• Energy Research

Abstract This report is the results of at-power reactor noise experiments conducted at JRR-4. The magnitude and the frequency spectrum of the fluctuation of neutron flux are analysed under various operating condition. Three characteristic results are obtained: (1) the magnitudes and the frequency components in the relative fluctuation of the neutron flux do not alter at various power levels; (2) decrease of flow rate makes the neutron flux fluctuation weak; (3) the change of the shim plate position also varies the frequency components in the fluctuation. Based on these results the cause of the neutronic noise of this reactor is inferred to be the flow induced reactivity perturbation especially due to the lateral vibration of a shim plate. Next, the vibration itself is measured simultaneously with the fluctuation of neutron flux under reactor operating condition. The correlation between these signals are examined. Very high coherence and extremely similar frequency spectra of both fluctuation are obtained and it is definitely shown that the cause of the noise in JRR-4 is the lateral vibration of the shim plate under forced circulation of coolant. It is noticeable that the frequency spectrum of the shim plate vibration have a broad spectrum at low frequency region in contrast to the common understanding that the frequency spectra of vibration of core components have only discrete frequency components. The method of inferring source of noise from the behavior of the neutronic fluctuation under several power levels, coolant flow rates and control rod positions is proved to be useful if it is difficult to measure the source of noise itself.

Abstract This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.

Design studies of supercritical-pressure light-water-cooled reactors (SCLWRs) have been carried out to pursue drastic improvement of the economy of nuclear power generation. The core is cooled by supercritical water which is superheated without the phase change. The cooling system is a once-through type; the whole core flow is driven by the feedwater pumps and is directly led to the turbine. No recirculation line is necessary. Besides, steam separators and dryers are not needed. Water rods are used to enhance the moderation and to increase the flow velocity around the fuel rods. The radial peaking factor is satisfactorily reduced by controlling uranium enrichment and gadolinia concentration as well as water rods. Flattening of the radial power distribution is important to enhance the thermal efficiency. This can be achieved by the coolant density feedback and the out-in refueling pattern. Orificing is also effective to enhance the thermal efficiency. The thermal efficiency is above 40% with stainless steel cladding. Plant control system and safety system are also designed. The core flow should be directly maintained due to the once-through direct cycle. Plant behaviors of large break LOCAs and loss of offsite power are analyzed. Safety criteria are satisfied in both cases. The feasibility of SCLWR is shown.

Abstract The steam-cooled fast breeder reactor with negative reactivity characteristics against voiding and flooding is feasible under the supercritical pressure. The breeding ratio is 1.04. A flat core with a zirconium hydride layer is adopted for mitigating the void reactivity. The thermal efficiency of the indirect cycle system is improved 9% relatively from current PWRs. The core should be cooled in 10 s after the large break LOCA. The coast down time should be larger than 30 s to overcome the LOF by the trip of all blowers.

Abstract Deterioration in heat transfer at supercritical water cooling in a vertical pipe is numerically analyzed. The calculation is based on a parabolic solver for steady-state equations in r−z two dimensions, a k−ϵ model for turbulence and a steam table library for physical properties of supercritical water. Calculation results agree with the experimental data of Yamagata et al. It is found that heat transfer deterioration is caused by two mechanisms depending on the flow rate. When the heat flux is increased much above the deterioration heat flux, a violent oscillation is observed in the temperature distribution.

Abstract Conceptual design studies of advanced power reactors are summarized. The concept of power reactors cooled by supercritical water was developed. The coolant system is once through type like a supercritical fossil-fired power plant. It attains higher thermal efficiency and simpler reactor system than the light water reactors. Breeding is possible in the tight lattice core. The coolant void reactivity of the fast reactors becomes negative by placing a thin zirconium hydride layer between seed and blanket. The basic safety principles follows those of LWRs. Heterogeneous cores of the liquid metal cooled fast breeder reactor was designed for assessing the effectiveness of coolant orificing. Overpower transients after withdrawing a peripheral control rod were analysed by a three-dimensional space-time kinetics code which was developed for the study. The space-time power variations were more influential on boiling behavior in the heterogeneous cores. Conceptual design of the fusion-fission hybrid reactors and the inertial confinement fusion reactor were described for comparing with the fission power systems. The differences in heat generation profiles and technical features are discussed.

Abstract Improvements of two-pass core design of Super critical-water cooled fast reactor (Super FR) are proposed. The core height has been shortened; assembly pitch is enlarged to take Control Rod (CR) design into account. Well-coupled fuel loading pattern is proposed, number of control rods location are limited only in second flow pass assemblies due to current flow pattern. Fuel shuffling is also confined within each own flow pass because of the different structure of assemblies. CR withdrawal is considered during the operation, satisfactory radial and axial power distribution can be achieved by implementing appropriate CR withdrawal strategy. Results suggest that all the design target, criteria and limits are satisfied in terms of average coolant outlet temperature, maximum linear heat generation rate (MLHGR), maximum cladding surface temperature (MCST) as well as core shutdown margin, negative coolant void reactivity coefficient and positive coolant density reactivity coefficient in all coolant density range.

Abstract Measurement was made of the reaction rate distributions for 238U(n, f), 235U(n, f), 238U(n, γ), 27Al(n, α) and 58Ni(n, p) in a large depleted uranium (DU) pile. The pile consisted of DU blocks forming a spherical shell of 45.72 cm radius and 40.64 cm thick. 14-MeV neutrons were generated at the center. Fast neutron leakage spectrum was also measured by an NE-213 spectrometer. In order to assess the 238U neutron cross sections of JENDL-2, the experiment was analyzed using the Monte-Carlo transport code MCNP with continuous energy cross sections. The agreement between the calculation and the experiment is generally unsatisfactory. The ratios of calculation to experiment of low energy reactions decreased with the thickness of the DU layer. The analysis of the Livermore pulsed sphere experiment for the small DU sphere revealed underestimation of leakage neutron spectrum around 10 MeV. The 238U cross sections of JENDL-2 should be improved for 14-MeV neutronics calculation.

Abstract The subchannel analysis with turbulent mixing rate law of supercritical pressure fluid (SPF) is carried out for supercritical-pressurized light water cooled and moderated reactor (Super LWR). It is different from the turbulent mixing rate law of pressurized water reactor (PWR), which is widely adopted in Super LWR subchannel analysis study, the density difference between adjacent subchannels is taken into account for turbulent mixing rate law of SPF. MCSTs are evaluated on three kinds of fuel assemblies with different pin power distribution patterns, gap spacings and mass flow rates. Compared with that calculated by employing turbulent mixing rate law of PWR, the increase in MCST is smaller even when peaking factor is large and gap spacing is uneven. The sensitivities of MCST on non-uniformity of the subchannel area and power peaking are reduced.