[en] Purpose: To enable to decrease the difference in the infinite breeding factor due to the difference in the void coefficient, in view of the fuel history, present in the upper and lower regions of the reactor core, flatten the power distribution in the axial direction of the reactor core, and minimize the change in the infinite breeding factor relative to an abrupt change in the instantaneous void coefficient. Constitution: Fuel rods containing nuclear fissile materials, nuclear fissile fertile materials and burnable poisons and fuel rods charged with thorium are arranged in combination. The ratio of mixing the thorium-charged fuel rods is set to less than 20%. The thorium-charged fuel rods are arranged at the outer circumference or, in the case where water rods are present, also on the periphery of the water rods. This cannot only enable the effective use of fuels but also provide an advantage in maintaining the fuel soundness and in view of the reactor core control. (Yoshihara, H.)

[en] A concrete body of a drip tray is square in the cross section and the floor surface has a moderate slope, and a liquid collecting pot is formed at the lowest portion. The inner surface of the concrete body is lined with stainless steels at intervals. Then, neutron absorbers are disposed, close to the lining, between the outer surface of the lining of the floor surface and the liquid collecting pot and the inner surface of the concrete body. As the neutron absorbers, neutron absorbing materials such as a boron incorporated aluminum alloy, a boron carbide powder, cadmium or hafnium are used. Alternatively, the lining of the floor surface and the liquid collecting pot is formed by materials applied with neutron absorbing materials. With such a constitution, critical safety against leaked radioactive solutions can be ensured while keeping the floor surface smooth. Accordingly, decontamination operation upon periodical inspections and upon leakage of radioactive fuel solutions is facilitated. (I.N.)

[en] Axial neutron emission rate distribution is determined on every spent fuel assembly, an axial burnup degree distribution is determined based on the value, and further, a nuclear constant for a neutron transportation calculation is determined based on the correlation between the axial burnup degree and previously corresponded nuclear constant for neutron transportation calculation. Then, an axial neutron flux distribution of a containing system during the step of containing spent fuel assemblies is measured. Neutron fluxes are calculated by neutron transportation calculation of a fixed neutron source mode by using the nuclear constant and the neutron emission rate. A portion of the nuclear constant is adjusted so that the calculated neutron fluxes are in proportion to the measured neutron fluxes, and neutron multiplication characteristics of the containing system are evaluated. A portion of the nuclear content is further adjusted axially so that the shape of the distribution as a relative change of the measured value of the axial neutron flux distribution can be realized by the calculation of the fixed neutron source mode. The subcriticality of a pool in the plant or a cask containing system can be evaluated appropriately in a case of considering credit of burnup degree. (N.H.)

[en] In a pulsative solution counter-current extraction column (pulse column), neutron detectors are disposed at axial positions suitable for detecting accumulation of nuclear fuel materials, based on an axial concentration distribution of nuclear fuel materials and accompanying fission products upon normal and abnormal states by an analysis of solvent extraction simulation codes. In addition, neutron detectors are also disposed at a plurality of radial positions. Further, a value obtained by dividing a counting rate of the neutron detectors at reference time in normal state by a counting rate at each time is monitored on time sequence, or an axial distribution of the counting rate of the neutron detectors is monitored. A concentration distribution and accumulation or presence or absence of plutonium are detected, thereby enabling to directly control criticality in a mixer settler, a pulse column and a storage tank by using the counting rate according to an inverse multiplication method and a distribution shape method. (N.H.)

No abstract available

[en] A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

[en] Purpose: In a reactor power distribution control device, to obtain correct power distribution in a reactor even if the measuring device corner gap is displaced from the center position. Constitution: A gamma radiation measuring device is installed in the axial direction of the center position of the space enclosed by the four fuel rod assemblies of a reactor core. The gamma ray radiation level is used as the reference to measure the average power of the multiple rods in the adjacent fuel assembly and then the reactor power distribution is computed from it. If the position of the measuring device installed in the tube in the corner gap, correct values are obtained by the gamma radiation measuring device and compared with the thermal neutron beam measuring device, high precision reactor power distribution can be obtained on a stable basis. Furthermore, computation of the average output of adjacent fuel rods from the gamma beam values can be obtained with a fairly high accuracy. This means that the reactor power monitoring device can be used much more effectively than when used with conventional devices. (Kawakami, Y.)

[en] Purpose: To monitor the local power peaking coefficients obtained by the method not depending on the combination of fuel types. Method: A plurality of representative values for the local power distribution determined by the nuclear constant calculation for one fuel assembly are memorized regarding each of the burn-up degree and the void coefficient on every positions and fuel types in fuel rod assemblies. While on the other hand, the representative values for the local power distribution as described above are compensated by a compensation coefficient considering the effect of adjacent segments and a control rod compensation coefficient considering the effect due to the control rod insertion relative to the just-mentioned compensation coefficient. Then, the maximum value among them is selected to determine the local power peaking coefficient at each of the times and each of the segments, which is monitored. According to this system, the calculation and the working required for the fitting work depending on the combination of fuel types are no more required at all to facilitate the maintenance as well. (Horiuchi, T.)

No abstract available

No abstract available