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*; *; *
PNC TN941 85-69, 201 Pages, 1985/04
Doppler reactivity has been calculated with uniform temperature within fuel rod for reactor design. Therefore, actual temperature distribution is close to parabolic, depending on the central and surface temperature, because the dependence of PuO
-UO conductivity with the temperature and nonuniformity of heat generation within the fuel rod affect the temperature distribution in opposite sense. Doppler broadening effect is considered mainly due to resonance absorption in 
U on the surface of fuel rod. In the present study, doppler reactivity effect due to nonuniform temperature within fuel rod was investigated by the RABBLE/WIMS-ATR code. We considered several temperature distributions in the cases of normal operation, transient and LOCA for a 36-rod cluster cell. The following results are summarized. (1)It is necessary to divide the fuel rod into about 10 concentric-annular sub-regions at equal intervals for calculations with nonuniform temperatures. (2)Doppler reactivities calculated with nonuniform temperatures in the cases of normal operation and transient are reduced by about 5% in comparison with ones calculated with uniform temperatures. In the case of LOCA, doppler reactivities calculated with nonuniform temperatures are about 6% larger than ones calculated with uniform temparatures.
and PuO-UO*; *; *
PNC TN941 79-19, 132 Pages, 1979/02
Loss-of-coolant reactivities in DO lattice with 28-fuel-rod clusters have been measured as function of plutonium enrichment in PuO-UO fuel, statistical weight of PuO-UO fueled region and 
B concentration in DO moderator. The measurements were done by using Deuterium Critical Assembly having a square lattice of a 25.0-cm pitch. PuO-UO clusters were progressively substituted for 1.2 wt% enriched UO clusters from the core center, as being surrounded by UO fueled region. The statistical weight of plutonium fueled region was changed from zero up to 0.6. The enrichments of PuO in PuO-UO fuel are 0.54 and 0.87 wt%; also, fissile content in each plutonium is about 91 wt%. The concentration of B in DO moderator of 99.45 mol% WaS zero or 3.9 ppm. Pulsed neutron source technique was applied to determine the reactivities. Prompt neutron decay constants were converted into reactivities in dollars by means of Simmons & King's formula. Changes in prompt generation time were corrected in the conversions. Calculational values of the reactivities were obtained by the codes of METHUSELAH-II and CITATION, and are in good agreement with experimental ones. Effect of anisotropic neutron diffusion was introduced into the calculation, on the basis of Benoist's theory. Perturbation theory was applied to resolve calculated reactivities into components concerning changes in neutron leakage, neutron spectrum and net production rate. The comparison between the experiment and calculation are given in the table. The following are concluded from the present study. (1)If PuO-UO and UO fuels are equal in content of fissile atomes, PUO-UO fuel is more effective in shifting the the loss-of-coolant reactivity to the negative side; on the other hand boron in DO moderator shifts the loss-of-coolant reactivity to the positive side. These results have the ...
*; *; *; Fukuda, Kenji*; ; *
PNC TN941 78-13VOL1, 266 Pages, 1978/01
PNC-TN941-78-13VOL1.pdf:18.12MB
Nuclear characteristics of 54-pin fuel assembly for commercial FUGEN were studied from the view points of reactor safety and fuel-cycle economy. Coolant void reactivity and power coefficient for the analysis of reactor safety, and burnup and fuel-cycle indicator for evaluating fuel-cycle cost were investigated in reference to lattice pitch, uranium and plutonium enrichment. From the present study, the following results are summarized. (1)Uranium and plutonium fuel can be used in a wide range of enrichment. (2)Natural uranium or low enriched uranium (
1wt% 
U) fuel mixed plutonium shows the best nuclear characteristics. (3)Nuclear characteristics are improved by more introduction of plutonium in consideration of local peaking factor and power mismatch.
*; *; *; Fukuda, Kenji*; *;
PNC TN941 77-75, 40 Pages, 1977/05
Liquid poison tube has been used for the fast shut-down system of reactor in the SGHWR and the zone control in the CANDU-PHW. In the present study of reactor control system for commercial Fugen, reactor control system using liquid poison tube was investigated from the view points of reactivity control, design and development. The fine and coarse reactivity control can be achieved with liquid poison level in double tubes similar to reactivity control by control rod system. Also, the fast shut-down of reactor is achieved with the same double tube in the fine and coarse reactivity control system without high-pressure head tank. Design evaluation of this system will be further discussed with development necessary for confirmation reliability and performance characteristics.
*; *; Fukuda, Kenji*; *; *;
PNC TN941 77-74, 34 Pages, 1977/05
Burnable poison in the fuel elements has been mainly used for decrease of power peak due to refueling in the Light Water Reactor. In the present study of core performance of commercial Fugen, core characteristics by introductions of Gd burnable poison was investigated from the view points of Keff, local peaking factor, mismatch factor and coolant void reactivity. The conclusions are summarized as follows. (1)Effective control or core reactivity is achieved by introduction of Gd burnable poison into a few fuel elements in inner layer. (2)Improvement of local peaking factor cannot be easily achieved by Gd burnable poison. (8)Mismatch factor is decreased from 6 % to 18 % by Gd burnable poison. (4)Coolant void reactivity shifts more negative by introduction of Gd burnable poison.
-UO fuel*; *; *
PNC TN941 76-09, 65 Pages, 1976/01
Void reactivities caused by loss of coolant have been measured in DCA core (which has a 22.5-cm lattice pitch) by changing the number of PuO (0.87 wt%)-UO fuel assemblies from 0 up to 25. The PuO-UO fuel assemblies were progressively loaded from the core center, being surrounded by the 1.2 wt% enriched UO fuel region. In order to find out the effect of the fissile content in plutonium on void reactivities, two kinds of mixed fuels were used in the experiment: one enriched with standard grade plutonium (91 wt% Pu fissile) and the other with reactor grade plutonium (74 wt%). Void reactivities were measured by means of pulsed neutron source technique. When evaluating void reactivities from measured prompt decay constants, corrections were made to the neutron generation time of a largely subcritical system, which accounted for up to 5% maximum. In the Table, measured void reactivities were compared with the values calculated by METHUSELAH-II and CITATION codes. The following are concluded from this study. (1)Void reactivities gradually shift to negative side by loading of more PuO-UO fuel. (2)The increase of fissile contents in plutonium also makes void reactivity shift to the more negative side. (3)Measured void reactivities are more negative than calculated ones. The discrepancy between calculation and experiment becomes greater with increase in fissile contents in plutonium.
*; *
PNC TN941 75-39, 53 Pages, 1975/05
Absolute thermal neutron flux in Deuterium Critical Assembly's core fueled with 0.87 w/o PuO-UO (reactor grade) and 1.2 w/o UO has been measured by means of Au foil activation for coolant void fraction of 0 %. Absolute activities of irradiated Au foils were measured with 
-
coincidence method and 4
-ray counting method, and both methods gave same results. The experimental error of absolute thermal neutron flux was estimated as 6% or lower. By using the absolute thermal neutron flux determined from this experiment, absolute thermal neutron flux distribution in the whole core was obtained from gross reaction rate distribution, which was measured with dysprosium foils.
*; *; Hachiya, Yuki*; Fukumura, Nobuo*; *
PNC TN941 74-26, 46 Pages, 1974/06
Intra-cell thermal neutron flux distributions in 0.54 w/o PuO-UO fuel loaded in 22.5 cm pitch lattice have been measured by means of dysprosium foil activation method for coolant void fractions of 0 %, 30%, 70% and 100%. These results are compared with values calculated by NOAH-II code. Thermal neutron flux distributions in DO moderator region were measureed with better accuracy (
1.5%) with a new experimental technique. Measurements at the inside and outside of a pressure tube and a calandria tube were made with a foil handling technique which had been developed in the present experiment. As a result, the detailed comparison of thermal neutron flux distributions in the pressure and the calandria tube with the calculation could be performed. The experimental result for coolant void fraction of 100 % agrees with the calculated value better than the results for 0 %, 30% and 70%.
*; *; *; *; Fukumura, Nobuo*;
PNC TN941 74-22, 52 Pages, 1974/05
Input data such as diameters and densities of fuel pellets which are used in nuclear calculatlon of DCA core are surmmarized in this file. The data are collected from score sheats and millsheets of DCA's reactor materials. Standared deviations from averaged values and results of chemieal analysis of the materials are also listed as a measure of accuracy of the data. Fuel materials and coolants summarized in this file are as follows;
and PuO-UO fuel clusters*; *; *; *
PNC TN941 74-15, 71 Pages, 1974/04
Void reactivities and their 1-st differential coefficients with respect to axial buckling have been measured as a function of void fraction and number of PuO-UO fuel clusters progressively substituted for UO fuel ones, using the pulsed neutron source. Table 1 lists the obtained results. Followings are concluded from this study. (1)Loading of PuO-UO fuel clusters increases the negative reactivity contribution to the void effect. (2)Neutron leakage in a core partially loaded with PuO-UO fuel clusters does not always produce the negative reactivity contribution to the void effect.