Kazuo Haga

Sodium boiling experiments in a 19-pin bundle under loss-of-flow conditions

Abstract An experimental study was conducted on transient sodium boiling in a 19-pin electrically heated LMFBR fuel subassembly mockup under loss-of-flow conditions. In each run the inlet flow was reduced or stopped at constant heater power. There was no strong effect of temperature ramp rate on incipient-boiling (IB) wall superheat. The observed coolant voiding was initially limited to the center subchannel because of steep temperature gradient in the bundle. The bulk pressure rise registered upon initial vaporization was markedly lower than the vapor pressure corresponding to the IB wall superheat. The pressure pulse generated at vapor bubble collapse correlated reasonably well with the re-entrant liquid velocity, but the measured value was very much smaller than the calculation by sodium hammer analysis.

Equilibrium and Nonequilibrium Partition Coefficients of Volatile Fission Products between Liquid Sodium and the Gas Phase

Two series of experiments have been conducted to obtain the gas-liquid equilibrium partition coefficient Kd and the nonequilibrium partition coefficient K’d of volatile fission products such as cesium, iodine, and tellurium between liquid sodium and the gas phase. In the equilibrium experiment, a sodium pool mixed with a fission product simulant was heated by an electric furnace, and the solvent of the vapors and aerosols trapped by filters was quantitatively analyzed. The results are as follows: 1. Cesium shows the largest Kd (20 to 100).2. The Kd value of iodine scatters as widely as 0.02 to 0.5 at 450°C and 0.3 to 0.8 at 650°C.3. The Kd values of cesium and iodine agree well with the theoretical ones reported by Castleman and Tang.4. If sodium telluride, which is harder to vaporize than pure tellurium, is assumed, the measured Kd value of tellurium agrees with the theoretical. The nonequilibrium experiment in which the temperature dropped relatively sharply in the cover-gas region shows that K’d was no...

Loss-of-flow experiment in a 37-pin bundle LMFBR fuel assembly simulator

Abstract An experimental study was conducted on transient sodium boiling in an LMFBR fuel subassembly mockup under loss-of-flow conditions. In the test section, an electrically heated 37-pin bundle was centered in a hexagonal tube. The measured maximum IB wall superheat was 36°C, and the effects of heat flux, temperature rise rate, and system pressure were unclear. Boiling was initiated at the end of the heated section, the bubble expanded mainly to the upstream central subchannels and to the downstream unheated section according to the expansion of the saturated temperature region. When the voided zone covered the whole flow cross-section, the void pattern changed to the one-dimensional slug ejection-type and the inlet flow decreased rapidly. Dryout occurred after the inception of flow reversal in the wide region of the bundle.

Sodium aerosol release rate and nonvolatile fission product retention factor during a sodium-concrete reaction

This paper reports on a series of tests conducted to study the mechanical release behavior of sodium aerosols containing nonvolatile fission products during a sodium-concrete reaction in which release behavior due to hydrodynamic breakup of the hydrogen bubble is predominant at the sodium pool surface. In the tests, nonradioactive materials, namely, strontium oxide, europium oxide, and ruthenium particles, whose sizes range from a few microns to several tens of microns, are used as nonvolatile fission product stimulants. The following results are obtained: The sodium aerosol release rate during the sodium-concrete reaction is larger than that of natural evaporation. The difference, however, becomes smaller with increasing sodium temperature: nearly ten times smaller at 400{degrees} C and three times at 700{degrees} C. The retention factors for the nonvolatile materials in the sodium pool increase to the range of 0.5 to 10{sup 4} with an increase in the sodium temperature from 400 to 700{degrees} C.

Experimental investigation of sodium boiling in partially blocked fuel subassemblies

Abstract Out-of-pile experiments were conducted to examine local sodium boiling in two wire-wrapped 37-pin bundles simulating LMFBR fuel subassemblies. The central 24 subchannels were blocked in the first bundle, and a 1 2 edge part of bundle cross-sectional area was blocked in the other. The boiling modes observed in the wake were irregular nucleate, oscillatory and stationary. Three different types of boiling transition were identified, which were characterized by the temperature gradient around the two-phase voiding region. Dryout occurred not in the irregular nucleate boiling mode, but in both the stationary and oscillatory boiling modes. The coolability margin is considered to be 20 to 30% in terms of power-to-flow ratio from the incipient boiling to the occurrence of dryout.

Experimental investigation of coolability degradation by fission gas release into flowing sodium in a fuel pin bundle

A series of experiments was performed to assess the thermal effect of a burst-type fission gas release from fuel pins. Simulated fission product gas was injected continuously and transiently from the central pin of a 37-pin bundle. The opposite pin surface impinged on by the released gas showed an extreme temperature rise under high coolant-flow conditions. Comparison of measured temperature change data with analytical results by a simple computer code revealed that the ratios of the heat transfer coefficient after gas injection to those of sodium single-phase flow were in the range of 0.05 to 0.15, irrespective of the magnitude of the gas plenum pressure and the nozzle diameter. The estimated pin-surface temperature increased by gas release in actual reactor operating conditions was less than the saturation temperature of sodium. The measured pressure pulse at the transient gas release was < 0.2 times the initial gas plenum pressure.

Experimental study on the effect of fission product gas release into blockage wake region using a simulated LMFBR fuel subassembly

Abstract The objective of this study is to evaluate temperature rise due to gas release in the wake region of LMFBR fuel subassemblies. The experiments were conducted in two sets of grid-spacer-type 37-pin bundles simulating LMFBR fuel subassemblies. In Test section 37GC, the central 24 subchannels were blocked by a stainless steel plate and in Test section 37GE one-half edge part (39 subchannels) of the total flow area was blocked by the same material. The experimental results were compared with data obtained in similar tests using a spacer wire-type pin bundle, designated 37WC. The temperature rises in 37GE and 37WC were nearly identical in value and effect of gas release rate. The marked agreement seems to imply that there is a limit in the content of released gas in the wake region. On the other hand, the temperature rise behind the central blockage in the grid-type bundle, where gas might easily flow out to the core flow region, was far less than in the other geometries.

Validation of RELAP5/3D Steam-Generator Analysis Model by Turbine Trip Test at the Prototype Fast Breeder Reactor MONJU

In order to develop a thermal-hydraulic analysis model of the SG (steam-generator) to simulate transient phenomena in the FBR (sodium cooled fast breeder reactor) MONJU, JNES (Japan Nuclear Energy Safety Organization) validated a SG analysis model using the RELAP5/3D code against the results of turbine trip test at a 40% power load. The modeling by using the code was performed to explain main observed behaviors of the pressure and the temperature of the EV (evaporator) steam outlet, and the temperature of water supply distributing piping (near the water supply chamber) until 600 seconds after the turbine trip. The analysis results of these behaviors showed good agreement with the test results on parameter study of the blow efficiency (release coefficient.). It was found that RELAP5/3D with a two-fluid model of water/steam and a sodium fluid model for tube outside predicted well the physical behaviors: the void of steam generated by the depressurization boiling moves upward in the down-comer tubes accompanied by the enthalpy, increases the temperature of the water supply chambers; and that the pressure change of a “shoulder” like shape is induced by the mass balance between the steam mass generated in the down-comer tubes and the steam mass blown from the SG. The applicability of RELAP5/3D to the SG modeling was confirmed by the well simulation of the actual FBR system.Copyright

RELAP5/MOD3 Analysis of Transient Steam-Generator Behavior During Turbine Trip Test of a Prototype Fast Breeder Reactor MONJU

In order to develop a thermal-hydraulic model of the steam-generator (SG) to simulate transient phenomena in the sodium cooled fast breeder reactor (FBR) MONJU, Japan Nuclear Energy Safety Organization (JNES) verified the SG model using the RELAP5/MOD3 code against the results of the turbine trip test at a 40% power load of MONJU. The modeling by using RELAP5 was considered to explain the significant observed behaviors of the pressure and the temperature of the EV steam outlet, and the temperature of water supply distributing piping till 600 seconds after the turbine trip. The analysis results of these behaviors showed good agreement with the test results based on results of parameter study as the blow efficiency (release coef.) and heat transferred from the helical coil region to the down-comer (temperature heating down-comer tubes). It was found that the RELAP5/MOD3 code with a two-fluids model can predict well the physical situation: the gas-phase of steam generated by the decompression boiling moves upward in the down-comer tubes accompanied by the enthalpy increase of the water supply chambers; and that the pressure change of a “shoulder” like shape is induced by the mass balance between the steam mass generated in the down-comer tubes and the steam mass blown from the SG. The applicability of RELAP5/MOD3 to SG modeling was confirmed by simulating the actual FBR system.Copyright