individuals in different milling sites. The present work helped in harmonising the collective dose estimation at different mining sites.
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.
Keywords: Exposure, radon, uranium mill, working level month
References
IAEA. Safety Reports Series No. 100. Occupational Radiation Protection in the Uranium Mining and Processing Industry. Vienna, Austria: International Atomic Energy Agency; 2020.ICRP. Protection against radon-222 at home and at work. ICRP Publication 65. Ann ICRP 1993;23.
Abstract - 22126: Reduction of normalised collective dose consumption per kCi of activity processed at high intensity cobalt-60 sealed source fabrication facility
T. M Ashraf, S. A Tariq, R. Sahu
Regional Centre – RAPPCOF, Board of Radiation and Isotope Technology, Kota, Rajasthan, India
E-mail: [email protected]
Introduction: Board of Radiation and Isotope Technology (BRIT) is the nodal agency under Department of Atomic Energy (DAE) to provide radiation and radio isotope technology for the larger benefit of the society as per the mandate. Sealed radioactive sources are widely used for beneficial purposes throughout the world in industry as well as in medicine. BRIT handles several Million Curies of Cobalt-60 activity per annum in a safe manner at RAPP Cobalt Facility (RAPPCOF) to supply Co-60 sources to healthcare and industrial user in India and abroad. The processing of reactor produced Co-60 is a challenging job in view of radiological protection concern. Recently the requirement of Co-60 sealed sources increased many folds and this paper discuss about the achievement of reduction in normalised collective dose consumption per kCi activity processed at RAPPCOF sealed source fabrication facility.
Materials and Methods: The fabrication of Co-60 sealed sources includes removal of adjuster from reactor, transportation to RAPPCOF, discharging of adjuster rod in storage pool, under water cutting of the adjuster rod, separation of sub-assemblies, estimation of activity, transferring of sub-assemblies from pool to hot cell, cutting of sub-assembly and recovery of Zr-capsules for sealed source fabrication, etc. These Zr-capsules are directly used after decontamination in irradiator sources and the Zr-capsules are again cut to recover Co-60 pellets/slugs for fabrication of Telethrapy sources, which is the most challenging job in view of radiological consequences. The new method adopted for Telethrapy source fabrication helped to restrict the radiological consequences arises from cutting of Zr-capsules.
Trained manpower along with strict radiation protection procedures followed in all the above steps to restrict the radiation exposure to the personnel and to reduce the collective dose consumption during source fabrication.{Figure 6}
Results and Discussions: A reduction in dose consumption of 0.0311 to 0.0126 Person-mSv/kCi was achieved. Maximum individual dose restricted below 9.0 mSv/year with the limited trained manpower available in the facility. Many folds increase in Co-60 processing from 1.0 to 6.7 MCi observed in last ten years due to the high demand in the national as well as in the international market, hence the collective dose consumption increased in the facility, which justifies the increased processing of Co-60 in the facility during the period.
The normalised collective dose consumption per activity processed achieved due to the modification of operational conditions from the lesson learned in past experience. Some of the remarkable modifications adopted in the facility mentioned below. New method adopted for activity estimation of Sub-assemblies in pool enables to pick right Sub- assembly for processing, by reducing men-power in each radiological operation which means effective utilization of trained manpower and experts, rotating the manpower for the high dose consumption jobs and effective training and adherence to the proper procedures. Conclusions: RAPPCOF improved in radiation protection aspects as well as in dose consumption aspects, over the years normalised collective dose consumption per activity processed reduced to less than 50%. Procedural modification with past lesson learned in demanding situations, effective utilization of trained manpower, adopting new methods, etc. will help to reduce the collective dose in the sealed source fabrication facility.
Keywords: Cobalt-60 sealed sources, collective dose reduction, radiation processing, radiation protection
Abstract - 22152: Substantiation of wind profiling exponent factor through air dispersion stability indices at BARC Vizag site using multi-level measurement facility of 30m micro-meteorological Tower
R. Jana, A. Vinod Kumar1, P. Chaudhury1
Radiation Safety Systems Division, Health, Safety and Environment Group, Bhabha Atomic Research Centre, 1Environmental Monitoring and Assessment Division, Health, Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
The study on atmospheric dispersion of radioactive pollutants has important role in nuclear industries to check the compliance with regulatory limits for release and also to assist in the emergency preparedness in case of an unlikely event of accidental release from stack or ground. In hourly mapping of various airborne pollutants from the source of origin, assignment of wind field at stack height is needed by wind profiling for the corresponding air dispersion stability index (ADSI) (popularly known as P-G class). These are two important variables which are driven by kinematic and thermal characteristics near the Earth's surface. In this context, various modification has been suggested in the past on determination of P-G classes/ADSI for which Turner's key fits best.[1],[2]
In the present study, best method is coined for deriving ADSI by amending Net Radiation Indices (NRI) as modified Turner's key for two closely comparable methods. NR method is based on nine categorization of measured Net Radiation (NR). Derived 1st Lifting Condensation Level (LCL) is coupled with NR method in NR_LCL method for categorization of nine NRIs. Further, six ADSIs (Unstable: A, B & C, Neutral: D, Stable E &F) are assigned based on nine NRIs and nine wind speed classes. Hourly averaged wind profile is predicted at a desired height at Z by determining the exponent factor (p) of wind speed with respect to the reference level of 3D Ultrasonic Anemometer installed at 4m height in 30m micro-met Tower (Eq. 1).
UZ = U4× (Z/4)p (1)
The p-factor (p) is established by minimising Converged Mean Bias Error (MBE) and Root Mean Square Error (RMSE) between hourly averaged predicted and measured wind speed at 10 m and 30 m height [Figure 1] using cup anemometers installed on the same 30 m micro-met Tower at BARC Vizag. Measurement period was from Aug.01 to Dec.31, 2015. The NR_LCL method is capable of determining p-factor for all ADSI [Table 1] and thus, this complex terrain site is designated under 'URBAN' type following US EPA, 2000 [Table 1].{Figure 7}{Table 7}
References
Turner DB. A diffusion model for Urban Area. J Appl Meteorol 1964;3:83-91.Jana R, Vinod Kumar A. Proceedings, A Method for Determination of Atmospheric Stability Index from Routinely Measured Meteorological Parameters. 31st IARP National Conference; 2014.US Environmental Protection Agency. Office of Air Quality Planning and Standards, Meteorological Monitoring Guidance for Regulatory Modeling Applications, EPA-454/R-99-005. USA: US Environmental Protection Agency; 2000.
Abstract - 22166: Review of collective dose consumption in Indian PHWR – A case study of Kaiga Generating Station-3 and 4
Veerendra Danannavar, S. S. Managanvi1, M. Seshaiah1, B. Vinod Kumar1
Health Physics Unit, Kaiga Generating Station-3 and 4, Nuclear Power Corporation of India Limited, 1Kaiga Generating Station-3 and 4, Nuclear Power Corporation of India Limited, Uttar Kannada, Karnataka, India
E-mail: [email protected]
Nuclear Power Corporation of India Limited (NPCIL) has designed, constructed, operating and maintaining Pressurised Heavy Water Reactors (PHWRs) to harness the energy needs of the country. At present, there are 17 PHWRs operating in India with the installed capacity of 4360 MWe.The first stage of the Indian Nuclear Power Programme has placed the country in an elite club of nations possessing advanced nuclear power technology. History of construction of PHWRs started with India's first PHWRs RAPS-1(1973) and RAPS-2 (1981) that paved the path to Indigenisation, standardisation, consolidation and commercialisation. Kaiga Generating Station, at Karnataka, has 4 PHWRs in operation. 2 among these reactors-3&4 (KGS-3&4), comprises of two 220 MWe PHWRs that started operation in the year 2007 and 2011 respectively. This paper highlights the improvements undertaken in the field of radiological protection of Indian PHWRs. Occupational exposure to ionizing radiation (External and Internal exposure) takes place during the process of Operation & Maintenance. The radiological exposures to the occupational workers are meticulously monitored and records are maintained.
The Collective Dose (CD) of the Nuclear Power Plants (NPP) is a yard stick to judge the overall health of the plant and effectiveness of radiological protection program. International Commission on Radiological Protection (ICRP) calls for keeping the magnitude of the individual doses, the number of people exposed and the likelihood of potential exposures As Low As Reasonably Achievable (ALARA). Initially, the CD in Indian PHWRs was relatively high[1] compared to prevailing industry standards. Many of the issues affecting CD in the older PHWRs were addressed in the new generation reactors. This was achieved by design improvements and improving the work practices. These improvements have significantly contributed to a reduction in CD from 4.0 p-Sv/reactor/ year in older stations to less than 1 p-Sv at KGS-3&4 (3). The focus was made on an overall reduction in both individual dose and CD thereby reducing radiation risk to the occupational workers. This was achieved through (a) Consideration for dose control measures at the plant design and engineering changes during plant operation. (b) Improved equipment reliability to avoid breakdowns and forced unit outages during operations. (c) Use of state of art technology which includes automation, remote tools, protective equipment and dosimetry devices. (d) Improving plant performance in terms of capacity factors and availability factors. (e) Development and usage of innovative mock up facilities during Biennial Shut Down (BSD). It can be observed that the CD per reactor per year of Indian PHWRs is on lower side as compared to world PHWRs and is presented in [Figure 1]. KGS-3&4 has achieved excellence in CD consumption as compared with other Indian PHWRs in Bench Mark Indices and Radiological Safety Performance Indicators. In addition to the measures mentioned in (a) to (e) above, following measures were implemented for reduction of CD. (i) Proper management of radiological conditions. (ii) Review of radioactive jobs likely to consume more than 1.5 p-mSv through ALARA committees and post job reviews. (iii)Development and implementation of radiological protection good practices. (iv)Sharing of operating experiences on radiological protection etc. (v) Fostering safety culture, Collective dose consumption of KGS-3&4 during 2012 to 2021 is given in [Figure 2].{Figure 8}{Figure 9}
Keywords: ALARA, bench mark indices, capacity and availability factors, collective dose, international commission on radiological protection, new generation reactors, nuclear power programme, operating experience, radiation risk, radiological safety performance indicators
References
Ramamritham B, Managanvi SS. Design improvements in modern PHWRs for reducing collective dose. J Radiat Protect Environ 2000.Ghopalan VP, Sunderarajan AR. Analysis of radiation exposures in Indian nuclear power plants. J Radiat Protect Environ 2000.Periodic Report. Safety Oversight Report on Radiological Aspects of NPPs for the Year. Mumbai: HS&E Group, NPCIL; 2021.
Abstract - 22184: Development of radiological safety performance indicators
Rashmi Prakash Sharma, V. Nagarajan, K. Venkata Raman, K. K. De
Directorate of Health Safety and Environment, NPCIL HQ
E-mail: [email protected]
Introduction: Paramount importance is given for safety in the Nuclear Power Plants (NPPs) in all stages right from design, siting, construction, commissioning, operation and decommissioning. Safety at NPPs include nuclear safety, radiological safety, environmental safety and industrial & fire safety. Radiological safety at NPPs ensures protection of people and environment. Quantitative assessment of station performance with respect to radiological safety is required to instil confidence about the effectiveness of radiation protection program implemented.
Radiological Safety Performance indicators (RSPIs) have been developed at NPCIL to meet this requirement of quantitative assessment of radiological safety at NPPs. RSPIs facilitate to monitor, evaluate and enhance the radiological safety performance of the stations for ensuring continual improvement and to gain additional perspective on performance relative to that of other nuclear power plants. These indicators provide an indication of the possible need to reorganize priorities and resources to achieve improved performance.
Currently, international trends are also seeking the possibilities of expanding the usage of safety performance Indicators.
Development of RSPIS: Comprehensive set of RSPIs in all the areas that directly or indirectly affect safety, bench marking of each RSPIs considering the past performance of all operating units of NPCIL and the methodology of assessment has been developed. To perform quantitative assessment, each RSPI has been assigned certain weightage considering their significance. The bench marking criteria and assignment of weightage for RSPIs have been derived considering design aspects / generic issues and current operational status of stations. These would evolve with time, to reflect the need to focus in greater depth on specific topics so as to ensure continual improvement to enhance safety in the identified areas. The actual values of the indicators are not intended to be direct measures of safety, although safety performance can be inferred from the results achieved.
Categorisation of RSPIs: RSPIs have been identified from each aspect of radiation protection program. This include efforts made by the station in controlling the source, planning and execution of works, controlling the system activities and effluent releases. In addition, RSPIs have been identified to assess the work practices followed by the radiation workers and efforts made by the station to enhance the awareness among the radiation workers regarding radiological safety. RSPIs developed are broadly categorised under the following heads.
Source controlRadioactive work managementWork practices at individual levelCollective dose statisticsPromotional activitiesSystem performanceOther RSPIs.{Figure 10}
Results and Discussion: Based on the methodology derived, radiological safety assessment is being carried out since year 2019. It has been observed that station's performance on radiological safety has continually improving.
Conclusion: Trend analysis of radiological safety performance indicators is intend to be utilized in identifying the areas for the improvement of radiation protection program so as to improve the performance. On the other hand, specific indicator trends over a period of time can provide an early warning to plant management to investigate the causes behind the observed changes.
Keyword: Radiological safety performance indicator
References
U.S. Nuclear Regulatory Commission, Reactor Oversight Process (NUREG 1649).CNSC Regulatory Document REGDOC-3.1.1, Reporting Requirement for NPPs.
Abstract - 22203: Removal of 137Cs from spent fuel storage bay water using Zeolite
Lokesh Kumar, Mangesh T. Valvi, Vivek Mishra1, Saurav, Ranjit Sharma
Health Physics Division, Bhabha Atomic Research Centre, 1Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
The irradiated fuel assemblies in heavy water moderated research reactor are stored in water-filled bays for removal of decay heat. Some of the fission product radionuclides are leached out from failed fuel assemblies, which lead to radioactive contamination of the bay water. The leached radionuclides in the bay water have half-lives ranging from few days (134Cs-2.06 y, 144Ce/144Pr-284 d, etc.) to years (137Cs-30.17 y, 90Sr-28.78 y, etc.). Among these radionuclide 137Cs, is considered as most hazardous radionuclide because of its intermediate half-life (30.17 years), high-energy decay pathways and high chemical reactivity.[1] The removal of Cesium from the bay water can significantly reduce the radiation field and the loose contamination in the bay areas. This study has been carried out to evaluate the 137Cs removal efficiency of commercially available Zeolite-13X from the bay water in a batch process. In this experiment, 5g of Zeolite-13X in beads form of size 1-2 mm was added to volumetric flask containing 250 ml of the bay water. The 137Cs concentration in bay water was estimated by plantcheting 1 ml of sample and its measurement with HPGe spectrometry system. The flask was shaken rigorously and samples were collected at regular intervals during the period of experiment. The Samples were collected at an interval of 1 hour for estimation of activity during the experiment. The 137Cs removal efficiency of Zeolite-13X has been calculated by the following relation.[2] Removal efficiency (%) = [INSIDE:1] Where A0 is the initial 137Cs concentration in Bq/ml and At is 137Cs concentration in Bq/ml after time t. The trend of reducing 137Cs concentration and the increasing 137Cs removal efficiency in the water with time is shown in the Graph [Figure 1]. The experiment shows that the final 137Cs concentration in 250 ml of the bay water reduces to 7.51 Bq/ml from 243.96 Bq/ml by adding 5 g of Zeolite-13X. The 137Cs removal efficiency of Zeolite-13X came out to 89.01 % after 4 hours and reached to 96.92 % after 96 hours. It concludes that the Zeolite-13X effectively removes the 137Cs from the bay water within few hours of mixing. The batch experiment shows that Zeolite 13X is having the good 137Cs removal efficiency. The Zeolite beads during experiment did not increase the turbidity of sample water indicating good mechanical strength. The experimental results indicate that Zeolite 13-X can be effectively utilized for activity reduction of 137Cs in the bay water. The lower activity of bay water will also reduce the radioactive contamination and radiation field in the bay areas.{Figure 11}
Keywords: 137Cs removal, bay water, zeolite 13X
References
Lee HY, et al. Selective removal of radioactive cesium from nuclear waste by zeolites: On the origin of cesium selectivity revaealed by systematic crystallographic studies. J Phys Chem C 2017;121:10594-608.Termsrirat U. Evaluation of Zeolite Efficiency for Removal of Cesium Ions from Seawater. The Asian Conference on Sustainability, Energy & the Environment; 2016.
Abstract - 22216: Preliminary experimental study on wet electrostatic technology for methyl iodine capturing
Ren Hongzheng, Li Yongguo, Li jiayu, Tian Lintao, Zhang Xueping
China Institute for Radiation Protection, Taiyuan City, Shanxi Province, China
Radioactive iodine is a kind of gas radioactive nuclide produced by nuclear fission. Because of its strong mobility, the iodine could participate in the circulation of food chain. If the radioactive iodine is released into the environment excessively, it would have a great impact on the environment and public safety. Radioactive iodine is released under operation, maintenance or accident conditions of nuclear facilities. There are two main forms of radioactive iodine, the one is elemental iodine, the other is organic iodine. Methyl iodine is the most difficult form of radioactive iodine to be removed due to its stable chemical properties among these forms. In order to capture the radioactive iodine, iodine adsorbers filled with impregnated activated carbon are used in nuclear air cleaning systems of nuclear power plants. But the performance of impregnated activated carbon would decline with moisture absorbing, poisoning and aging. The replacement of iodine adsorbers results in a large amount of solid waste and increases the cost of the nuclear waste disposal.
Compared to the iodine adsorbers, wet electrostatic technology produces less solid waste and has advantages of smaller pressure drop and longer utilization period, therefore it is widely used in civil market. It's worth noting that methyl iodide can react with ozone to form iodine oxide particles. The particles could be captured using an electrostatic field, thus the wet electrostatic technology have great potential for radioactive iodine cleaning. In this study, a prototype of the wet electrostatic methyl iodine capturing equipment was developed. Methyl iodine with a certain concentration was used as the test gas to evaluate the cleaning effect of the prototype.
The prototype is composed of ozone reactor, high voltage electrostatic field, flushing module and other accessories. In order to ensure the cleaning efficiency, two high voltage electrostatic fields are used in series. Ozone and methyl iodine mix and react in the reactor. Products of the reaction in form of aerosol particles enter two-stage series high voltage electric fields with carrier gas and are charged in the high voltage electrostatic fields. The oxyiodide particles are collected on the metal wall under the action of electrostatic force. Flushing module could wash out contaminants into the waste liquid collecting box. NaOH solution was used as the spray solution in order to clean the metal wall sufficiently. The design capacity of the prototype is 6 Nm3/h.
A test system was built to measure the cleaning efficiency of prototype. After testing, the cleaning efficiency of methyl iodine can reach more than 99% under the design capacity. The cleaning efficiency of prototype was tested in different electrostatic voltage, temperature, humidity, reaction time, system air volume and other experimental conditions to clarify the influence of parameters on the cleaning efficiency. The experimental results show that: the cleaning efficiency of the prototype for methyl iodine is positively correlated with the electrostatic voltage, and the efficiency increases slightly when the voltage value is greater than 12kV. Air temperature and relative humidity have little effect on cleaning efficiency. The cleaning efficiency is negatively correlated with the air volume of the system, and the maximum processing capacity of the prototype was about 10 Nm3/h. The cleaning efficiency is negatively correlated with oxidating reaction time, and the cleaning efficiency of methyl iodine can reach more than 99% when oxidating reaction time was more than 15 minutes. Meanwhile, increasing the number of high-voltage electrostatic field modules can improve the processing capacity to a certain extent.
For purpose of figuring out the chemical composition of oxidating reaction products between methyl iodine and ozone, the testing gas was sampled by filter membrane and the sampled membrane was analyzed under XPS and SEM. The result shows that the products of reaction are solid and most of them were existing as I(V), possibly I2O5. The radioactive CH3131I was used to test the cleaning efficiency of total iodine including gaseous iodine and iodine aerosol. Experimental results show the cleaning efficiency of total iodine can reach more than 99% under the design capacity.
The experimental results show that the performance of the wet electrostatic prototype for methyl iodine basically satisfies the practical requirements. With obvious advantages, there is a great application prospect in nuclear facility accidents and other scenarios. If the technology would like to apply in nuclear facilities, in-depth tests and studies are also needed including the testing in actual operating conditions of nuclear facilities, equipment maintenance, and external environment adaptability testing.
Keywords: Methyl iodine, nuclear air cleaning, ozone, wet electrostatic
Acknowledgments
This work is financially supported by the National Key Research and Development Program of China (No. 2017YFE0115100).
References
Vikis AC, Macfarlane R. Reaction of iodine with ozone in the gas phase. J Phys Chem 1985;89:812-5.Kärkelä T, Holm J, Auvinen A, Ekberg C, Glanneskog H, Tapper U, et al. Gas Phase Oxidation of Elemental Iodine in Containment Conditions, International Conference on Nuclear Engineering, Proceedings, ICONE 2; 2009. p. 719-27.Gouello M, Hokkinen J, Karkela T, et al. Development and qualification of an innovative wet electrostatic precipitator in view of gaseous iodine filtration on laboratory-scale. Nucl Eng Des 2018;327:7-21.Kalinin NN, Kolyadin AB, Metalidi MM, et al. Decomposition of radioactive methyl iodide in an electric discharge field. Radiochemistry 2011;53:202-5.Fahr A, Nayak AK, Kurylo MJ. The ultraviolet absorption cross sections of CH3I temperature dependent gas and liquid phase measurements. Chem Phys 1995;197:195-203.Stanford JP, Rooyen NV, Vaidya T, et al. Photodecomposition of methyl iodide as pretreatment for adsorption of radioiodine species in used nuclear fuel recycling operations. Chem Eng J 2020;400:125730.
Abstract - 22414: Role of various hydrodynamic instabilities in jet breakup during molten fuel coolant interactions
Ajay Rawat, A. Jasmin Sudha, V. Subramanian, B. Venkatraman
Aerosol Transport and Biodiversity Section, Radiological Safety and Environment Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Following a severe accident in a Sodium Cooled Fast Reactor (SFR), in-vessel retention of core debris is essential to contain the radioactive materials and chemically reacting sodium to minimize the risk to the public. A core catcher is placed within the main vessel to collect the core debris and aid in effective decay heat removal by natural circulation of sodium. The heat removal mainly depends on the porosity, debris particle size distribution and decay power. Before settling on the core catcher, molten core material enters into liquid sodium as liquid jets, and because of the relative motion between the jet and coolant, the fuel jet breaks up and gets fragmented. The jet breakup behaviour is governed by the surface tension, inertial and gravitational forces which manifest as various fluid instabilities like Rayleigh-Plateau, Kelvin-Helmholtz (KH), Rayleigh-Taylor (RT) and Boundary Layer Stripping (BLS). The present study aims at finding out the probable droplet/debris sizes resulting from these hydrodynamic instabilities. The molten jet profile captured by performing Volume-of-Fluid (VoF) simulations in Fluent code is shown in [Figure 1].[1] VoF is a numerical scheme to track the interface between a liquid-liquid or liquid-gas system. In the safety analysis of an SFR, it is necessary to study the role of fluid instabilities and particle sizes in the fragmentation of the corium jet as it influences the heat removal from the core catcher.[2] While the VoF simulations can capture the instabilities, it is tough to predict the droplet diameter; thus, it is estimated using the linear stability theory of the jet. In linear stability analysis we impose perturbation on any variable η in the form [INSIDE:2], where [INSIDE:3]. For the KH and RT instabilities, the most unstable wavelength giving rise to the maximum growth rate of the surface perturbation is given by dispersion relations which are available in the literature.[3] The present analysis considers hydrodynamic effects, and the thermal effects are neglected. Results pertinent to our problem of jet fragmentation leading to the formations of droplets for different velocities of corium jet are presented. Following Saito's recommendations, we have considered the relative velocities between the jet and the coolant medium from 0.5 m/s to 50 m/s.[2] The most unstable wavelength is used to provide an estimate of fragment size during Molten Fuel Coolant Interactions (MFCI), as given in [Table 1]. The size of these droplets is estimated using linear stability theory, wherein the mean droplet size is of the order of most unstable wavelength. As shown in [Table 1], for a relative jet velocity of 1 m/s the probable droplet sizes predicted by RT, KH and BLS for MOX in liquid sodium are about 26 mm, 3.75 mm and 0.95 mm respectively. KH and BLS predicted drop sizes fall within the observed size distribution of SFR core debris in the experimental data reported by Hiroshi where the particle sizes range between 8 mm and 60 microns.[4]{Figure 12}{Table 8}
Keywords: Fluid instabilities, jet fragmentation, molten fuel coolant interaction
References
Rawat A, Jasmin Sudha A, Subramanian V, Venkatesan R. Internal Report; 2021.Saito S, Abe Y, Koyama K. Nucl Eng Des 2017;315:128-43.Drazin PG. Introduction to Hydrodynamic Stability. Cambridge: Cambridge University Press; 2002.Hiroshi M. J Nucl Sci Technol 1974;11:480-7.
Abstract - 22435: ALARA efforts for reactor recirculation loop replacement works at TAPS-1&2
Ramdas Mharse, Shailesh Yadav, B. K. Parida, B. K. Rautela, P. S. Raveendran, Hitesh Shivhare1, D. De1, V Parashar, K. Venkata Ramana2, K. K. De2
Health Physics Unit, TAPS-1&2, 1Mechanical Maintenance Section, TAPS-1&2, 2NPCIL, Anushaktinagar, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Tarapur Atomic Power Station-1&2 (TAPS-1&2), India's first nuclear power reactor played significant role for five decades in establishing the efficacy of country's nuclear energy program.
TAPS-1&2 houses twin units of Boiling Water Reactors. It consists of reactor recirculation system to provide forced recirculation of primary coolant for improving heat transfer and to overcome the power density limitation of natural recirculation. There are two recirculation loops in each reactor. Reactor cleanup system ensures the removal of radionuclides generated due to fission and activation and maintains good chemistry control. Despite this, some amount of radionuclides may result in the deposition on the internal surface and lead to radioactivity buildup. Periodical In-Service Inspection program of reactor systems ascertains the integrity of the system.
Problems: In the recent past, indications of Inter Granular Stress Corrosion Cracking (IGSCC) was observed on one of the weld joint at TAPS-1 reactor recirculation loop-A. Subsequently, station is planning to replace both the reactor recirculation loops made up of SS-316 with IGSCC resistant material (SS 316LN). Being a first of its kind activity and high dose intensive, considerable efforts are being made to minimize the radiological impact.
ALARA Efforts: Assessment of radiation levels on internal surfaces of the loop was by introducing the semiconductor based detector of high range monitor with the guiding hose through the flange opening. Smears collected from the opening and guiding hose were used for qualitative analysis. Sample coupon was prepared from the recirculation loop flange and subjected to chemical decontamination involving three sequential steps namely oxidation, neutralization and reduction. Decontamination factor of about 4-7 was achieved. In view of difficulties in total isolation of reactor from the entry of chemicals, chemical decontamination of the loops has been dropped. Loops were flushed with high pressure water jet. Concerted efforts were made to estimate the collective dose by analyzing the different phases of activity by considering the ALARA measures planned for implementation. To minimize air borne activity in the primary containment during the work, provision to exhaust the air from the work place by providing an enclosure and though the HEPA filters is planned to be introduced. For continuous assessment of radiological conditions, continuous air monitoring instruments and area radiation monitors with audio visual alarms are planned to be used. To prevent spread of contamination, additional rubber stations with adequate contamination monitoring arrangements and suitable trays for collecting the fine dust will be in place. Specially designed metal boxes for disposing the piping and procedures for waste handling are being drawn. Mock up arrangements for various phases of the work are being established. Efforts to minimize the number of field joints are being worked out to minimize the time spent in active areas. In order to gain first-hand experience and test the preparedness, replacement of small diameter piping of primary system and removal of sample coupon from the reactor recirculation loop for metallurgical studies were carried out. Custom made shielding arrangements as per layout is intended to be used. ALARA measures as discussed above were in place during the sample coupon removal for metallurgical studies and extent of small diameter piping replacement carried till now. There were no cases exposure exceeding investigation levels and contamination.{Figure 13}
Conclusion: Works carried out till now has given confidence in carrying out the work safely without any radiologically significant incidences.
Keywords: ALARA, Inter Granular Stress Corrosion Cracking
Abstract - 22543: Study on estimation of radioactivity in insulation materials of neutron detectors
T. Raghunath, V. Ramakrishna, G. Ganesh, M. S. Kulkarni
Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
In-core detectors play a crucial role in providing the neutron flux data in nuclear reactors during start-up. These detectors are placed in a dry thimble well inside the reactor core where the temperature may reach beyond 300°C. These detectors need to be protected from such a high temperature for their durability and longevity. This is achieved by encapsulation with ceramic insulators which will protect the detector thermally and electrically. For this purpose, two types of materials are widely used viz. ceramic spacers and adhesive ceramic tapes shown in [Figure 1]. As these materials are irradiated in high neutron flux (of the order of 1012 and above), will get activated and become radioactive. When the surveillance checks of such irradiated detectors are required, these are taken out in a well planned manner. The knowledge regarding the activation products in these materials will help the health physicist for a better and effective radiological surveillance while handling. An irradiation study of the above two types of materials can generate data of activation products and hence an effective radiological surveillance during their periodic maintenance. Therefore, a sensitive and reliable method based on the use of available KAMINI reactor's Instrumental Neutron Activation Analysis (INAA) facility was chosen for Elemental analysis and the results were obtained[1] as shown in [Table 1] and [Table 2]. The elemental composition is used as input in ORIGEN code and the task of estimation of radioactivity in these two materials is accomplished and given in [Table 3] and [Table 4] respectively. From the estimation study done it is found that there are some signature radio nuclides forming in both the materials which will help in identifying the issues like leak in the dry thimble nozzle. Presence of 182Ta and 46Sc along with regular activation products in air activity samples will help in the investigation when Ceramic spacer and Ceramic tape are used respectively.{Figure 14}{Table 9}{Table 10}{Table 11}{Table 12}
Keywords: Estimation of radioactivity, insulation materials, irradiation study, neutron detectors
Reference
Sharma R, et al. Development of indigenous insulation material for superconducting magnets and study of its characteristics under influence of intense neutron irradiation. IOP Conf Ser Mater Sci Eng 2017.
Abstract - 22545: Development and testing of a passive set-up for drying tritiated cotton waste
T. Raghunath, V. Ramakrishna, G. Ganesh, M. S. Kulkarni
Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
Tritiated cotton waste forms as a major component of solid radioactive waste in the facilities which are handling tritiated liquids. As it falls in low active waste category (on the basis of Radiation dose rate), it is generally disposed in earthen trenches of near surface disposal facilities at the waste management sites causing chances of increase in tritium activity of ground water. Drying before such disposal of the cotton waste would reduce the activity of tritium in ground water. A passive setup for drying the cotton waste was designed such that it has neither a pump nor heating arrangements. As there is no pump used to give air flow and no arrangement to heat, it is required to study its feasibility with a prototype in the field. It is tested for the drying performance firstly with inactive wet cotton and then with Tritiated cotton which was spiked with known activity of tritium. It is found to be drying the cotton within 2-3 days and tritium activity in the cotton found to be reducing by several times. The description of the drying set-up is given in [Figure 1] and [Figure 2]. The procedure followed for the analysis of the residual activity present in the cotton after drying is detailed in this. The dried cotton is soaked in water and tritium activity in water is measured using LSS 300SL of Hydex make1. The decontamination factor is calculated from the results obtained after test drying the wet cotton spiked with various levels of tritium activity as given in [Table 1].{Figure 15}{Figure 16}{Table 13}
Conclusion: Decontamination factor as observed from the test results is ranging from 1500-2500. More than 1000 times of reduction in the tritium activity of the cotton could be achieved in this passive set-up. This type of set up would be most useful in PHWRs in reducing the ground water activity around those sites as this setup is converting the ground water release to air borne dispersal.
Keywords: Drying set-up, solid radioactive waste, tritiated cotton, tritium
Reference
Operating Manual of LSS 300SL of Hydex make.
Abstract - 22593: Study of iodine number of charcoal filter granules of different sizes for ambient, lower and higher relative humidity conditions
D. N. Sangeetha, Y. Ramani, S. Viswanathan, M. Menaka, V. Subramanian
Safety, Quality, and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: In nuclear facilities, the ventilation system are equipped with High Efficiency Particulate Air (HEPA) Filters and impregnated activated charcoal adsorbers for ultimate removal and retention of particulates and different chemical forms of airborne radio-iodine respectively. Activated charcoal (AC) produced from coconut shell is a common adsorbent material for harmful substances including organic vapour due to its excellent adsorption capacity and cost advantage. The adsorption capacity is affected by temperature, flow rate, concentration of organic vapour, and humidity. The AC in Iodine filter assembly is being exposed continuously by air flow at the rated flow of the system under prevailing environmental conditions. In this context, a study on adsorption of iodine in AC for various grain sizes by Surface Area Analysis, Thermo Gravimetric Analysis, and iodine adsorption analysis was carried out in various humidity conditions to optimum grain size.
Methodology: Activated Charcoal (AC) granules of different sizes <0.5 mm, 0.51-0.7 mm, 0.71-1.0 mm and 1.1-2.0 mm and > 2.0 mm are taken for evaluation of surface area, the effect of exposure to temperature, estimation of iodine number at 68%RH (ambient), 30% and at 95% relative humidity conditions.
Surface Area (SA) Analysis: Surface area analyzed for various regeneration temperature revealed that the sample analysed at 300oC for 60 minutes gives more SA. Hence regeneration temperature is fixed at 300°C and SA of various grain sizes was evaluated. As granule sizes decreases the SA increases as shown in [Figure 1].{Figure 17}
Thermo Gravimetric Analysis: The mass change of charcoal granules is monitored as a function of temperature. The entire sample specimen is subjected to a controlled temperature program in a controlled atmosphere. The mass change for granule size is -19.56%, -18.83%,-17.23%, -12.80% and -7-71% w.r.t <0.5, 0.5 to 0.71, 0.71 to 1, 1to 2, & >2.0 mm. From the [Figure 2] >2.0 mm granules size show least mass change in the selected grains sizes.{Figure 18}
Iodine Number: It is a measure of the iodine adsorbed in the pores. The ASTM standard D4607-14 procedure is followed for determination of iodine number. The experimental set-up with titration facility is shown in [Figure 3]. A leak tight SS chamber is used for maintaining various RH conditions. After placing the charcoal granules inside the chamber, RH is increased by purging steam and with continuous monitoring to rise upto 95%.{Figure 19}
The 30% RH exposure is done by heating the chamber using heating coil and continuous monitoring with thermocouples as shown in [Figure 3]. The charcoal granules are exposed at room atmosphere where the ambient environment prevails with 68% RH. The experiments were conducted for the exposure period of 120 h. Iodine adsorption capacity is determined and the results are presented in [Figure 1]. It is observed from [Figure 1], that the charcoal of lower granule size has higher iodine adsorption capacity i.e. higher the surface area the higher the adsorption capacity. It is also found that adsorption capacity does not change with RH condition [Figure 4] for the tested exposure period.{Figure 20}
Conclusion: Hence the experimental results of surface area analysis, TGA and Iodine adsorption analysis show that the activated charcoal of granules sizes of 1mm has optimum surface area and iodine number having less change in TGA which is found to be fit in nuclear industry for long life even at low or high humidity condition.
Keywords: Anisokinetic sampling, efficiency test, HEPA filter
Abstract - 22595: Development and validation of a discrete ordinate method based neutron transport model for criticality safety and shielding analysis in nuclear facilities
R. Kondala Rao, A. Jasmin Sudha
Indira Gandhi Center for Atomic Research, A CI of Homi Bhabha National Institute, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Neutron transport simulation is essential for criticality safety and shielding analysis of fissile systems such as nuclear reactors and fuel reprocessing facilities. The Boltzmann neutron transport equation is solved for computing the distribution of neutrons in the medium by considering various sources and reactions of neutron. A discrete ordinate (Sn) method based numerical model developed for solving the steady state integro-differential neutron transport equation in 3D XYZ geometry is presented in this paper. The solution method adopts a diamond difference scheme with negative flux fix-up for spatial discretization and a discrete ordinate method for angular discretization. Level symmetric quadratures are used for the admissible directions and their weights (up to S8). The fixed source problem is solved using source iteration (SI) technique and the criticality problem is solved using traditional outer-inner iteration strategy.[1] A computer code has been developed and employed to reconstruct the neutron flux and keff results of different fixed source and criticality benchmark problems. The model predictions for the neutron flux in different zones of IAEA-EIR2 benchmark problem are shown in [Table 1]. The geometry of the problem (96 cm x 86 cm x 96 cm) involves absorber (zones 1, 2, 3 and 4) and scattering dominant (5) media with neutron source (1 and 3) distributed non-uniformly and computation takes more time to converge spatially.{Table 14}
In [Table 2], keff values predicted by the code for different benchmark problems[4],[5] are compared with reference results and are observed to be in reasonable agreement. The neutron flux distribution evaluated for Iron-water benchmark problem is shown in [Figure 1]. The number of iterations taken by the present code (1386) for achieving the same convergence (10-4) with SI technique is of the same order (~1000) as that of the 3D nodal method based code[6] for iron-water problem. These results demonstrate the applicability of the code for criticality safety and shielding analysis of nuclear systems. The code is being improved further by implementing advanced iterative techniques for achieving faster convergence. It is also planned to improve the anisotropic scattering treatment in the solution algorithm. The present work is an initial attempt towards developing an advanced neutron transport solver (such as ATTILA and DENOVO) for criticality and shielding problems.{Table 15}{Figure 21}
Keywords: Criticality safety, neutron transport, nuclear facility, numerical modelling, radiation shielding
References
Lewis EE, Miller WF. Computational Methods of Neutron Transport. John Wiley and Sons; 1984.Khalil H. A nodal diffusion technique for synthetic acceleration of nodal Sn calculations, Nucl Sci Eng 1985;20:263-80.Ciolini R, et al. Simplified PN and AN methods in neutron transport. PNE 2002;40.Takeda T, Ikeda H. 3-D neutron transport benchmarks. J Nucl Sci Technol 1991;28:656-69.Warin X. Recent results about SN nodal methods in neutron transport. Note HI-72/96/026/0; 1996.Kokonkov NI, Nikolaeva OV. Consistent P1 synthetic acceleration of inner transport iterations in 3D geometry. Keydish Institute Preprints, No; 2015. p. 28.
Abstract - 22596: Development of methodology for measurement of carbon-14 activity: In stack exhaust air at Kaiga Generating Station-3&4
G. S. Salunke, S. S. Managanvi1, Veerendra Danannavar1, M. Jayasudha1, Ashok Bhatia2, M. Seshaiah2, B. Vinod Kumar2, N. Karunakara3, G. K. Nagaraja4
Chemistry Control Laboratory, 1Health Physics Unit, 2Kaiga Generating Station-3&4, Nuclear Power Corporation of India Limited, Uttara Kannada, 3Center for Advanced Research in Environmental Radioactivity, Mangalore University, 4Department of Chemistry Mangalore University, Mangalagangothri, Karnataka, India
E-mail: [email protected]
Introduction: Carbon-14 is a radionuclide of considerable interest innuclear power production because of its long half-life 5730 years and due to it's high mobility in the environment. Indian PHWR's (Pressurised Heavy water reactors) are fueled with natural uranium and moderated and cooled by heavy water (D2O).The routine operation of this type of reactor and its auxiliary process systems results in the production of variety of solid, liquid and gaseous radioactive wastes. The design and layout of the plant including reactors and associated process systems ensure that release of liquid and airborne wastes are minimized. Nevertheless, very small quantities of these wastes are released. Notable airborne radioactive emissions include Tritium, Radiocarbon (herein after designated C-14 or C14), noble gas radionuclides, radioiodine and particulates, C14 and Tritium are naturally occurring radioisotopes produced continuously in atmosphere by cosmic ray neutron interaction with Nitrogen and Hydrogen respectively and are also produced as a by-product or special products in nuclear reactor systems. C14 is a pure beta emitting isotope of Carbon that is formed in nature as well as inreactors depend on the reactor type, being the highest in PHWRs (Pressurised Heavy Water Reactors) and lowest in Fast Breeder Reactors (FBRS). This paper presents the study done on analytical method. Development of C-14 analysis in stack exhaust air for monitoring of C-14 generated, using liquid scintillation Techniques both time resolved liquid scintillation counting (TR –LSC) instrument and LSC based on TDCR (Triple and double coincidences method. This study was carriedoutatKGS3&4nuclearpower plant (NPP) at Kaiga, India. Experimental setup: The LSC cocktail used in the experiment is commercially available Gold Star (DIN (Di-isopropyl naphthalene isomer)) solvent. All samples were prepared in 20 ml low potassium high performance borosilicate glass vials and counted in Packard Tricarb 2900 TR LSAs (2 No. PE LSA-1 and PE LSA-2) and Hidex 600SLLSA). In Perkin Elmer make LSA (Model: TR-2900) a quench curve was established with a set of C-14 standards. Transformed Spectral Index of External Standard (tSIE) is used as the quench indicating parameter. tSIE is expressed for unquenched sample as maximum of 1000 and is independent of count rate or accumulated counts. In Hidex 600 SL LSA, TDCR method is used for quench correction. In TDCR triple and double coincidences are measured and the ratio of these coincidences is calculated. Triple coincidences are more affected by quench (chemical and color) than double coincidences (or the total of all coincidences). The method is based on a free parameter model describing the processof light emission and detection in a scintillation counter. Background samples were prepared for sample to scintillation solution ratio of 1:5 and 3:15 of DM water and scintillation solution. In order to assess the influence of sample load on counting efficiency, samples were prepared in stack exhaust air. Samples were prepared in the ratio 1:5 and 3:15. Immediately after preparation, the samples were transferred to the sample chamber compartment of LSAs and allowed for one hour. Each sample was counted for one min. Background and efficiency were assessed for each LSA and Figure of Merit (FOM) was calculated. FOM determines the detection limit of any measurement by the signal-to noise ratio. In LSAs this signal-to noise ratio is expressed as a function of counting efficiency (E) to intrinsic background (B). Results are as follows:
[INLINE:2]
FOM is comparable in all the three LSAs with sample to scintillation solution ratio of 1:5 and 3:15. Hence at KGS-3&4,the ratio of1:5 is adopted for analysis of samples. Optimization of analysis technique: Optimization of analytical method was done by studying parameters like counting efficiency, intrinsic background contribution, sample load capacity to arrive at required detection limits, acceptable measurement concentration, sample to scintillation solution volume used for routine measurements keeping in view the liquid waste generation.
Results and Discussion: Results arrived with different sample to cocktail ratio and different LSAs is given in [Table 1]. From the data it is revealed that activities measured with sample to cocktail ration of 1:5 and 3:15 are in good agreement. Hence the ratio of 1:5 can be adopted, which helps in minimize the consumption of scintillation solution. The results arrived in all the three LSAs are comparable. The releases measured at KGS-3&4 were compared with the established values of Advanced Research in Environmental Radioactivity (CARER), Mangalore University, which were in the range of 0.1 to 0.6 TBqGWe-1a-1. Activity measured at KGS-3&4 were in good agreement with the results arrived atCARER.{Table 16}
Keywords: Analytical method, Di-isopropylnaphthaleneisomer, FOM, intrinsic background contribution, liquid waste generation, quench curve, TDCR
References
Approaches for reducing carbon-14 stack emissions from Korean CANDU nuclear power plant. J Nucl Sci Technol 41:235-46.Yim SM, Caron F. Life cycle and management of carbon14 from nuclear power generation. Prog Nucl Energy 2006;48:2-36.
Abstract - 23172: Radiological surveillance during the replacement of expansion bellow at the DFRP stack
K. C. Ajoy, Sunil, Abhishek Maurya, K. Jothi, P. Devesh Ramanan1, K. Kamaraj1, G. Elaiyaraja1, A. Dhanasekaran, R. Santhanam, R. Mathiyarasu, M. Dhananjaya Kumar1, D. Ponraju
Health and Industrial Safety Division, Safety Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, 1Reprocessing Maintenance Division (Mechanical), Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Demonstration Fuel Reprocessing Plant (DFRP) shares a common stack with KAlpakkam Reprocessing Plant (KARP), having two exhaust ducts attached from either side at the same elevation. The connection between the stack and the duct is made through a bellow to reduce the stress to the civil structure created by thermal expansion, vibration, or any abrupt changes in the system pressure.[1] The KARP duct is connected to the stack via a SS bellow, while a canvas expansion bellow was used in the DFRP duct. The bellow was found damaged due to aging and decided to be replaced. Given the potential of contamination, the job was treated as a radioactive job and followed all the necessary radiation protection procedures. The Health Physics surveillance provided during the stack bellow replacement at an elevation of nearly 10m level is discussed in this paper. Initially, continuous air monitoring was done from the duct for nearly three months to confirm that blow-back of particulate activity was not happening from KARP. Safety clearance was obtained from KARP Local Safety Committee (LSC) and decided to execute the job during the shutdown of KARP operations. Industrial safety clearance was obtained and the maintenance group framed a step-by-step procedure. The Health Physics unit prepared a checklist for radiological surveillance during the job. Radiation survey and contamination survey was done to establish the background. The areas were covered with PVC sheets and temporary shoe barriers were erected. Necessary receptacles were also kept to collect used gloves and contaminated materials. A monitoring station was operational at the terrace to quickly assess radiological status. Personnel involved were subjected to pre-operational bioassay before issuing TLDs. A battery-operated air sampler was fixed close to the bellow for air sampling and an improvised telepole arrangement was made for swipe sampling from the inner surface of the duct. Alpha and beta hand and foot monitor was installed near the site and arrangements were made for personnel decontamination. After ensuring the suspension of the process from KARP, work was started by closing the dampers near the stack and shutting down the exhaust systems. After removing the screws at one side, a small opening was made to take a swipe sample which showed BDL for alpha and beta contamination. Air samples were collected continuously and filters were replaced every two hours till the completion of the job. [Figure 1] show the steps involved in the replacement job. The radiological status of the bellow was found to be within background levels. It was covered with PVC, tagged and disposed of as solid waste. Swipes were taken from the inner side of the duct and on the damper plates and found to be BDL. Air samples were tested for a week and no residual activity was detected in the filter paper. The external dose for the individuals was below the reportable level. The bellow was successfully replaced without any radiological consequence within two days. The job was meticulously planned and well-executed, following all the safety precautions. An additional SS hood is used to cover the bellow to mitigate the adverse effects of weathering.{Figure 22}
Keywords: Bellow, contamination, duct, surveillance
Reference
Vinoth A, et al. A review on application of bellows expansion joints and effect of design parameters on system characteristics. Indian J Sci Technol 2016;9.
Abstract - 23176: Airborne radioactivity concentration measurement in CORAL reprocessing facility during off ventilation condition
A. Dhanasekaran, Sunil, Abhishek Maurya, K. C. Ajoy, R. Santhanam, R. Mathiyarasu, D. Ponraju
Health Physics Section, Health and Industrial Safety Division, Indira Gandhi centre for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Compact Reprocessing of Advanced fuels in Lead cells (CORAL) is situated in Kalpakkam and reprocesses the spent fuels discharged from the Fast Breeder Test Reactor (FBTR). The effluents from the facility are released at 116000 m3/h through a 75 m stack. The stack has aviation lamps in platforms at every 25 m height and all the plant exhausts is shut down when maintenance work is taken up during 61st reprocessing campaign at 75 m height. All the exhaust fans were switched off at 1000 hrs and restored on at 16:15 hrs after completing the work. The facility had only the draft created by the stack effect during the work period. The continuous air monitors (CAM) and Alpha spectrometry CAM readings are vigilantly monitored. At 0645 hrs, a baseline air sample (AS) is collected in the OPA and after that, AS are collected periodically in the OPA with suitable protective gears. Finally, an AS is collected approximately after 1 hr after restoring the ventilation systems. All the AS and CAM samples are counted at different counting intervals of > 8 h, > 24 h and > 200 h to measure the long-lived radionuclide concentration. The stack flow meter showed 40000 m3/h after switching off the exhaust. There is a swift increase in readings of the CAMS located in all plant areas. The highest rise is observed in the OPA CAM. [Figure 1] shows the trend of the OPA CAM readings in terms of Eq. DACh and Eq. DAC. The online alpha spectrum in ASCAM had natural radon, thoron daughter peaks only and is shown [Figure 2]. The maximum DACh value observed in the ASCAM, which calculates DACh using artificial alpha region counts, is 2.5 DACh, in contrast to conventional gross alpha CAM, which registered around 400 DACh. The baseline air sample collected in the OPA showed 0.4 Bq/m3 of 212Bi concentration, reached a maximum of 23 Bq/m3 and reduced to 0.2 Bq/m3 within an hour of ventilation restoration. The size-selective air sampler showed 0.5 Bq/m3 when the large area sampler had the highest concentration. The collection efficiency of the size-selective air sampler for natural aerosol is 2-3%. The OPA's ACH estimated based on the 212Bi concentrations measured during ventilation off and on conditions is 3.5- 7. [Table 1] shows the delay counting details of the 7 CAM and ASCAM filter papers. The standard two-count method is applied to measure the long-lived activity. It is observed that the residual activity in the filter papers during the second counting interval is within the statistical fluctuation of the expected activity in the filter paper calculated using the first count. All the CAM, ASCAM and spot air samples showed below detectable level of activities (MDA=0.098 DACh) when counted > 200 hrs of delay. One important observation from this study is that the draft due to the stack effect is enough to maintain the containment's integrity during off ventilation condition. Further, the trend of 212Bi concentration demonstrates that the ventilation systems can provide a protection factor of 100.{Figure 23}{Figure 24}{Table 17}
Keywords: Air monitoring, centripeter, grab sampling, thoron
Abstract - 23194: Particle size distribution characteristics of cesium nitrate [CsNO3]), Strontium nitrate [Sr(NO3)2] and standard test aerosols using atomizer
G. Ganesh1, A. M. Shinde1, M. S. Kulkarni1,2
1Health Physics Division, BARC, 2Homi Bhabha National Institute, Anushakti Nagar, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: The radioactive aerosols generated during various operations in nuclear fuel cycle facilities such as reprocessing and waste management facilities mainly consist of Cs and Sr nitrate compounds in the form of CsNO3 and Sr(NO3)2. Inadvertent inhalation of these contaminants may lead to intake and internal exposure to the radiation workers. Intake expressed in terms of ALI (Annual Limit on Intake) depends on the particle size. The dust respirator used for preventing internal exposure in nuclear facilities plays an important role in radiation protection. The respirator filter media should have the ability to trap the specified contaminant to the desired level which ultimately depends on the particle size distribution of the aerosols. Hence the present study aims to carry out the particle size distribution of Cs and Sr nitrate aerosols in the range of 0.25 to 32 μm. and compare the results with the standard test aerosols.
Materials and Methods: The aerosol study setup as given in [Figure 1] is designed to carry out the study of particle size distribution for selected aerosols using aerosol generator which is connected to Optical Particle Counter (OPC) via a 5 litre sampling chamber. The major components of this aerosol study setup are Aerosol generator cum Atomizer, Sampling chamber and Optical Particle Counter (OPC) as given below in [Figure 1].{Figure 25}
Experimental: In the experiment, the aerosols of CsNO3 and Sr(NO3)2 compounds, standard test aerosols of 2 % NaCl,[1] Di-Octyl Phthalate (DOP) and Ultra-pure water generated using aerosol generator are fed to the sampling chamber from which the aerosols are drawn into OPC for counting. The data acquired in an OPC (Grimm Make Dust Monitor Model 1.109 Version 12.30 by M/s GRIMM AEROSOL TECHNIK GmbH & Co. KG, Germany) is analysed for different particle sizes of aerosols of CsNO3, Sr(NO3)2, Ultra-pure water, Sodium Chloride (NaCl) (Bureau of Indian Standards, IS 9473:2002) and DOP in the range of 0.25 to 32 μm. For all experiments, 5 minutes is considered for the attainment of equilibrium concentration.[2] For each type of aerosol, 50 data points obtained in each experiment to arrive at the average concentration (no. of particles/litre) in each size range and its percentage to the total counts. Water aerosols were considered due to use of steam in the facilities which often carries activity in the form of condensed water in air activity releases. The [Figure 2] presents a graphical plot of percentage contribution versus particle size of NaCl, DOP, Nano pure water, Sr(NO3)2 and CsNO3 aerosols.{Figure 26}
Results and Discussion: The studies indicate that the aerosols of CsNO3 and Sr(NO3)2 compounds show concentration of 48.12 and 44.26 % respectively in the range of 0.25 to 0.3 μm whereas the concentration is about 90% observed in the range of 0.25 to 0.5 μm. This is important from respiratory protection point of view for testing various filters which exhibit maximum penetration of aerosols in the range of 0.28 – 0.3 μm.[3] The size distribution of the aerosols of CsNO3 and Sr(NO3)2 is comparable to that of standard NaCl aerosols unlike DOP aerosols.
Keywords: Aerosol, Cs, particle size, respiratory protection, Sr
References
Bureau of Indian Standards, IS 9473:2002, Respiratory Protective Devices, Filtering Half Masks to Protect Against Particles.Ganesh G, et al. J Radiat Protect Environ 2019;42:77-83.Perry JL, et al. NASA/TM–2016–218224, Submicron and Nano-Particulate Matter Removal by HEPA-Rated Media Filters and Packed Beds of Granular Materials; 2016.
Abstract - 23196: Contribution of natural background radiations from human body during 238U lung content measurement
A. Y. Balbudhe, D. Praveen, K. Vishwa Prasad, S. K. Jha1, M. S. Kulkarni1
Health Physics Division, Bhabha Atomic Research Centre, Hyderabad, Telangana, 1Health Physics Division, Bhabha Atomic Research Centre and HBNI, Trombay, Mumbai, Maharashtra, India
E-mail: [email protected]
Nuclear Fuel Complex (NFC) manufactures Natural Uranium (NU) fuel for the nuclear power reactors. All the three natural isotopes of uranium emit mainly alpha radiation and may pose internal radiation hazard. Internal contamination monitoring of individual radiation workers is the important aspect of effective radiation protection program. Realistic background correction is essential for in vivo estimation of NU. The paper presents data of unexposed persons. 203 mm (dia.) x 12.7 mm (thick) NaI(Tl) detector with single PMT associated with plug on MCA with InterWinner 7.0 gamma analysis software inside graded shielding steel room facility (203mm steel + 3mm Pb + 2mm Cd) is placed for lung monitoring. 234Th, immediate progeny of 238U emits gamma of 63.3 keV(3.8%), 92.4 keV(2.7%) and 92.8 keV (2.7%); which are used for the measurement of 238U. NU is estimated from 238U activity knowing the fact that 48.8% activity is due to 238U in NU. Considering the resolution of the system counts in the region of 40-120 keV (ROI) are considered for measurement of 238U. Earlier similar studies were carried out for 127 mm x 12.7 mm NaI(Tl) detector system.[1] Apart from background due to cosmic and terrestrial radiation 40K present in the body of individual being monitored is an important source of background in lung monitoring. The average 40K content in a human body weighing 70 kg is 4.4 kBq[2] which emits a gamma of 1460.8 keV (10.7%). The Compton continuum due to 1460.8 keV is the major and variable interfering factor for in vivo measurement of 238U. 419 persons who were never exposed to any radioactivity either occupationally or for medical purpose were monitored. The distribution of mean count rate (cps) of persons in different weight group and gender is shown in [Table 1]. The plots in [Figure 1]a and [Figure 2]a shows cps in the ROI against person's weight and weight/height (W/H) ratio, respectively. The [Figure 1]b and [Figure 2]b shows the variation of cps against the midpoint of weight group and W/H group, respectively. It is observed from the figures that cps in the ROI increase with person's weight as well as W/H ratio. Variable cps in the ROI is mainly contributed by 40K in the person's body. Potassium body content was observed to increase with weight and body build index (W/H) by Thulasi Brindha et al. (2006)[3] at Kalpakkam, India. Similar trend of cps in 238U ROI is observed in present study. cps in the 238U ROI for unexposed persons found to vary from 4.69 to 6.41 cps with geometric mean 5.43 cps and geometric standard deviation of 1.059. The data of cps in the ROI with the relevant physiological parameters of the individuals will be useful in realistic correction for 238U lungs content measurement.{Figure 27}{Figure 28}{Table 18}
Keywords: In vivo, lung, uranium
References
Padma Savitri P, Balbudhe AY, Praveen D, Srivastava SK, Vishwa Prasad K, Ravi PM, et al. 32nd IARP Conference, Kalpakkam (India), IARPIC-2016, Background for In vivo Estimation of Natural Uranium; 2016.Rao DD. Radioactivity in human body and its detection. Radiat Prot Environ 2012;35:57-8.Thulasi Brindha J, Rajaram S, Kannan V. Comparative study of body potasium content in males and females at Kalpakkam (India). Radiat Prot Dosimetry 2007;123:36-40.
Abstract - 23198: Trend analysis of long-lived airborne contamination in a reprocessing facility
A. Chakraborty, Neeraj, D. K. Pandey, O. Sahu, Pankaj Kumar, J. P. N. Pandey, G. Ganesh
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: The major sources of airborne radioactivity during Operation and Maintenance activities in a reprocessing facility are 137Cs, 90Sr and α emitting radionuclides. The regulatory body stipulates that the long-term trends for external radiation levels and airborne contamination levels of a radiation facility should be well identified and documented, so that deterioration in radioactive material handling equipment or change in operating conditions can be promptly identified and remedial actions could be initiated. The trend analysis of gamma radiation levels is straightforward real-time data acquisition systems. A different approach is required for airborne contamination, since it involves analysing results of delayed counting of Continuous Air Monitor (CAM) GFA filter papers (FPs). This paper makes an attempt to study the trend of long-lived airborne contamination in a reprocessing facility and develop a methodology to draw conclusions about the integrity of radioactive handling equipment and engineered safety measures to control the airborne contamination. Materials and Methods: The airborne contamination data for two-year period i.e., 01-Jan-2020 to 01-Jan-2022 was used for this study. This involved collating 4-day delay counting activity measurements of CAM FPs for 45 installed monitors in 18 distinct locations (denoted as L-1 to L-18), and statistical analysis of 63,000 data points. Using the flow rate of CAMs (70 lpm), the sampling time (24 h) and efficiency of α and β counting systems (25% and 10% respectively), the Airborne Particulate Radioactivity (APR) concentration (in mBq/m3) of various locations of the facility have been estimated. The airborne α activity concentrations in different locations are given in [Figure 1]. Similar bar-plots have been plotted for β activity concentration also. The timeseries (indexed by date) for α and β activity concentrations in all locations have been plotted. The outlier observations (3σ away from the mean) have not been considered for trend analysis. The normalized values of the observations w.r.t. minimum and maximum concentrations for location L-16 are shown in [Figure 2] as an example.{Figure 29}{Figure 30}
Results and Discussions: The APR concentration of a location varies over time, and observing the timeseries is not a quantitative method to ascertain the rising trend of air activity in a location. Hence, the time series of α and β airborne activity concentration in all locations were subjected to Augmented Dickey-Fuller (ADF) test.[1] The timeseries of airborne α and β concentration were found to be stationary with significance of 1% (i.e., 99% confidence of stationarity). A non-stationary timeseries for airborne concentration would have indicated a source of airborne activity.
Conclusion: Our study shows that APR concentration in controlled areas of the facility is stationary i.e., it has a constant mean and constant variance which indicates that there is no source of airborne activity. Similar analysis can be carried out for other facilities to identify the deterioration of safety measures or ventilation systems.
Keywords: Airborne contamination, stationarity of time series
Reference
Leybourne SJ, Mills TC, Newbold P. Spurious rejections by Dickey–Fuller tests in the presence of a break under the null. J Econom 1998;87.1:191-203.
Abstract - 23202: Particle size distribution of Pu aerosols during reconversion operations in a reprocessing plant
D. K. Pandey1, A. Chakraborty1, Neeraj1, Pankaj Kumar1, J. P. N. Pandey1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Center, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: At reprocessing plant, Plutonium nitrate solution is converted into PuO2 through oxalate precipitate route in reconversion laboratory. During handling of Pu there is a potential for release of Pu aerosols of variable size in the working environment due to breach of containment or ventilation disturbance. Presence of Pu aerosols in working environment may lead to internal hazard due to inhalation. For estimation of internal dose due to Pu intake, 5 μm AMAD is commonly accepted as the default particle size in the dosimetric calculations for occupational exposure. However, it is recommended to use site specific parameters of radioactive aerosol wherever possible.[1] Therefore, a study was carried out for the characterization of Plutonium aerosols during reconversion operations at a typical reprocessing plant. This study will help in the assessment of internal dose due to intake of Pu for occupation workers.
Materials and Methods: Sampling of Pu aerosols was carried out at reconversion lab using Particle Aerodynamic Size Separator (PASS) at a flow rate of 45 lpm. PASS segregates particles in size ranges of 0.53-8.95 μm in seven class intervals with a backup filter for collection of submicron particles.[2] Glass fiber filter (GFA) papers of diameter 2.95” were loaded in collection plates at all stages for collection of particulate activity. Sampling time was optimized to 48 hours to get sufficient long-lived activity collection. Large number of samples were collected however only fifteen such samples showed measurable activity which were considered for evaluation of AMAD. Alpha activity of Pu aerosols collected on each stage was estimated using ZnS (Ag) detector with 25% efficiency. The MDA of the detector was 0.828 mBq/m3 for sampling time of 48 h and counting time of 500 s.
Results and Discussion: Average alpha activity of Pu aerosol of each stage of PASS was worked out. Results are presented in bar graph for interpretation of distribution pattern as shown in [Figure 1]. AMAD and GSD of Pu aerosols for all 15 samples were estimated by plotting a graph of cumulative percentage less than stated size versus effective cut off diameter (ECD) on log probability scale as shown in [Figure 2]. For each sample, the linear regression line was evaluated y = a + bx where y is “cumulative percentage less than stated size” and x is the ECD. AMAD for each sample was calculated by evaluating x at y = 50. GSD for each sample is calculated by using ratio of x at y = 50 to x at y = 16. The AMAD of Pu aerosol in Pu reconversion laboratory is found in the range 3.06 μm to 8.98 μm and the average AMAD representative for the reconversion area is 5.25 ± 0.48 μm with a GSD of 4.02 ± 0.20. Aerosol activity is widely distributed with respect to aerodynamic diameter. Higher values of GSD may be due to handling of mixture of compounds in different chemicals form.{Figure 31}{Figure 32}
Conclusion: Particle size distribution during Pu reconversion operations showed mean AMAD of 5.25 ± 0.48 μm with GSD 4.02 ± 0.20. Based on these AMAD and GSD values, the dose coefficient and ALI values can be derived for use in such operations which will help in an accurate estimate of the intake in case of any internal exposure to radiation workers.
Keywords: AMAD, effective cut off diameter, gauntlets, GSD, impactor, plutonium
References
ICRP. Human respiratory tract model for radiological protection. ICRP Publication 66. Ann ICRP 1994;24.Vishwa Prasad K, et al. Aerosol size distribution in a uranium processing and fuel fabrication facility. Radiat Prot Dosimetry 2010;140.
Abstract - 23213: Particle size characterization of aerosols generated during dismantling of contaminated cell equipment
Ashish Kumar Singh1, U. V. Deokar1, A. R. Khot1, Prabha Mathew1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Objective of the current study is to measure the activity distribution of the aerosols generated during the dismantling of vitrification cell equipment. The removal of piping and components from vitrification cell is currently underway. Air samples were taken utilising a Particle Aerodynamic Size Separator (PASS) cascade impactor during the grinder cutting of radiologically contaminated high level waste process equipment in the vitrification cell. The experimentally determined particle distribution was compared with default AMAD value for occupational workers to see the inhalation risk during dismantling operation. A study was also performed to determine the isotopic composition of the aerosols generated in the vitrification cell. The generation of radiologically contaminated aerosols during the dismantling and demolition of vitrification cell equipment presents a number of challenges to the health physics professional. Primary among these is the risk of internal deposition of radioactive particles due to inhalation by occupational workers. Secondarily is the risk of the spread of contamination to the public and environment from the uncontrolled release of radioactive dusts and aerosols.[1] The intent of present study was to determine the activity median aerodynamic diameter (AMAD) and geometrical standard deviation (GSD) of aerosols sampled during grinder cutting of vitrification cell equipment. Assuming a log-normal particle size distribution, the following formula is used to calculate the particle size Geometric Standard Deviation (GSD) (σg):
[INLINE:3]
Aerosols were sampled using 7-stage PASS cascade impactor developed by EAD, BARC. The PASS impactor operates at a flow rate of 45 lpm and segregates particles in the size range of 0.53-8.95μm in seven class intervals. In this work, glass-fiber filter papers are used because they have a higher pressure drop than cellulose filters and provide good filtration efficiency (> 99%). To obtain a representative sample, the impactor was placed approximately 1 metre from the dismantling equipment using an in-cell crane and maintained in a downwind location. The average sample collection time was 1-2 hours, depending on the work scope for the day. During the cutting of pipes and support structures, three sets of samples were collected. The obtained sample was analysed for its isotopic composition using a High Purity Germanium (HPGe) detector. During the demolition, Cs-137 was found to be the dominant radionuclide in the vitrification cell. The activity data sets were developed collectively by combining the three sets. The results were used to develop aerodynamic diameter distribution table [Table 1] and AMAD and GSD were then determined by plotting the data on a log-probability plot [Figure 1]. The AMAD was determined to be 5.6 μm with its GSD equal to 1.6. The 137Cs distribution obtained in HPGe confirms predominant radionuclide in the facility. The obtained AMAD value is consistent with the default AMAD value for occupational workers i.e. 5 μm as referenced in ICRP 66.{Table 19}{Figure 33}
Keywords: Dismantling, particle size, vitrification
Reference
MacMillan WJ, Brey RR, Harris JT. Particle size characterization of aerosols generated during surface contaminated concrete demolition. Health Phys 2013;104:S83-6.
Abstract - 23221: Shielding evaluation of mini radiological facility for storing radioactive samples
Lopa Basak1, Brij Kumar1, P. Bhargava1, K. D. Singh1 and M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Radioactive beamline is required at bending magnet port at a synchrotron source to carry out investigations on radioactive samples such as post-irradiation fuel pin samples etc. using XRD, XRF and XAS techniques. The beamline hutch will also have a mini radiological facility to handle and store radioactive samples. A study on the radiation shielding effectiveness for the said radiological facility and experimental hutch has been conducted for safe operation and to limit the dose within regulatory limit.{Figure 34}{Figure 35}
Materials and Methods: Maximum rated spent fuel bundle/pin with a burn-up of 10000 MWd/t and cooling period of 1 year after the irradiation is considered in the present case for the estimation of the source within fuel pin.[1] The methodology in brief involves estimation of combined 18-group gamma ray energy spectrum in the spent fuel pin, using the ORIGEN2 code.[2] This is followed by making use of this spectrum in the estimation of gamma dose rates using fluence to dose conversion coefficients (ANSI/ANS–6.1.1, 1979) at representative receiver locations on the outer surface of the shielded walls, using the point-kernel method based code QAD-CGGP.[3] The total dose rate so calculated must meet the acceptance criteria for occupancy of that area which is 0.1 mR/hr for full time occupancy (AERB/NPP-PHWR/SG/D-12, 2005).
Results and Discussion: Post-irradiated fuel pin samples of radius 3.6mm and thickness 2mm (corresponding mass ~ 4g and total activity 1.58E+11 photons/sec) has been considered in the estimations of shielding thicknesses and dose rates at the required receiver locations. The distances from the source to detector are approximately same (~ 120 cm) in all the locations from A to F. The estimated shielding thicknesses and the dose rates at different locations are shown in [Table 1].{Table 20}
Conclusion: It can be seen from [Table 1] that Lead thickness required to achieve the acceptable dose rate criteria is maximum (12 cm) at the center of the sample and reduces with the elevation.
Keywords: Dose rate, ORIGEN2, QAD-CGGP, radioactivity, shielding
References
Bajaj SS, Gore AR. Nucl Eng Des 2006;236:701-22.Croff AG. Nucl Tech 1983;62.QAD-CGGP Manual. CCC-0493/01; 1989.AERB Guide. AERB/NPP-PHWR/SG/D-12; 2005.American Nuclear Society. ANSI/ANS – 6.1.1. American National Standards Institute; 1979.
Abstract - 23222: Shielding Effectiveness Analysis of different materials for active pipelines duct
Lopa Basak1, Brij Kumar1, P. Bhargava1, K. D. Singh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Waste evaporation and storage block is designed to concentrate the raffinate waste generated from the reprocessing and store the concentrated liquid waste before transferring to other plants. There is a matrix of 4.4m long pipe lines having radius 1.25cm located inside the duct for transferring radioactive process liquids which passes through the inactive corridor. These lines are required to be shielded to keep radiation exposure within regulatory limit inside the corridor. A study on the effectiveness of different shielding materials for the duct carrying active pipe lines has been conducted and results are presented in the present paper.
Materials and Methods: The methodology in brief involves estimation of the shielding thicknesses of the duct carrying active pipe lines using different shielding materials, followed by calculation of the gamma dose rate at a given representative point on the outer surface of the duct due to the all active pipe lines through which radioactive process liquids flow. The total dose rate so calculated must meet the acceptance criteria for occupancy of that area which is 0.1 mR/hr for full time occupancy (AERB/NPP-PHWR/SG/D-12, 2005). Dose rate at the designated locations outside the shield is calculated by using the QAD-CGGP code (QAD-CGGP, 1989) for 1 Ci of Cs-137 activity passing through the each pipeline A, B and C. QAD-CGGP is a FORTRAN 77 code which uses the point-kernel method, in which the original source volume is divided into a large number of small volumes of voxels and the dose rates from all voxels are summed up to give the dose rate due to the entire volume of the source. The formula of dose rate calculation is given below,
[INLINE:4] = point at which gamma dose rate is to be calculated,
r̄′ = location of source in volume V,
V = volume of source origin,
μ = total attenuation coefficient at energy E,
[INLINE:5] = distance between source point and point at which gamma intensity is to be calculated,
K=flux-to-dose conversion factor; B=dose buildup factor. This formula has been implemented within the QAD-CGGP code by the developers.
Results and Discussion: Dose rates were estimated at the surface, 1 m and 2 m distance from the surface of the duct for different materials which is shown in [Table 1] and at a time only 3 pipes (A, B and C) will be operational which is shown in [Figure 1].{Table 21}{Figure 36}
Conclusion: It can be seen from [Table 1] that for 1 Ci of active liquids, the required shielding thicknesses are 19 cm for MS, 7.2 cm for Lead and 60.8 cm for concrete to meet the acceptance criteria which is less than 0.1 mR/hr.
Keywords: Curie, dose rate, QAD-CGGP, radioactivity, shielding
References
AERB Guide. AERB/NPP-PHWR/SG/D-12; 2005.QAD-CGGP Manual. CCC-0493/01; 1989.Martin JE. Physics for Radiation Protection: A Handbook. Weinheim; Wiley-VCH Verlag GmbH & Co. KGaA; 2006.American Nuclear Society. ANSI/ANS – 6.1.1. American National Standards Institute; 1979.
Abstract - 23265: Assessment of radiation level and exposure control during in-cell crane maintenance work in vitrification cell
P. Mathew, U. V. Deokar, Ashish Singh, A. R. Khot, G. Ganesh
Health Physics Division, BARC, Trombay, Mumbai, Maharashtra, India
E-mail: [email protected]
Advanced Vitrification System is used for the vitrification of High Level Waste (HLW) using Joule Heating Ceramic Melter (JHCM). JHCM is located in a Hot Cell having dimensions 7115 mm (L) x 4000 mm (W) x 8000 mm (H) and in-cell crane is utilized for all in-cell operations.[1] The in-cell crane experienced a malfunction in its LT motion during the earliest stages of the plant's operation. This had resulted in hindrance to all in-cell operations and stoppage of vitrification. Vitrification cell was having very high inventory of activity (5275 TBq) and estimated general background of the hot cell was 27- 30 Gy/h. Because of this high radiation level inside the cell it was not possible to do manual intervention for maintenance of in-cell crane. So, it was planned to do maintenance of the crane remotely. For remote maintenance openings were made for the installation of master slave manipulator (MSM), material movement and insertion of camera for remote viewing. Radiation field inside the cell at the inner and outer openings at four locations were theoretically estimated using IGSHIELD code for the calculation of collective dose budgeting. Joule Melter with dimensions of 1450 mm (L) x 1450 mm (W) x 2024 mm (H) and located 1.3 m above the cell floor was considered as major source for theoretical calculations using IGSHIELD. Online continuous air monitoring was used to keep track of the air activity during drilling work. Extensive radiological surveillance was provided for the safety of occupational workers. In addition to that, provision of movable shielding and mockup trial was also given. In-Cell crane maintenance work was planned to be carried out remotely. Provision of Continuous Air and radiation monitors were made on the working platform. Openings 1 and 2 with diameters of 250 mm each were used for MSM installation, while openings 3 and 4 with diameters of 75 mm and 80 mm, respectively, were used for camera insertion and material movement and were located at heights of 7025 mm, 7650 mm, 7775 mm, and 5000 mm above the cell floor. These opening locations were made available with sliding shielding arrangement whose shielding thickness was estimated using the IGSHIELD code. The concrete debris removed while drilling was having contamination of 3-5 Bq/gm. After making openings, radiation survey was carried out at different locations by inserting in-cell monitor. All the estimated radiation level using IGSHIELD was based on the activity inventory of vitrification cell at the four drilled opening locations were in good agreement with measured radiation level [Table 1]. The movable shielding provided on the drilled openings has helped in reducing personal exposure considerably. The work was planned properly and number of inactive trials and mock ups were made to reduce personal exposure which further contributed in the reduction of collective dose.{Figure 37}{Table 22}
Measured dose rate at all four locations are matching within 10% of estimated value by code IGSHIELD. Significant reduction in collective dose could be achieved by good planning, conducting mock-up trials, training, provision of movable shielding, online radiation and air activity monitoring and by using personal radiation protective equipments. Day to day work was covered under special work permit for effectively implementation of proper safety measures. All the above efforts leads to the reduction of no. of trails that's why work was completed within dose consumption of less than 10% of the allocated dose provided by the ALARA committee.
Reference
Deokar UV, Mathew P, Kulkarni VV, Deshpande MD. Radiological Safety Aspects During Operation of Advanced Vitrification System. Tarapur: 28th IARP National Conference on Management of Nuclear & Radiological Emergencies, ORP-1; 2008. p. 160-3.
Abstract - 23267: Design and development of calibration facility for health physics instruments
A. R. Khot1, U. V. Deokar1, P. Mathew1, Ashish Singh1, Sonali Khurana1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, BARC, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Calibration of radiation detection instruments is required to ensure that the measurements performed with these instruments are reliable and accurate. Hence, a calibration facility in shielded cubical for portable radiation instruments as well as Area Gamma Monitor (AGM) having the range up to 1 mGy/hr and 100 mGy/hr was set up. The calibration facility is equipped with features such as master slave manipulators fitted to the shielded cubical, remotely operated trolley and radiation shielded viewing window. Inside shielded cubical, two Cs137 sources of strength 0.814 and 74 GBq in shielding casks are kept. Radioactive Plant comprise of large number of installed and portable monitors to alert plant personnel about the radiation status. A network of Area Gamma Monitors (AGM), connected with Centralized Radiation Protection Console (CRPC) best serves this purpose. It is a statutory requirement that all instruments used for the monitoring of controlled or supervised areas are to be tested and calibrated on a regular basis.[1] The calibration facility established in shielded cubical have Radiation Shielding Window (RSW) and two master slave manipulators. The thickness of the cubical wall is 200 mm CS. The inside dimension of the shielded cubical is 1500 mm X 1500 mm. Radioactive source kept inside the lead shielded casks fitted with remotely operated motorized door to expose the radiation source for calibration of the instrument. The casks have a lead shielding of thickness 65 mm and 100 mm for 0.814 and 74 GBq sources respectively, as evaluated by point kernel computer code.[2] Remotely operated trolley arrangement was fitted for allowing the movement of instrument to be calibrated with respect to the source. To measure the radiation fields for different relative distances between the source and the instrument, Programme logic Control (PLC) is incorporated. With PLC controlled mechanical trolley, the radiation measurement is possible for different input distance from the source automatically. An extended probe AGM is installed outside the door of the cubical with its detector probe kept inside the cubical having audio-visual alarm set at 10 μSv/hr. Additional safety feature is incorporated by interlocking the cask door with cubicle door so that cask door can be open only if cubicle door is closed and AGM radiation level ≤ 10 μSv/hr. To keep exposure consumption as low as reasonably achievable, it is ensured that the radiation field out the cubical is always ≤ 10 μSv/hr. The result obtained at calibration facility as per theoretical calculations for three distances are in concordance with inverse square law. For validation of calibration facility, AGMs of different ranges calibrated at our facility were send to accredited Laboratory Radiation Standards Section, RSSD, BARC Trombay for re-checking. The results of calibrations obtained at both the facilities are in good agreement within the Relative Intrinsic Error (RIE) of less than ± 10%.{Figure 38}{Figure 39}{Table 23}
Keywords: Area gamma monitor, calibration, point kernel method, shielded cubical, validation
References
International Atomic Energy Agency. Calibration of Radiation Protection Monitoring Instruments. Safety Reports Series No. 16. Vienna: International Atomic Energy Agency; 2000.Subbaiah KV, Sarangapani R. IGSHIELD: A new interactive Point kernel Gamma Ray shielding code. Ann Nucl Energy 35:2234-42.
Abstract - 23284: Planning, preparation and decontamination of an alpha handling fume-hood
V. Hemachandar, Gopal P. Verma, D. K. Patre, P. Ashok Kumar, M. S. Kulkarni
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Fume hoods are one of the most important equipment applied to reduce the potential of airborne contamination in laboratory environments. A Number of radioactive solutions are being handled during radiochemistry experiments related to actinide chemistry. Some of them are, pre-concentration by evaporation, column separations, mixing of radioactive solutions, sample planchatings etc. These procedures lead to generations of radioactive fumes which get settled on fume-hood's inside surface and increases radioactive contamination levels on the walls, duct, inner glass surface and the floor of the fume-hood. The present paper discussed about the planning preparation and decontamination procedure of an alpha handling fume-hood having high contamination levels on the inner surfaces.
Material and Methods: The fume hoods are design to give clean environment to the persons handling radioactive solutions during experiments. It has a box structure with net operating area of about ~1 m2 and volume of 1m3. It has front opening fitted with a laminated safety glass panel with two circular cut outs to perform operations in the fume hoods. The hood wall and back surfaces are made of mild steel coated with anticorrosion paints which provides smooth surface reasonably good for removal of surface contaminants. The fume hood is connected to laboratory exhaust with laminar air flow rate of 100 - 120 linear feet per min through the openings. Over the years, the experiments in this hood included compound mixing and heating in acids and bases, working with volatile compounds and various gases. These activities result in accumulations of radioactive materials and higher levels of inner wall surface contaminations of a fume-hood. The first step in the decontamination was to conduct a reasonable assessment of the surface activity concentration within accessible areas of the hood. Which showed an approximate “surface alpha contaminations” of 30 to 100 Bqcm-2. The dose rate at the glass panel was 10-15 mRh-1. The fume hood cleaning job was planned after discussion in safety committee (PLSC). A SOP was prepared and approved which was meticulously followed during the entire campaign. A tent was erected in front of the fume hood and with double rubber station. The cleaning campaign lasted for fifteen days. The job was done using fresh air lines and double protective suits. The active materials accumulated inside the fume hood, were shifted in different batches inside glove boxes. The qualitative analysis of each batch was carried out using Cadmium Zinc Telluride (CZT) based portable gamma spectrometer, the results are given in the [Table 1]. The smear survey of all surfaces of the hood was conducted before and after the decontamination.{Table 24}
Results and Discussion: The activity on the back and top surfaces of the hood was higher than the other areas. Several attempts of decontaminations were taken to clean the surfaces. The decontamination factor achieved in each attempt is listed in the [Table 2]. The used smear papers were preserved for the recovery of actinides if any. Gamma spectrometry analysis of the recovered sample solutions mainly showed presence of Transuranic isotopes, 137Cs, and 60Co. The cleaning campaign was successfully completed without any overexposure or internal contamination.{Table 25}
Keywords: 137Cs, airborne, alpha, fume hood
References
Remark JF. A Review of Plant Decontamination Methods, Final Report, EPRI NP-6169. Marietta, Georgia: Applied Radiological Control, Inc.; 1989.O'Dou TJ, Bertoia J, Czerwinski KR. Decontamination of a technetium contaminated fume hood in a research laboratory. Health Phys 2011;101 Suppl 2:S124-30.
Abstract - 23303: Shielding analysis for contactor cell and hull monitoring set up of DFRP
Pew Basu1,2, R. Sarangapani1, M. Menaka1, V. Subramanian1,2, B. Venkatraman1,2
1Safety, Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, 2Homi Bhabha National Institute, Training School Complex, Anushaktinagar, Mumbai, Maharashtra, India
E-mail: [email protected]
In demonstration of fast reactor fuel reprocessing plant (DFRP) various operation of reprocessing of spent fuel from FBTR and PFBR are planned. There are two contactor cells (177 & 175) in DFRP where operations such as solvent extraction, solvent washing, clarification of washed solvent, maintenance of centrifugal extractors, maintenance of centrifuge and waste disposal etc. are planned to be carried out. The cells are made of ordinary concrete of thicknesses of 1 m at all sides. There are four numbers of Centrifugal Extractors (CE) such as HA, HC, RA and RC located at 177 cell and 1A, 1C, 2B and 2C located at 175 cell and placed in a containment box. In the east side and south side of the 177 cell, instead of 1 m concrete equivalent lead of 0.25 m is present. There exist a gap of 4 mm between the concrete structure and lead walls at four locations. The CEs have capacity of 3 litres corresponding to the fission product activity of 1.48E+03 Ci under normal condition. The modeled geometry of the contactor cell in IGSHIELD[1] code is shown in [Figure 1] along with the array of detectors (D1 to D14) at the operating area. The 4 mm gap between the lead wall and concrete structure is assumed as a rectangular duct of length 25 cm (L), width 4 mm (W) and breadth 2.35 m (B). The formula used for the calculation of dose rate at the duct exit due to an isotropic disk source is
[INLINE:6]
{Figure 40}
where, D0 is the field at the duct mouth (source side), D is the field on the other side of the duct (operating area side) which is to be calculated. The maximum gamma dose rate obtained at the east side of the operating area is estimated to be 0.83 μSv/h and 0.90 μSv/h for PFBR and FBTR spent fuel spectrum respectively which is within the continuous occupancy limit of 1 μSv/h set by AERB. However, the streaming dose rate just outside the 4 mm gap between lead wall and concrete structure is calculated to be 2.34E+04 μSv/h. To reduce the dose rate to 1 μSv/h, 25 cm thick lead or equivalent shielding has to be provided in the 4 mm gap area to avoid streaming of photons. Gamma based hull monitoring technique is adopted in DFRP to estimate the activity of Plutonium in leached hulls after dissolution. Traces of fission products present in the hulls causes a major problem in hull monitoring. The essential requirement is to have a minimum radiation background at the detector place and the dose rate limit is set to be 10 μSv/h. The background activity is mainly from dissolver handling fission product activity corresponding to 40 spent fuel pins of activity 4120 Ci. Therefore, background activity around the hull monitoring set up is optimized with the help of special arrangement of lead shielding around the hull basket and detector. Shielding analysis for this hull monitoring set up has been carried out using IGSHIELD code. The diagram of hull monitoring set up is shown in [Figure 2]. It is observed that with 180 mm of thick lead instead of 250 mm (designed thickness) surrounding the detector dose rate criterion is meeting provided the height of the lead shielding around the hull basket is kept fixed.{Figure 41}
Keywords: Contactor cell, DFRP, hull monitoring set up, IGSHIELD
Reference
Subbaiah, et al. Ann Nucl Energy 2008;35:2234-42.
Abstract - 23305: Gamma ray backscattering estimation towards in-situ calibration of area gamma monitor
Pew Basu1,2, M. Menaka1, V. Subramanian1,2, B. Venkatraman1,2
1Safety, Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, 2Homi Bhabha National Institute, Training School Complex, Anushaktinagar, Mumbai, Maharashtra, India
E-mail: [email protected]
An in-situ AGM calibrator is planned to be set up at Radiological & Environmental Safety Division (RESD), Indira Gandhi Centre for Atomic Research (IGCAR). The system allows the calibration set up to be taken to the installed location of the AGMs and hence, avoid the labor for removing and refixing the instruments and facilitates uninterrupted active jobs. Operating the system from a distance of about 1 meter is possible that results in the minimization of exposure to operating personnel. The complete portable setup and its different view are shown in [Figure 1]. A significant and vital aspect of in-situ AGM calibrator is that the calibration will be performed with the AGMs mounted on the wall and the gamma ray scattered from back wall will contribute to the dose rate. Hence, the gamma ray wall scattering component has been estimated in the present study using Monte Carlo simulation for the AGMs with external probe (detector is outside) which will help to obtain more accurate results. The schematic diagram of the modeled geometry is shown in [Figure 2]. Concrete walls of varying thicknesses such as 5 cm, 10 cm, and 15 cm are considered in order to understand the effect of concrete thickness on wall scattering. The effect of gamma ray energy on scattering has also been studied for 0.662 MeV gamma from Cs-137 and 1.25 MeV gamma from Co-60. The detector is assumed to be mounted on the wall and its position is kept fixed, whereas the source position has been varied from a distance of 10 cm to 100 cm in step of 10 cm. The scattering factor (SF) for Cs–137 varies from 1.08–1.13, 1.08–1.15, and 1.08–1.15 for 5 cm, 10 cm, and 15 cm thick concrete wall respectively. The SF for Co–60 varies from 1.04–1.08, 1.05–1.09, and 1.05–1.09 for 5 cm, 10 cm, and 15 cm thick concrete wall respectively. Therefore, it is evident that the SF saturates after a certain thickness of concrete and no rise in reflection component has been observed if the thickness is beyond the saturation thickness. It is also observed that the SF is lower for Co-60 which has photon energy of 1.25 MeV than Cs-137 which has photon energy of 0.662 MeV. Therefore, Co-60 is a better choice for the AGM in-situ calibrator due to the lower contribution in wall scattering component as compared to Cs-137. The plot of the SF with varying distance between source and detector is shown in [Figure 3]. The SF is found to be lowest when the source is closest to the detector and the scattering contribution increases with distance between source and detector. However, the increase is not significant with distance. A maximum increase of only 6.5% and 4% in the SF is observed with varying distance from 10 cm to 100 cm between source and AGM for Cs-137 and Co-60 sources respectively.{Figure 42}{Figure 43}{Figure 44}
Keywords: AGM in-situ calibrator, back scattering, gamma ray, Monte Carlo simulation, scattering factor
Abstract - 23307: Identification of potential abnormal events in a clean reject oxide recycling facility using functional failure mode and effects analysis technique
Arti Sachin Mhatre, Gitender Singh1, S. Chitra, Kapil Deo Singh, D. B. Sathe1, R. B. Bhatt1, M. S. Kulkarni
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, 1Fuel Fabrication-INRP(O), Nuclear Recycle Board, Bhabha Atomic Research Centre, Tarapur, Maharashtra, India
E-mail: [email protected]
Radiological consequence analysis of a nuclear or radiological facility requires that all potential abnormal events (AE) which could have significant impact on-site and off-site be identified and evaluated to ensure safety of worker and public. Failure Mode and Effects Analysis (FMEA) is an inductive method of analyzing a system to find out what can fail and what will be the effect of the failure on the system (MIL-STD-1629A, 1980).[1] During fabrication of mixed oxide (MOX) i.e. (Pu,U)O2 fuel, around 10-15% of pellets are likely to be generated as rejects at various stages.[2] The rejects are broadly divided into two categories; clean reject oxide (CRO) and dirty reject oxide (DRO) depending on the physical and chemical characteristics of the pellets respectively. For recovery of valuable fissile or fertile material from clean rejects, microwave heat process based dry and wet recycling schemes[2],[3] are utilized in CRO recycling facility. Dry recycling scheme consists of air oxidation, attritor milling, reduction, stabilization of fuel in the microwave furnace. Whereas, the wet processing for CRO recycling consists of dissolution in nitric acid, concentration and crystallization, dehydration, denitration, calcination in microwave applicators. The MOX CRO and DDUO2 (deeply depleted Uranium oxide) CRO both can be recycled in this facility. This paper presents the scenarios in a CRO recycling facility which may lead to abnormal events. Functional FMEA has been applied in the present analysis. The different sections of the facility where hazards could exist are identified. Functioning of equipment and the processes involved during facility operation are considered for postulation of scenarios. Functional FMEA method is applied for identification of abnormal events. It is carried out in the following stages: a) Development of Functional Structure of the system or process, b) Identification of Failure Modes, c) Identification of Effects of Failure and Abnormal Events. Present work considers internal events with respect to plant. The process-related potentially hazardous events considered for evaluation[4] are: Loss of confinement (dispersal) of nuclear materials; Fire events; Load handling events; Explosive events; Criticality; Flooding which are considered as AE. One among the various postulated scenarios for CRO recycling facility has been presented in a structured manner in [Table 1]. Total 47 scenarios (26 in MOX recycling and 21 in DDUO2 recycling) have been postulated for the CRO recycling facility using Functional FMEA method. Among these, 42 scenarios are identified as abnormal events. These scenarios are essentially required for consequence analysis to evaluate radiological impact. This study also provides feedback for improving/modifying the design or process with respect to radiological safety.{Table 26}
Keywords: Abnormal event, clean reject oxide, FMEA, microwave furnace, radiological consequence analysis
References
Procedures for Performing a Failure Mode Effects and Criticality Analysis MIL-STD-1629A. 1980.Khot PM, et al. Development of recycling processes for clean rejected MOX fuel pellets. Nucl Eng Design 2014;270:227-37.Singh G, et al. Microwave based oxidation process for recycling the off-specification (U,Pu)O2 fuel pellets. J Nucl Mater 2017;484:81-90.USNRC. Final Safety Evaluation Report on Construction Authorisation Request for the MOX Fuel Fabrication Facility at Savannah River Site, South Carolina, NUREG-1821. USNRC; 2005.
Abstract - 23316: In situ 222Rn exhalation study over the Banduhurang open-cast mine and impact assessment on the surrounding the environment
B. K. Rana1, Samim Molla1, S. Ranjan2, S. Kumar2, S. K. Jha1,3, M. S. Kulkarni1,3
1Health Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Mumbai, Maharashtra, 2Health Physics Unit, UCIL, Turamdih, Jharkhand, India
E-mail: [email protected]
Banduhurang Open-cast uranium mining has been operating since 2009 in the East Singhbhum district of Jharkhand. The uranium mineralization at this deposit is confined within chlorite and felspathic-chlorite schist while the sericite-quartz schist is devoid of any radioactivity and appears to act as a footwall marker horizon of uranium mineralization. During the mining activities, 222Rn is exhaled from the ore and waste rock and released into the environment. 222Rn exhalation is a complex physical process that depends on several factors viz. grain size, grain shape, moisture content, mineralogy of host rock, and activity of 226Ra in the grain.[1] The 222Rn exhalation rate from the ore, waste yards, and mine surface is assumed to be much higher than the natural soil because of the higher 226Ra activity in the ore and waste rock. Continuous release of radon from the mining activities can give an additional dose to the public over the natural background. 222Rn and its progeny contribute to 55% of the total radiation dose received by the population from the natural environment.[2] In the present investigation, the in-situ radon exhalation rate over the mine pit, ore yards, and the waste yard was studied to understand the radon exhalation potential and the radiological impact on the surrounding environment of the Banduhurang open-cast mine. The accumulation chamber method was used to estimate the in-situ radon exhalation rates. The gradual buildup of the radon concentration in the chamber was measured upto 60 minutes with 10 minutes cycle by a continuous radon monitor (AlphaGUARD), which has sensitivity of 1 counts per minutes (CPM) at 20 Bq.m-3 and a wide dynamic measurement range (2 Bq.m-3–2×106 Bq.m-3). 222Rn exhalation rate (Js) was calculated from the slope obtained by linear regression analysis of measure radon concentration data (Ct) vs. elapsed time (t) using Eq. 1.[1]
[INLINE:7]
Where, Ct is 222Rn concentration (Bq m−3) at a time 't' (s), Js is 222Rn exhalation rate (Bq m−2 s−1), A is the surface area (m2), V is the effective air volume in the setup (m3), C0 is the 222Rn concentration (Bq m−3) in the chamber volume before enclosing i.e. t = 0. A specimen graph is shown in the [Figure 1]. The overall 222Rn exhalation rate was varied from 0.03–0.37 Bqm-2s-1 with a mean of 0.15 Bqm-2s-1. The variation in exhalation rate may be attributed to the difference in the moisture content, grain size, compactness, and 226Ra activity in the matrix. The 222Rn exhalation rate over these areas were in a similar range [Table 1]. Therefore, the mean 222Rn exhalation rate over the three areas was used to predict the 222Rn concentration at different distances from the mining site. Prediction of 222Rn at different distances from the source was carried out by the UNSCEAR (2000)[2] recommended dilution factors. The 222Rn concentration in the downwind direction at different distances of 0.5, 1, 2, 5, 10, and 20 km from the mine site were calculated to be 130, 71, 34, 9.5, 3.4, and 1.2 mBq m-3, respectively. The addition of 222Rn at different distances from the mining site was insignificant as compared to the background 222Rn concentration of this region. The estimated dose to the members of the public was estimated to be 7.2 μSv.y-1 at 0.5 km distance and subsequently less at far distances from the mine site.{Figure 45}{Table 27}
Keywords: 222Rn, open-cast mine, radon exhalation, uranium mining
References
Ishimori Y, et al. Measurement and Calculation of Radon Releases from NORM Residues, IAEA, TRS No. 474. 2013.UNSCEAR. Sources and Effects of Ionizing Radiations. Vol. I. New York: United Nations; 2000.
Abstract - 23317: Assessment of equilibrium factor between 222Rn and its progeny in Turamdih underground uranium mine
B. K. Rana1,, Samim Molla1, P. Kumar2, Gopal Verma1, S. K. Jha1,3, M. S. Kulkarni1,3
1Health Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Mumbai, Maharashtra, 2Health Physics Unit, UCIL, Jamshedpur, Jharkhand, India
E-mail: [email protected]
222Rn is an inert noble gas produced by radioactive decay of 226Ra. 222Rn is released continuously into the mine atmosphere through the exhalation process from uranium-bearing rock and through percolated mine water in the mine. The cracks and fissures present in the rock and the different mining activities such as blasting, fragmentation of ore, and mucking, also trigger the release of 222Rn into the mine atmosphere. In an underground mine, the dilution of 222Rn gas entirely depends upon the forced air ventilation and its circulation. The mine has three intake airways (decline, main shaft, and central ventilation shaft) and two return airways called ventilation shafts (E &W). Presently, the two ventilation fans are working with a combined capacity of about 240-250 m3s-1. The low-grade ore is mined out from the Turamdih mine for the extraction of uranium and the average grade is about 0.035 % (as U3O8). As per practice, the 222Rn concentration in the mines is measured and equilibrium equivalent 222Rn (EER) is estimated by using ICRP recommended equilibrium factor (F) (ICRP, 1993). The present study was focused to assess the actual equilibrium factor and EER concentrations in the mine. The generated data will be useful for a true dose assessment of the radiation workers of the mine.222Rn concentration was measured by using pulse ionization chamber-based AlphaGUARD. EER measurements were carried out by passivated ion-implanted silicon detector-based AlphaPM. Temperature and relative humidity (RH) in the sampling locations were monitored by AlphaGUARD. The instruments were programmed in flow mode with a cycle of 10 minutes, acting as a continuous monitor. The AlphaGUARD and AlphaPM were connected simultaneously and deployed in different location of the mine for the measurement of 222Rn and EER concentration. The sampling locations were chosen based on the places where mine workers were getting more exposure time. The equilibrium factor (F) was estimated by using Eq. 1.
[INLINE:8]
ZnS(Ag) coated Lucas scintillation cell was also used at the same time to measure 222Rn concentration for quality assurance of the data.[1]
222Rn, EER concentration and equilibrium factor were varied from 0.04-0.95 kBqm-3, 0.04-0.68 kBqm-3, and 0.03-0.96 with respective mean 0.34 kBqm-3, 0.15 kBqm-3 and 0.43. Since, underground mine is a dynamic environment and widely varied in terms of ore grade, radon exhalation rate, mining activities, ventilation rate, depth etc., thus, wide variation of 222Rn, EER, and F were observed. The estimated average equilibrium factor was comparable to the ICRP recommended value of 0.4. The Pearson correlation coefficient between the data measured by Lucas cell and AlphaGUARD techniques was estimated to be 0.91, indicating good correlation between these techniques. The estimated equilibrium factor and EER concentration were found to be increasing with increase in depth of the mine. The average equilibrium factors were 0.32, 0.47, and 0.60 in the 1st, 2nd, and 3rd levels, respectively. Higher equilibrium factor in the 3rd levels may be attributed to more residence time of the air as compared to the 1st and 2nd levels. For the 4th level and below, the average equilibrium factor was (0.37) less than the 2nd and 3rd levels, which may be attributed to the less mining work and higher ventilation rate in these areas as shaft is connected to the 4th level. Major mining activities are presently being carried out at the 1st, 2nd, and 3rd levels of the mine. Temperature and RH were varied from 17–36°C and 51–99%, respectively; and showed positive correlation with radon, EER and F, which may be attributed to increasing depth of mine. The above facts indicated that with increase in depth of mine and mining activities can govern higher EER concentration and equilibrium factor. Thus, less equilibrium factor and EER concentration can give low dose to the radiation workers that can be achieved by continuous augmentation of ventilation at work places. Radon exhalation rate is also increases by reduction of size of the ore and elevated 226Ra content in it. Thus, broken ore and waste rocks which are in reduced sizes need to be removed frequently from underground mine after each blasting activities for reduction of exposure to the workers.
Keywords: 222Rn, equilibrium equivalent 222Rn, equilibrium factor, uranium mining
Reference
Panigrahi, et al. Assessment to 222Rn and gamma exposure of the miners in Narwaphar underground U mine, India. Radia Phys Chem 2018;151:225-31.
Abstract - 23320: Optimisation for dose reduction program in bagjata underground uranium mine
R. L. Patnaik1,2, V. N. Jha1,2, M. K. Singh1,2, D. Rana1,2, S. K. Jha2,3, M. S. Kulkarni2,3
1Health Physics Unit, RPS(NF), HPD, BARC, Jaduguda, Jharkhand, 2Health Physics Division, BARC, 3HBNI, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: In low grade uranium mines large area has to be excavated to exemplify the sustainable economics. Controlling the individual or collectives dose in low grade underground uranium mine is quite challenging due to the complexity in source term distribution such as geo-genesis of ore body, grade of the ore, cracks and fissures in the ore, mine water dynamics and radon dispersion profile. The optimization exercise for effective implementation of ALARA for low grade uranium mines mostly focus on mitigating the challenges posed by radon (222Rn) and its short lived progeny. They are the principal radiological hazard of uranium miner through the inhalation route. The engineering solutions such as adequacy of ventilation requirement, auxiliary fan installation, sealing the hot-spots and seepage water transfer/ control mechanism are provided for reducing the internal exposure. Although economics and societal factors allow only limited scope for reduction in external exposure, the optimization is achieved through mechanization-automation and distancing from the source. In either case reduction in occupancy period is effective method of optimization of individual exposure. The dose reduction program through change in ventilation plan and exercising the other options mentioned is discussed in the paper.
Materials and Methods: Individual dose assessment in Bagjata mine has been carried out through the ambient dosimetry technique using the radon concentration, equilibrium factor and external gamma dose rates and individual occupancy details (number of days of underground work / visit and daily average period of stay inside the mine). The equation below is used to calculate the individual dose
[INLINE:9]
Where A, K, WLi, Ti, Gi represent the occupancy time, conversion factor from exposure to radiation dose, working level, gamma dose rate respectively. 170 h is being used as conversion factor for working level to exposure parameter.
Results and Discussion: Work management measures aim at optimizing occupational radiation protection in the context of the economic viability of the operation. Work profile management eventually led to a reduction of occupational exposures that can comply with the ALARA principle. The results of Dose reduction program at Bagjata mines are described in [Table 1]. The Results shown in [Table 1] indicated that there has been a reduction in all the dose parameters since year 2019 onwards. This is due to the engineering modification (ventilation plan) implementation of work management plan inside the mine. Earlier two low capacity ventilation fans were used in Bagjata mine. During the year 2018 Eastern Ventilation fan with higher capacity was commissioned. Further, new system as West ventilation fan construction was commenced in during 2019. After this modification in ventilation circuit reduction in radon profile thereby reduction in inhalation dose has been observed. Further optimization is being considered through automation and mechanization so that total average dose can further be reduced in the range of 2-3 mSv y-1.{Table 28}
Conclusion: Modification in the engineering control methods (revision in ventilation plan) has been effectively able to reduce the internal dose at Bagjata underground uranium mines by an approximate order of two. Mechanisation and automation plan is being worked for further reduction of total exposure below 3 mSv y-1.
Keywords: Dosimetry, internal dose, miners, optimization
Reference
Khan AH, Jha G, Srivastava GK, Jha S, Jha VN, Srivastava VS, et al. Assessment of radiation exposure of uranium mine workers in India. IAEA N; 2003.
Abstract - 23321: A comparative study of 226Ra activity in the uranium mill tailings estimated by NaI(Tl) gamma spectrometry and radon emanometry techniques
Ranjit Kumar1, B. K. Rana1,2, Samim Molla1, Gopal Verma1, S. K. Jha1,2, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: 226Ra is a naturally occurring radionuclide, originating from its parent 238U and present in the soil, rock, water, etc. at a wide range of concentration. In uranium ore, the activity of 226Ra is many times higher than in environmental soil and rocks. Uranium mining and milling produce almost all processed ore as waste called tailings, which contain a trace of unrecovered uranium and major radioactivity that was associated with the low grade ore. Accurate estimation of radioactivity in the tailings is important for proper environmental management and long term impoundment of the uranium tailings. At the Turamdih site, about 3000 tonnes of low grade ore are processed per day and 70% of the generated tailings are discharged to the tailings pond for permanent impoundment, and the rest of the coarse fraction of the tailings is used as back filling material in the generated voids of the mines. In this study, the 226Ra activity present in the mill tailings was estimated and compared.
Materials and Methods: To quantify and compare the 226Ra activity in tailings (14 nos.), NaI(Tl) gamma spectroscopy and emanometric techniques were used. Dry tailings samples were packed in airtight vial (250 ml) and kept for one month to achieve secular equilibrium between 226Ra and its daughter products. The samples were counted for 10000 s in a 5“×4” NaI(Tl) detector, which was shielded by a 12 cm thick circular lead. Its Energy calibration was carried out using 137Cs and 60Co sources, while efficiency calibration is carried out using RGU-1, RGTh-1 and RGK-1 standards, received from IAEA, Austria. The activities of 226Ra, 232Th and 40K were evaluated by considering the 1764 keV peak of 214Bi, the 2614 keV peak of 208Tl and the 1460 keV peak of 40K.[1] Activity in the sample was estimated by using Eq. 1.
[INLINE:10]
Where C is the net count for the counting time T (in sec), γ (%) is the percentage yield of gamma radiation, ε (%) is the percentage efficiency, and w is the sample weight (kg). In emanometric technique, a known quantity of dried tailings sample was digested by using concentrated HNO3-H2O2 mixture. The entire digested sample volume was reduced to about 70 ml and transferred into the radon bubbler and sealed airtight, allowing the radon to build up inside the sample solution for about 28 days. The built-up 222Rn in the bubbler was collected in an evacuated Lucas scintillation cell. After a post-sampling delay of 3 hours, the scintillation cell was counted by a programmable radon counting system for a period of 10 minutes, and 226Ra activity is estimated using the Eq. 2.
[INLINE:11]
Where, C is net counts for counting period T (min), t is counting delay (min), E is efficiency of the counting system, λ is 222Rn build up period in the bubbler (min), λ is the decay constant of 222Rn (min-1), and W is sample weight (kg).
Results and Discussions: 226Ra activity in composite tailings samples estimated by NaI(Tl) gamma spectroscopy and emanometric technique were found to be 2745 ± 10–4166 ± 13 and 2504 ± 145–4325 ± 189 Bq kg-1 with respective means of 3182 ± 462 and 3292 ± 423 Bq kg-1, respectively with a variation within ± 5%. However, individual estimated activity was found to vary from ± 2% to ± 14%. The wide variation may be attributed to the small sample quantity (5 g) taken for the analysis in the emanometric technique. However, the overall performance of the two measurement techniques was found to be in good agreement with each other. The Pearson correlation coefficient for the two sets of data was estimated to be 0.85. A paired 't' test performed on the two sets of data showed that at the 0.05 level, the difference between the two estimated means was not significantly different from each other.{Figure 46}
Keywords: Gamma spectroscopy, radioactivity, radon emanometry, scintillation cell
Reference
Sahoo SK, et al. Radiat Prot Dosimetry 2010;140:281-6.
Abstract - 23323: Assessment of radioactivity and speciation of uranium in mill tailings generated by the alkaline leaching process at the Tummalapalle mill
B. K. Rana1,3, Samim Molla1, Ranjit Kumar1, Gopal Verma1, A. Halder2, S. K. Jha1,3, M. S. Kulkarni1,3
1Health Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Mumbai, Maharashtra, 2Health Physics Unit, UCIL, Jamshedpur, Jharkhand, India
E-mail: [email protected]
The higher demand for uranium has spurred the exploration, mining, and processing of uranium-bearing ores, that generate a large quantity of uranium mill tailings. Uranium and the associated decay products may represent a potential radiological and toxic hazard to the environment.[1] These radioactive elements can be migrated and transformed through a series of physical and chemical processes, such as dissolution-precipitation, redox, adsorption-desorption, and complexation reactions, and may remain in the environment for a long time. As tailings are stored in the environment for a longer period, various environmental conditions may influence the physicochemical properties of uranium and toxic elements associated with the tailings may result in a significant impact on the environment. The potential impact of uranium and its decay products present in the tailings on the environment can be assessed by quantifying their activity in the tailings. The present study focused on the distributions of 238U, 226Ra, 210Pb, 235U, 232Th, 228Ra, and 40K activities in the mill tailings. The bioavailability, mobility, and toxicity of the metals to the biota depended on the physical and chemical forms of uranium and toxic elements present in tailings. Thus, the speciation of uranium was carried out by a sequential extraction procedure.
Materials and Methods: 24 nos. of mill tailings were collected from the tailings pond at Tummalapalle, Andhra Pradesh, where the alkali leaching process is followed to extract uranium from dolostone-type uranium ore. High purity germanium (HPGe) detector-based gamma-ray spectrometer with a relative efficiency of 50% with respect to a 3“×3” NaI(Tl) detector, was used for the assessment of 238U, 226Ra, 210Pb, 235U, 232Th, 228Ra, and 40K in the uranium mill tailings. The HPGe detector was shielded with a 10 cm thick circular lead disc with an inner lining of 1.5 mm Al, 1 mm Sn, and 1.5 mm Cu to reduce background counts. The energy calibration was performed by using standard 60Co and 137Cs sources. The efficiency calibration for each of the radionuclides was performed by using certified reference materials (RGU-1, RGTh-1, and RGK-1) obtained from IAEA, Vienna. The energy resolution was calculated in terms of full-width half maxima (FWHM) and found to be 2.1 keV at 1332 keV gamma energy of standard 60Co source. Tessier's sequential extraction procedure was also followed to understand the distribution of uranium in different chemical phases.
Results and Discussions: 238U, 226Ra, 210Pb, 235U, 232Th, 228Ra, and 40K activities were varied from 1091 ± 6–1851 ± 8, 2457 ± 16–3958 ± 11, 3754 ± 13–5153 ± 16, 64 ± 1– 98 ± 1, 10 ± 1–19 ± 1, <1.1–15 ± 1, and 393 ± 6–555 ± 7 Bq kg-1, with respective means of 1413 ± 230, 3014 ± 378, 4187 ± 376, 80 ± 9, 15 ± 2, 6 ± 5, and 439 ± 34 Bq kg-1. The abundance of 232Th and its daughter products in the mill tailings were found to be insignificant in comparison to the radioactivity originating from the radionuclides of the uranium series. The primordial radionuclide 40K is present in significant quantity with a mean of 439 Bq kg-1. The mill tailings contain almost all the uranium-series radionuclides including unrecovered uranium. The activity ratio of 210Pb/238U and 210Pb/226Ra were varied from 2.1–4.4 and 1.2–1.5, respectively, which indicates disequilibrium among U-series radionuclides. Tessier's sequential extraction showed that the mean order of distribution for uranium in different phases was: residual (51.1%)> exchangeable (19.8%) > Fe-Mn oxide bound (19%) > carbonate bound (8.9%) > oxidizable (1.3%). Half of the uranium (51.1%) was mainly distributed in the residual phase indicating they are strongly bound to the mineral lattice. A considerable amount of uranium is present in the weakly bound exchangeable phase (19.8%). Total Uranium concentration obtained by sequential extraction process in the tailings was varied from 1085–1483 Bqkg-1, which is relatively less as compared to the ore and hence poses insignificant impact to the surrounding environment.
Keywords: Mill tailings, NORM, sequential extraction, uranium
Reference
Tuovinen H, Vesterbacka D, Pohjolainen E, Read D, Solatie D, Lehto J. J Geochem Explor 2015;148:174.
Abstract - 23327: Study on the response of beta particulate monitor for Krypton-85
V. Ramprasath, K. Kannan, A. B. Chakne, Hemant Sharma, Anshuman S. Roy, G. M. Rao, Vikram Singh, Lakshmi Narayan Pulgam, B. B. Murmu, A. Ashok Kumar, Rakesh Tiwari1, Prem Chandra1, U. P. Shriwastawa1
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, 1INRP(K), NRB, BARC Facilities, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: Krypton-85 (85Kr) is chemically an inert radioactive gas with a half-life of 10.7 years. It decays to 85Rb emitting a beta of energy 0.687 MeV (Emax). In a spent fuel reprocessing facility, 85Kr gaseous release through the stack is used as an index of good and complete dissolution of spent fuel in the acid.[1] The long half-life allows the 85Kr gas to mix thoroughly in the atmosphere. Ionisation caused by 85Kr leads to uncertain effects in atmospheric phenomenon as well.[2] Hence it is vital to monitor the release of 85Kr both from the point of view of plant operation as well from the point of view of environmental concerns. In this paper, study done on estimating 85Kr release from the response of stack Beta Particulate Monitor (BPM) is discussed.
Measuring Devices: The sampling system as shown in [Figure 1] consists of a pipeline originating from the stack through which air is drawn at a known rate and routed through the glass fiber filter papers (having efficiency of 99.97% for 0.3 micron particles) of the Alpha and Beta Particulate Monitors (APM & BPM) for evaluating the particulate activity releases. The outlet air from particulate monitors, which is free from particulates, then passes through a shielded Krypton Monitor (KM) which has a GM tube placed in a SS chamber of volume 5 litres for evaluating the 85Kr activity. End window GM tube (LND 72314) is used for the Beta Particulate Monitor and thin walled GM tube (LND 719) is used for Krypton Monitor. Efficiency testing using standard gaseous 85Kr source has shown efficiency of about 1% for KM. Every minute data of monitors are logged in a centralised radiation monitor console panel through Ethernet.{Figure 47}
Response of BPM to 85Kr: The data of both the BPM and KM reading have been analysed for many batches of dissolution operations over a period of ten years. The background in the monitoring area is about 0.1 – 0.2 μGy/h only. The regular counting of particulate filters have not shown any activity deposition which confirms that the response of BPM during dissolution is due to the passage of 85Kr through it. [Figure 2] shows typical identical response trend of both BPM and KM during one of the dissolution processes. It is observed from [Figure 3] that the Krypton Monitor's response is found to be about twice that of the Beta Particulate Monitor during dissolution process.{Figure 48}{Figure 49}
Results and Discussion: The ratio of the response of Krypton Monitor to that of Beta Particulate Monitor is found to vary between 1.9 and 2.5 with an average of 2. This factor will be useful for indirectly confirming the proper response and integrity of Krypton Monitor and also for the estimation of 85Kr release from the data of Beta Particulate Monitor, in case of non-availability of Krypton Monitor.
Keywords: Krypton monitor, particulate monitor, reprocessing, stack monitoring system
References
Dey PK. Spent fuel reprocessing: An overview. Kalpakkam: INSAC; 2003.Harison RG, et al. Atmos Environ 1994;28:637.
Abstract - 23330: Estimation of radioactivity concentration in steam condensate based on the response of a special-purpose gamma monitor
Sarswati Devi1, V. Ramprasth1, K. Kannan1, A. B. Chakane1, Lakshmi Naraynan Swamy Pulagam1, G. M. Rao1, E. S. Rao2, Sadhu Khan2, J. K. Gayen2, G. Ganesh1, M S Kulkarni1,3
1Health Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Mumbai, Maharashtra, 2INRP-K, Nuclear Recycle Board, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: In a nuclear spent fuel reprocessing plant, Low Pressure (LP) steam is used for evaporation of high active liquid wastes for volume reduction. As steam is used for evaporation purpose, there is a possibility for active liquid getting mixed up with steam condensates which will render management of Low Level Liquid Waste (LLW) difficult. Hence an attempt has been made to estimate the activity concentration of LLW steam condensates) based on the response of Special-purpose Gamma Monitor (SGM) installed on the LP steam condensate header. This will help to isolate the systems promptly and to arrest breach of activity through LLW.
Method and Calculation: The Low pressure steam condensate from an evaporator is first collected in a buffer tank i.e. Condensate Monitoring Tank (CMT) of volume about 1.5 m3. The contents are then sent to Hold Up Tank (HUT) of volume about 25 m3 and then finally sent to Delay Tank (DT) compartment of volume about 200 m3 for pumping to waste management facility for final disposal. SGM is installed on the steam condensate headers routed to CMT to promptly annunciate and alert about any breach of activity from the evaporator to the steam condensates. The dimension of condensate header considered for the study is 80NB SS pipe line with an inner diameter of 77.9 mm and thickness of 5.49 mm. Using IGHIESD code,[1] calculations have been carried out to estimate the dose rate and concentration in CMT, HUT & DT corresponding to 75% and 100% of the SGM scale, i.e. 0.75 mGy/h and 1 mGy/h respectively. An optimum length of 200 mm corresponding to 952 ml of steam condensate has been assumed to be under the vicinity of the SGM probe. Calculations have been done for i) typically encountered condensate flow rates of 250 lph and 1000 lph in a plant and ii) for different time duration of persistence of SGM reading (10 sec to 3600 sec). Cs-137 is considered as the predominant radionuclide for the simulation. The βγ activity concentration in the header under the vicinity of SGM probe is estimated as 60600 Bq/ml and 81554 Bq/ml corresponding to SGM reading of 0.75 mGy/h and 1 mGy/h respectively. Minimum activity concentration by the SGM for the dimension of header under study is about 37 Bq/ml. Based on the mix-up of this activity with the contents of tanks in the LLW route, βγ activity concentration (Bq/ml) and dose rates (mGy/h) in respective tanks have been calculated. The calculated concentrations / dose rates are tabulated in [Table 1]. As tabulated above, the Delay Tank samples will always be within Category II range (<37 Bq/ml) under the following scenarios: (i)Annunciation of 0.75 mGy/h by SGM for duration of about 30 minutes when condensate flow rate is 250 lph or for duration of about 7 minutes when condensate flow rate is 1000 lph. (ii) Annunciation of 1 mGy/h by SGM for duration of about 20 minutes when condensate flow rate is 250 lph or for duration of about 5 minutes when condensate flow rate is 1000 lph.{Table 29}
Conclusion: Based on the response of SGM on a LP steam condensate header, βγ concentrations in the tanks in the LLW route have been estimated for different flow rates of steam condensates and for different duration of persistence of SGM readings at 0.75 mGy/h and 1 mGy/h. This will be of immense help to the operators of reprocessing plants to initiate prompt action so as to ensure that concentration in the DT is contained well below Category II range (37 Bq/ml) for safe disposal by waste management facility.
Keywords: Activity concentration, delay tank, dose rate, low level liquid waste, SGM
Reference
Subbaiah KV, Sarangapani R. Ann Nucl Energy 2008;35:2234-42.
Abstract - 23346: Estimation of activity content of hull drum
B. C. Muduli1, Gitansh Arora1, T. K. Mandal1, K. Kannan1, S. K. Nayak1, Lakshmi Narayan Pulagam1, Rabindra Nath Juine1, V. Ramprasath1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: The head end process in a PHWR spent fuel reprocessing plant involves chopping of fuel bundles followed by dissolution using HNO3. The solution is filtered and the cladding pieces (called hull pieces) are tilted into an SS drum (called as hull drum). Activity estimation of hull drum is mandatory for disposal at waste management facility. In this paper, estimation of gamma activity of hull drum is done with IGSHIELD and Monte Carlo simulation.
Code and Computation: Hull solution and pieces are placed in a cask after the head end process and a simplified picture of the cask is shown in [Figure 1]. The hull drum is a right circular cylindrical drum made of stainless steel of 3.0 mm thick. For each batch of spent fuel process, on an average 150 kg of hull is produced. Hull drum dimensions are 800 mm in diameter and 840 mm in height. After loading hull into the drum, the effective volume of hull is 0.420 cubic meters. Monte Carlo simulations have been done for estimation of gamma dose rate on contact, 0.5 meter, 1 meter and 2 meters from the surface of hull drum. The results were compared with computation done with IGSHIELD,[1] a gamma ray shielding code which estimates gamma doses using the point kernel technique. Only comparisons were made between the two methods, no attempt at benchmarking was made due to lack of standards to compare the results. The following inputs and assumptions are taken into consideration.{Figure 50}
Gamma activity is uniformly distributed throughout the hull drum till the actual height of the hull. The gamma spectrum of hull material consists of primarily 137CsEffective density the hull is 0.357 g/cc of zirconium and dose rate is estimated by using ICRP-74 photon flux to dose conversion factor in the simulation.
Results: The results of the Monte Carlo simulations are tabulated in [Table 1]. Dose rate and activity conversion factors (ACF) have been calculated per unit radiation field of 1mGy/h for four cases: on contact with the drum, at 0.5 m, 1m and 2m distances from the surface of the drum. Conclusion: The computation by IGSHIELD code shows conversion factor of 2.35E+03MBq/(mGy/h)and by the Monte Carlo simulation shows a factor of 2.19E+03MBq/(mGy/h) (On contact radiation field). The results are differing due to build-up factors (BUF) generated by the two simulations. Monte Carlo gives a closer value to 137Cs content (40 Ci) as obtained from spectroscopic means.{Table 30}{Table 31}
Keywords: Activity conversion factor, hull drum, Monte Carlo
Reference
Subbaiah KV, et al. IGSHIELD: A new interactive point kernel gamma ray shielding code. Ann Nucl Energy 2008;35:2234-42.
Abstract - 23370: Estimation of skin dose and deep dose equivalents due to exposure situations during handling of uranium
K. S. Varshitha, S. Manilal, G. Ganesh
ORPR Section, Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
During handling of uranium compounds, radiological protection of the workers has to be ensured. For appropriate assessment of radiological risk, data concerning doses for different practical irradiation scenarios is necessary. Radionuclide and Radiation Protection Data Handbook[1] contains dose rate coefficients for different exposure scenarios containing uranium isotopes. These coefficients were calculated using analytical codes VARSKIN 2 and Microshield 4.10. Limitations of analytical approach include non-inclusion of Bremsstrahlung contribution and range straggling effects when the skin lies in range of primary electrons. As an improvement, Amato et al.[2] used Monte Carlo Code GAMOS for estimating skin dose and deep dose for different medically relevant radioactive isotopes but not for isotopes of Uranium and its daughter products. In the present study, FLUKA,[3] a Monte Carlo code for simulation of radiation interaction and transport, is used to estimate skin (70 μm) and deep (10 mm) dose equivalents in tissue for different configurations as shown in [Figure 1] for isotopes U-238, U-235 and their immediate daughter products Th-234, Pa-234m, Th-231 and Pa-231. In FLUKA, USRBIN card is used to score the energy deposited in the detector volume which is divided by the mass of the detector to obtain the dose deposited. Gamma and beta emission of each isotope is handled separately. Gamma and beta energy with abundances for each isotope is stored in an external file and read using source.f. Number of simulation histories is chosen according to the isotope and geometry so that the statistical error is less than 5%.{Figure 51}
Calculated skin dose and deep dose for different isotopes are tabulated in [Table 1] and [Table 2].{Table 32}{Table 33}
Pa-234m being a beta emitter with Emax of 2.27 MeV (97.6%) energy, results in large skin dose. Deep dose is highest for the case of U-235 as the isotope has many gamma energies of significant abundances (143.8 keV - 10.96%, 163.33 keV - 5.08 %, 185.7 keV - 57.2 %, 205.3 keV - 5.01%).
Keywords: Contamination, deep dose, skin dose, uranium
References
Delacroix D, Guerre JP, Leblanc P, Hickman C. Radionuclide and radiation protection data handbook 2nd edition (2002). Radiat Prot Dosimetry 2002;98:9-168.Amato E, Italiano A, Auditore L, Baldari S. Radiation protection from external exposure to radionuclides: A Monte Carlo data handbook. Phys Med 2018;46:160-7.Battistoni G, et al. Overview of the FLUKA code. Ann Nucl Energy 2015;82:10-8.
Abstract - 23371: Radiological safety challenges during Hotcell decontamination, cleaning and reinstallation of radiation shielding glass windows of a hotcell facility
P. T. Ghare, A. Bhaktivinayagam, Atul Govalkar, D. P. Rath, P. Ashok Kumar, Anil Bhandekar1
Health Physics Division, BARC, 1Post Irradiation Examination Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Post Irradiation Examination Division (PIED) at Bhabha Atomic Research Centre, Trombay, over the past forty-four years, has been used for examination of different types of experimental as well as power reactor fuels and irradiated components. High density concrete shielded hot cells are used to handle up to 3700 TBq of fission product activity.[1] Over the years, opaqueness developed in the radiation shielding glass windows rendering the operational activities difficult. Hence it was planned to remove, clean and reinstall the radiation shielding glasses. As a pre-requisite, it was essential to remove radioactive waste material from the hotcell for safe disposal adhering all radiological safety parameters. This operation posed radiological safety challenges and required extensive planning and execution as it was of its first kind since the inception of PIED.
Methodology: The entire operation was executed in phased manner under administrative control. Standard Operating Procedures (SOPs) and Emergency Operating Procedures (EOPs) were prepared. Visual inspection of the hotcell through periscope was carried out for segregation of various radioactive components. Equipment viz. Special vacuum cleaning machine with shielded collection drum, Remotely operating cable cutting machine etc. were fabricated in-house. Thereafter, radioactive components were removed using appropriate tools under stringent health physics surveillance. Initial cleaning was carried out by vacuum cleaning machine.[2] Segregated rad-waste was remotely transferred to MS drums and shifted to interim storage for safe disposal. Remote tools were used to decontaminate hotcell. After significant reduction in the radiation levels, personnel entries to the hotcell with adequate PPEs were carried out for further decontamination[3] with continuous radiological assessment. For removal of the glass assemblies, a metal platform was installed and a PVC tent was erected in hotcell operating area. The mineral oil between shielding glass was collected in SS drums. The glass assemblies were removed, placed on the platform with the help of monorail and aluminium sheets were installed in their place. Assemblies were reinstalled back after decontamination using alcohol based glass cleaner.
Discussion: Rad-waste removal and repetitive decontamination operations reduced the radiation level inside the hotcell significantly from 200 mGy/h to 4 mGy/h [Figure 1]. Maximum beta-gamma airborne activity during the operation was 7 DAC with respect to 90Sr. The major radionuclides detected in the air samples were 137Cs and 90Sr. Continuous exhaust monitoring during this job did not show any release of activity to the environment. Due to extensive health physics coverage & meticulous planning, the collective dose for the operation was restricted to 32.10 person-mSv with no case of internal exposure and overexposure.{Figure 52}
Conclusion: Removal of high active material and decontamination of hotcell was indispensable to minimize the exposure of working personnel. The operation being first of its kind provided a benchmark experience for planning and execution of similar operations in future.
Keywords: Hotcell, radiation dose rate profile, radiation shielding glass, radiological safety challenge
References
Anantharaman S, et al. Post irradiation examination of thermal reactor fuels. BARC Newsl 2010;315.Cheever CL, et al. Decontamination of Hot Cells K-1, K-3, M-1, M-3, and A-1, M-Wing, Building 200: Project Final Report Argonne National Laboratory – East ANL/D&D/TM-96/2. 1996.Weaver P, et al. Decommissioning of Hot cell facilities at the Battelle Columbus Laboratories. Tucson, AZ: Battelle Memorial Institute; 2003.
Abstract - 23381: Radiation shielding evaluation for water storage pool of 60Co sources
Deepa Anilkumar, Tanmay Sarkar, Shreenivas Vaikuntam, Amit Acharya1, A. K. Pradhan1, Pravin Kumar2, S. Anand, Kapil Deo Singh, M. S. Kulkarni
Health Physics Division, Bhabha Atomic Research Centre, 1Product Development Division, Bhabha Atomic Research Centre, Mumbai, 2 Board of Radiation and Isotope Technology, Navi Mumbai, Maharashtra, India
E-mail: [email protected]
Sealed radioactive sources are widely used for beneficial purposes in industry and in medicine. The availability of alternate storage facilities for the temporary storage of very high intensity 60Co sources during refurbishment & repairs of major irradiators has acquired a lot of importance recently. This is due to certain major irradiator facilities are taking up these activities in order to meet regulatory requirements as well as go for facility up-gradation. A theoretical study for one such DAE facility was carried out to assess the radiological safety parameters for storage of new and disused 60Co pencil sources in a water storage pool during normal and abnormal conditions. Water storage pool of dimensions 1370 mm (L), 2410 mm (W) and 6200 mm (H) is assumed to store disused and new 60Co pencil sources. Study was carried out to estimate the dose rate on the water surface of the pool and operating locations around the pool when both type of sources are being placed for varying water depths during normal operation and in abnormal situation arising from change in dose rate due to change in pond water activity in the pool. Storage rack inside the pool is a rack system of 4 rows x 12 columns and has a pitch of 45 mm and 4 racks can be stored. Both new 60Co pencils and used 60Co pencils sources have specific activity of 1110 TBq and 55.5 TBq respectively. Dimensions of the old and new pencil are taken as 38.1 mm in diameter and 700 mm in length. Optimisation studies were carried to study the temporal variation of dose rates with different combinations of old and new 60Co pencils in pool in racks during normal operations for varying water heights in pool. The allowable dose rate outside the pool and in the operating areas is 10μSv/h(supervised areas). In this paper, results of the one of the studies is presented, where it is assumed that both the sources are mixed equally and placed in racks with a maximum inventory of 1.85E4 TBq and 3.7E4 TBq in normal operations. In abnormal situations, the pool water activity concentration can vary during operations in storage pool or minute defects in the pencil sources. Effect of change in pool water concentration was studied and was found to be a major dose contributor. The geometry of water pool containing 60Co sources and surrounding areas are modelled using QAD-CGGP code.[1] The point kernel method in this code involves representing the source volume by a number of point isotropic sources and computing the line-of-sight distance from each of these source points to the detector point. Dose rates at various locations are presented in [Table 1] and [tablw 2]. Studies show that during normal storage conditions (10 Bq/ml of 60Co activity in the pool water and for a maximum inventory of 3.7E4 TBq of 60Co pencils having 5.5 m thickness of water shielding), the dose rates at the surface of the storage pool and pond bridge are 6.57 μSv/h and 0.5 μSv/h respectively and are within allowable limits.{Table 34}{Table 35}
Keywords: 60Co, radiological safety, shielding, storage pool
Reference
QAD-CGGP Manual. CCC-0493/01; 1989.
Abstract - 23398: Study on radiological conditions during glove boxes cleaning at MOX fuel fabrication facility
S. Pattanayak, R. N. Dubla, S. K. Tibrewala, G. Ganesh1, M. S. Kulkarni1
HPD, Bhabha Atomic Research Centre, Health Physics Unit, Tarapur, 1Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: The MOX fuel consist of oxides of fissile & fertile radio-nuclides used for nuclear power production. MOX fuel fabrication uses powder metallurgy route which involves operations such as mixing, milling, compaction, sintering and grinding followed by loading of pellets in tubes, end plug welding etc. These operations are assisted by quality control at every stage.[1] MOX fuel involve handling of actinide which are alpha, gamma and neutron emitters, thus possess internal & external hazards. Therefore all the MOX fuel fabrication operations are being carried out in glove boxes (GBs). To control the internal hazard, enough safety is taken care at design stage while placing fuel fabrication equipment's inside GBs which contains the MOX powder. However most of the operation of fuel fabrication is carried out on contact of GBs through gauntlet ports. With due course of time, MOX powder gets deposited on inner surface of GB walls, floor and on the equipment's during fuel fabrication. Accumulation of MOX powder results in gradual rise in gamma and neutron radiation level on the GBs surface and hence the collective dose of plant. Therefore periodic cleaning of the GBs was recommended to reduce the radiation levels on GBs and hence the collective dose of plant. Periodically cleaning of GBs and its effect on GBs radiation level as well as the collective dose of plant are presented in this paper.
Material and Methods: Measurement Method: Gamma and neutron radiation level measured by GM and 3He detector based survey meters respectively. Personnel doses are measured using TLDs and SSNTD badges.
Glove boxes cleaning methodology: During cleaning process, initially all the compressible material of GBs such as PVC bags, tissue papers, damaged gauntlets etc. used for fuel fabrication were packed in PVC bags & bagged out from the GBs as alpha bearing waste.[2] The MOX powder deposited on inner surface of the GBs and the equipment's installed in GB are collected in containers using soft brush and bagged out for recycling. Finally inner surface of the GBs and equipment's are cleaned by using tissue paper and non- abrasive cloth which are also bagged out from the GBs as alpha bearing waste. Through-out the cleaning process, anti-piercing gloves are used over the gauntlets to avoid any wound from sharp edges or wires present in GBs. Radiation level on all the GBs were periodically measured before cleaning and at every stage of cleaning process. The collective doses of radiation worker working on these GBs also analyzed.
Results and Discussion: From [Table 1] it is observed that, periodic cleaning of GBs results in 42 -51 % reduction in gamma radiation level & 40 - 45 % reduction in neutron radiation level on GB surfaces. Around 0.4 m3 alpha bearing waste generated and only 0.2% of plant collective dose budget was consumed in cleaning process. To assess the effectiveness of GB cleaning, collective dose data of radiation workers involved in the fabrication process was analysed. The collective dose for year 2014 was normalised to 100% and trend of collective doses after implementing periodic cleaning of GBs for same work over the successive year is shown [Figure 1].{Table 36}{Figure 53}
Keywords: Fissile radionuclide, glove box cleaning, MOX fuel, radiation level
References
Bhatt RB, et al. Proceedings of the Theme Meeting on the Journey of BARC Safety Council for Strengthening Safety Culture in BARC Facilities; 2000-2017, Mumbai, July 2017. p. 70-3.Rath DP, et al. Radiat Prot Environ 2010;33:180-2.
Abstract - 23406: Study on tritium concentration in low level liquid waste of a fuel reprocessing plant
Letha Sebastian1, Sarswati Devi1, A. B. Chakane1, Hemant Sharma1, V. Ramprasath1, S. Babu2, Sadhu Khan2, J. K. Gayen2, G.Ganesh1, M. S. Kulkarni1,3
1Health Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Mumbai, Maharashtra, 2INRP-K, Nuclear Recycle Board, BARC Facilities, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: The presence of tritium in the nuclear fuel, formed by ternary fission of uranium, presents a problem in the management of waste from nuclear reprocessing plants. Tritium can be neither separated nor concentrated by conventional procedures used in waste treatment. During the processing (PUREX) of spent fuel in a reprocessing plant, the fission product tritium retained in the fuel is released in the forms of hydrogen (HT) and water (HTO). In an LWR fuel reprocessing plant using the PUREX process, after chopping the fuel elements and fuel dissolution in nitric acid, about 60% of the total amount of tritium is retained in zircaloy cladding.[1] The bulk of the remainder is in the form of HTO & nitric acid which is distributed through various streams. Major source of tritium in low level liquid waste stream (LLW) of reprocessing plant is from the evaporator condensates generated during volume reduction of High-Level Liquid Wastes (HLW). The present study is to evaluate the activity of tritium in Low Level Liquid Waste (LLW) in a PHWR fuel reprocessing plant and to assess the activity in terms of total tritium inventory in spent fuel.
Materials and Methods: Scintillation technique is the most common analysis method for H-3 measurement.[2] Perkin Elmer make Liquid Scintillation Analyzer LSA 2910 TR is used for analysis of H-3 in LLW samples. 1 ml of the sample is mixed with 10 ml of cocktail (Ultima Gold/SRL) and counted. Tritium counting window 0 – 18.6 keV is used in MCA for calibration & sample counting. Since chemical quenching is always present in reprocessing effluent samples, quench correction with quench curve carried out for accurate measurements. Quench calibration curve of efficiency Vs tSIE is generated using quanta smart software & tritium quench standards supplied by M/S Perkin Elmer, USA. Counting efficiency of the sample is computed using the tSIE value of sample from the quench calibration curve. Using CPM value & counting efficiency, tritium activity concentration of the sample (Bq/ml) is calculated. Total amount of tritium in LLW is calculated from specific activity & volume of waste generated. The presence of tritium in LLW in terms of % of total inventory present in the spent fuel is then calculated. The inventory of tritium in the spent fuel has been generated using ORIGEN code for various burn-ups and cooling periods.
Results & Discussions: [Table 1] shows the annual average tritium in LLW per ton of fuel processed for a period of seven years and the activity in terms of percentage of actual inventory in the spent fuel. During this period, burn up of fuel processed varied from 6700 MWD/Te to 10000 MWD/Te with cooling period from 10 years to 15 years. Theoretically calculated values of tritium inventory in PHWR spent fuel vary from 2.22 to 3.84 TBq/Te. Conclusion: In a PHWR spent fuel reprocessing plant that uses PUREX process, the tritium activity in LLW is observed to be about 35% of total tritium inventory in the spent fuel.{Table 37}
Keywords: Liquid scintillation, percent of inventory, PUREX process, reprocessing, tritium
References
Henrich E, et al. The Concentration of Tritium in the Aqueous and Solid Waste of LWR Fuel Reprocessing Plants, Management of Gaseous Wastes from Nuclear Facilities. Proceedings of the International Symposium on IAEA, Vienna; 1980.Jakonic I, et al. Study on quench effects in liquid scintillation counting during tritium measurements. J Radioanal Nucl Chem 2014;302.
Abstract - 23488: Operational health physics experience gained during the special operations at post irradiation examination
R. Akila, R. Mathiyarasu, D. Ponraju
Health and Industrial Safety Division, IGCAR, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Post Irradiation Examination (PIE) of the irradiated fuel, blanket materials, control rods and structural materials are carried out in Radio-metallurgy Lab (RML) hot cells. Health Physics (HP) surveillance was provided for the various operations of RML. Some of the major operations include receipt of irradiated fuel subassemblies from FBTR, X-ray and neutron radiography of these fuel pins, transfer of fuel pins for reprocessing, receipt and analysis of experimental test assemblies containing structural materials, ferro-boron capsules, density measurement on failed fuel pin. Apart from these, some special operations like study on the radiation induced damage and the failure analysis of the components from other facilities are also carried out. This paper discusses about the Health Physics experience gained during these operations. Dimensional measurement on W and WC pellets irradiated in FBTR was carried out after opening the HotCell#4. The dose rate on the pellets was 600 mGy/h. The collective dose for dimensional measurement on 9 pellets was estimated to be around 4.5 PmSv based on the dose rate at chest level (5 mGy/h). In order to reduce the dose rate, lead shielding of 200 mm thick wrapped in polythene sheet was erected inside the HC#4. The dose rate reduced to 100μGy/h. Apart from this the personnel used lead apron and lead gloves to further minimize the dose. The collective dose incurred was 100PμSv. Samples of SS316LN and SS304 LN structural material to be used in future FBRs were irradiated at FBTR for mechanical property evaluation. The dose rate on the samples ranged from 400 mGy/h to 1Gy/h. In order to reduce the dose to the working personnel, a special shielded facility with MSM (for handling the samples remotely) and viewing window was designed, fabricated and erected. The dose rate at different locations of the shielded facility was computed using monte carlo method and the shielding adequacy was checked using 60Co standard reference source. Mock up was conducted with dummy samples for familiarization. Testing of samples was carried out in this facility under strict HP surveillance. The personnel also used lead apron. The collective dose for this operation was 0.12 PmSv as against the estimate of 1.8 PmSv. After completion of the work, the facility was dismantled and the tong and sample holder which was contaminated was removed and packed as solid waste. There were no cases of personnel contamination. Air borne activity was detected during the gauntlet changing operation of glove box attached to HotCell#6. While changing the gauntlet, a sudden increase in the counts was observed in the alpha channel of continuous air monitor (CAM) by HP who alerted the crew. Immediately the work was stopped and the glove port was closed with a bung. All the personnel were evacuated. Air sample taken during the operation showed 17.2 DAC of 239Pu and the CAM filter paper showed 4.3 DACh. After one hour, the air activity reduced to normal. All the personnel were referred to bioassay and there was no internal contamination. As a pre-emptive measure, the personnel were wearing full face mask. The prompt evacuation of personnel and the use of PPE helped in averting the internal dose. Due to the air borne activity, the floor and the equipment surrounding the glove box were found to have contamination of alpha: 0.4 – 1 Bq/cm2 which was decontaminated thoroughly and normalcy was restored after 3 weeks. The contamination was attributed to the accumulation of particles in grooves of the glove port. Failure analysis of irradiated KKNP Double Check Valve (DCV) was carried out for first time in RML. The dose rate on the component was 70 – 100 μGy/h. The gamma spectrum on the component revealed the presence of 54Mn and 60Co. It was recommended to handle the sample in fume hood while examining the sample, there were no cases of area or personnel contamination. The above said operations were unique and first of its kind in RML. Designing of temporary shielding with MSM and viewing window, was a new experience and learning process. This temporary shield is designed in such a way that it can be dismantled and stored and can be reassembled as and when required. The mock up conducted helped in incident free operation. Initial dose and contamination assessment made on the samples was the key factor in deciding the shielding and PPE required for the specific operation. The experience gained from these special operation, gives confidence in examining irradiated samples with high dose rate in the future. Modifications in the standard operating procedure for glove box operations were done to avert any air borne activity in future which includes wet wiping of grooves prior to gauntlet changing. The strong and strict implementation of radiation protection program combined with HP surveillance contributed in minimizing the dose accrued.
Keywords: ALARA, collective dose, health physics, post irradiation examination
References
Hazard Evaluation Report for Radiometallurgy Lab Hot Cell Facility. Vol. 1; 2008.AERB Safety Manual Rev 4.
Abstract - 23493: Computation of dose on terrace and the surrounding area of proposed new shielded facility at post irradiation examination
R. Akila, Ashish Kolhatar1, R. Sarangapani, R. Mathiyarasu, D. Ponraju
Health and Industrial Safety Division,
SQRMG, IGCAR, 1Post Irradiation Examination Division,
MMG, IGCAR, Kalpakkam,
Tamil Nadu, India
E-mail: [email protected]
A new shielded facility is proposed to be established for the Post Irradiation Examination of irradiated structural materials and fuel from FBTR at Radiometallurgy Lab (RML). The mechanical properties of irradiated structural materials are characterized in the Mechanical Property Evaluation Cell (MPEC) facility and the micro analytical studies on different fuel test irradiated at FBTR is characterized in Shielded Facility for Micro Analytical Characterization (SFMAC). MPEC facility consists of two containment boxes shielded with lead for housing the instrumented impact testing machine and acreep testing machine. SFMAC facility consists of two separate shielded lead cubicles (SC) which houses the SEM & XRD. A detailed analysis on the amount of activity of different composition that can be handled and the shielding was carried out and submitted to AERB for approval. As a part of AERB recommendation, the dose rate outside the facility is to be computed for the different source terms and this paper gives the details of the same. In SFMAC facility, the cubicles are shielded with 200 mm lead on the sides and 180 mm steel on top. The cubicle housing XRD is taken as the reference for computing the dose rate at the roof. The source position for computation is taken as the sample location at XRD which is at height of 1.4 m from the ground level. The height of the cubicle including the steel is 2.8 m from the ground. The roof of the facility is at a height of 5.35 m from the ground and is made of 150 mm thick concrete. MPEC facility is a shielded cubicle with 200 mm thick lead shielding on all the three sides and on the fourth side it has 300 mm thick steel. This cubicle is shielded with 80mm thick steel on top. The source is placed at height of 1.4 m from the ground level and at 1.4 m from the western wall. The height of the cubicle including the steel is 2.57 m from the ground. The roof of the facility is at a height of 8.0m from the ground and is made of 150 mm thick concrete. [Figure 1] and [Figure 2] shows the schematic of the facilities SFMAC and MPEC. In the figure the source position and detector position are marked as “S” and the “D#” respectively. The computation of dose rate on the roof and the sitting rooms near the shielded facilities were computed using IGSHIELD an Interactive Gamma-ray Shielding code developed in Visual Basic and validated with benchmark problems using monte carlo method and analytical methods. The computational methodology is based on the point kernel technique. The source term taken for calculation is 14.7 Ci (0.6 MeV gamma) for SFMAC facility. The computation is also performed for 5.2Ci and 3.7Ci. The source term for MPEC facility is 2Ci of 60Co. The dose rate on the roof of SFMAC is 0.2μSv/h, 0.076μSv/h and 0.054μSv/h for the source term 14.7Ci, 5.2Ci and 3.7Ci respectively. The dose rate in the rooms surrounding the Shielded facility is found to be <1μSv/h. The dose rate at different elevation of MPEC facility for different roof shielding thickness is given in [Table 1]. It was observed that for 80mm thick steel roof the dose rate on terrace was found to be 80μSv/h. Hence it was recommended to increase the roof thickness to 150mm wherein the dose rate reduced to 9.0μSv/h and also recommended that access to the area should be administratively controlled.{Figure 54}{Figure 55}{Table 38}
Keywords: Impact machine, dose computation around the facility, lead shielded facility for SEM, XRD
Reference
Subbaiah KV, Sarangapani R. IGSHIELD Ann Nucl Energy 2008;35:2234-42.
Abstract - 23500: Radiological perspective for the utilization of uranium mill tailings
S. K. Jha1,2, M. S. Kulkarni1,2, D. K. Aswal1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: “Tailings” is one of the waste products of the ore processed by the UCIL to produce uranium concentrate in the form of 'Yellow Cake' to meet the requirement of nuclear power generation. As per the international/AERB norms, Uranium mill tailings are required to be confined in order to exercise proper source term control. At Jaduguda, tailings after suitable chemical treatment and coarse-fines separation are impounded in an engineered tailings pond (T.P.). The concern arises because of the very large amounts of U tailings need recycling or disposal. Other largest NORM waste stream is coal ash, followed by phosphogypsum, fertiliser phosphate. with million tonnes arising globally each year, and carrying U-238 and all its non-gaseous decay products, as well as Th-232 and its progeny.
Results and Discussion: Samples were analysed by gamma spectrometry using HPGe detector (with a relative efficiency of 50% coupled to a PC-based spectrum stabilized 8K multi-channel analyzer. Sample were subjected to overnight counting. IAEA certified reference materials (CRMs) were used for the QA/QC of the analysis. 226Ra was estimated through (186.2 keV), and its daughter products 214Pb (351.9 keV) and 214Bi (609.4 keV and 1764.5 keV). As uranium concentration in these samples was reasonably high, 238U activity was estimated from 1001 keV of 234mPa having less abundance (0.837%) but negligible interference from other peak. 232Th activity concentration was estimated by using the gamma line 2614.5 keV (35.85%) of 208Tl. daughter of 228Th. The results of analysis of Uranium mill tailings & radium equivalent are given in [Table 1]. To understand its applicability as building material radium equivalent is calculated using following equation [Eq. 1].{Table 39}
Raeq = CRa-226 + 1.43CTh-232 + 0.077CK-40 Eq. 1
where C is activity conc. in Bq/kg.
The Radium equivalent (mean) of Indian U mill tailings were found in the range of 2331 to 3039 which comes under class III when used as building materials.[1] Building materials are classified into several different types: class 1 – materials for new private houses and public buildings the effective activity parameter of which does not exceed 370 Bqkg-1; class 2 – materials used for road constructions and industrial buildings the effective activity parameter of which does not exceed 740 Bqkg-1; class 3 – materials used for road constructions far away from living regions the effective activity parameter of which does not exceed 2800 Bqkg-1.[2] The current IAEA Basic Safety Standards (BSS) and AERB specify 1 Bqg-1 as the clearance levels for natural radionuclides in the U-238 series in secular equilibrium with their progeny, and the same for those in the Th-232 series. Conclusion: There is a general recognition that at some level, society is now expecting practices that put less pressure on natural resources and entail the recycling and reuse of materials. In view of the presented data, the Indian U mill tailings can be used for the different building materials in construction industries. The same classification system is recognized in Austria, Finland, Slova-kia, Latvia, Norway, Israel and Russia.
Keywords: Radiological concern, recycling, uranium mill tailings
References
Construction and Decoration Materials used in Moscow Region Radioactivity Database. Available from: http://www.radiation.ru/eng/project/Bbase.htm.Turhan S, Arıkan IH, Demirel H, Gung N. Radiat Phys Chem 2011;80:620-5.
Abstract - 23505: Computed dose map of a fuel material handling facility using radiation transport
P. S. Sharma1, S. Anand1,2, K. D. Singh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Fuel cycle facilities handle a large amount of fissile and radioactive substances leading to external and internal radiation exposure. It is necessary to minimize the dose received without affecting the productivity of the facility. The estimation of radiation dose in areas with limited or full occupancy is an important requirement during the design and planning stage of such a facility. Computer programs provide a convenient tool to estimate the radiation dose. Three main categories of computer codes are used for dose estimation namely Point Kernel method based codes, Monte Carlo radiation transport based codes and deterministic radiation transport codes. In the present study we are using a Monte Carlo radiation transport code to estimate neutron and gamma dose inside a fuel material handling facility. The fuel material in the facility is placed in two regions inside the facility. It acts as a source of neutrons which are emitted from α-n reactions and spontaneous fission and as a source of gammas due to n-ϒ capture reaction and fission products present in the fuel. The neutron and gamma ambient dose equivalent was estimated in the facility using the Monte Carlo radiation transport code in various regions of the facility. The gamma source spectra used corresponded to a typical fission product spectra obtained in fuel cycle facilities whereas the neutron spectra used was a watt fission spectrum. The contribution to gamma dose from n-ϒ was not included as it was found to be negligible in comparison to fission product gamma dose. The total source strength used for neutrons was 9.24x107 n/s and for gammas was 2.52x1014 ϒ/s. The output of this computation was fed to a code written to plot the output dose values in various regions of the facility and present a visual display of the dose. The output plot was such that the colour value of a region represented its ambient dose equivalent value H*(10mm) in units of μSv/hr. The plots were obtained in the same plane perpendicular to the Z axis, although the plotting program was written in such a way that the plots can be obtained in X, Y or Z planes by appropriate selection of program input data. The final output plot of neutron ambient dose equivalent is presented in [Figure 1] below which is a plan view with room length along horizontal and breadth along vertical. The walls of the room are made up of 30cm ordinary concrete of density 2.35g/cc. It indicates a very large dose above 2000 μSv/hr in the source region and a rapid decrease as we move away from the sources. The plot is obtained in a plane passing through the sources. The dose in regions close to the sources decreases from 1400 μSv/hr to 800 μSv/hr whereas in faraway regions of the facility doses are below 400 μSv/hr. The doses are below 200 μSv/hr near the wall regions of the facility. Similarly, the final output plot of gamma ambient dose equivalent is presented in [Figure 2] below. It shows a dose equivalent of more than 2900 μSv/hr in the source region which quickly decreases below 400 μSv/hr in the regions close to the sources. It is below 50μSv/hr in most regions of the facility and below 1 μSv/hr near the wall regions. The neutron dose equivalent was found to be above regulatory limit of 1 μSv/hr outside the facility walls, hence additional local neutron shield of 10cm borated polyethylene was suggested to bring down the dose equivalent within regulatory limits.{Figure 56}{Figure 57}
Keywords: Computer program, dose mapping, Monte Carlo
References
Conversion coefficients for use in radiological protection against external radiation. Adopted by the ICRP and ICRU in September 1995. Ann ICRP 1996;26:1-205.Jaeger RG. Engineering Compendium on Radiation Shielding-Vol.1. Springer-Verlag; 1968.
Abstract - 23526: Prediction of continuous air monitor (CAM) count rate using mathematical modelling
Vaibhav Bhardwaj, Joy Chakraborty, Ashish Arvind, Amit Bhatnagar, M. K. Sureshkumar
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Continuous air monitors (CAM) are installed in various nuclear/radiological facilities for the early detection of radioactive air contamination at work place. Many nuclear and radiological facilities still continue to use the gross α/β counter in CAM for the quantification of the collected activity on the filtering media. Natural Radon/ Thoron (Rn/Tn) progenies present in air leads to build-up of short-lived activity on the CAM filter. At the same time, if the same filtering media is used for a prolonged period, a reduction in flow rate due to filter clogging is expected. The net α/β gross count rate on a CAM filter on continuous operation is an interplay of these competing factors. In present work, we tried to model the gross β count rate of a CAM in continuous operation for 8 days, without in-between renewal of the sample collection medium. The count rate variation with time depends on background concentration of Rn/Tn progeny in air, which in turn is governed by factors such as Rn/Tn emanation rate from the wall, ventilation rate, flow rate of the CAM and atmospheric conditions. Radioactive decay and convective dilution of Rn and its progenies (assuming negligible Tn concentration in the room) can be represented by the following differential equations,
[INLINE:12]
[INLINE:13]
where Q0 = Rn atoms per m3 in air, Qi= Rn progenyatoms per m3 in air for I = 1 to 3, Ki = decay constant in min-1,I = air changes per min and S = source term for Rn from emanation from the wall and environmental Rn coming into the air due to ventilation. The deposition rate of Rn progeny on the CAM filter paper at any time can be represented as:
[INLINE:14]
[INLINE:15]
where Ni is the number of atoms (i) deposited on the filter paper at time t, fis the air suction rate of CAM, € is the efficiency of counter which is 20% and Cb is the gross β count rate on CAM at time t. For standard operating conditions of laboratories, a constant ventilation and Rn emanation rate can be assumed. Under such conditions, flow rate f is the variable that affect the observed count rate on CAM. For understanding the time dependant flow rate reduction in a typical case, a Pla PAM341AB model CAM was continuously run for 8 days. Glass fibre filter paper having 2” dia and 0.4 mm thickness is the collection medium used in this instrument. Observed reduction in sampling flow rate as a function of time is given in [Figure 1]. Shielding in dust is ignored due to higher energy ranged β particles. Applying this observed flow rate function into Eq. 3, and solving Eqns. 1, 2, 3 and 4 simultaneously result in Eq. 5.{Figure 58}
[INLINE:16]
[Figure 2] represents the temporal evolution of gross β count rate on CAM by experiment and modelling. Individual peak on the experimental curve are due variation in ambient Rn/Tn-progeny concentration as a result of variation in day-night temperature.[1] Though model and experiment roughly match, modelling with huge data set need to be generated before using the results for predictive purpose and alarm setting for CAMs placed at specific locations in a nuclear facility. This is due to several factors such as daily variation in ambientRnconcentration with change in temperature, humidity etc., as well as disturbances in room ventilation.{Figure 59}
Keywords: Continuous air monitors, flow rate, radon progeny, ventilation rate
References
Ashoke GV, et al. Radiat Prot Environ 2011;34:235-9.Das T, et al. Nucl Technol 2021;207:596-603.
Abstract - 23553: Estimation of site-specific atmospheric dilution factors for nuclear fuel complex, Hyderabad
Amrit Pal Singh, P. Bhargava1, S. Chitra1, Praveen Kumar1, Kapil Deo Singh1, M. S. Kulkarni1, K. Vishwa Prasad1, V. V. Mahesh Kumar, Dinesh Srivastava
Nuclear Fuel Complex, Hyderabad, Telangana, 1Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
The Nuclear Fuel Complex (NFC) manufactures the fuel for the pressurized heavy-water reactors (PHWR) and boiling water reactors (BWR) operating in India. The routine operation of fuel fabrication facilities of NFC may lead to the release of small amounts of uranium particulates into the atmosphere that gets diluted and dispersed depending on the prevailing meteorological conditions at the site. The atmospheric dilution factors (X/Q) at various heights are used in radiological impact assessment due to the release of effluents in the particulate or gaseous form from nuclear facilities. In the present study, X/Q values were estimated for NFC Hyderabad site using the meteorological data of five years starting from 2016 to 2020. The atmospheric stability category for each hour was determined by the Sigma Theta (σθ) method based on the wind direction fluctuation in each hour from the five-minute average data of each hour. The data was processed into annual Triple Joint Frequency Distributions (TJFDs) of three parameters, namely Pasquill stability category, wind direction, and wind speed using an in-house developed computer code. Six Stability categories (Pasquill A through F), 10 wind speed classes (calm, 3-5, 6-11, 12-19, 20-29, 30-38, 39-50, 51-61, 62-74, 75-87, 88-98 in km/hr) and 16 directions starting at North and moving clockwise at 22.5° angular spacing were used in the calculations. The annual TJFD data is input to the in-house computer code “TRINOR“, which outputs the atmospheric dilution factor (X/Q) for each sector and for different stack heights based on the ECPDA Report.[1] The code models the long-term time-integrated concentration at various distances assuming the annual source rate to be constant and is based on the formulation given in Hukkoo et al.[2] The X/Q values were computed among all 16 sectors for 10 m and 50 m heights corresponding to rooftop and stack release at distances from 0.15 km up to 30 km. The predominant wind direction and wind speed during 2018 is represented as the wind rose diagram in [Figure 1]. The predominant wind direction is from West-Northwest (WNW). The X/Q values for releases from roof top and 50 m stack at 300 m distance are 32.7 and 1.60 μs/m3 respectively. It was observed that the X/Q values are highest in the West direction followed by the WNW direction [Figure 2]. It was also observed that for the rooftop releases the X/Q values decrease with increasing distances. The estimated X/Q values can be used for public dose assessment for recommending dose constraint for the facility.{Figure 60}{Figure 61}
Keywords: Atmospheric dispersion, stability category, uranium, wind rose
Acknowledgements
The authors are grateful for the encouragement and support provided by Dr Dinesh Srivastava, Chairman & Chief Executive of Nuclear Fuel Complex.
References
Report on Methodology for Computation of Public Dose and Dose Apportionment for DAE Facilities-Atmospheric and Aquatic Pathways. AERB, ECPDA; December 2020.Hukkoo RK, Bapat VN, Shirvaikar VV. Manual of Dose Evaluation from Atmospheric Releases, BARC-1412, Health Physics Division. 1988.
Abstract - 23554: Immobilization of uranium contaminated soil in cement matrix and estimation of leaching fraction for immobilized waste
Swaroopa Lakshmi, Amrit Pal Singh, K. Vishwa Prasad, V. V. Mahesh Kumar, Dinesh Srivastava
Nuclear Fuel Complex, Hyderabad, Telangana, India
E-mail: [email protected]
Nuclear Fuel Complex, Hyderabad is involved in manufacture and supply of nuclear fuel and core structural components for all operating PHWRs across the country. During the dismantling of various civil structures in radioactive plants, significant amount of civil debris/soil is generated which is radioactive in nature and cannot be disposed as exempt level waste. Cementitious materials are being widely used as solidification/stabilisation and barrier materials for a variety of chemical and radioactive wastes. Cement has many favourable properties, both chemical and physical, making it a desirable barrier or matrix for the encapsulation of radioactive and toxic wastes. The retention properties result from various mineral phases in hydrated cement that possess a high density and diversity of reactive sites for the fixation of contaminants through a variety of sorption and incorporation reactions. Experiments were conducted on immobilization of soil contaminated with uranium in cement matrix. A dye was made for preparing cylindrical blocks of about 32 mm dia and 30 mm height with different compositions of soil and cement. Different composition of contaminated soil and cement have been used to make blocks and a composition of 50 % cement and 50 % contaminated soil was finalized based on the physical strength of the blocks. Blocks with a composition of 50 % cement and 50 % contaminated soil were subjected to leaching in aqueous media at different contact intervals. Leaching characteristics of uranium immobilized in the solidified matrices was investigated through the semi-dynamic leaching test. The leaching behaviour was evaluated by several leaching parameters and predicted by one-dimensional radionuclide decay model based on the Fick's second law. The leach rate coefficient for uranium at neutral pH condition was found to be 2.35×10-6 cm/d. The total uranium leached from the sample after the end of the experiment was 5.75 percent. Authors are grateful to the Lab staff of Health Physics Unit for their constant support during the analysis of samples.{Figure 62}{Table 40}
Keywords: Cementation, leachability, radioactive waste, uranium
References
Atkins M, Glasser FP. Application of Portland Cement-Based Materials to Radioactive Waste Immobilization. Department of Chemistry, University of Aberdeen, Meston Walk, Old Aberdeen, AB9 2UE, Scotland.Bart F. Ce´line Cau-dit-Coumes Frizon F, Lorente S, editors. Cement-Based Materials for Nuclear Waste Storage.Zhang W, Wanga J. Leaching Performance of Uranium from the Cement Solidified Matrices Containing Spent Radioactive Organic Solvent.Venkobachar C, Iyengar L, Mishra UK, Chauhan MS. Release of u(w) from spent biosorbent immobilized in cement concrete blocks.
Abstract - 23564: Determination of dose rate-to-activity conversion factor for off-gas filter by gamma spectrometry and theoretical means
Ram Sharma, B. Suresh, S. Selvaganapathy, S. Murugan, G. Ganesh1, M. S. Kulkarani1,2
HPU, WIP, HPD, Bhabha Atomic Research Centre Facilities, Kalpakkam, Tamil Nadu, 1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Off-gases are generated during radioactive waste treatment and contain moderate to significant amounts of radioactive particulates in them. These gases effluents are treated in order to retain the radioactivity within the system and minimize the release of activity to the environment. The treatment involves passing the off-gas through a series of equipment such as scrubber, chiller, demister, heater, cyclone separator, filter, etc. Of these, the off-gas filter would require replacement at certain intervals depending upon the differential pressure across the filter (owing to dust loading) or a certain threshold surface dose rate on the filter. Estimation of activity loaded in the filter is necessary before suitable disposal/storage. This study presents the methodology adopted in the quantification of activity content in the filter matrix at any given point of time based on the measurement of surface dose rate on the filter by identification of gamma-emitting radionuclides in the off-gas stream and estimation of dose rate-to-activity conversion factor by using Monte Carlo Method. Materials and Methods: Experiment: The composite off-gas filter is made up of micro-glass (borosilicate) fiber medium of dimension 62 cm (L) x 62 cm (W) x 61 cm (H) enclosed in a 3 mm thick SS 304 frame. A specially designed air monitoring system was made use of for collection of air sample from upstream of the filter ensuring adequate radiological protection during sample collection. Air sample were drawn from the stream and subjected to gamma spectrometry (using a p-type HPGe detector with 45% relative efficiency). The spectrum analyses revealed the predominant presence of 137Cs (661.65 keV) as shown in [Figure 1]. Simulation: Monte Carlo FLUKA[1] code was used to model and simulate the gamma transport taking into account the dimensions and composition of the filter medium and assuming the activity was assumed to be uniformly distributed within the filter medium. The dose rate was calculated using the fluence determined with a track length estimator (USRTRACK card) and fluence to dose conversion factors based on ICRP-74 which folds the particles fluence in to dose. Number of histories used in this code was 108 to keep the statistical errors below 5%. Estimation of dose rates on the surface of the filter and at a distance of 1 meter from the surface was tallied.{Figure 63}
Results and Discussion: Dose rate on the surface of the filter and at a distance of 1 meter from the filter surface were tallied corresponding to unit activity are presented in [Table 1]. From the measured data, the extrapolated activity in the filter per unit dose rate is derived and also presented. Thus, the dose rate-to-gamma activity factor for the composite filter is estimated to be 1.14E+03 MBq per mGy/h.{Table 41}
Conclusions: The study of off-gas stream using gamma spectrometry has confirmed the presence of 137Cs as the predominant radionuclide. The dose rate-to-activity conversion factor established helps evaluate the gamma activity content present in the filter medium at the time of filter disposal.
Keywords: Activity, conversion factor, dose rate, filter, off-gas
Reference
Ferrari A, Sala P, Fasso A, Ranft J. FLUKA: A Multi-Particle Transport Code. 2005.
Abstract - 23623: Study on the dynamic regulation of emission limits of radioactive airborne effluents
J. Kang1,2, B. Lian1,2, H. L. Chen1,2
1China Institute for Radiation Protection, 2Key Laboratory of Radiation Environment and Health of the
Ministry of Ecology and Environment,
Shanxi, China
E-mail: [email protected]
According to the national standard GB13695 or operating experience, the regulatory authorities will propose the limit values to manage the normal operation of the nuclear facilities for the effective supervision the operating emissions of the nuclear facilities. Manage limits consist of two components, the normal annual radioactive emissions and possible emissions from expected operating events. With the stable operation and good practices of nuclear facilities, the annual emissions of facilities are well below the limit value. The role of existing management limits in actual operation is not significant. With the safe and stable operation of China's nuclear facilities in the past two decades, it is necessary to optimize the radioactive emission limit value from the perspective of management strategy. In order to optimize radiation protection and feed back good operating practices in the operation of nuclear facilities to the safety regulatory authorities, based on the situation of multiple operating facilities and a single discharge outlet, a dynamic regulation method for the optimization of radioactive emission limit value is established. The influencing factors established in the methodology include monthly emission statistics, normalized emissions, regional meteorology, local meteorology, and expected operating events. By the method, a nuclear facility could implement the dynamic regulation of the dynamic radioactive emission limit value in [Figure 1], and finally characterizes it in the form of public dose and collective dose. In addition, by strengthening the level of nuclear safety culture in the industry, it can further optimize the emissions of facilities and improve the level of refined management and control of radiation environment safety in nuclear bases.By the method, the expected dose of a nuclear facility during the normal operation of a nuclear facility to the representatives is reduced by 15%. Finally, by the method, more space for development of the nuclear base is obligated.{Figure 64}
Keywords: Emission limits, headroom amount, optimization method, statistical analysis
Abstract - 24107: Electro oxidation of cerium (III) for radioactive decontamination of stainless steel surface
V. S. Yalmali, J. G. Shah1
Human Resource Development Division, Bhabha Atomic Research Centre, 1Process System Development Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
The experiments were conducted to demonstrate that Ce(IV) can effectively decontaminate Stainless Steel surface before disposal as metallic waste having alpha radioactivity. As Ce(IV) is powerful oxidizing agent in nitric acid.[1] Spray of Ce(IV) on contaminated surface attacks the oxide layer below metal surface to remove a 0.5 to 3 μm layer.[2] The kinetics of electro oxidation wrt various aspects like cell design, electrode geometry membranes were studied.
Experimental: Electro oxidation of Ce(III) to Ce(IV). In 'H' shaped glass tube, 150 ml of 4M nitric acid was taken and in other side, 0.4M Cerium(III) nitrate in 4M HNO3 was taken. Both solutions were separated by sintered disc G3 at midst (pore size 16–40 μm) [Figure 1]. The electrodes used were SS plate as anode and platinum strip as cathode having surface area 25 cm2. The electrodes were connected to the digital DC power supply. The SS plate used as an electrode which would be decontaminated. To the electrodes voltage was applied gradually in the range of 0.1 to 30 Volts and current produced was up to 2 amperes. The dissolution kinetics of SS plate is in [Table 1]. The experiments were carried out at different current densities from 0.08 to 0.02 amp/cm2 for 2 hours. The solution samples were collected after every 15 minutes and tested for Ce(IV) concentration. The analysis was carried out by iodometric titration using standard sodium thiosulphate. The results were calculated for the current efficiency at different time intervals for that current density. The summarized results are in [Table 2]. The detailed results of each current density wrt to time up to 2 hours are in [Figure 2].{Figure 65}{Table 42}{Table 43}{Figure 66}
Results and Discussion: It is observed that the kinetics of the oxidation reaction depends on resistance of membrane. If resistance is more, the kinetics is slower. The resistance of the membrane increases with the lower voltage but applied, generated current increases temperature and reaction kinetics. The SS surface gets dissolved faster with rising concentration of Ce(IV). [Figure 2] shows highest Ce (IV) conversion kinetics at 0.05 current density in 2 hours; hence 1.5 ampere current is optimized for oxidation process which will remove maximum layers from the metal surface.
Keywords: Cerium ions, decontamination, electro oxidation, solid waste disposal, surface contamination
References
Shah JG, Dhami PS, Janardanan C, Yalmali V, et al. SESTEC-2012. Mumbai: p. 121.Kurath DE, Bray LA, Jarrett JH. Beneficial Reuse '97' (PNNL) Decontamination Of SS Using Ce(IV): Material Recycle and Reuse.
Abstract - 24157: Quantitative evaluation of conservatism included in parameter settings for clearance level calculations: A case study in Japan
M. Sasaki, T. Kimura, T. Hattori11†
Biology and Environmental Chemistry Division, Sustainable System Research Laboratory, Central Research Institute of Electric Power Industry, 1Nuclear Damage Compensation and Decommissioning Facilitation Corporation, Tokyo, Japan
2†The author belonged to the Biology and Environmental Chemistry Division during the time of his involvement in the present study at the Central Research Institute of Electric Power Industry
E-mail: [email protected]
In the evaluations of numerical criteria in radiation safety standards, several possible exposure scenarios and parameters are assumed to calculate the doses to the public. The 'representative person' as defined by the International Commission on Radiological Protection (ICRP) is used for this purpose. Particularly for very low doses, excessive conservatism may not only increase the financial burden of society but also hinder our utilization of resources for recycle/reuse. For example, dose assessment for clearance levels may be a case of excessive conservatism because activity concentrations for clearance levels are extremely low and their impact on human health is negligible. Coats has pointed out and discussed five conservatism factors included in the clearance process. Regarding conversion to activity concentration, he noted that the number of parameters varies between a minimum of 4 or 5 to a maximum of around 12 (especially when environmental transfers are involved), and pointed out that the overall conservatism would be roughly a factor of 3-15, assuming the factors of conservatism for each parameter compared with realistic values were in the range of 1.2-2.[1] In this study, we attempted to quantitatively evaluate the factor of conservatism for parameters used in the dose calculations for an actual case of clearance level calculations in Japan. For major nuclear facilities in Japan, clearance levels were calculated for metals and concrete in the scenarios of disposal and recycle/reuse, and reported by the Special Committee on Safety Standards for Radioactive Waste of the Nuclear Safety Commission (NSC) in 1999.[2] In the NSC report, realistic/representative/conservative parameters were chosen accordingly for several tens of plausible exposure pathways, and clearance levels were evaluated by comparing the annual dose of the highest exposure scenario (critical pathway) with 10 μSv/year. Furthermore, the probability distribution of annual doses was calculated by giving normal, uniform, log-normal, and log-uniform distributions to the parameters, and it was confirmed that the 97.5%ile of the resultant annual dose distributions did not exceed 100 μSv/year by referring to IAEA TECDOC 855. The quantitative evaluation of conservatism was represented as the ratio of the parameter settings actually applied in the clearance level calculation to the expected values evaluated from the probability distribution. As a result, it was found that the parameter that had the highest conservatism was the aquifer thickness of the groundwater migration pathway for the disposal scenario, with a factor of 7.2. For the recycling and reuse scenario, the most conservative factor was 4.5 of the ratio of the amounts of clearance objects to clearance and nonradioactive objects in concrete materials. As mentioned, although Coats roughly assumed that the factors of conservatism for each parameter were in the range of 1.2–2 without explicit explanation for derivation in his paper, the values of the factor of conservatism obtained in this study were not substantially different overall. The regulatory clearance level is not necessarily evaluated solely by a simple multiplication of these parameters; conservatism is also considered for element- and radionuclide-dependent parameters. By recognizing of the existence of such conservatism, it is considered that reasonableness can be applied to practice in radiation protection in accordance with the risk level.[3]
Keywords: Case study, clearance level, conservatism, parameter
References
Coats R. Prudence and conservatism in radiation protection: A case study. Radiat Prot Dosim 2017;173:100-3.Nuclear Safety Commission. Clearance level for major nuclear facilities. 1999.Hattori T. Trend of Strengthening Clearance Regulation in Japan and Concerns about its Worldwide Effects on Regulations for Natural and Artificial Radionuclides. Vol. 49. Proceedings of the Fifth International Symposium on the System of Radiological Protection; 2021. p. 98-112.
Abstract - 24230: Leaching behavior of Tc99 in cement-vermiculite matrix
Sonali Khurana, U. V. Deokar, G. Ganesh, M. S. Kulkarni1
Health Physics Division, 1Homi Bhabha National Institute, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
Radioactive sludge produced from chemical treatment of Low-Level Waste (LLW) is fixed with cement-vermiculite matrix for storage into the underground Reinforced Concrete trenches. It's not been long that the presence of long-lived radionuclide, Tc-99 with half-life of 215,000 years have been detected in the LLW. One of the primary radiological safety issues of concern is the failure of concrete encasement in case of long term storage and the probability of leaching of these radionuclides from the active waste product and their further transport along the various environmental pathways that may reach the general public. This may result in water intrusion and consequent mobilization of the radionuclides from the encased waste in concrete by mass flow and/or diffusion into the surrounding subsurface environment. Therefore, it is necessary to conduct an assessment of the performance of the concrete encasement structure on laboratory scale and the ability of cement to confine the radionuclide's leaching process within the trench. In this study, Transport phenomena involved in the leaching of the radioisotope from a cement composite matrix are investigated using an empirical method employing a polynomial equation. Leaching tests were carried out in accordance with a method recommended by IAEA.[1] It will help in managing the long term storage of radiological waste and provide the input data for radiological impact assessment. Lab-scale leaching studies were carried out on cylindrical blocks prepared with waste solution containing known activity of Tc-99. Cement, vermiculite and waste solution were mixed in the ratio 1:0.1:1 to fabricate blocks. Semi-dynamic method of leaching was followed for 300 days. Rain water (RW) stored in open RCC trenches are employed as leachant to simulate the rain water ingress into the trenches. Demineralised water (DW) is also used for comparative measurements. Leachate samples collected are subjected to physico-chemical and radiometric analysis. Plot of cumulative leach fraction with square root of time [Figure 1] shows the system has reached saturation within the study period. The data fits into the three degree polynomial equation that explains the possibility of combination of one or more leaching mechanisms till saturation.[2] The leaching period based on the rate of leaching as per the plot is divided into two regions; fast leaching region (Day 1-7) and slow leaching region (From Day 8).[3] Leaching parameters including first order kinetic rate constant, diffusion coefficient and dissolution velocity have been evaluated. Index of agreement (IOA), a statistical parameter is used to find the best fit model that explains the leaching of Tc-99 from cement-vermiculite matrix. Variation of pH and Conductivity of Leachate samples with leaching period shows good correlation with the activity of Tc-99 leached out. Similar trend is observed for all three parameters showing higher release in the first 7 days and slower release for the rest of leaching period bifurcating the leaching period into two regions, namely, fast leaching region and slow leaching region. Based on the statistical indices of the parameters obtained considering each model individually, first-order rate kinetics, i.e., surface wash-off shows the best fit for the first 7 days in both leachants whereas for the later period, dissolution is the controlling process.{Figure 67}
Keywords: In situ fixation, leaching studies, source term modeling, techentium-99, Waste management
References
Hespe ED. Leach testing of immbolised radioactive waste solids. IAEA 1970.Plecas I. Comparison of mathematical interpretation in radioactive waste leaching studies. J Radioanal Nucl Chem 2003.Patra AC, Sumesh CG, Mohapatra S, Sahoo SK, Tripathi RM, Puranik VD. Long-term leaching of uranium from different waste matrices. J Environ Manage 2011;92:919-25.
Abstract - 24245: Decontamination methodology of aluminium fins tubes used in uranium handling facility
Samsul Arefin, T. I. Khan, S. K. Suman, R. V. Kolekar
Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
Aluminium is used as cladding material in research reactor fuel. The fuel has a form of metallic natural uranium fuel pin enclosed by aluminium fins tube. Uranium fuel pin is inserted inside aluminium fins tubes in canning operation. In this process uranium comes into contact of aluminium. Although there is no loose contamination on bare uranium fuel pins but it may possible to transfer some contamination on inner surface of fins tubes due to frictional forces exerted between uranium pin surface and inner surface of fins tubes. If the cladded fuel pins fail in quality control, then it was de-canned to reuse uranium. De-canned aluminium fins tubes generated in this process was treated as active solid waste without prior knowledge of contamination level of inner surface of de-canned fins tubes. To assess whether the de-canned fins tubes will be treated as active waste or will be exempted as non-active reusable material, contamination of inner surface of fins tubes is quantified. For this purpose, 15 cm part of each fins tube was cut from different portion and each cut piece was divided vertically. Then its contamination was measured directly by alpha scintillation detector with 95% confidence limit. Measured contamination of each fin tube's cut piece was divided by its inner surface area to get contamination in Bq/cm2 unit and measured contamination of each fin tube's cut piece was divided by its mass to get contamination in Bq/g unit. Measured data of contamination of fins tubes was given in [Table 1]. MDL value of the detector system is 0.015 Bq/g. Although contamination is expressed in Bq/g unit, but contamination is not homogenously distributed throughout the solid volume of fins tubes, rather contamination was tightly bound on the inner surface. Normal swipe did not remove contamination from inner surface of fins tubes. For decontamination purpose, ultrasonic decontamination with different medium was tested and factor of decontaminations was estimated. Few fins tubes were monitored before placing it for overnight soaking in 0.5% NaHCO3 solution. Later it was placed in ultrasonic bath for 10 minutes. Observed decontamination factor for this method was 42%. Another few fins tubes were soaked overnight in 2 N HNO3 solution then it was kept in ultrasonic bath for 10 minutes. Observed decontamination factor in this method was 100%. Average background count of the detector in this study is 20 counts in 300 seconds. For recycling of aluminium it is very important to abide by guidelines of regulatory body and suggestion of international agency. The levels of exemption of moderate amounts of material without further consideration of different nuclide is mentioned in IAEA-GSR part-3.[1] The exemption level of 238U and 234U activity concentration in aluminium for recycling purpose is 2 Bq/g.[2] Since most of the activity (more than 97%) in natural uranium is contributed by these two isotope, so exemption level of natural uranium in aluminium can also be assumed as 2 Bq/g. Loose contamination should further be restricted to 0.37 Bq/cm2 for beta and 0.037 Bq/cm2 for alpha.[3] From this study it can be concluded that activity limit in de-canned aluminium fins tubes are below than exemption limit, so it can be recycled for use in another purpose and in any case if it is needed to decontaminate aluminium by ultrasonic decontamination then one should use acidic medium for better decontamination.{Table 44}{Table 45}
Keywords: Decontamination procedure, disposal limit, uranium contamination
Acknowledgement
Authors acknowledge the contribution of Dr. M.S. Kulkarni, Head-HPD, Shri S.K. Jha, Head-AFD, Shri Pankaj S Damke, UED and Ashma Khot HP/AFD.
References
International Atomic Energy Agency. Radiation Protection and Safety of Radiation Sources: International Basic Safety Standard. IAEA Safety Standards Series No. GSR Part 3. Vienna: International Atomic Energy Agency; 2014.International Atomic Energy Agency. Clearance Levels for Radionuclides in Solid Materials. IAEA-tecdoc-855. Vienna: International Atomic Energy Agency; 1996.Bhabha Atomic Research Centre. Radiation Protection Manual for BARC facilities. BSC Safety Manual No. BSC/SM/2020/3 Rev.0. Mumbai: Bhabha Atomic Research Centre; 2020.
Abstract - 24278: AQUADOS: An indigenously developed code for carrying out radiological safety assessment of aquatic releases
Praveen Kumar, Pradeep Bhargava, Kapil Deo Singh, M. S. Kulkarni
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Public doses of radioactive discharges, both atmospheric and aquatic, from nuclear facilities into the environment needs to be evaluated for the purpose of evaluating suitable discharge limits and to allow comparison with the relevant dose limiting criteria specified by the relevant regulatory authority. AQUADOS, a code in python language, is being indigenously developed that can be used for public dose assessment of aquatic releases into surface water bodies. Radionuclides which are released into the surface waters undergo a series of physical and chemical processes during their transportation in the surface water body. Some of these processes are: advection and turbulent dispersion; sediment processes; other processes, including radionuclide decay and other mechanisms that will reduce radionuclide concentrations in water. Radionuclide transport in surface waters is estimated by employing one of the three basic types of models: numerical models which transform basic radionuclide dispersion equations into finite difference or finite element forms; box type models which divide the entire water body or sections of a water body into homogeneous compartments; analytical models which solve the basic radionuclide dispersion equations using simplifying assumptions. In developing AQUADOS, third approach as taken from SRS-19 (SRS 19, 2001) is used to estimate the radionuclide transport in surface waters. In general, the radionuclide transport in surface waters is modelled as three-dimensional advection-diffusion equation:
[INLINE:17]
where Cw,tot is the radionuclide concentration [INSIDE:4]; U,V,W, are the flow velocities in the x, y, and z directions, respectively [INSIDE:5] is the radionuclide addition or subtraction, for example production of a daughter product (Bqm-3s-1); t is the time (s); x,y,z are the longitudinal, lateral and vertical directions, respectively, in Cartesian co-ordinates (m); εx,εy,εz are the dispersion coefficients in the x, y and z directions, respectively [INSIDE:6] is the radionuclide decay constant (s-1). Various simplifications are applied to this general governing advection-diffusion equation to obtain the appropriate solutions for radionuclide concentrations in the different types of surface water bodies. Some of these simplifications being, discharge of radionuclides from the nuclear facility is assumed to be steady and continuous, and steady flow conditions in each of the surface water body. The analytical solutions to the governing equation thus obtained in SRS-19 (SRS 19, 2001) under the above simplifications are modelled using python language into AQUADOS code. Standard dose assessment models given in SRS-19 (SRS 19, 2001) are also included in this code for estimating the doses from the concentration obtained as a solution to the analytical models. As a verification of the AQUADOS code, the results obtained in an example study from SRS-19 (SRS 19, 2001) are compared with results obtained using this code, in [Table 1]. In this example study, Strontium-90 is discharged into a small estuary and the resulting concentration is estimated in the estuary 1 km upstream. Even though the doses resulting from this concentration are not reported in SRS-19, just to serve as an illustration of the kind of radiological safety analysis that can be performed using the code, the adult doses obtained from different exposure pathways are reported in [Table 2]. These doses were obtained assuming standard reference values (SRS 19, 2001) for different food items consumed by the reference population.{Table 46}{Table 47}
Keywords: Advection-diffusion equation, AQUADOS, aquatic releases, surface water
Reference
IAEA. Generic Models for Use in Assessing the Impact of Discharges of Radioactive Substances to the Environment, Safety Report Series19. IAEA; 2001.
Abstract - 24319: Speciation of toxic metals in uranium mill tailings by sequential extraction method
Samim Molla1, B. K. Rana1, S. K. Jha1,2, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Mill tailings consist of uranium with their progeny in the respective decay series and contain several toxic metals, such as iron, manganese, lead, cadmium, chromium, etc. Uranium and the associated metals may represent a potential radiological and toxic hazard to the environment. These metals can be migrated and transformed through a series of physical and chemical processes, such as dissolution-precipitation, redox, adsorption-desorption, and complexation reactions, and may remain in the environment for a long time.[1] Tessier's sequential extraction procedure is a classical method allowing the chemical separation of major mineral ingredients. The present study focused on the distributions of Co, Zn, Mn, Fe, Pb, Cr, Cd, Cu, and Ni in different phases of mill tailing samples to understand the fate and transport of toxic metals in the environment.
Materials and Methods: Tailings samples (4 nos.) were collected from the tailings pond at Turamdih, Jharkhand, India. Mill tailings are generated by sulphuric acid-assisted leaching of uranium ores. Tessier's sequential chemical extraction procedure was followed to fractionate elements in the tailings samples.[2]
Exchangeable fraction (Exc.): 40 mL of a 1M MgCl2 solution (pH = 7.0) was mixed with 5g of sample (TMD-1). The mixture sample was shaken for 1h and centrifuged for 20 min at 8000 rpm. The residue was labeled as TMD-2 for the next steps.
Carbonate fraction (Car.): 40 mL of a 1M sodium acetate solution (pH = 5.0) was used to leach the residue (TMD-2). The mixture was shaken for 5h at room temperature and subsequently centrifuged. The residue was labeled as TMD-3.
Fe/Mn oxide fraction (Red.): TMD-3 residue was treated with 100 mL of 0.04M NH2OH.HCl solution with a 25% acetic acid solution at 96 °C for 6h, the mixture sample was centrifuged. The residue (TMD-4) was retained for subsequent use.
Oxidizable fraction (Ox.): TMD-4 was allowed to react with 15 mL of 0.04M HNO3 and 25 mL of a 30% H2O2 solution for 2h at room temperature and subsequently heated at 85 °C for 3h. Finally, 25 mL of a 3.2M ammonium acetate solution (pH=2) were added, and the mixture was shaken for 2h, after which the mixture was centrifuged and the residue was labeled as TMD-5.
Residue fraction (Res.): The last residue (TMD-5) was digested with 10 mL HClO4 and 10 mL of HCl until complete dissolution. Finally, the residue was dissolved in 12N HCl and diluted to 50 mL.
Results and Discussion: The order of distribution of Co in different phases was: residual (83%)> exchangeable(6.6%)> carbonate bound> reducible> oxidizable. The order of distribution of Zn in different phases was: residual> exchangeable> carbonate bound> reducible> oxidizable. The uranium ore is mainly composed of silicate minerals, thus the high abundance of cobalt and zinc in the residual phase is justified. Willemite (Zn2SiO4) and hemimorphite (Zn4(Si2O7)(OH)2.H2O) are well-known silicate minerals that host zinc. The major portions of Mn were found in the reducible phase(44%) due to the existence of the manganese as oxides and oxide-hydroxide. Iron was mainly found in the residual phase(99.8%), which indicates its low bioavailability. The order of distribution for lead (Pb) in different phases was: reducible> residual> carbonate bound > exchangeable> oxidizable. 50% of the Pb was found in the Fe/Mn oxide bound phases, which indicates that Pb cannot be mobilized under oxidizing environmental conditions. About 73% of the chromium was present in the residual phase and only about 3% was in the exchangeable phase. Therefore, it may be assumed that the chromium complexes are quite stable under general environmental conditions. The total cadmium concentration in the tailings was in the range of 1.5–1.6 mg kg-1, 16% of which was in the exchangeable fraction. The high occurrences of the Cu with the oxidizable phase (84%) may be attributed to the presence of chalcopyrite (CuFeS2) in the tailings. Nickel was mainly associated with residual (93%) and Fe-Mn oxide bound phase (4%) which may be attributed to the tendency of nickel to be associated with limonite and Fe-silicate minerals.
Keywords: Sequential extraction, uranium mill, uranium tailings
References
Zachara JM, et al. Geochemical processes controlling migration of tank wastes in Hanford's vadose zone. Vadose Zone J 2007;6.4:985-1003.Tessier A, et al. Sequential extraction procedure for speciation of particulate trace metals. Anal Chem 1979;51:844-51.
Abstract - 24395: Uranium sorption using phosphate functionalized chitosan and silica/chitosan composite
P. Singhal, B. G. Vats1, V. Pulhani
Environmental Monitoring and Assessment Division, Bhabha Atomic Research Centre, 1Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected], [email protected]
Uranium is one of the important elements present on the earth and used in nuclear industry. It is also a well known nephrotoxic element with a maximum permissible contamination level of 30 ppb in drinking water.[1] Numerous methods have been developed in past to extract uranium.[1] However most of the techniques are expensive, need pre-treatment of the sample, large contact time, non-selective and have low sorption capacity. Therefore it is essential to design cost effective and environmental friendly methods for uranium extraction. Nowadays several nanomaterials, such as carbon-based composites, metal oxides have been used.[1],[2] They exhibit excellent sorption capacity but their high dispersibility in aqueous solutions makes separation from the matrix tedious which restricts the application in real water bodies. Nowadays the most popular sorbents for uranium extraction is amidoxime-based materials but the method has several shortcomings such as high cost, large equilibration time (>24 hrs), and high adsorption of other seawater cations (e.g., vanadium).[1] Such limitations prompt further research in this direction that can lead to feasible technologies for extracting uranium. Herein, we have synthesized chitosan and silica/chitosan composite functionalized with phosphate group and uses them for uranium extraction from water. Chitosan can be further coupled into polymer sheets and beads with an ultimate aim to check the feasibility of these materials in large scale samples. The phosphate functionalized chitosan was synthesized by Mannich-type reaction. In brief formaldehyde, orthophosphoric acid and chitosan was taken in a reaction flask and heated at ~100°C overnight. The product is separated from the reactants by centrifugation and repeated washing by alcohol. To synthesize silica/chitosan composite first silica NPs were synthesized by hydrolysis reaction of tetra ethylorthosilicate (TEOS) in alcohol. 5 ml of TEOS is added dropwise in 500 ml of alcohol and the pH of solution was adjusted to 9. The solution was stirred overnight and particles were separated by centrifuge followed by washing with alcohol. This silica NPs were added to reaction mixture as describe above to synthesize phosphate functionalized silica/chitosan composite. The material was characterized by XRD, FTIR, TG and BET measurements. Sorption studies were carried out to understand the sorption properties. Initially pH dependent experiments were carried out to understand the working pH range of the sorbent. The results are shown in [Figure 1]a and [Figure 1]b.{Figure 68}
It is interesting to observe from [Figure 1] that uranium sorption occurs on both the NPs surface however the maximum sorption was observed on the composite material. Also, the working pH range of the material is different. It is due to the presence of phosphate group on NPs surface that enhances the sorption capacity and interact with UO22+ ions prevailing at pH 4-5 where the sorption using this material was observed to be the maximum. Time dependent studies have been carried out to determine the time required to achieve the maximum sorption capacity. It was observed that the maximum sorption in composite reach within 15 min. The material was finally fabricated onto sheets and beads using polypropylene and tested in lab. It was observed that this material sorb uranium however with time the capacity is decreasing and the material is leaching out from the surface and further work is being carried out to modify it. To summarize, we have synthesized and characterized phosphate functionalized chitosan/silica composite and use for the separation of uranium. The material was observed to remove >99% of uranium however to use it commercially more studies are needed.
Keywords: Nanomaterials, sorption, uranium
References
Singhal P, Vats BG, Pulhani V. J Ind Eng Chem 2020;90:17.Singhal P, Vats BG, Pulhani V. J Hazard Mater 2020;384:121353.
Abstract - 24429: Kinetics and adsorption isotherm for migration of caesium and sorption on soil
Sukanta Maity, P. Sandeep, S. Mishra, C. B. Dusane, D. K. Chaudhary, P. Padma Savitri, J. Sudhakar, Anilkumar S. Pillai, A. Vinod Kumar
Environmental Monitoring and Assessment Division, Bhabha Atomic Research Centre, Trombay, Mumbai, Maharashtra, India
E-mail: [email protected]
Due to environmental processes, the engineered disposal modules of radioactive materials may lose their structural integrity and may release some radioactivity to the geo-environment.[1] Accordingly, released radionuclides transportation may occur through saturated porous media and reach the far field.[2] During nuclear accidental scenario like Fukushima the most concern radionuclide was the caesium (Cs) for wide environmental spread. For transportation of radionuclides, one of the most important parameters is the solid liquid distribution coefficient (Kd). Geochemical as well as transport behavior of Cs is necessary for radiation protection and dose assessment for the members of the public around upcoming and operational radioactive disposal sites. Adsorption kinetics and adsorption isotherms play important role in migration of Cs in soil. Soil and groundwater samples were collected from two locations across upcoming nuclear facilities, new BARC site, Vishakhapatnam. Collected soil samples were dried and then sieved through an electromagnetic sieve shaker to collect soil particles of size <2 mm. Suspended solid particulates were removed from collected groundwater through filtration using 0.45-micron filter paper. Sorption experiments were carried out using the laboratory batch method. All the experiments were carried out in triplicates. Cs was analyzed using Atomic Absorption Spectrometry (AAS) (model GBC Avanta). Blank was prepared for each set of experiments following the same procedure as that followed for samples. During the sample analysis, the instrument was calibrated with four points using fresh Cs standards. In between the sample analysis, known standards were analysed to check the accuracy of calibration. Cs concentration in the studied soil samples varied from 3.17±0.09 mg Kg-1 to 4.15±0.12 mg Kg-1. Sorption equilibration time was determined through conducting the experiment at different time intervals (10 min to 4320 min) in two soil samples and was found to be 1980 minutes or 33 hours in both the studied soils. But the rate of sorption was different in the two soil samples. Initially, the rate of sorption is very high and then gradually decreases and reaches equilibrium. Different mechanisms predominate like diffusion, mass transfer, or chemical reactions in sorption kinetics. To understand the sorption mechanism, the time variable sorption data were fitted with both pseudo-first-order and pseudo-second-order rate equations. The plot between t/qt (qt represents Cs adsorbed at time t) and time (t) in the pseudo-second-order kinetics model is represented in [Figure 1]. In the present study Cs sorption for both the soils followed the pseudo-second-order rate equation indicating a chemisorption phenomenon. The sorption study was carried out using variable Cs concentrations (1.66 mg L-1 to 33.33 mg L-1) to understand the role of initial Cs concentration during the sorption process. By increasing the initial concentration of Cs from 1.66 mg L-1 to 20.00 mg L-1, a linear relationship was observed between Cs adsorbed per unit mass of soil and the initial Cs concentration in solution, which indicated that up to 20.00 mg L-1 of Cs solution concentration as tracer can be used for performing sorption experiment without saturation. The above experimental data (Cs adsorption with change in Cs initial concentration) were fitted with linearized forms of different isotherm models, namely Langmuir, Freundlich and Dubinin-Radushkevich (D-R) to understand the Cs adsorption behavior on both studied soils. Here the experimental data fitted well with Freundlich and D-R isotherm models. The Cs adsorption process in collected two soil samples followed a pseudo-second order rate equation with rate constants of 0.00073 g μg-1 min-1 and 0.00025 g μg-1 min-1 at 250C, respectively. The generated data fitted well with Freundlich and D-R isotherm models, indicating Cs sorption in the studied soil is complex and chemical in nature mostly ion exchange type.{Figure 69}
Keywords: Equilibration time, experimental conditions, initial Cs concentration, soil solution pH, sorption
References
Zhaorong S. International Conferences Safety of Radioactive Waste Disposal, IAEA-CN-135/02. Vienna, Austria: International Atomic Energy Agency; 2005. p. 01-4.Yim MS, Simonson SA. Prog Nucl Energy 2000;36:1-38.
Abstract - 24450: Heavy metal detection and removal from aqueous solutions using radiotracer technique and gamma-ray spectrometry
S. A. Shaikh, H. K. Bagla
Department of Nuclear and Radiochemistry, Kishinchand Chellaram College, Mumbai, Maharashtra, India
E-mail: [email protected]
Heavy metals are deemed to pose the greatest threat to the environment and the human body of all pollutants documented. The current approaches to minimize heavy metal pollution are often expensive and have a limited efficacy for practical applications.[1] Thus, biosorption was selected over the conventional techniques. Adsorption and desorption processes are highly dependent on the accessibility and mobility of metal ions in the surrounding environment. The operating parameters were investigated by batch equilibrium experiments. Humic acid (HA), a green biosorbent, was used in this study to remove and recover Ag(I) and Zn(II) ions from aqueous solutions. Ag-110m (658 keV) and Zn-65 (1.116 MeV) radiotracers with a level of radioactivity 0.024mCi and 0.0115mCi respectively, were used to detect the elements using radiotracer technique. This method offers numerous distinct benefits over conventional technologies, including high accuracy, simplicity, non-destructive and relatively inexpensive nature. Humic acid was examined using EDAX, FTIR, and SEM techniques. A well-type gamma-ray spectrometer (GRS) (PSP649/N, PEA GRS-301) with a NaI(Tl) detector was utilised for elemental detection and was calibrated by using sealed sources of 137Cs, 60Co, 22Na procured from BRIT (Board of Radiation Isotope and Technology). GRS detection is time and cost efficient. It identifies even minimal concentrations of metal ions. Its reliable spectral data analysis offers a way to characterise vast areas of contamination. This feature has made gamma-ray spectroscopy the go-to method for large-scale radio-contaminant mapping and for hot particle detection. Study of mathematical modelling techniques for metal biosorption has also been done. To locate and quantify sorption sites and evaluate the mobilisation process, isotherm models were fitted. Recovery of Ag(I) and Zn(II)ions from loaded humic acid is necessary for disposal and reuse of the adsorbate. Mineral acids (HCl, HNO3), organic compounds (Acetic acid and Citric acid) and EDTA were very efficient in this respect and desorbed almost 70-90% Ag(I) and Zn(II) ions from the loaded humic acid as against only 4-6% by ultra-pure water or distilled water [Table 1]. Under optimum conditions, 200 mg of HA can remove around 15 mg of Ag(I) ion from aqueous solution. Whereas, for the removal of Zn(II), 50mg of HA was more than sufficient. Optimum adsorption occurred at the solution pH of 1.0 for Ag(I) and 6.0 for Zn(II), however, temperature had no significant effect. The adsorption process followed Freundlich isotherm model. The process was rapid and within 5 min 75-78% of Ag(I)[2] and 61-65% of Zn(II) uptake was completed following pseudo-second order rate kinetics. Thus, the interaction of Ag(I) and Zn(II) with HA has industrial applications for the removal of pollutants and recovery of valuable metals from effluents. Therefore, this technique has been used to the fullest potential to maximize the amount of metal uptake from solution and has also proved its significance in optimizing the variables that influence the adsorption and desorption process in reducing both time and cost.{Figure 70}{Table 48}
Keywords: Adsorption, kinetic studies, radiotracer technique, scintillation detector, silver
References
Williams CJ, et al. Water Res 1998;32:216-24.Shaikh SA, Bagla HK. J Radioanal Nucl Chem 2019;322:225-30.
Abstract - 24451: Recovery of 137Cs from simulated reprocessing wastes using humic acid
A. N. Khan, H. K. Bagla
Department of Nuclear and Radiochemistry, Kishinchand Chellaram College, Mumbai, Maharashtra, India
E-mail: [email protected]
137Cs is a major fission product of 235U, and is present in low-level wastes (LLW) along with 90Sr, salts like NaNO3 and NaI, and radionuclides which originate from the structural materials of a nuclear reactor. The Chernobyl and Fukushima Power Plant incidents led to the release of ~85PBq and ~17PBq of 137Cs respectively. With its long half-life, high solubility and mobility in the environment, and chemical similarity to K, the radiotoxic nature of 137Cs has been well-documented. LLW contains large volumes of diluted wastes with radiation less than 1mCi/L.[1] Adsorption was applied using humic acid (HA) as it forms an important part of soil which is a natural barrier in geological waste repositories or in accidental releases. Radiotracer technique was used for elemental detection with 137Cs (initial activity 0.01mCi) using a well-type gamma-ray spectrometer with NaI(Tl) detector and the spectra obtained were analysed as seen in [Figure 1]. Characterization of the adsorbent was carried out using SEM and EDAX. SEM analysis at pH 7.0 revealed clumped and layered structure of HA at 10000x magnification (20μm scale). EDAX spectra gave the native metal ion composition of HA which included Fe, Al, Si, Ti, Ca, and K, among other elements. Experimental parameters for a batch adsorption process were optimized and 88±2% Cs was adsorbed within 10 min, in the pH range of 7.0 to 8.0, with 30 g/L of HA at R.T., 303K. As reprocessing LLW contains higher salt content, the effect of increasing concentration of NaNO3 has been evaluated. At 170 g/L NaNO3, 70±2% Cs adsorption was obtained. The range from 1 to 170g/L can be seen in [Figure 2]. Radiation stability of HA was studied by subjecting it to 100 and 200 kGy of γ-radiation from a 60Co source at the rate of 1.5 kGy/hr. Experiments performed post-irradiation at 100 kGy revealed that there was no change in adsorption, whereas at 200 kGy, 2 to 3% decrease in adsorption was noted. FTIR analysis pre- and post-irradiation indicated a shift in the O-H and C=O stretch. Desorption investigations are important as they dictate the final disposal procedure required. The use of desorbing agents like 1M citric acid and 1M HCl resulted in 60±3% and 69±2% Cs desorption respectively. Compared to Cs adsorption from simulated reactor waste, which was previously studied,[2] it is seen that 91±2% of adsorption and 82±2% of Cs desorption was obtained. Thus, it is easier to decontaminate reactor waste as compared to reprocessing waste. A contributory factor could be the high salt concentration and its interference with comparatively low Cs concentration. Post desorption, HA was washed and reused for three adsorption-desorption batches, with adsorption efficiency dropping from 70±2% to 62±3% in the third cycle at 170g/L NaNO3 concentration. This research documents the use of commercially available geomaterial- humic acid, for the recovery of 137Cs from simulated reprocessing wastes. The low cost of humic acid, short reaction time, rapid detection using radiotracer technique, desorption and sorbent regeneration, and recycling, makes the process efficient and green.{Figure 71}{Figure 72}
Keywords: 137Cs, adsorption, humic acid, reprocessing waste, separation
References
Valsala TP, et al. J Hazard Mater 2011;196:22-8.Khan AN, Bagla HK. J Trace Elem Miner 2022;1.
Abstract - 24452: Application of tracer technique for sorption-desorption of radionuclides from simulated low level waste
S. Sayed, H. K. Bagla
Department of Nuclear and Radiochemistry, Kishinchand Chellaram College, Vidyasagar Principal K. M. Kundnani Chowk, Mumbai, Maharashtra, India
E-mail: [email protected]
Management of low level waste is a significant concern of the present day; as the use of radionuclides increases, waste containing radioactive substances increases. The present study deals with removal of Cs(I) and Co(II) metal ions from simulated low level waste by using an eco-friendly adsorbent dry cowdung powder.[1] The biosorbent does not require any pre-treatment and is enriched with adsorptive ability due to the presence of humic acid, fulvic acid, bile pigments, proteins, etc. Low level radioactive wastes contain small amount of long lived radionuclides in which Cs-137 and Co-60 with half-life of 30.17 and 5.27 years respectively have a major part. To remove these radionuclides from simulated low level waste, radiotracer technique has been applied in which small amount of Cs-137 (0.662 MeV) and Co-60 (1.17 and 1.33 MeV) radiotracers with initial activity of 0.01 mCi and 0.028 mCi respectively were added to the system containing Cs(I), Co(II), Sr(II), NaI, and NaNO3. Percentage sorption and desorption was analyzed based on the amount of radioactivity present in the aliquot after the process. The analysis of radiotracer was done using Single Channel Gamma-ray spectrometer with NaI(Tl) well type detector. The experiments were done in triplicate to understand the reproducibility of the process. 60±3% and 90±2% biosorption of Cs(I) and Co(II) was successfully achieved in 5 mins of contact time at pH 8.5 at room temperature of 303 K with 500 mg and 200 mg of adsorbent respectively. Biosorption of both metal ions was exothermic in nature due the negative value of ΔH° from the thermodynamics study. The positive value of ΔS° revealed that the degrees of free active sites increased at the solid–liquid interface during the biosorption process. The biosorption data for Cs(I) and Co(II) follows Ho and McKay Pseudo Second Order kinetics with maximum adsorptive capacity of 1.41 mg/g for Cs(I) and 4.60 mg/g for Co(II). For the complete removal of metal ions from metal-loaded DCP, desorption studies were done using mild and eco-friendly acid like citric acid and acetic acid from which approximately 90% of adsorbed Cs(I) and Co(II) was successfully desorbed in 5 mins of contact time with 5 mL of desorbing agent and 0.1 M concentration at room temperature of 303 K. The interference of different ionic salts was also investigated for the selective separation of metal ions from simulated low level waste. Salts like chloride, nitrite and bromide were selected for separation studies.[2] The biosorptive behaviour of DCP in presence of these salts was explained using Hard and Soft Acid Base (HSAB) Theory. By following the maximum percentage biosorption parameters of Co(II), these salts suppress Cs(I) biosorption from 56% to 10%, 15%, and 18% by chloride, nitrite, and bromide salts respectively, and does not affect Co(II) biosorption. The process follows rapid kinetics and green principle of generating minimal sludge. The application of DCP for treatment of aqueous radioactive waste is an eco-friendly approach.{Figure 73}{Table 49}
Keywords: 137Cs, 60Co, biosorption, desorption, low level waste, radiotracer technique
References
Bhatt ND, Bagla HK. J Environ Biotechnol Res 2016;6:168-78.Sayed S, Bagla H. Res J Chem Environ 2021;25:1-6.
Abstract - 24486: Calibration of WACT System in segmented gamma scanning mode: Establishing isotope-specific calibration and detector efficiency curves
R. Ramar, R. Mathiyarasu D. Ponraju, B. Venkatraman
Health and Industrial Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
A Waste Assay Computed Tomography system (WACT) is indigenously designed and developed for the assay of low and intermediate-level alpha-bearing wastes stored in standard SS drums. This system works in Tomographic Gamma Scanning (TGS) and Segmented Gamma Scanning (SGS) mode. SGS is slice scanning with the rotational motion of the drum, whereas TGS is voxel scanning with translation and rotational motion and pinpoints the location of the radionuclide in the drum. In SGS mode, the calibration curves are essential to get the layerwise activity distribution of radionuclide in the drum from the emitted gamma rays. WACT system [Figure 1] is calibrated as per the ASTM-C1133[1] standard. The isotope-specific (mass) calibration for estimating the mass of 239Pu and the efficiency curve (energy) calibration for assessing the activity of fission and activation products inside the drum are established. A simulated waste drum filled with cotton waste and different loading masses (89 mg to 2177 mg) of the PuO2 sample were used for Isotope-specific mass calibration of the WACT system in SGS mode. WACT system is collimated using a 50 mm height collimator. The waste drum (580 mm dia. & 900 mm height) is split into six segments, each with an effective scanning height of 150 mm. The layerwise SGS were taken by rotating the drum and using external multi gamma energy emitting 152Eu source is used to obtain the segment-wise average attenuation correction factor. The emission counts of 239Pu for the 129 keV energy are used to estimate the mass of 239Pu. There is a spectroscopy challenge in detecting and estimating the net area count of 129 keV gamma-ray peak. This low yield peak submerges in the Compton continuum spectrum and backscattered peak of high energy photon. This photo peak at 129 keV energy is extracted from the emission spectrum after applying SNIP and in-house developed peak fitting algorithms. The extracted peak counts were corrected for rate-related losses and attenuation by the drum's matrix content and wall. The total Corrected Count (CC) rate summed over the entire segment is proportional to the isotope mass in the drum. The calibration curve is plotted using the total CC rate and the corresponding mass of 239Pu in the drum. Isotope specific calibration curve for 239Pu is shown in [Figure 2]a. This Calibration curve was validated with a blind test sample. In the blind test, the system estimated the mass of 239Pu up to 100mg with an error of ±20%. The efficiency calibration Curve was established by loading known activities of 133Ba, 137Cs, and 60Co sources in the drum. SGS detection efficiencies(ε) are calculated from the CC of particular energy (E) of radioisotopes of known activity(S) using the following formula.{Figure 74}{Figure 75}
[INLINE:18]
Br(E) is the branching ratio of the isotopes for energy(E). The SGS efficiency of the system is calculated from the CC of various energies of loaded sources. The established efficiency curve is shown in [Figure 2]b. Unknown activity of radioisotopes in the drum can be estimated using the efficiency curve. The efficiency curve is validated with the known activity of 137Cs (1 mCi). The system could measure the activity with an error of 11%. Established calibration curves are used to get the mass of 239Pu and the activity of fission products in radioactive waste drums.
Keywords: Efficiency calibration, non-destructive assay, radioactive waste, segmented gamma scanning
Reference
ASTM International. Standard Test Method for Nondestructive Assay of Special Nuclear Material in Low-Density Scrap and Waste by Segmented Passive Gamma-Ray Scanning. ASTM C1133-10p. ASTM International; 2018.
Abstract - 24501: Correction factor for gross alpha activity measurement in wastes generated from Tummalapalle uranium processing plant
Ranjan Prakash1, Abhigyan1, V. Kumaraswamy1, B. Naresh1, S. K. Jha1,2
1Health Physics Division, Bhabha Atomic Research Centre,
2Homi Bhabha National Institute, Mumbai,
Maharashtra, India
E-mail: [email protected]
Radioactivity is widely present in various kinds of water bodies, such as seawater, river, drinking water, groundwater, and wastewater. The measurements of gross alpha and gross beta activity concentrations are among the most effective methods for preliminary screening and evaluating the total radioactivity contents in samples.[1] The waste generated during the process extraction of Sodium Diuranate, uranium concentrate from ore is pumped to tailing ponds in the form of slurry. The solid part of the slurry is contained in the tailings pond and the liquid part of the slurry is decanted and pumped back to the plant for reuse. It has been observed that without the application of self-absorption factors the measured values of alpha specific activity were underestimated significantly. Therefore, a study has been carried out in order to obtain the correction factors corresponding to the waste matrices present in the facility. Fresh samples of U mill tailings were collected and different volumes of liquid part of the wastes were plancheted on aluminum planchet. After drying, evaluation of sample thickness was determined from weight of the sample. The planchet was counted for gross alpha counts using Alpha Beta Radiometer UMF2000 for low level activity measurement. The starting volume as well as Total Dissolved Solid of liquid in waste is so low that self-absorption factor corresponding to the measured thickness is negligible and hence this value of specific activity is taken as the best reportable value. Correction factors are determined with respect to the best reportable value and measured values at different thicknesses. The solid part of the waste, which is finely powdered (<74 μm), is of very low specific activity. So, different masses of solid were spiked with known activity of sodium diuranate powder and the samples were spread as evenly as possible on planchet. The thickness of sample in the planchet was determined from weight of the sample. Correction factors are determined with respect to the known value of activity of SDU at different thicknesses of solid waste samples. The correction factor for liquid and solid waste as a function of thickness has been shown in [Figure 1] and [Figure 2]. The measured values are fit into the curve of second degree polynomial CF = 0.065x2 - 0.063x + 0.959 and straight line CF = 0.388x + 1.542 for liquid waste and solid waste respectively where CF is the correction factor and x is the thickness in mgcm-2. For liquid waste CF is varying from 1 to 2.2 over the measured thickness range of 0.2 to 5 mgcm-2. There is not considerable self absorption till the thickness of 2 mgcm-2. For higher thicknesses correction factors are becoming more and more important. It can be seen that CF for solid waste spiked with SDU is varying from 1.9 to 5.3 over the measured thickness range of 1.25 to 9.25 mgcm-2.{Figure 76}{Figure 77}
Keywords: Alpha activity, correction factor, radioactive waste, tailings, uranium
Reference
Li X, et al. Simultaneous determination of gross alpha/beta activities in water by liquid scintillation counting and its applications in the environmental monitoring. Nat (Sci Rep) 2022.
Abstract - 24530: Radiation protections lessons learned during refurbishment of an old radioactive liquid waste storage tank
Anjan K. Singh, S. Susheela, S. Singh, S. S. Vichare, S. M. Ojha, Pratibha Kamble, A. P. Jakhate1, G. Anand1
Health Physics Division,
Bhabha Atomic Research Centre,
1Waste Management Division,
Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: This report present the radiological safety parameters observed during refurbishment of a Low level radioactive liquid waste (LLW) storage tank which has been in use for nearly 60 years. This underground concrete tank was provided with 6 to 8 mm thick mild steel (MS) lining. Due to prolong use, the MS lining got corroded and need replacement. The major radiological challenge was the high contamination on the lining, which may result in significant external exposure and/or internal/external contamination to the workers.The radiological safety parameters during this work could be of importance to future refurbishment activities of similar nature.[1]
Materials and Methods: Radiation field measurement was carried out usingMGP make dosimeter (DMC 3000). α and β transferable contamination on the MS lining was carried out using portable Pan-cake G M contamination monitor ofmeasured efficiency 23% for 137Cs and ZnS (Ag) scintillation detector with 25% efficiency for 239Pu respectively. Air sampling is done with glass fiber filter paper as the collection medium at a flow rate of 50l pm.
Results and Discussion: Measured dose rate close to the surface at different elevation from the bottom of the tank varied form 1.6 mGy/h to 0.1 mGy/h. During the operational life of the tank, radioactive sludge mostly accumulate at the bottom of tank which resulted in higher contamination and dose rate at the bottom. Radiochemical analysis of the sludge samplefound on the surface was carried out using techniques like gamma spectrometry and liquid scintillation analysis. Major radionuclide were 137Cs (2666 Bq/gm), 90Sr (1300 Bq/gm), 99Tc (800 Bq/gm) with trace amount of other radionuclides like 125Sb, 144Ce and134Cs.After thorough decontamination also significant transferable contamination was observed on the MS lining. The measured activity on the swipe from different locations on the tank was found to be in the range of 0.5 – 2.5 Bq/cm2 (gross α) and 236 – 472 Bq/cm2 (gross β) with contamination level gradually decreasing from bottom to top of the tank. For assessing the radiological hazards during the dismantling of lining, two 600 mm x 200 mm size portion was cut usinggrinding machine. Air sample was taken at 2 feet away from the cut positions as represented in [Figure 1]. [Table 1] provide the measured air contamination during the cutting operation. Despite the fact that the maximum measured transferable contamination was 2.5 Bq/cm2 (α), alpha air contamination generated during the cutting work demand respiratory protection for the worker. Due to long aging and subsequent corrosion, alpha activity might have been embedded in the lining which may not be available for measurement using swipe procedure.{Figure 78}{Table 50}
Conclusions: The radiological monitoring data generated during the representative operations helped in safe completion of the refurbishment of the LLW tank.Measured air contamination during the trial run indicated that half mask respirator could provide sufficient protection from intake during the work.β/αratio in swipe were of the order of 100, while that in the air contamination is found to be ~20. Such anomalies need to be addressed to avoid surprises and to prevent personal exposures while carrying out work on radioactive legacy systems.
Reference
IAEA. Safety Assessment for Decommissioning. IAEA Safety Report Series No 77.
Abstract - 24588: Estimation of distribution coefficient (Kd) of soil for the radionuclide 90Sr at Kaiga Site, Karnataka
T. L. Ajith, Joshy P. James, M. S. Vishnu, I. V. Saradhi1, A. Vinod Kumar1
Environmental Survey Laboratory, EMAD, BARC, Kaiga Site, 1Environmental Monitoring and Assessment Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
This paper presents the distribution coefficient (Kd) value estimated for the radionuclide 90Sr for soil from various depths at Near Surface Disposal Facility (NSDF), Kaiga site, Karnataka. Kd is an important parameter applied to study the migration of radionuclides and is defined as the ratio of the concentration of radionuclide sorbed on a specific solid to the concentration of a radionuclide in liquid phase at equilibrium. (IAEA, 2010).[1] The present study is important because low and intermediate level radioactive solid wastes are disposed in NSDF and the assessment of migration of radionuclides from these facilities to the environment is essential for environmental impact studies at Nuclear power stations in India. Samples were collected from four locations of NSDF area at various depths of 0 to 4 m as given in [Table 1]. Extraneous materials like roots, leaf litters, pieces of gravel and stone were removed from the soil sample, weighed, dried in oven at 1100C for two days, dry weight is noted, crushed in to fine powder, homogenized and sieved for 180 μm particles using analytical sieving machine. 2 gm of sieved soil was mixed with 60 ml of filtered bore well water and 90Sr standard solution. On completion of the shaking period for seven days the supernatant was collected. Samples were stored for a minimum of 35 days prior to counting to allow for secular equilibrium with 90Y to be obtained, and for any unsupported aqueous 90Y present to decay below detection limits.[2] A known volume is transferred to an aluminum planchet, dried and counted in a low beta counting system. Kd is calculated by the equation.[3]{Table 51}
Kd (L.kg-1) = [INSIDE:7] where Ci and Cf are the initial and final amounts of radionuclide in the aqueous phase (Bq.L-1), v is the experimental volume (L) and m is the mass of the soil (Kg). The results are presented in [Table 1]. Kd value of soil for the radionuclide 90Sr for at different depth intervals of each location at NSDF, Kaiga site was varying from 91.8 to 650.7 L.kg-1 with a mean value of 255.8 L.kg-1 and results are comparable with the Kd value reported in IAEA report (range; 4.0 × 10-1 to 6.5 × 103 L.kg-1, mean; 52 L.kg-1)
Keywords: 90Sr, distribution coefficient, Kaiga site, Kd, NSDF
References
IAEA. Technical Reports Series No.472. Vienna: IAEA; 2010.Wallace SH, Shaw S, Morris K, Small JS, Fuller AJ, Burke IT. Appl Geochem 2012;27:1482-91.Twining JR, Payne TE, Itakura TJ. Environ Radioact 2004;71:71-87.
Abstract - 24590: Estimation of dose rates of vitrified waste product canisters by passive dosimetric methods
S. Selvaganapathy, B. Suresh, O. Annalakshmi1, G. Suneel2, G. Ganesh, M. S. Kulkarni, J. K. Gayen2, K. V. Ravi2
Health Physics Division, Bhabha Atomic Research Centre, 2Nuclear Recycle Board, BARC, Mumbai, Maharashtra, 1Radiation Dosimetry Section, EAD, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: High Level liquid Waste (HLW) generated during reprocessing of spent fuel from power reactor contains most of the radioactivity present in the entire nuclear fuel cycle resulting in need for its containment, isolation and surveillance for extended periods of time.[1] HLW from reprocessing plants having an activity concentration of around 100 Ci/l is vitrified into borosilicate glass and the molten mass of Vitrified Waste Product (VWP) is poured in stainless steel (SS) canisters. These canisters contain upto 0.6 million Curie of activity. The limited range of available radiation measuring instruments pose a constraint on measurement of surface dose rate on canisters, necessary in order to qualify the canisters for their interim storage.[2] This paper describes the estimation of surface dose rate on canisters using passive measurement techniques.
Materials and Methods: Around 7-8 mg of CaSO4:Dy (0.2 mol%) thermo-luminescence (TL) phosphors synthesised by co-precipitation technique and loaded in brass capsules were used. TL measurements were carried out in RISϕ TL reader, calibrated using the TL phosphor synthesised in the same batch and irradiated to gamma dose using 137Cs source up to 1 kGy. Electron Paramagnetic Resonance (EPR) based Alanine chemical dosimeter was also used to measure the dose. The concentration of free radicals in the irradiated alanine dosimeter was measured using a Bruker EMX X-band EPR spectrometer and converted into dose using the already established calibration curve for doses up to 22 kGy. Three TLD & Alanine dosimeters each were enclosed inside a polythene pouch and secured to a SS rod. With the canister engaged using in-Cell crane at the desired location, the dosimeters were inserted through the survey port and positioned at the point of interest (canister surface and 1-metre away from canister). The capsules were subjected to short-term exposure (30 minutes for background measurement and 5 minutes for canister measurement) with due care to prevent contaminating the capsules in the process. The dosimeters were then processed to estimate the integral exposure recorded. In arriving at the dose rate from the integral dose, it is assumed that the radiation dose registered by the dosimeters during their course of travel from the survey port to the point of reference and back (5 seconds each) is insignificant compared to their residence time at the point of reference and hence ignored.
Results and Discussion: The values obtained from measurements on two of the eight canisters are presented in [Table 1].{Table 52}
The average normalised surface dose rate on canister per unit gamma activity for the eight canisters is estimated to be 2.9 ± 0.5 mSv/h-Ci. The values for the two reference canisters are presented in [Table 2]. The dose rates measured using two different passive dosimetry techniques are in close agreement and correspond with the radioactivity content of the canisters. The study demonstrates that dose rates of canister can be measured with reasonable precision using passive dosimetry.{Table 53}
Keywords: Dose rate, high-level liquid waste, vitrification, vitrified waste product canister
References
IAEA. Design & Operation of High-Level Waste Vitrification & Storage Facilities. Technical Report Series No. 339. IAEA; 1992.Helger, et al. Principles of Product Quality Control of German Radioactive Waste Forms: Vitrification, Compaction & Numerical Simulation, WM2012 Conference. Phoenix, AZ; 2012.
Abstract - 24604: Consideration of a new technology for the large scale clean up of Fukushima wastes
B. M. Cassels
Department of Engineering, University of Melbourne, Melbourne, Australia
E-mail: [email protected]
The current approach to disposal of radioactive waste essentially combines two major elements: immobilisation/conditioning of the wastes, followed by emplacement of immobilised wastes into an environment that further prevents radionuclide migration. Geological properties of the surrounding ground have been important limiting factors for such waste isolation due to the long-term breakdown of encapsulation material. The containment properties of the immobilised waste package combined with the environmental performance of the surrounding natural geological strata need to meet the overall selected dose constraint for disposal of the wastes. Under the current approach, finding a site with the right geological properties to further inhibit radionuclide migration is a limiting factor. It is also politically difficult to site radioactive waste disposal facilities when suitable sites are identified potentially leading to selection of a sub-optimum site requiring increased technological enhancement such as engineered barriers to meet performance requirements. Permanent isolation to prevent radionuclide movement consists of two parts: 1. Physical immobilisation of the wastes, and 2. Placement of immobilized wastes into an environment that further inhibits long-term radionuclide migration. Separating these components by using a technology that permanently immobilises everything it contains and does not require further specific environmental isolation of the encapsulated wastes would negate the requirement to find specific geological areas that inhibit radionuclide migration. Many waste companies around the world have been considering synthetic encapsulation for decades e.g. Synroc in Australia. Some companies have also been working on the use of polymers to provide simpler cost- effective options for encapsulation. An Australian company has successfully developed a polymer that achieves the objective of permanently encapsulating radioactive materials. It is this technology that can provide the solution for the clean-up of large volumes of radioactively contaminated soils. Australian company Hazprotect has created an innovative, proprietary composite encapsulant suitable for exactly this purpose. Their patented[2] composite polymer binds with hazardous wastes to create a stable long-term waste immobilization solution. The encapsulant material is a combination of LDPE polymer and other additives to create a hydrophobic compound that eliminates water absorption essentially eliminating polymer degradation and providing near-zero leaching. Laboratory tests exceed US DOE leachate, structural integrity and radiation degradation requirements. This technology effectively removes the requirement to provide further geological isolation of immobilised wastes, allowing them to be safely transported and disposed to a standard landfill facility or for permanent totally safe placement anywhere ensuring that hazardous chemicals will not leach into the environment. Considering the waste immobilisation requirements for the millions of bags of contaminated soils wastes from the Fukushima-Daiichi Accident it is contended that the use of this new technology would provide improved immobilisation outcomes and subsequently offer the flexibility of time to consider eventual placement of these wastes.{Figure 79}
Keywords: Disposal, new technology, radioactive waste
References
Cassels BM. The Determination of Three Selected sites for the Construction and Operation of a Low Level Radioactive Waste Storage and Burial Facility. Available from: https://eprints.qut.edu.au/37051/.Patents: EU, Australia, Japan, India, Singapore, China, USA, Canada, Israel, Saudi Arabia, UAE, Russia.
Abstract - 24619: Engineering application and research progress of low-level radioactive waste incineration technology in China
Liu Qun, Zheng Bowen, Chu Haoran, Chang Sicheng
China Institute for Radiation Protection, Taiyuan, China
E-mail: [email protected]
China Institute for Radiation Protection had independently developed a multi-purpose radioactive waste pyrolysis incineration technology and built 3 incineration facilities in China, which were mainly used to treat solid waste and waste oil from nuclear facilities such as NPPs. The composition of solid waste included paper, cloth, plastic, rubber, etc. In order to solve the problems in the early operation of incineration facilities, targeted improvements had been made in the aspects of equipment anti-corrosion capability, secondary waste generation and system safety. The improved facilities had treated a large amount of low-level waste and operated for more than 15 years. The stability and reliability of the incineration system were verified, and the advancement of pyrolysis incineration technology was proved. Considering the current situation that the proportion of plastics in low-level waste is increasing, the process was optimized so that more plastics and resin can be incinerated. The optimized incineration system still showed good adaptability when the proportion of plastic in the waste composition over 60%. The research on miniaturization and mobile technology of incineration were continuously carried out for the small reactors or small nuclear facilities, so as to further improve the economy. Compared with the incineration facilities, the mobile incineration technology can reduce the floor space by more than 90% and cut the construction cost by more than 75% under the same capacity and meeting the emission requirements.
Typical waste types tested and verified are
75% fabric + 10% PE (polyethylene) + 5% PVC (polyvinyl chloride) + 10% rubber;40% fabric + 25% PE (polyethylene) + 15% PVC (polyvinyl chloride) + 10% rubber + 10% resin (dry).
The alkane gas produced by pyrolysis is fully mixed with the heated air and then enters the combustion furnace at 850°C for complete combustion. The premixed combustion method can completely burn the pyrolysis gas, and the off gas after combustion can meet the emission standard after being cleaned. [Table 1] shows the content of off gas compounds at the outlet of the high-efficiency filter of the incineration facility. The test method refers to GB-18484.{Table 54}
Keywords: Engineering application, incineration, pyrolysis, radioactive waste
References
Bowen Z, Hong Z, Jiasheng R, et al. Experimental study on explosion venting simulation of radioactive waste incinerator. Radiat Prot 2016;36:200-5+217.Bo Z, Weihaoran X, Xiaowen L, Liguo Y, Zhang, Lili Y, et al. Design and verification of a simple low-level solid waste incineration process device. Radiat Prot 2020;40:365-37.Kiss G, Marfiewicz W, Riegel J, et al. Thermoselect recovery of energy and raw materials from waste. In: Schweitzer FJ, editor. The Thermoselect Process for the Degasification and Gasification of Wastes [M]. Berlin: EF-Verlag; 1994.Chateauxvieux H, Guiberteau P, Longuet T, et al. The IRIS Incinerator at the CEA Valduc Research Center: Assessment after One Year of Active Incineration [C]. Proceedings of the International Conference TopSeal'99, October 10-14, Antwerp, Belgium; 1999.Xiaohai L, Lianquan Z, Peiyi W, et al. Establishment and commissioning of combustible radioactive waste incineration device. Radiat Prot 2004;24:46-50.China Environmental Monitoring Station, University of Science and Technology of China. Pollution Control Standard for Hazardous Waste Incineration: GB 18484-2020[S]. Beijing: China Standard Publishing House; 2020.Bowen Z, Wei X, Lili Y, et al. Design improvement and verification of flue gas purification system of low-level waste incineration plant. Radiat Prot 2014;34:206-11.
Abstract - 24625: Method for establishing reference biosphere of Beishan high-level pre-selected site and public radiation dose evaluation
Bing Lian, Yan Wang, Jie Yang, Yang Jun Zhao
Department of Nuclear Environmental Science, China Institute for Radiation Protection, Taiyuan, China
E-mail: [email protected]
The end point of the safety assessment for high-level radioactive waste geological disposal facility is the biosphere assessment. Its main purpose is to study the migration and transformation process of radionuclides released from the repository in various media of the biosphere (such as water, soil, animals and plants, etc.), and to estimate the dose produced to humans through various exposure routes. Because the time scale of the assessment is very long (usually more than ten thousand years), the living conditions, eating habits and the ecosystems that interact with them in the future of human society will have great changes, and the uncertainty of the assessment scenario will lead to greater uncertainty in the assessment results of long-term radiation dose. Therefore, the selection of the assessment scenario needs to describe and analyze the selected biosphere. In order to accurately and rationally predict the migration process of radionuclides in the environment of high-level radioactive waste repository, this study is conducted in the Beishan site of China's geological repository with the environmental characteristics survey of this site, the reference biosphere of Beishan site is established. Beishan pre-selected site is located in Jiuquan area in the west of Gansu Province, which is deep in the hinterland, windy, unusually dry, with large evaporation, rainy in summer and dry in winter. There is no perennial surface water system in Beishan area, where the ditches are caused by seasonal floods. The groundwater is mainly supplied by precipitation infiltration. According to the geological and hydrological survey report of Beishan area, the groundwater in this area flows from west to east and finally discharges to the Heihe River basin. Heihe River originates from Qilian Mountain in Qinghai Province and is the largest inland river in Hexi Corridor. The main river channel is 821 km long, the drainage area is 130,000 km2, and the annual runoff is 3.81 billion m3. The population around the river basin is about 60000, mainly involved in farming and animal husbandry, which means the groundwater in Beishan area is used for agricultural irrigation and drinking. [Table 1] shows the amount of main long-lived radionuclides in high level radioactive vitrified waste in a single waste package, which contains Se-79, Zr-93, TC-99, Pd-107, Cs-135, Sn-126, Np-237, U-236, U-238. The software Ecolego developed by Facilia AB, Sweden, was used to evaluate the assessment of the radioactive waste disposal system. When simulating the waste disposal facility, Ecolego divided the disposal repository system into a series of compartments, and assumed that each compartments has a certain boundary space. Once the nuclides enter the repository compartment, they will be mixed immediately, so that the pollutant concentration in the entire repository room will be mixed evenly. The migration ratio is used to represent the migration process. Migration ratio is the ratio of the activity of a certain nuclide lost or gained by migration to the total activity of the nuclide in the chamber at that time. The migration of geosphere mainly considers the radionuclides in the fractured zone and surrounding rock. The dominant water fissure migration and the dominant water fissure nuclide were calculated. It is assumed that after the closure of the repository, the radionuclides in Beishan repository will release from the unsaturated zone into the aquifer, migrate with the groundwater, and finally discharge to the Heihe River basin. [Figure 1] shows the rate of nuclides in crevice water. In the evaluation, internal exposure caused by drinking Heihe water, ingesting agricultural products irrigated by Heihe water, and ingesting poultry and livestock products fed by crops irrigated by Heihe water was considered. The evaluation model and parameters refer to the 30-year quality evaluation report of China's nuclear industry. The maximum individual dose and occurrence time of each age group caused by each nuclide are shown in [Figure 2]. The assessment results show that about 1.5 million years after the closure of high-level waste repository, the dose caused by the public is the largest, about 10-11 Sv/a, and the key nuclide is Cs-135. The assessment results show that the radiation dose to the public after the closure of Beishan repository is very small.{Table 55}{Figure 80}{Figure 81}
Keywords: Beishan, radiation dose, reference biosphere
References
Ishihar Y, Makino H, Ohi T. Inventory Evaluation of High Level Radioactive Vitrified Waste. JNC TN8400 99-085. 1999.Stockholm Facilia AB. Ecolego3 User Guide; 2008.Avila R, Broed R, Pereira A. Ecolego-a toolbox for radioecological risk assessment [C]. Int Conf Prot Environ Eff Ionizing Radiat 2003.Pan Z. Anthology of Radiation Environmental Quality Evaluation of China's Nuclear Industry in the Past 30 Years. Atomic Energy Press [M]; p. 47-61.
Abstract - 25261: Decommissioning of old active buildings in a monazite processing facility
S. Ajesh Kumar, Sujata Radhakrishnan, S. K. Jha
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Rare Earths Division of IREL (India) Limited was set up in the year 1952 at Udyogamandal, Kerala for processing Monazite to produce composite rare earth chloride, trisodium phosphate, a variety of rare earth compounds, and thorium concentrate. Thorium oxalate produced was stored in closed RCC structures over the ground (Silos). Due to safety concerns and proximity to river Periyar, Thorium oxalate from Silo1-3 was shifted to other silos and taken up for decommissioning. The dominant radionuclides contributing to radiation hazards in the building were Th228, Th232, Ra228, Ra224, and Rn220 and its progeny. The assessment of radioactivity in Silos was done by radiation field mapping, in-situ contamination monitoring, and air activity measurements. Samples from walls and floor were collected from different depths at selected locations after chipping these locations to ascertain the extent of impregnation of radionuclides in the concrete matrix. It was found that activity extended up to 30 mm thickness of the walls and up to 40 mm on the floor of silos. Hence wall surface & columns were chipped up to 30 mm thickness and floor finish concrete was chipped up to 40 mm thickness to remove fixed contamination after initial loose contamination removal. The variation of contamination levels along the height from floor was also studied. Maximum Surface contamination was observed up to 3 m from the floor of the building, and this was decontaminated first. The airborne contamination inside the building was negligible as the plant has sufficient ventilation and was not handling any active material for long time. Floor contamination was 2-5 Bq.cm-2 in Silo 1, 3-8 Bq.cm-2 in Silo 2 and 3-10 Bq.cm-2 in Silo 3 respectively in the beginning of decontamination and demolition activities and it was brought to 0-0.7 Bq.cm-2 after the decontamination of the floors. [Table 1] shows the comparison of radiation field before and after decontamination of Silos. All decontamination and demolition works were carried out under radiological work permit system where the exposure of individual is controlled by restricting the duration of work by assessing the radiological conditions. All works were carried out 4 hours per day per individual for slurry recovery works and 6 hours for wall and floor decontamination works. A total of 93 radiological safety work permits were issued. Radiation dose data for the contract workers engaged in decontamination and demolition works are given in [Table 2]. The effective dose incurred in the work was 0.19 Person mSv per ton of thorium Oxalate retrieved and is very small fraction of the institutional dose during regular thorium operations. Effective dose incurred during the demolition work was less than 5% of the total dose as most of the operations were carried out from a distance using machines like hydra with extendable grabbers and the radiation levels were reduced by removing out layers first. Waste generated in decontamination works are mainly from the active sludge recovered from Silo no.3, active concrete waste from wall and floor chippings and general non-active wastes from dismantling of mechanical and concrete structures. Active wastes were disposed in RCC trenches at disposal yard and non-active wastes were used as backfills in waste disposal yard. All the active washings generated during the cleaning of floors, beams, walls, and columns were collected separately, activity content measured and pumped to ETP for further treatment. A total of 170.43 tons of waste having an activity of 201.08 GBq was generated and disposed in RCC trench. The entire decontamination and demolition work was carried out in a radiologically and industrially safe manner adhering to all regulatory practises and controls. The decommissioning of large Thorium Slios is being done for the first time and the experience gained would be helpful in further optimising the man-rem exposure and reducing the radioactive waste in similar facilities.{Table 56}{Table 57}
Keywords: Decommissioning, decontamination, demolition
Reference
Pillai PM, Maniayan CG, Paul AC. Health Physics Experience on the Decommissioning of the Rare Earth Plant at the Indian Rare Earths Limited, BARC Report; 1998.
Abstract - 25304: A study of neutron activation of steel alloys for nuclear reactor decommissioning
Chitra Subramanian, Riya Dey, Arti S. Mhatre, A. K. Deepa, Tanmay Sarkar, Kapil Deo Singh, M. S. Kulkarni
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Steel alloys such as stainless steel and carbon steel are widely used in manufacturing PHWR components like calandria vessel, end shield, end fitting, feeder pipes, primary heat transport system pipelines, fuelling machine head, moderator system components etc. During reactor operation, in the presence of neutron flux, elements present in steel alloys get activated and various short-lived and long-lived radionuclides are formed. The present paper focuses on the study of neutron activation of various steel alloys for radionuclide characterization and their impact on decommissioning planning. Activation products formed in steel alloys is computed using the ORIGEN2 code.[1] Inputs required include nuclear reactor type, elemental compositions of the material, mass, neutron flux, period of irradiation and cooling period after reactor shutdown. Sample of a cut un-irradiated SS304L [Figure 1] collected from an Indian NPP site was analysed to obtain the elemental composition and these values were then compared with the standard composition specified in ASTM [Figure 2]. Build-up and decay of radioactivity due to neutron activation in one kg of three steel alloys (SS304L, SS403 and carbon steel) is computed and studied.{Figure 82}{Figure 83}
Good agreement is seen between the measured value of elemental composition and ASTM specification [Figure 2] especially for Co and Ni (major contributors to long-term radioactivity). Major radionuclides contributing to radioactivity in one kg of SS304L are shown in [Figure 3]. Build-up and decay of total radioactivity in the three steel alloys (1kg) is shown in [Figure 4]. In case of SS304L, reduction in radioactivity over the years is not very sharp because of substantial contribution from long lived radionuclides (63Ni, 59Ni and 14C). Ni content is highest for SS304L. In carbon steel the total radioactivity is mainly determined by 60Co both in the initial and later stage, hence reduction is substantial. An interesting observation is seen regarding the radioactivity of 14C. Carbon content in carbon steel (0.2%) and SS403 (0.15%) is higher than that of SS304L (0.03%). But higher amount of 14C is formed in SS304L because, the major contribution to formation of 14C is from (n, p) reaction with 14N. Nitrogen content in SS304L is 0.1% and not specified for carbon steel and SS403 (taken as zero). As an application of the study, total radioactivity generated in calandria shell of a typical 220 MWe PHWR was estimated as 6.85 x 102 TBq after a cooling of 50 y. Study of neutron activation of steel alloys has been carried out successfully. It is seen that a cooling period of 50 years after reactor shutdown results in substantial reduction of radioactivity. Dismantling activities can be taken up there after. Thus the study helps in planning and implementation of decommissioning strategies for nuclear power plants.{Figure 84}{Figure 85}
Keywords: Activation products, decommissioning, radionuclide characterisation, steel alloys
Reference
Croff AG. ORIGEN2: A Versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials. Nucl Technol 1983;62:335-52.
Abstract - 25503: Inhalation hazard of radioactive aerosol generated from dismantling operation of in-core structural components during decommissioning of PHWR
Tanmay Sarkar1,2, S. Chitra 1, S. Anand1,2, Kapil Deo Singh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Decommissioning of nuclear power plants takes the facility out of operation with adequate protection of the worker, the public, and the environment. Decontamination, dismantling, demolition of structures, and processing and disposal of waste are steps in decommissioning. India has adopted a deferred decommissioning option for NPPs. After final shutdown and safe enclosure period (~50y) is over a significant reduction in external dose rate is observed from the active components which facilitate the dismantling process. In the dismantling process, all equipment in the reactor building is removed and involves cutting of various active components leading to the release of airborne activity hence inhalation hazard. Coolant channel assemblies are the first metallic components which are to be removed from the Calandria vault of PHWR. At the end, civil structures like concrete or other steel components will be dismantled. In the present study, a preliminary study is carried out to calculate the inhalation dose from cutting activity of pressure tube.
Materials and Methods: There are various cutting methods such as mechanical, laser cutting, oxy-acetylene cutting, plasma cutting, etc. The activity generated in a pressure tube (Zr-2.5%Nb alloy) after 20 years of irradiation and 50 years of decay is calculated using the ORIGEN2 code.[1] The external dose rate at various distances is calculated using Monte Carlo method. The plasma cutting technique is the best choice of a cutting method generating the least aerosol and internal radiation hazard to workers.[4] The major radionuclides which can become airborne during the cutting process and their content in a single pressure tube are given in [Table 1]. The aerosol concentration for each radionuclide during the cutting process is calculated using Eq.1.[3] Typical values of parameters chosen for calculation are given in bracket.{Table 58}
[INLINE:19]
C̅i is air activity concentration for ith radionuclide, zx is kerf width (0.35 cm), l is kerf length (33.8 cm), Ai is surface activity, f is released-respirable mass fraction (0.019), t is cutting time (11 s), Qv air flow from room to outside [INSIDE:8]. The inhalation dose to the worker is calculated using Eq.2.
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BR is breathing rate, tinh is time during which worker breaths from radioactive cloud (8 hrs), PF is protection factor for respirator used by worker (0.01), DCFi,inh is committed effective dose co-efficient for ith radionuclide (GSR Part 3, 2014), C̅i is taken from Equation 1. It is assumed that pressure tube is cut in two locations.
Results and Discussion: [Table 2] gives comparison of total collective dose for 8 hours cutting duration. It is observed that internal dose contributes ~1% of total dose. The present study gives methodology to assess internal dose during dismantling operation of an active structural component of PHWR.{Table 59}
Keywords: Aerosol, decommissioning, inhalation
References
Croff AG. ORIGEN2: A versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials. Nucl Technol 1983;62:335-52.General Safety Requirements (GSR) Part 3. Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. Vienna: International Atomic Energy Agency; 2014.Simonis A, et al. Modelling of radiation doses during dismantling of RBMK-1500 reactor emergency core cooling system large diameter pipes. Ann Nucl Energy 2015;85:159-65.Chae N, Lee MH, Choi S, Park BG, Song JS. Aerodynamic diameter and radioactivity distributions of radioactive aerosols from activated metals cutting for nuclear power plant decommissioning. J Hazard Mater 2019;369:727-45.
Abstract - 25508: Release of the decommissioned reactor building for restricted use
K. S. Babu, M. Swarnkar1, Lalit Vajpyee1, Saurav1, Ranjit Sharma1, Rakesh Ranjan1
Health Physics Division, Bhabha Atomic Research Centre, 1Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
The research reactor Apsara was permanently shut down in the year 2009. Subsequently reactor fuel was transferred to a storage facility, auxillary systems were dismantled and pool water was disposed as part of decommissioning. The pool structure was not dismantled. It is desired to release part of the reactor building from regulatory control for general purpose use. The sites and buildings which are once used for various practices involving radioactivity can be released from regulatory control if it can be established that the radiological status has been restored to the acceptable levels. The release criteria is based on the prospective dose to occupants of the released sites or buildings. The prospective dose is believed to be due to residual radioactivity in the form of external radiation exposure, elevated airborne activity and ingestion of activity due to surface contamination. IAEA has recommended 1 μSv/year as the dose criteria for unrestricted release of sites and buildings. A radiological status survey was carried out in Apsara reactor building to evaluate the conditions with an aim to release the building partially for general purpose use.
Methods: The reactor building houses the reactor pool structure, water treatment area, laboratories and office rooms. The rooms and area outside the reactor pool are planned to be released for general purpose use. The status of these areas was evaluated and compared with surrounding areas. Surface contamination levels were measured. Sources of airborne activity inside the reactor pool were studied. Comparison of the radiation levels inside the reactor building with surrounding environment was done using sensitive environmental TLD measurements. TLD measurements were taken in all the rooms and accessible areas of the building. The non-impacted reference areas around the reactor in all directions were also covered with TLDs for a period of 2 months. The range and distribution of radiation levels inside and outside the reactor building were compared. The data shows that there is no difference between the radiation levels at accessible areas inside the reactor building and surrounding environmental radiation levels outside the reactor building. The surface contamination was below detection levels and there are no sources of airborne activity in the accessible areas.
Results:
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Conclusion: The radiological status of the reactor building apart from the pool structure, office rooms and laboratories shows that there is no difference in radiation levels between surrounding environment and the proposed areas of release. The office rooms and other inactive areas can be released from regulatory control and utilized for other purposes. The measured radiation levels inside and outside the reactor building are in the range of general environmental radiation levels at the site.
Keywords: Decommissioning, remediation, site release, status survey
Abstract - 25510: Radiological assessment during decommissioning of reactor pool structure
Saurav, K. S. Babu, Lalit Vajpyee, Lokesh Kumar, Ranjit Sharma, Rakesh Ranjan1
Health Physics Division, Bhabha Atomic Research Centre, 1Reactor Operations Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
The research reactor Apsara was permanently shut down in the year 2009. The reactor core and other removable active material were removed and sent to storage and waste management sites. The reactor pool is 18' deep concrete trench which also acts as biological shield. The reactor pool is provided with steel lining. The reactor core was positioned at 1 m elevation from pool bottom. There are horizontal beam holes and a thermal column provided for experiments. During the operation of reactor, the beam hole openings, pool wall lining and the wall concrete developed induced radioactivity (mainly 60Co due to impurities in structural components) due to neutron flux irradiation. The location of induced activity, measurement of radiation levels and hot spots was essential for decommissioning planning. The detailed radiation mapping was planned for decision making on the method of decommissioning, decontamination and waste management.[1] The pool wall is 8 m x 8 m on control room side and graphite storage shed side. The wall has dimensions 3 m x 8 m on shutter side and thermal column side. The wall is provided with SS lining. Beam hole openings extend into the pool through the wall. Radiation mapping on the wall was conducted using high range radiation survey meter. The mapping was done on a grid of 1m x 1 m area covering the whole wall surface. It was observed that the radiation levels drastically reduce after an elevation of 3 m from pool bottom. The upper elevations of the pool structure might be containing low levels of radioactivity. In order to determine the actual radiation levels due to induced activity at the higher elevations of the pool wall, the contribution from higher active material from bottom elevations was required to be shielded. This was achieved by using a partially shielded detector which is exposed only on the side facing the wall and remaining area covered by 2“thick lead shield. The detector response is due to locally induced activity on the wall. The detector response with and without the partial shield was compared along the wall at all elevations. Similar measurements using a partially shielded detector were used to estimate the extent of induced activity along the thermal column MS lining plates. The maximum radiation level measured on the pool walls at bottom elevation was 0.08 mGy/h and at higher elevation radiation level was 1.50 – 2.40 μGy/h. The radiation level at pool top was found to be 0.05 μGy/h and measured using partially shielded detector. The fixed contamination levels on the inner section of MS lining of the thermal column is 2000-3000Bq/cm2 and on the outer section it is 10-15 Bq/cm2 measured by a shielded detector. The measurement in [Table 1] shows that the induced activity is higher at the bottom elevations of the pool, close to reactor core position. At higher elevations, far from core locations, the levels of radiation due to induced activity is low. The measurements on the thermal column shows that the inner sections of the thermal column are highly active due to higher neutron exposure whereas the outer section of the thermal column MS lining has less radioactivity. The data would help in planning for appropriate remediation action and extent of surface area to be shielded.[2]{Table 60}
References
Abelquist EW. Decommissioning Health Physics: A Handbook for MARSSIM Users. 2013.Joerg K, Boris B. Radiation Protection during Decommissioning of Nuclear Facilities: Experiences and Challenges.
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