[en] The proceeding contains papers presented in the Third Scientific Presentation on nuclear Fuel Element Cycle held on 4-5 Nov 1997 in Jakarta, Indonesia. These papers were divided by three groups that are technology of exploration, processing, purification and analysis of nuclear materials; technology of nuclear fuel elements and structures; and technology of waste management, safety and nuclear fuel cycle. There are 38 papers indexed individually. (ID)

[en] Fuel technology development program pf the nuclear fuel element centre is still devised into two main pillars, namely the research reactors fuel technology and the power reactor fuel technology taking into account the strategic influencing environment such as better access to global market of fuel cycle services, the state of the art and the general trend of the fuel technology in the world. Embarking on the twenty first century the fuel development program has to be directed toward strengthening measure to acquire and self-reliance in the field of fuel technology in support to the national energy program as well as to the utilisation of research reactor. A more strengthened acquisition of fuel cycle technology, in general, and particularly of fuel technology would improve the bargaining power when negotiation the commercial fuel technology transfer in the future

[en] This research has been done to solve it. The work was done through making variation on the fine powder composition from 0% up to 50% and variation on compacting pressure from 1 MP up to 5 MP. Each variation combination was replicated eight times. The result showed that the data were within confidence interval of the density and the ANOVA showed that the parameters affected the density. The affect of them was showed in the various correlation. The maximum density was reached at the range of fine powder composition between 20% and 30%. When Heckel theory was applied to the data, the conclusion showed that the fine powder composition did not affect the mechanical strength of material, but when Smith theory was applied, the conclusion showed that the fine powder composition affect the compressibility of powder. The maximum compressibility of powder was reached at the range of fine powder composition between 20% and 30%

[en] A computer program for steady state thermohydraulic evaluation of the straight plate fuel element for MTR has been made. The evaluation includes axial temperature distributions along the plate (I.e. coolant, surface cladding and fuel center temperature), surface cladding temperature, and heat flux for onset of nucleate boiling (ONB), pressure drops and flow rate through the coolant channel, the condition for flow instability and the critical velocity as an upper margin at which the fuel plate will not be collapse, and heat flux for the departure from nucleate boiling (DNB). The program was made based on equations recommended in the IAEA - TECDOC No. 233. A preliminary verification of the program for the RSG-GAS fuel elements showed appropriate results

[en] High level liquid waste is generated from the first step extraction in the nuclear fuel reprocessing. The waste is immobilized with boro-silicate glass. A certain composition of glass is needed for a certain type of waste, so that the properties of waste glass would meet the requirement either for further process or for disposal. The effect of waste loading on either density, thermal expansion, softening point and leaching rate has been studied. The composition of the high level liquid waste has been determined by ORIGEN 2 and the result has been used to prepare simulated high level waste. The waste loading in the waste glass has been set to be 19.48; 22.32; 25.27; and 26.59 weight percent. The result shows that increasing the waste loading has resulted in the higher density with no thermal expansion and softening point significant change. The increase in the waste loading increase that leaching rate. The properties of the waste glass in this research have not shown any deviation from the standard waste glass properties

[en] Radiation safety analysis of the 3 facilities in the nuclear fuel element center (NFEC) for assessing possible implementation of the ICRP-60 standard has been done. The analysis has covered the radiation dose received by workers, dose rate in the working area, surface contamination level, air contamination level and the level of radioactive gas release to the environment. The analysis has been based on BATAN regulation and ICRP-60 standard. The result of the analysis has showed that the highest radiation dose received has been found to be only around 15% of the set value in the ICRP-60 standard and only 6% of the set value in the BATAN regulation. Thus the ICRP-60 as radiation safety standard could be implemented without changing the laboratory design

[en] The fuel plates containing U₃Si₂-AI dispersion fuel having respective loading of 3.55; 4.20; and 4.80 g/cm³ were prepared by dispersing certain amount of U₃Si₂ powder in the AI powder as matrix. The weight ratio of U₃Si₂ and AI at different loading was chosen based on the 19.23 cm³ volume basis fuel core calculation. Each fuel mixture was pressed into a fuel core having dimension of 100.20 x 60.35 x 3.15 +- (0.05) mm, which was then cut into mini fuel core having dimension of 16 x 8 x 3.15 +- (0.05) mm. The mini plates were prepared by picture and frame technique using AIMg2 as cladding material. The mini plates have been tested for blister, homogeneity, white spots, surface defects and their cladding thickness, revealing that out of 74 mini plates, they are ten (10) mini plates that have to be rejected due to blisters and white spots, thus of 64 mini plates can be further fabricated as samples for irradiation test

[en] Uranium ore body which has been discovered at Eko-Remaja Hill belongs to the vein type system, but later on, after mining test study the oer body might not be as ideal as the vein from; at the southern part they show deviation from the original direction or even discontinue. From the view of the mining work these veins seem to from another type of ore body, which is called a stock work. The stock work consists of groups of closed vein, and the stock work at Eko is assumed to have 6 groups corning from 16 vein system. The stock work ore body seems to fit and to be supported also by the uranium geology condition at Eko hill. According to this assumption, the quality and the quantity of the uranium ore body will change, and the result of the evaluation will have a total of 4.32 million ton of ore, containing 531,56 ton uranium and it can be recovered at as much 0.18 Kg U₃O₈/ton ore. The mining system for these stock works vein system at Eko Hill is underground bulk mine, which is characterised by higher production capacities compared to underground shrinkage stopping mine

[en] Batan as a facilitator in achieving the nuclear programme, has to start to introduce the nuclear science and technology to the publics and announces the nuclear activities results. By this mean time, most of people is still quite sure considering that the use of fossil fuel is better than that of nuclear fuel, even though there will be no other choice ultimately to use nuclear energy. This contradiction appears due to three reasons i.e., public's unbelieveness crisis dealt with less of knowledge of public about nuclear sciences and technology, their confident of nuclear-danger risks and the broadly miss-information sprout by the opposition. The main challenge created from the crisis of the safety culture of human activities. Some points to anticipated those situation are to establish infrastructure, maintain co-operation, links and coordination within points in the infrastructure, also to create the working culture system based on professional development and high quality safety management, and to increase publics information by a birds point approach through regional, cultural, social and political communication. Publics knowledge and understanding will bring about their supports and acceptances to the nuclear programme and its development (Nuclear Power Plant as a particular)

[en] An extraction process for uranium (U) separation from ruthenium (Ru) by tributyl phosphate (Tbp) as a solvent with kerosene as a diluent and surfactant the Span-80 as emulator has been performed. A sodium carbonate solution having ph 10-11 was used as the internal phase. the feed contained U and Ru in HNO₃ solution, while butanol was used as the membrane sp liter. the membrane used for extraction had a composition of 5% vol surfactant, 10% vol Tbp, 35% vol kerosene and 50% vol Na₂CO₃ solution. The mixing time and speed were varied in value, so were the acidity levels, ranging from 0.5 M, 0.75 M up to 1 M. The result of the experimental separation of a mixture containing 5000 ppm U and 325 ppm Ru showed that a molarity of 0.5 M HNO₃ in the feed, 15 minutes mixing time, and 800 rpm mixing speed were found to be the best conditions for the extraction. Such condition had resulted in achieving the value of stripping distribution coefficient K_d for U= 0.5660 with 49.73% U recovery efficiency and a practically zero separation factor from Ru