Gas phase in experimental fuel elements with compact uranium dioxide,irradiated in the SM-2 reactor

Conclusions A complex procedure has been developed for the study of gas release from nuclear fuel, including reactor measurements and post-reactor determination of the amount and composition of the gas medium in the fuel elements at room and elevated temperatures. In fuel elements with compact uranium dioxide (density 10.0–10.43 g/cm³), in addition to gaseous fission products and the helium introduced, Ar, H₂, O₂, CO, CO₂, and N₂ are present, and after irradiation their quantity exceeds the initial quantity, measured for unirradiated fuel elements, by a factor of several.The yield of Xe and Kr under the can of the fuel elements during irradiation of uranium dioxide in the SM-2 reactor amounts to 30–50%, but the measured ratio of Xe/Kr exceeds the calculated ratio by a factor of 1.2, because of the reaction¹³⁵Xe(n, )¹³⁶Xe. The content in the fuel of adsorbed helium is equal to 0.004 n.cm³/g UO₂. The data obtained can be used for physics and technological calculations, and also for refining the procedure for the determination of gas release.Translated from Atomnaya Énergiya, Vol. 57, No. 2, pp. 91–95, August, 1984.     

Structural analysis of reactor fuel elements

An overview of fuel element modeling is presented that traces the development of codes for the prediction of light-water reactor and fast breeder reactor fuel element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations.Current efforts include modeling of alternate fuel systems, analysis of local fuel cladding interactions and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.     

Physical and operating characteristics of the SM-2 reactor

Conclusion Operating experience with the SM-2 reactor verified the correctness of the basic physical and engineering ideas incorporated in the design:Translated from Atomnaya Énergiya, Vol. 17, No. 6, pp. 452–463, December, 1964Paper No. 320, presented by the USSR at the Third International Conference on the Peaceful Use of Atomic Energy, Geneva, 1964.     

Facility for the irradiation of fuel elements in the SM-2 and MIR reactors with variable operating regimes

Conclusions The procedures and irradiation facilities developed for testing fuel elements at a variable power have received experimental verification in the SM-2 and MIR reactors, and have shown satisfactory operating efficiency of all units. During the experiments, no appreciable effect of the facilities was noted on the operation of the reactors into which they were loaded, and on the testing conditions in other channels. These procedures at the present time are being used in investigations conducted in the SM-2 and MIR reactors, for verification of the operating efficiency of the fuel elements of power reactors in different variable operating cycles.Translated from Atomnaya Énergiya, Vol. 58, No. 2, pp. 97–100, February, 1985.     

研究堆用U3Si2-Al燃料板的热稳定性试验

U3Si2-Al板状燃料组件是一种推广应用的新型燃料元件,在国内首次应用。燃料组件的各项性能,特别是热稳定性必须通过实验验证。通过对铀密度为3.02 g/cm3的U3Si2-Al燃料板的热稳定性试验,得到:热稳定性试验会使燃料板的体积略有增大;120℃及250℃的热循环下,燃料板无明显变形,表面无变化,400℃的热循环下,燃料板略有弯曲,个别芯体裸露的燃料板表面有起泡现象;循环温度越高,芯体中U3Si2颗粒开裂越严重等实验结论,为该燃料组件的结构设计、安全分析、加工工艺提供了关键参数,并为该组件的堆内运行提供了借鉴。