No abstract available

[en] In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

No abstract available

[en] The methodologies of life-cycle analysis are discussed. The system boundaries have to be adequately defined, which is illustrated with the example of coal-fired electricity generation. The input/output method of LCA is discussed, including the incorporation of material recycling in the analysis. Also discussed is the linkage of engineering and economic approaches together with the necessary improvements of MARKAL in order to integrate the indirect processes. Finally examples are given of the analysis of the direct and indirect CO₂ emission from a pressurized-water reactor and its fuel cycle. The analysis shows that the life-cycle CO₂ emission coefficient is 25.7 g CO₂/kW.h in case of gas-diffusion enrichment, whereas in case of centrifuge enrichment this emission coefficient amounts to 7.9 g CO₂/kW.h only. (author). 9 refs, 8 figs, 3 tabs

[en] The effects of fuel designs on plutonium recycle in a large boiling water reactor are investigated using the LEOPARD, SKAT-4 and BOLERO code. Some of the limitations within which any substitute fuel must operate in an existing reactor are imposed on the design of plutonium-bearing fuel. These are: the external dimensions of the substitute fuel assembly must be the same as the original assembly to avoid vessel modification; the void reactivity, the initial excess reactivity and the burn-up reactivity swing must be of the same order as the original one to avoid control rod modification; the heat transfer condition must be similar to the original one and the reactor must always be capable of rated power output. The design to satisfy these limitations and to give near optimum moderation is found to employ the smaller diameter rods or higher water-to-fuel ratio for the plutonium recycle case. The fuel element of constant surface area has a tendency to possess a positive void reactivity coefficient at the end of burn-up. However, the smaller fuel rod diameter and slightly increased heat transfer area of the fuel element can improve this tendency. This design has the advantage that less fissile inventory is needed to give the same burn-up level and the disadvantage of incremental fabrication costs of the smaller diameter rods. The plutonium value analysis is calculated on the selected fuel designs on the basis of fuel cycle cost compared with that of ²³⁵U fuel core. It appears from these results that the plutonium recycle in a large boiling water reactor can be economical, based on the present analysis of plutonium value. (author)

[en] Direct application of nuclear process heat could be one of the vital problems in future energy systems. The authors made a system analysis on process-heat application and a economical feasibility study on supply technologies. This paper presents those results and gives a brief description of the recent RandD activities

[en] The possible types of a large HTGR (thermal output 3,000 MW, outlet gas temperature 1,000⁰C) that might be integrated into a plant complex containing the direct reduction-iron making process and electric power generation plants were discussed. Process flow diagrams were made for the plant complex, paying attention to the optimization between the process heat utilization and electric power generation. Design studies were concentrated on the HTGR and its components; this effort ephasises the need to overcome the technical problems of very high temperature. (author)

[en] The heavy-water-moderated power reactor has been studied mainly at Japan Atomic Energy Research Institute (JAERI) since 1963, and the boiling light-water-cooled type was selected for further development in Japan in 1966. After this decision many investigations have been made on the characteristics of the above reactor at JAERI and, based on the results, the five nuclear power consortia in Japan have made preliminary designs individually. A report is made on the results obtained from the investigations and designs. The study and investigation based on the principle of using natural uranium led to the adoption of a Pu self-sustaining cycle, from the standpoint of fuel-cycle cost and power cost. This cycle has the following advantages: it is equivalent to a slightly enriched fuel cycle giving much flexibility in the core design, and about 15 000 MWd/t of burnup might be expected with natural uranium when enriched by Pu produced in its own spent fuel. The positive void coefficient, which decreases the stability and safety of this reactor type, would no longer be a major problem when using Pu enrichment. However, it must be noted that a slightly enriched uranium would be loaded in the initial cycle, which might have a positive void coefficient. The problem can be partially overcome by placing inter-lattice tubes in the core as in the SGHW in order to adjust the moderator to fuel ratio, making a lower ratio for the U-loaded core and a higher ratio for the Pu-loaded core. This would give rise to a higher quality of reactor exit steam or higher dryout limit in the Pu-loaded than in the U-loaded core. On-power refuelling is one of the major problems to be solved for the reactor. Two refuelling methods (access from the reactor top and from the reactor bottom) were investigated and the corresponding refuelling machines were designed. An outline of some reshuffling schemes, stability and safety analysis and plant design (stressed to the reactor structure) are also given. (author)

No abstract available

[en] The marginal values of the investment cost were analyzed for each of five types of SMSNRs to penetrate into the energy system optimized by minimizing the total system cost. The results showed the upper values of the investment cost as 1.4, 1.35, and 1.45 times that of large-scale LWR for medium-sized LWR, small-sized LWR, and LWR for cogeneration, respectively. While, the economic conditions for SMSNRs to be adopted by electric utilities were analyzed through comparison of financial performance of a Japan's typical electric utility in two cases of selecting large-scale LWR and small-sized LWR. The results indicated that the upper limit of the construction cost is 1.4 times that of large-scale LWR. (author)