[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 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)

[en] A method based on Pulsed laser deposition followed by Isotope dilution mass spectrometric method is evaluated towards the possibility of direct measurement of burn up of nuclear fuel and also to find out spatial distribution of burn-up along the pellet. The wave length dependent results show larger error with 1064 nm, compared to 532 nm laser beam. Much less error is expected with shorter wave length and shorter pulse width laser beam. Further work is being carried out in this direction

[en] The theme meeting focussed on perspectives of the chemistry of materials relevant to fast reactors, novel materials and their applications, novel extractants and resins, applications of XPS and AES, inorganic matrices for special applications, thermochemical properties of alloys and compounds, computations and modelling, ultratrace analysis and advanced separation techniques etc. Papers relevant to INIS are indexed separately

[en] The chemistry of materials pertaining to fast reactors is both fascinating and challenging considering the nature of materials involved such as the fuel, coolant, control and shielding materials in addition to the interactions between the structural materials and the fuel/coolant depending on the nature and conditions involved. The different chemical forms of fuel materials, the need to operate up to high burnups with consequent interactions of the fuel with clad materials, the need to close the fuel cycle by recovery of the fuel materials from spent fuels for refabrication and the necessity to manage the waste, throw a host of challenges which make their study scientifically interesting and technologically important. The use of liquid sodium as coolant in fast reactor heat transport systems combined with its inherent chemical reactivity opens up an interesting branch of chemistry involving liquid sodium especially in contact with structural materials during normal operation of the reactor and with fuels in the event of fuel pin failure. The phenomenon of sodium wetting and the associated corrosion of structural materials in contact with it combined with the need to carryout decontamination of such materials make it interesting to examine and evaluate their suitability for reuse without compromising on their structural integrity. Boron being the material of choice for control and shielding applications in fast reactors with varying isotopic enrichment and the technological challenge to produce large quantities of boron carbide makes it unique. Some of these aspects are addressed in this paper. (author)