Mobile system for the management of fuel assemblies from research reactors

Interim storage in transport and storage casks of the CASTOR type, and later the final storage of these casks are planned for the management of spent fuel assemblies from German research reactors.A mobile transfer unit is used for loading the casks with fuel assemblies on the reactor sites. Key components of the mobile transfer unit are a transfer cask, the recharging lock, and an air-cushion transport system. By means of the air-cushion transport system, the whole equipment, as well as the CASTOR casks, is transported into the reactor building. Thus, handling of the 16 t CASTOR casks is possible even on reactor sites within sufficient crane capacity. A 20 ft container accommodates the mobile transfer unit and all accessories so that the whole equipment can be transported to the reactor sites by truck.     

Measurement of the subcriticality of power-stressed reactors with an analog reactimeter

Conclusions A technique has been developed for measuring the effects of reactivity in a subcritical reactor with an analog reactimeter. It is based on the compensation of the current applied to the reactimeter input from the neutron detector. The compensation of current produced by neutrons of the subcritical multiplying assembly formalizes the algorithm for reactivity calculation, making it an adequate model of a reactor with a source and making it possible to determine the subcriticality without prior entry into the critical state. In this case the measurements are made in the presence of neutron sources characteristic of power-stressed reactors. The regular devices of the control and safety system could be used to produce unsteady variation of the neutron flux.All of this permits the proposed method to be extended to zero-power reactors and to power-stressed reactors. Once the subcriticality has been measured an analog of the neutron source is introduced into the reactimeter. This instrument measures the effects of reactivity in the subcritical state without the reactor being previously put into the critical state, monitors the entry of the reactor into the critical state by checking the reactivity, and makes all the measurements usually made with analog reactimeters. If the intensity of the source does not change during measurements (5–10 min) when chambers sensitive to rays (e.g., KNK-56 chambers) are used as neutron-flux detectors, then the accuracy of -ray compensation does not affect the results of the measurements.Translated from Atomnaya Énergiya, Vol. 45, No. 5, pp. 375–376, November, 1978.     

A study of critical assemblies with a beryllium moderator

Experiments are described which were carried out with critical assemblies of beryllium and uranium to determine the square of the moderation length up to an energy at which the moderation spectrum undergoes transition to a thermal neutron spectrum; also to determine the fast neutron multiplication factor, taking into accountffission and the Be(n, 2n) reaction and allowing for absorption as a result of moderation. The critical assemblies are calculated by a multigroup method and by the use of the parameters obtained. There is good agreement between calculated and experimental data.Translated from Atomnaya Énergiya, Vol. 16, No. 3, pp. 228–233, March, 1964     

Advanced mixed oxide fuel assemblies with higher plutonium contents for pressurized water reactors

Optimizing fuel cycle costs by increasing the final burnup leads to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in mixed oxide (MOX) fuel assemblies (FAs) reduces further the amount of plutonium which has to be disposed of in final storage. Increasing the final burnup requires higher initial enrichments of uranium fuel to be matched by an advanced design of MOX FAs with higher plutonium contents. The neutronic design of these MOX FAs has to consider the licensing status of nuclear power plants concerning the use of MOX fuel. The Siemens Nuclear Fuel Cycle Division, with more than 20 years' experience in the production of MOX fuel, has designed several advanced MOX FAs of different types (14 × 14 to 16 × 16) with fissile plutonium contents up to 4.60 w/o.     

A simple approximation to estimate the reactor period in critical assemblies

A simple approximation is proposed to estimate the reactor period in critical assemblies for which the reactivity parameters, such as the fissile material, neutron-energy spectrum, and effective fraction of delayed neutrons, are characterized. This approximation is based on the period-reactivity equation, in combination with the delayed-neutron parameters corresponding to the fissile material and neutron-energy spectrum of the critical assembly, and Monte Carlo computer-code calculations of the criticality factor, prompt-neutron lifetime, and average number of neutrons per fission. The proposed approximation is validated with experimental results for the Lady Godiva fast metal and the Solution-High Energy Burst Assembly (SHEBA) thermal solution critical assemblies that operated at Los Alamos National Laboratory. In addition, the results of the approximation are also compared with those reported for experiments conducted at the Kyoto University Critical Assembly. The results from this analysis indicate that the proposed approximation can be used to estimate the reactor period in critical assemblies.     

Effects of low heterogeneity in fast critical assemblies

Translated from Atomnaya Énergiya, Vol. 66, No. 1, pp. 13–17, January, 1989.