[en] Fission products that emit considerable decay heat and radioactivity, such as ¹³⁷Cs, have a large impact on waste management. Small and high-performance extractor is desirable for separating such nuclei. In this study, we implemented the continuous extraction of Cs from nitric acid in a single liquid-liquid countercurrent centrifugal extractor with Taylor Vortices by calix arene-bis(t-octylbenzo-crown-6)(BOBCalixC6) as an extractant with trioctylamine(TOA) as a suppressant and with 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (Cs-7SB) as a phase modifier. Because of slow extraction kinetics of this process, extraction with multiple theoretical stages by just replacing conventional extractors into the single centrifugal extractor is difficult. Hence, we improved the dispersion of organic phase by an inner rotor made of lipophilic epoxy resin and elevating the solution temperature to lower the viscosity. Higher temperature was not appropriate from the aspect of chemical equilibrium in this process, but extraction with multiple theoretical stages was found to be possible. (author)

[en] The traditional shielding method was mostly to use a thicker shielding of lead plates, and cast reinforced concrete, etc., mainly by reducing the neutron speed and preventing the passage of neutrons. However, the problem of making or machining the thick protective layers cannot meet the development needs of the security of the nuclear power industry. Currently, most widely used nuclear protective materials are polyethylene plastic plates adding boron carbide, because of polyethylene containing a relatively high content of hydrogen atoms that is an effective neutron moderator by virtue of its scattering power, and boron carbide that is also a good thermal neutron absorber by means of huge thermal neutron absorbing cross section. In this regard, in the present work, polymer flame spraying coatings of neutron-absorbing boron containing polymer composite powder is developed for application in the field of spent nuclear fuel. Changes in the microstructure of the coating layer are discussed with respect to the content of boron carbide and the thickness of the coating layer in view of the neutron absorbing efficiency. In this work, polymer flame spraying coatings of neutron-absorbing boron containing polymer composite powder was developed for application in the field of spent nuclear fuel. From the observation of coating layer, B4C particles were distributed uniformly in the polymer matrix and the LDPE-B4C composite coating layer was joined well with Al substrate without any detachment. The thermal neutron absorbing property is enhanced with an increase in the coating thickness. A flame spraying coating method of boron-containing polymer composite powder is very effective way for the application in a spent nuclear fuel facility

[en] Thorium as a nuclear fuel is receiving renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTR development employed thorium together with high-enriched uranium. After 1980, most HTR fuel systems switched to low-enriched uranium. After completing fuel development for AVR and THTR with BISO coated particles, the German program expanded efforts on a new program utilizing thorium and high-enriched uranium TRISO coated particles for advanced HTR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of LTI inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with HTI-BISO coatings. The improved performance of the HEU (Th,U)O₂ TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 °C in normal operations and 1600 °C in accidents, with burnups up to 13% FIMA and fast fluences to 8 × 10²⁵ m⁻² (E > 16 fJ), the results exceed the design limits on manufacturing and operational requirements for the German HTR Modul concept, which were: <6.5 × 10⁻⁵ for manufacturing; <2 × 10⁻⁴ for normal operating conditions; and <5 × 10⁻⁴ for accident conditions. These performance statistics for the HEU (Th,U)O₂ TRISO fuel system are in good agreement with similar results for the LEU UO₂ TRISO fuel system

[en] The inundation area of Ploučnice river, Czech Republic, has been contaminated by natural radionuclides during the early mining of the uranium ore deposit in the region of Stráž pod Ralskem. A study of temporal changes of contamination was based on a comparison of airborne gamma-ray spectrometric data from 1991 - 1993 and from 2005. After that, detailed ground gamma dose rate measurements were performed at several chosen areas. The results indicate a decrease of contamination with time in a majority of contaminated areas. Advantages and disadvantages of both approaches to the evaluation of the level of contamination (airborne gamma-ray spectrometry, ground gamma dose rate measurements) are described. (author)

[en] The the salient features of the ore processing flowsheet for uranium recovery from the Tummalapalle ore which was the fore-runner for the commercial plant coming up at Tummalapalle in Andhra Pradesh with a slated capacity to treat 3000 tonnes of ore per day using state-of-art alkaline pressure leach process technology are presented

[en] Decladding of spent PHWR fuel is achieved by mechanically chopping the spent fuel employing special purpose shearing machine. Spent Fuel Chopper based on progressive feeding, clamping and chopping has been in operation in the present operating reprocessing plants located in Tarapur and Kalpakkam. Valuable experience has been gained in operation and maintenance aspects of this equipment over the years. In order to increase the productivity and reduce the maintenance down time, a new spent fuel chopper based on gang chopping concept has been developed incorporating the good features of the existing model. The first spent fuel chopper designed and manufactured as per new concept has undergone cold commissioning in ROP Tarapur and hot commissioning is in progress. (author)

[en] High level Liquid Waste (HLW) is generated during the reprocessing of spent nuclear fuel which is used to recover uranium and plutonium. More than 99% of the radioactivity generated during the burning of nuclear fuel in the reactor is present in HLW. For the efficient management of HLW either by vitrification in the suitable borosilicate glass matrix, or partitioning and transmutation (P and T) of the minor actinides and long lived fission products, it is desired to assay the HLW for its constituent stable elements as well as radioactive content. The present article gives a brief account of an exercise carried out recently to characterize the HLW from PREFRE, Tarapur. (author)

[en] It is demonstrated, that the fluoride can be separated from LIQUID WASTES using pyrohydrolysis method. The routine method for separation of halides from ceramic samples was modified for radioactive liquid wastes. Subsequently, after separation by pyrohydrolysis, fluoride was determined from the pyrohydrolysis distillates by ion chromatography. Total time taken to determine fluoride is about 45 min including 30 min for the pyrohydrolysis and 15 min for ion chromatography. The results of recovery tests ranged 98% or above. The limit of detection for fluoride is 0.5 mgkg^-1. (author)

[en] The ever increasing demand on power requirement in the country has opened up need for exploring use of nuclear fuels that could meet such demands. This makes the mission of the department to shift from the first stage of nuclear programme employing natural uranium in PHWRs to the second stage of deploying a large number of fast reactors with plutonium based fuels capable of realising high breeding ratios in addition to energy production. The transition to fast reactors with advanced fuels, capable of higher breeding ratio, opens up a number of scientific and technological challenges in design and operation of such fast reactors. In the Indian context, after successful demonstration of natural uranium based PHWRs, the performance of U-Pu based carbide fuel, as a unique experience in the world, has been demonstrated in FBTR at Kalpakkam. This paper deals with the performance of carbide fuel in FBTR and the programme on development of metallic fuels with appreciably high breeding ratio that would result in considerable reduction in doubling time thereby addressing the increasing demands of power production as well as pave way for introduction of a large number of such fast reactors to provide energy security to the country. The advantages of introduction of metallic fuels as well as the scientific and technological challenges to be faced in doing so and the ongoing efforts towards metallic fuel development are also described in the paper. (author)

[en] Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have conventionally been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents, such as Am, are problematic in a conventional injection casting method. As an alternative fabrication method, low pressure gravity casting has been developed. Casting soundness, microstructural characteristics, alloying composition, density, and fuel losses were evaluated for the following as-cast fuel slugs: U-10 wt% Zr, U-10 wt% Zr-5 wt% RE, and U-10 wt% Zr-5 wt% RE- wt% Mn. The U and Zr contents were uniform throughout the matrix, and impurities such as oxygen, carbon, and nitrogen satisfied the specification of total impurities less than 2,000 ppm. The appearance of the fuel slugs was generally sound, and the internal integrity was shown to be satisfactory based on gamma-ray radiography. In a volatile surrogate casting test, the U–Zr–RE–Mn fuel slug showed that nearly all of the manganese was retained when casting was done under an inert atmosphere. (author)