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*; Takemura, Morio*; *; *; Fukumura, Nobuo*
PNC TN941 83-65, 51 Pages, 1983/05
Critical experiments on DCA cores with aseismatic plates have been done in order to clarify the value of the reactivity worth of aseismatic plate and the behavior of the detailed power distribution or the thermal neutron flux distribution around the aseismatic plate. The aseismatic plate was arranged in the heavy water moderator of the nine central channel region of the DCA core. The thickness of the plate is 5mm, 20mm, 40mm or 80mm. The axial position of the plate was a half of critical heavy water level. The DCA core has a 25cm square lattice pitch and 97 fuel clusters with 1.2w/o uranium enrichments. The heavy water includes 
B(
3ppm). The coolant is light water. The reactivity worth of the plate was measured by use of the critical heavy water level between the cores with and without the plate. Axial power distribution was obtained by measuring the 
-ray intensity of the typical fission product (
La) of irradiated fuel rods. Axial thermal neutron flux distribution was obtained by measuring the activities of irradiated copper wires inserted in the central fuel channel or the moderator regions with and without the plate. The followings are found from the present experiments. (1)The reactivity worth of the aseismatic plate is monotonously increasing with the increase of the thickness of the plate. The values obtained by the present experiments change from about 0.1% 
K to 1% 
K. (2)The axial power distribution depression by the plate with a 40mm thickness in the fuel cluster is increasing toward the outer from the inner in the fuel cluster. The value of the ratio of the depression to the maximum power is about 5% in the inner rod and about 10% in the outer rod. (3)The depression of the thermal neutron flux distribution around the plate in the moderator or in the fuel channel is increasing with increase of the plate thickness. The obtained value of the depression for the core with a 40mm thickness plate is ...
Takemura, Morio*; *; *; *; *
PNC TN941 83-67, 96 Pages, 1983/04
Pressure tube type heavy water reactor has an advantage of load following operation compared with light water reactor. It is planned to prove the possibility of load following operation in the Fugen type demonstration reactor using stainless control-rod (SUS control-rod). Changes in local power distribution and thermal flux distribution due to a small withdrawal of SUS control-rod have been measured for the purpose of confirming the soundness of fuel pin from the viewpoint of the load following operation and fuel design. SUS control-rod (74mm
) with the same dimension as those of the demonstration reactor was inserted into the D
O moderator of the central region of 0.54wt% plutonium mixed-oxide fuel lattice in 25-cm pitch DCA core. The lower end height of the control-rod inserted was changed from 505mm to 605mm (about 40 change in reactivity). Experimental results for the local power change were compared with calculations obtained from WIMS-D and CITATION codes. The following were concluded from the present study. (1)The maximum power change in outer layer pins of fuel cluster due to a small withdrawal of the control-rod occurs in the nearest fuel pin at the middle position of the axial displacement of the control-rod. (2)Outer layer fuel pin power after 100mm withdrawal of control-rod is (1.12
0.03) times in maximum as large as that before withdrawal. (3)Local pin power change due to withdrawal of the control-rod occurs mainly in the fuel pins of the nearest and the second nearest fuel clusters to the control-rod. Even in the nearest fuel cluster to the control-rod, power change in the back side fuel pin to the control-rod is very small below 2%. (4)Calculated value of maximum ratio for the outer layer fuel pin power due to 100mm withdrawal of the control-rod overestimate experimental one about 4%.
C control rod; Measurements of change in power distribution in fuel pin close to control rodTakemura, Morio*; *; *; *; *; *
PNC TN941 82-93, 202 Pages, 1982/04
Changes in power distribution and thermal-neutron flux distribution due to the movement of control rod have been measured for the purpose of confirming the advantage of load following operation using control rod in a heavy-water-moderated, light-water-cooled, pressure-tube-type thermal reactor (HWR) from the viewpoint of fuel design. A BC control rod with the same dimensions as those of the Fugen was inserted into the DO moderator region of the central plutonium fuel lattice about half way. Experimental results were compared with calculations with use of codes WIMS-D and CITATION. It has been clarified that the maximum power change due to the movement of the control rod occurs in the nearest fuel pin at the axial middle point of the displacement of the control rod. In a lattice of 25.0 cm pitch, the maximum value of power change due to pulling out of the control rod by 10 cm has been estimated to be about 23 %. The axial power distribution in fuel pin in the nearest position of the control rod was predicted well by the codes WIMS-D and CITATION. The calculations in reference to the maximum value of power change are in good agreement with experimental results within 3 %.
Takemura, Morio*; *; *; *
PNC TN941 81-208, 55 Pages, 1981/10
Reactivity worths of BC, SUS and Cd control-rod have been measured in Deuterium Critieal Assembly (DCA) core of a 25-cm pitch lattice with non-voided HO coolant. In the core, twenty-five assemblies of 0.54 wt% Pu enriched mixed-oxide fuel were loaded in the central region and seventy-two assemblies of 1.2 wt% enriched UO fuel in the surrounding region. The fissile plutoniu content of PuO-UO fuel used in the experiment is 91 wt%. Experimental values of reactivity were obtained by integrating level reactivity coefficient of moderator between critical levels in withdrawal and in insertion of control-rod. The calculation of control-rod worths in the full insertion was carried out by use of WIMS-CITATION and accuracy obtained by using of these nuclear design codes was evaluated. Calculation of the lattice cell consisted of the control-rod region and its adjacent ordinary cell was done by means of the multi-cell option of the WIMS code. Fourteen group constants obtained by the multi-cell option of each region were rondensed into four group ones through one-dimensional diffusion calculation centering around the control-rod. Calculation of control-rod worths was made by two-dimensional X-Y geometry in the CITATION code using those four group constants. The calculational values of BC, SUS and Cd control-rod worth in the full insertion agree with each experimental one within 2 % uncertainty as following table. It was found that the introduction of spatial effect for neutron flux distribution in the group condensation procedure by use of the above diffusion calculation was found to lead to improvement of calculational accuracy of control-rod worth by 10 % compared with the ordinary condensation method. The control-rods having several kind of structure and absorber material were investigated to clarify their nuclear characteristics in comparison with the BC control-rod used in FUGEN. Those control-rod worths were ...
