Chemicals are useful and indispensable in every walk of our life. Nuclear industry is no exception to this fact. A number of chemicals are in use in the nuclear fuel cycle. Thus, hazards due to exposure to chemicals coexist with radiological hazards in a nuclear fuel cycle industrial or research set-up. In the realms of control of chemical exposure, limits on concentration of chemicals in workplace are prescribed. These are known as occupational exposure limits. Threshold Limit Value (TLV) is one such occupational exposure limit. American Conference of Governmental Industrial Hygienists (ACGIH) prescribes TLVs as guideline values for various chemicals and also for physical agents. The discussion is confined to philosophy of chemical TLV, its applicability, and its limitations.

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The origin of the thorium deposits on the West Coast of India is not well established though a number of studies on the various aspects of these deposits have been carried out for more than 50 years. The deposits are believed to be washed down from highlands through water streams. It is proposed that atom ratios of various radioactive and stable products in the uranium and thorium series as a function of time could be effectively used for the purpose. Accordingly, the various ratios have been calculated as a function of the age in the time range of 100-70,000 years. The stable products are included for the first time.Pb-208/Th-232 in Th-232 series, Pb-207/U-235, Pb-207/Pa-231, and in U-235 series and Pb-206/U-238, Pb-206/U-234, Pb-206/Th-230, and Pb206/Ra226 in U-238 series, are presented. The ratios show significant variations in the time range studied. The ratio Pb-206/Pb-207 from the two uranium series is of interest and is examined. This ratio shows a variation of up to 100 times whereas, the natural isotopic ratio is around unity. It is suggested that an extensive measurement of these ratios and their correlation in samples from the hypothesized source in highlands, in river beds, and on the beach sand might give some idea of the transport path for these minerals. Further, it is suggested to measure the ratio of all the lead isotopes from the three series with Pb-204 which is purely of primordial in origin which would also be useful to indicate the contribution from radiogenic sources. In order to undertake such a project which, of course, involves many disciplines large efforts in sampling, in modeling of possible transport paths and strategies for analysis are required.

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Radiation dosimetric methods are used for the estimation of dose absorbed by radiation in a detector material. These methods are required for estimation of absorbed dose in various applications of radiation, such as personnel and environmental dosimetry, retrospective/ accident dosimetry and medical applications of radiation. The use of thermoluminescence (TL) as a method for radiation dosimetry of ionizing radiation has been established for many decades and has found many useful applications in various fields, such as personnel and environmental monitoring, medical dosimetry, archaeological and geological dating, space dosimetry. Several high sensitivity TL phosphor materials and thermoluminescent dosimeters (TLDs) are now commercially available in different physical forms. There are many commercial TLD systems which are being used for various dosimetric applications and even presently, TL is a major player in the field of radiation dosimetry, particularly in personnel dosimetry. In the last two decades an alternative technique, optically stimulated luminescence (OSL), has been developed, as the optical nature of the readout process does not involve problems of blackbody radiation and thermal quenching. Due to this and some other advantages OSL is also being used for various applications in radiation dosimetry, such as personnel and environmental dosimetry, retrospective/ accident dosimetry and medical dosimetry. The development of Al ₂ O ₃ :C TL/OSL phosphor by Akselrod et al. and later investigation of its suitability for personnel dosimetry using pulsed OSL (POSL) technique of stimulation by Akselrod and McKeever, resulted in the development of a personnel dosimetry system based on Al ₂ O ₃ :C OSL phosphor. Therefore, thrust of modern luminescence dosimetry development is more towards OSL. The main advantages of the small size optic fiber based OSL dosimeter over the currently available radiation detectors, such as TLD, used in clinical applications, are the capabilities of measuring both real-time dose rate (using radioluminescence, RL) and absorbed dose (using OSL). Although radiophotoluminscence (RPL) dosimeters were developed in parallel with TLD systems during 1960s, but high pre-dose and photon energy dependent detector material prevented major breakthrough of the glass dosimetry. Therefore, RPL glass dosimeters were used as an emergency dosimeter in accident situations. However, in mid 1980s introduction readout systems using a pulsed UV stimulation, in place of conventional mercury UV lamps, helped in reducing pre-dose by a factor of 100 (from mSv to a few μSv). Use of pulsed stimulation permits electronic discrimination of the signals from the pre-dose and absorbed dose on account of their different fluorescence decay times. This development resulted in the manufacture of improved RPL glass dosimeters and fully automatic RPL reader systems capable of measuring doses in the range 10μSv to 10Sv. In 2001, silver activated phosphate RPL glass dosimetry system has been introduced as the major personnel monitoring service in Japan marketed by Chiyoda Technol Corporation. Some of these developments in the field of TL, OSL and RPL dosimetry are reviewed.

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Carotenoids are the most common naturally occurring bioactive terpenoid pigments, which are commonly produced by a wide variety of plants and microbes. The present study was aimed to evaluate the antioxidant, antimicrobial and radio-protective properties of carotenoid pigments isolated from ultraviolet (UV)-C resistant Micrococcus spp. The UV-C resistant Micrococcus roseus and Micrococcus luteus were isolated from the soil samples of Savandurga hills region, Karnataka (India), and their pigments were identified as carotenoids based on spectral analysis. The UV-protective efficacies were determined by cling-film assay. Further, the antioxidant activities of pigments were evaluated by 2,2-diphenyl-1-picrylhydrazyl assay, and antibacterial activities by disc diffusion and broth microdilution assays. The optimum growth and pigment production by M. roseus and M. luteus were observed at temperature ranged between 35°C and 37°C, pH 7.0-8.0, NaCl 5.0-7.0%, and sucrose as major carbon and KNO ₃ as major nitrogen sources. In the present investigation, the isolated carotenoid pigments of M. roseus and M. luteus showed significant UV protective activity along with antioxidant (IC ₅₀ 3.5-4.5 mg/mL) and antibacterial (minimal inhibitory concentration 0.25-2.0 mg/mL) properties.

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The impact of gamma irradiation on Terminalia arjuna (T. arjuna), one of the potent medicinal plants for cardiac disease is described in this article. The seeds of T. arjuna were irradiated with different doses of gamma radiation ranging from 0 to 200 Gy using the ⁶⁰ Co source. The effect of gamma radiation on the growth and biochemical constituents were compared with the control plants. Germination speed at 25 Gy was found to be 0.65, which was double compared to the un-irradiated seeds. An increase in germination percentage, vigor index, and relative growth rate, in terms of dry weight was noticed at lower doses of the radiation treatment. The proline content increased with increasing doses. The chlorophyll content was found to have increased to 12.2 mg/g FW at 100 Gy compared to the control level of 8.44 mg/g FW. Increased phenolic content and radical scavenging capacity was observed at 25 and 150 Gy. Hence, lower doses of radiation treatment may be used to increase the germination, growth, and vigor, and also the enhancement of plant metabolites like proline and phenolics in T. arjuna.

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Medical cyclotrons are now constructed as turnkey facilities at nuclear medicine clinics, specialised particle therapy facilities and radioisotope production centres. Most medical cyclotrons usually accelerate protons to high energies and could be divided mainly in two categories: (a) Low energy (E _P = 15-30 MeV) machines, dedicated for medical positron emission tomography and single photon emission computer tomography radioisotope production and (b) High energy (E _P = 100-250 MeV) machines, predominantly used for radiotherapy of malignant tumours. Parasitic gamma and neutron radiation are produced during the operation of medical cyclotrons. Furthermore, high level of gamma radiation produced by the activated cyclotron components could impose radiation exposure to maintenance crew. Hence, radiation safety is imperative to safe and reliable operation of medical cyclotron facilities. A sound operational health physics procedure assures the minimisation of radiation exposure to patients and members of the public abiding the regulatory guidelines. This paper highlights the important radiation safety aspects related to safe operation of proton therapy and radioisotope production medical cyclotrons.

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In general, lead is used as shielding material for protection against radiation. In spite of its high density, lead is toxic and lead aprons are very heavy for personal shielding. Thus, there is a need for nontoxic, light, and environmental friendly radiation-shielding materials. Polymers cannot be effective against gamma radiation on their own. High-density metal wolfram carbide could be useful against gamma radiation, and boron carbide could also be useful for neutron shielding. In this study, high-density polyethylene, boron carbide, and wolfram carbide can be mixed in certain amounts and composite discs can be obtained in this way. According to results, a new shielding material is efficient for gamma radiation.

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Background: Two isotopes, ²²² Rn and ²²⁰ Rn of the natural radioactive gas radon are generally of radiological importance. About half of the total radiation effective dose to the general public is due to the irradiation of the lungs by alpha particles following the inhalation of radon decay products. In view of this, ²²² Rn and ²²⁰ Rn measurements were carried out for a period of 1 year in about 50 dwellings belonging to different residential areas of Bangalore city. The results obtained were systematically analysed and discussed. Materials and Methods: ²²² Rn and ²²⁰ Rn measurements were carried using Solid State Nuclear Track Detectors. The gamma exposure rate measurements were also carried out in all the dwellings using a G M tube based Environmental Radiation Dosimeter. Results: The concentration of ²²² Rn and ²²⁰ Rn was found to be highest in the dwellings having mud flooring (Bare) and lowest in the dwellings having mosaic flooring. The geometric mean values of the concentrations of ²²² Rn and ²²⁰ Rn were found to be 24.1 ± 8.3 and 24.5 ± 10.8 Bq m ^-3 respectively. The annual mean value of the gamma absorbed dose rate is found to be 175.8 nGy y ^-1 . Conclusion: The mean value of radon concentration is well within the indian average (42 Bq m ^-3 ) and world average (40 Bq m ^-3 ) values. The mean effective dose was found to be 1.19 mSv y ^-1 and is well within the action level as prescribed by ICRP-65.

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Measurement and verification of strength of monomodal high-dose rate (mHDR)¹⁹² Ir source supplied by the vendor is a major part of quality assurance program. Reference air kerma rate (RAKR) or air kerma strength (AKS) is the recommended quantity to specify the strength of gamma emitting brachytherapy sources. Physicist in our institution performed the source calibration as soon as each ¹⁹² Ir new source was loaded on the mHDR afterloading machine. The AKS accurately measured using a physikalisch technische werkstatten (PTW) re-entrant chamber-electrometer system in a scatter-free geometry was used to compute the air kerma rate (AKR) at one-meter distance in the air. To ensure accurate dose delivery to brachytherapy patients, measured AKS or RAKR should be entered correctly in both HDR treatment console station (TCS) as well as treatment planning system (TPS) associated with it. The clinical outcome mainly depends not only on the accuracy of the source strength measurement in the hospital but also on the correct source strength entered into both TCS and TPS software. A retrospective study on 22 mHDR V2 sources supplied by the vendor for the period of 10 years was taken up to access the accuracy of source strength supplied to the Radiotherapy department. The results are analyzed and reported. The accuracy in measured RAKR of all 22 sources supplied by vendor was well within the tolerance limits set by the national regulatory body and international recommendations. The deviations observed between measured RAKR versus manufacturer's quoted RAKR were in the range from −1.71% to +1.15%. In conclusion, the measured RAKR have good agreement with vendor quoted RAKR values.

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A prototype two element eye lens dosimeter badge based on highly sensitive α-Al ₂ O ₃ :C optically stimulated luminescence dosimeters (OSLDs) was designed and developed. The badge consists of a plastic card in which two thin α-Al ₂ O ₃ :C optically stimulated luminescence (OSL) discs are placed. The OSLDs in the plastic card (when inserted into plastic cassette) are covered with energy compensation filters made up of 0.3 mm thick Cu and 1.35 mm thick Teflon discs positioned symmetrically on both sides of the dosimeter. The OSLD badge is useful for monitoring doses from photons and beta particles. In this paper, theoretical studies using Monte Carlo method as well as using the analytical technique have been performed to study the energy response of the bare α-Al ₂ O ₃ :C based disc dosimeters. These dosimeter discs have been found to exhibit over-response by a factor of ~ 3.4 at ~ 33 keV photon energy, whereas, beyond 80 keV photon energy, the response is nearly energy independent. Studies have also been performed to find the energy response of the α-Al ₂ O ₃ :C disc dosimeters under different metal filters, viz., Al, Cu, Sn etc., and under various thicknesses of Teflon. From theoretical simulations, it has been found that 0.3 mm thick Cu is sufficient to correct the over-response in lower energy region within acceptable limits. Further, Teflon disc (DuPont, USA) having thickness of 1.35 mm is found to be the optimized choice as filter for the second dosimeter disc. It is worth mentioning that the ratio of the response of the OSL disc under Teflon to that under Cu filter indicates average energy of X-ray photons and same is used to correct the over-response as well as to estimate the quantity H _p (3). Also for higher photon energy region, the readout of the dosimeter disc under Teflon filter directly measures the quantity H _p (3). Same holds true for beta particles having maximum beta energy, E_max beyond 0.7 MeV.

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The National Disaster Management Authority (NDMA), ever since it's formation as an apex policy making body for the country in the field of disaster management, has formulated a set of guidelines to assist the various ministries, states and stakeholders in preparing their plans to handle different types of disasters. The guidelines on management of nuclear and radiological emergencies assume great importance in the present context, as our country has very ambitious programme to exploit nuclear energy for peaceful uses. Though, we have an enviable and impeccable record of safety and virtually fail-safe operations in all our nuclear establishments, the possibility, however, remote it may be, of human error, systems failure, sabotage, earthquake, floods, terrorist attacks etc leading to the release of radioactive material in the public domain, cannot be entirely ruled out. With this view, it was decided to prepare the national guidelines by NDMA to manage any nuclear/radiological emergency in public domain. Through these guidelines, we aim to further strengthen our existing nuclear/radiological emergency management framework and generate public awareness, which will go a long way in allaying misapprehensions, if any, amongst the public about the country's nuclear programme. Like in all our guidelines for handling of different types of the disasters, in these Guidelines also, maximum emphasis has been laid on the prevention of nuclear and radiological emergencies, along with a detailed consideration of all other elements of the disaster management continuum. The National guidelines have been prepared and a consensus was arrived on various issues, after wide-spread consultations and elaborates discussions amongst experts as well as stakeholders. It is assumed that once these guidelines are implemented by the stakeholders and converted into action plans followed by SOPs that will further reduce the chances of accidents in the nuclear arena.

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Radiation safety is reviewed for electron accelerators, mainly synchrotron radiation facilities. Accelerator radiation safety systems consist on safety interlock system, radiation shielding and radiation monitoring. These systems depend strongly on the characteristics of machines such as the maximum electron energy. In this paper, conceptual safety systems and radiation sources for synchrotron radiation facilities are overviewed including the evaluation methods of shielding.

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Activated carbon (AC) is widely used in various industries as an adsorbent material. It is used in gas cleaning industries, because of its low cost and high efficiency for removal of pollutants from effluents. It can be produced from a wide range of agricultural activities as by-products. There are different methods for producing the AC. The most common methods are physical and chemical activation that includes heat treatment, amination and impregnation. In this study, the effect of three impregnates, i.e., NaOH, KI and ZnCl ₂ on the quality of AC for ¹³¹ I removal, was investigated. Our results show that the sodium hydroxide impregnated AC is more effective for ¹³¹ I removal. Also, the quality of the impregnated AC was decreased in the case of impregnants material percentage exceeding 2%wt.

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Uranium is a naturally occurring radioactive element that is both radiologically and chemically toxic. The presence of uranium in the aquatic environment is due to the leaching from natural deposits, release in mill tailings, the combustion of coal and other fuels, and the use of phosphate fertilizers that contain uranium. Intake of uranium through air and water is normally low, but in circumstances in which uranium is present in a drinking water source, the majority of intake can be through drinking water route. The uranium concentrations in ground water samples from Vishakhapatnam, India were estimated using laser fluorimetric technique and were observed to range from 0.6 to 12.3 ppb. The laser fluorimetry technique was found to be an excellent tool for direct measurement of uranium concentration in water samples at ultra-trace levels. The annual effective dose, cumulative dose for 70 years and the lifetime excess cancer risk from drinking of this water were calculated. The risks were low averaging only 10.6 × 10 ^-6 as none of the samples were observed to exceed the WHO recommended uranium concentration limit of 30 ppb.

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A large number of medical cyclotrons of various types and makes are now in operation in India and their number is ever increasing. A careful analysis of the various safety issues particularly in case of possible accidental conditions is required for a realistic evaluation of their impact. Apart from the external dose involved, internal dose issues under normal operation and in case of abnormal operational conditions such as target rupture, accidents, spills, etc., which are generally neglected also need evaluation. There is a need for carrying out worst scenario analysis and the possible dose consequences to the operating staff as well as the public due to releases through the stack. The paper carries out an analysis for a typical release of activity into the vault environment and the dose implications. Along with any measured air activity measurements in the vault of operating cyclotrons this would resolve the issue one way or other. During radio pharmaceutical processing a substantial fraction of the volatile positron emitting radiopharmaceuticals are released into the atmosphere. In some cyclotrons a provision of an air trap for holding the air is mentioned. Analysis of possible dose to a member of the public using typical release rates is also presented and shown to be not negligible. A short review of such analysis in literature is carried out to show that the possible internal dose consequences cannot be ruled out and need to be addressed to in the safety analysis of these facilities for regulatory controls. Methods for proper calibration of stack monitors are indicated. In case of location of medical cyclotrons in crowded areas replete with high rise buildings, it may be necessary "to insure engineered safety features to ensure zero discharges from the machines."

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Accelerator driven systems (ADS) have the potential to generate nuclear power by coupling a high energy high current proton accelerator to a sub-critical nuclear reactor. The system has several advantages such as inherent safety from a runaway chain reaction, independence of the reactivity from delayed neutrons, resistance to nuclear proliferation, ability to transmute long lived minor actinides and fission products and, to convert thorium to ²³³ U. The radiological safety aspects of a typical ADS are discussed in which lead bismuth eutectic (LBE) is the spallation target as well as the coolant. In the irradiated target, residual activity is a major concern, particularly the amount of ²¹⁰ Po. The time evolution of the major isotopes present in LBE is studied along with the prevailing residual gamma dose rates. Few issues that may surface while coupling a reactor to an accelerator are also discussed.

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Shielded LINAC hall has been built to accommodate a radio frequency quadrupole (RFQ) linear accelerator (LINAC) based pulsed neutron generator at RSD, IGCAR. The concrete neutron shield wall thickness for the LINAC hall is finalized by adopting NCRP-51 methodology. The dimensions of the trenches and labyrinth are decided based on the neutron generator installation requirements. As per the AERB safety criteria, the radiation dose levels in the occupancy area should be less than 1μSv/h. However, the refined dose estimate with the inclusion of actual trench and labyrinth dimensions indicates an increased dose rate of ~3.10 μSv/h in LINAC control room. Hence, the control room is declared as controlled area. Additional shielding has been proposed for the trenches and labyrinth to make the control room as full occupancy area. For designing the additional shielding, experimental and theoretical analyses are needed to estimate the radiation streaming through the trenches and labyrinth. To start with, the neutron streaming through trenches have been studied using 185 GBq (5Ci) ²⁴¹ Am-Be neutron source for qualifying the trenches. The ²⁴¹ Am-Be source position is selected at the trench entrance such that it contributes the same neutron flux as that of neutron generator. Neutron dose rate and spectral measurements have been carried out at five locations along the trench from the entrance (LINAC hall side) to the exit (Control room side) of the trench.The experimental results are validated with the theoretical calculations using Monte Carlo N particle (MCNP) code. The analysis shows that the trenches are having a dose reduction factor better than 800 with respect to that of entrance dose. The observed dose rate at the trench exit is found to be less than 450nSv/h. In this paper, both the theoretical and experimental neutron streaming analyses through the trenches of LINAC hall are presented.

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This overview gives a brief summary on the experimental results on three topics: (1) Thick target neutron yields produced by protons, deuterons, He and heavier ions having wide energy range from MeV to GeV, (2) spallation products production data together with induced activities by proton to U ion and (3) benchmark experiments on neutron shielding using various accelerators of MeV to GeV energies. These three items are essentially important for radiation safety of accelerator facility.

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³²P is preferentially eliminated from the body in urine and is estimated by in situ precipitation of ammonium molybdophosphate (AMP) in the urine followed by gross beta counting. The amount of AMP formed depends on the amount of stable phosphorus (P) present in the urine. Hence, the present study was undertaken to estimate daily urinary excretion of P by the spectrophotometry method. P forms a colorless complex (phosphomolybdate) with molybdic acid, which on reduction produces deep-blue-colored complex called molybdenum blue. The intensity of this blue color is directly proportional to the amount of P present in the sample. 24 h urine samples collected from radiation workers were analyzed for stable P, and its range was observed to be between 0.4 and 1.4 g/day. This information was valuable in finalizing volume of the urine sample required for analysis of ³²P in bioassay sample by gross beta counting technique.

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The exposure of human beings to ionizing radiation from natural sources is a continuously inescapable feature of life on earth. Direct measurement of absorbed dose rates in air has been carried out in many countries of the world during the last few decades. Such investigations are useful for the assessment of public dose rates. Indoor and outdoor gamma dose rates were evaluated along the coastal regions of Kollam district, Kerala, through direct measurement using portable gamma dosimeter, and analysis of soil sample for activity of ²³⁸ U, ²³² Th, and ⁴⁰ K concentration was carried out using gamma spectroscopy. Indoor and outdoor exposure rates, the annual effective dose (AED), and lifetime cancer risk of residents along the coastal regions of Kollam district, Kerala, were evaluated. The reduction coefficients were also calculated for the region. The mean indoor effective dose due to background gamma along the coastal region of Neendakara panchayath was found to be 7.56 mSvy⁻¹ which is larger as compared with the worldwide average of the AED of 0.48 mSv y⁻¹  and the outdoor mean effective dose of 4.83 mSvy⁻¹ . Estimated excess lifetime cancer risk (ELCR) from indoor AED equivalent ranges from 22.56 to 26.46 × 10⁻³ and ELCR from outdoor ranges from 14.95 to 16.65 × 10⁻³ . Excess average lifetime cancer risk estimate from all the values is found to be 20.56 × 10⁻³ , which is larger compared with the resulting worldwide average 0.25 × 10⁻³ .

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Monitoring of radiation workers and maintaining their dose data is an important part of radiological protection program. As per RPR-71 (and presently RPR-2004), it is a regulatory requirement. National Occupational Dose Registry System (NODRS) at Bhabha Atomic Research Centre (B.A.R.C.) is maintaining and updating annual and lifetime dose data of all monitored radiation workers since the inception of Directorate of Radiation Protection in late sixties. The registry has a database of over 4,00,000 radiation workers. Over the years, the registry underwent several structural, software and hardware changes. Recently, it has been upgraded to network all monitoring labs located at NPP sites and other DAE units with the main database server at BARC through NPCNET and ANUNET. The new system provides online information on dose data, previous dose history of radiation worker which is an important requirement for monitoring units. Capability of NODRS to store biometric information of radiation workers enhances the effectiveness of the system. This paper gives the outline of NODRS along with some analysis of the data accumulated over the years.

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