We studied a new detection algorithm for the oscillation power range monitor (OPRM) system to suppress postulated neutronic/thermal-hydraulic instability events in boiling water reactors (BWRs) before exceeding the fuel cladding integrity limit. The new method evaluates the standard deviation for OPRM oscillatory periods instead of monitoring them directly. Instability is detected when the standard deviation remains lower than a setpoint for a certain elapsed time. This method aims at excluding the impact of the perturbation caused by OPRM noise components for each OPRM cell to the detection performance, utilizing statistics that represent the oscillatory characteristics for the instability criterion.
Toshiba's BWR system analysis code TRACT simulated the OPRM signals for an up-rated Advanced Boiling Water Reactor (ABWR) regional instability event to examine the applicability of the new algorithm to this new reactor. Since there are no plant instability data for ABWR, we estimated noise components based on the available stability test data and superimposed them on the simulated oscillation data. The new algorithm showed that the trip signal could be generated for the simulated regionally instable condition before the fuel cladding safety limit was reached.

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In a sodium-cooled fast reactor (SFR), if a heat transfer tube in the steam generator (SG) is failed, high pressurized water vapor blows into the liquid sodium and sodium-water reaction (SWR) takes place. SWR may cause damage to the neighboring heat transfer tubes by thermal and chemical effects. Therefore, it is important to clearly understand the SWR for safety assessment of the SG. From recent study, sodium (Na)-sodium hydroxide (NaOH) reaction as secondary surface reaction of the SWR phenomena in a SFR was identified by ab initio (first-principle) method. However, kinetics of this reaction is a still open question. Thus, it is important to obtain quantitative rate constant of sodium monoxide (Na₂O) generation by Na-NaOH reaction because Na₂O may accelerate the corrosive and erosive effects. Differential thermal analysis (DTA) provides us with the valuable information on the kinetic parameters, including activation energy, pre-exponential factor (frequency factor) and reaction rate constant. In this work, kinetic study of Na-NaOH reaction has been carried out by using DTA technique. Based on the measured reaction temperature, the first-order rate constant of Na₂O generation was obtained by the application of the laws of chemical kinetics. From the estimated rate constant, it was found that Na₂O generation should be considered during SWR. The results can be the basis for developing a chemical reaction model used in a multi-dimensional sodium-water reaction code, SERAPHIM, being developed by the Japan Atomic Energy Agency (JAEA) toward the safety assessment of the SG in a SFR.

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A lab scale trickle bed reactor (TBR) has been employed to investigate the effect of particles and liquid flow rate on the phase distribution in TBR. The TBR module, made up of 100 mm inner diameter acrylic column, was randomly packed with inert porous alumina particles of different average diameters that are used in actual reactor. Water was fed from the top of the column by “point feed” and “homogenous feed” strategies and cross-sectional liquid distributions were captured at several axial positions from the top to the bottom of the column by employing electrical resistance tomography (ERT) technique. ERT is a non-invasive cross-sectional imaging technique that provides the cross-sectional conductivity distribution by injecting current and measuring voltages between the several electrodes (16 electrodes in our case) that are attached around the column. The cross-sectional conductivity thus obtained represents the liquid hold up and degree of maldistribution of the liquid. In the experiment, quicker and more homogenous distribution of liquid was obtained for the particles with smaller diameters. That is due to capillary force that cancels the randomness of packing. Electrical resistance tomography seems to be reliable non-invasive instrumentation technique to optimize the design and operations of the trickle bed reactors.

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In order to establish a safety evaluation method of a steam generator of sodium-cooled fast reactors, a computer program called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction under tube failure accident has been developed. In this study, a numerical model for liquid droplet entrainment from an interface of the gaseous jet and its transport was newly constructed to evaluate the environment of the liquid droplet impingement erosion. The applicability of the SERAPHIM program which incorporates the droplet entrainment / transport model was investigated through the analysis of vertical discharging of water vapor in the liquid sodium pool under the actual condition of the steam generator. The analysis reproduced appearance of the underexpanded vapor jet. The calculated peak temperature agreed with the experimental results well. Also, appearance of the dispersed phase of droplets in the reacting jet and its acceleration by the supersonic gaseous flow were represented successfully.

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The transient critical heat fluxes (transient CHFs) in a SUS304-circular tube with twisted-tape insert are systematically measured for mass velocities (G=3997.79 to 13419.8 kg/m²s), inlet liquid temperatures (T_in=293.55 to 300.85 K), outlet pressures (P_out=825.19 to 860.95 kPa) and exponentially increasing heat inputs (Q=Q₀exp(t/τ), τ=26.85 ms to 8.42 s) by the experimental water loop comprised of a multistage canned-type circulation pump controlled by an inverter. The SUS304 test tube of inner diameter (d=6 mm), heated length (L=59.4 mm), effective length (L_eff=49.4 mm), L/d (=9.9), L_eff/d (=8.23) and wall thickness (δ=0.5 mm) with average surface roughness (Ra=3.89 µm) is used in this work. The SUS304 twisted-tape with width (w=5.6 mm), thickness (δ_T=0.6 mm), total length (l=372 mm) and twist ratio, y [=H/d=(pitch of 180° rotation)/d], of 3.37 is used. The transient CHFs for a SUS304-circular tube with twisted-tape insert are compared with authors' steady-state CHF data for a SUS304-circular tube with various twisted-tape inserts, their transient CHF data for the empty SUS304-tube and the values calculated by authors' steady-state CHF correlations for the test tubes with various twisted-tape inserts and their transient CHF correlations for the empty test tubes. The influences of twisted-tape insert, heating rate and swirl velocity on the transient CHF are investigated into details and the widely and precisely predictable correlations of the transient CHF for the test tube with twisted-tape insert are given based on the experimental data. The correlations can describe the transient CHFs for a SUS304-circular tube with twisted-tape of y=3.37 obtained in this work within -27 to 7.9 % difference.

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