A new computer code PREC was developed to solve neutron and nuclide density distributions at the equilibrium cycle of pebble bed reactors. The PREC code has the following special advantages :
(1) To provide a direct solution of the equilibrium cycle
(2) To fix the effective multiplication factor as an input
(3) To treat continuous fuel movement
(4) To treat r-z two-dimensional geometry, leading in turn to the following special advantages :
(4-1) Ability to treat the cavity at the top of the core
(4-2) Ability to treat the curved fuel stream-line.

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Effect of cell simplification on cell average constants are evaluated by a Monte Carlo method in order to provide a simplified very high temperature reactor (VHTR) design method. First, the applicability of the Monte Carlo method to the present work is verified by the experiment analysis of the VHTR mock-up critical core. Then a technique to separate the volume and surface terms of resonance integral based on the Monte Carlo method is developed by introducing Marchuk theory. The percentage contributions to the resonance integral of the VHTR fuel assembly are 60% from the surface term and 40% from the volume term. Finally, due to simplification of the hexagonal SHE-14 cell to an annular cell with equal fuel volume, the K_∞-value is enlarged by 0.8% Δk/k as the result of a 7% decrease of surface area, i.e. 0.7% increment of the resonance escape probability. The contributions of individual resonances of ²³⁵U and ²³⁸U to these terms are also discussed.

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In order to develop the hydraulic model of the ECC water bypass and the refill period at a PWR-LOCA, flashing transient and the downcomer Counter Current Flow Limit (CCFL) experiments were conducted with the large scale Cylindrical Core Test Facility (CCTF). The ECC water was bypassed by the two phase mixture swelled from the lower plenum to the downcomer at the flashing transient experiments. This swelling behavior was predicted well by using the void fraction correlation proposed by Okabe & Murao. The CCFL correlation based on Battelle experiments predicted well the bypass of the ECC water and its penetration into the lower plenum at the CCFL experiment.
These swelling and CCFL models were combined to form the best estimate analytical model which was applied to a large break LOCA of a commercial PWR. Calculated results showed that (1) the proposed model predicted the existence of water at the initiation of the refill, (2) the present licensing calculation was conservative because it assumed no water at this time, and (3) this conservativeness mainly came from the neglect of the water in the downcomer at the ECC water bypass period.

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Reflooding tests were conducted in a rod bundle geometry at the maximum pressure of 12 MPa to investigate thermal-hydraulic behavior during a small break loss-of-coolant accident (SBLOCA) in a nuclear reactor. The test conditions ranged 0.6 12 MPa for pressure, up to 920 K for rod surface temperature, up to 20 cm/s for bundle inlet flow velocity and up to 2 kW/m for linear power input. The principal objective of this paper is to investigate the onset condition for liquid entrainment by steam flow in the relatively high pressure reflooding phase. Experimental results showed a tendency that the liquid entrainment decreased with increase in pressure when the other parameters such as an inlet flow rate and rod temperature were fixed. A new correlation for the onset criterion for liquid entrainment was derived from the experimental results and an analysis of a force balance for a liquid droplet. Effects of pressure on liquid entrainment in the reflooding phase were made clear from the experimental and analytical results.

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Burnout characteristics were experimentally studied using uniformly heated tube and annular test sections under rapid flow reduction conditions. Observations indicated that the onset of burnout under a flow reduction transient is caused by the dryout of a liquid film on the heated surface. The decrease in burnout mass velocity at the channel inlet with increasing flow reduction rate is attributed to the fact that the vapor flow rate continues to increase and sustain the liquid film flow after the inlet flow rate reaches the steady-state burnout flow rate. This is because the movement of the boiling boundary cannot keep up with the rapid reduction of inlet flow rate. A burnout model for the local condition could be applied to the burnout phenomena with the flow reduction under pressures of 0.5 3.9 MPa and flow reduction rates of 0.6 35%/s. Based on this model, a method to predict the burnout time under a flow reduction condition was presented. The calculated burnout times agreed well with experimental results obtained by some investigators.

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The effect of the decay heat on the reflood phenomena in a PWR-LOCA were investigated under a large-scale facility, by using the Cylindrical Core Test Facility (CCTF). Two tests were performed at the initial average linear power of 1.06 and 1.4 kW/m.
It was found that the core inlet boundary condition (the core flooding mass flow rate, the pressure and the subcooling) being important for core cooling was little influenced by the variation of the linear power for the entire reflood period. It was also found that the core cooling behavior (the heat transfer coefficient) was little influenced for the early reflood period (until the turnaround time), and the heat transfer coefficient was lower in the higher power test for the later reflood period, indicating that assuming the higher power is conservative for safety evaluation.
The reason of the nearly identical core flooding mass flow rate and the nearly identical heat transfer coefficient for the early period regardless the different linear power is that the in-core steam generation rate is contributed mostly by the release of the initial stored energy rather than the supplied linear power and that the initial stored energy was identical between both tests.
The insensitiveness of the core flooding mass flow rate and the heat transfer coefficient for the first period against the decay heat could be extrapolated to a PWR scale, since the insensitiveness was commonly observed in the present large scale reflood experiments and the small scale reflood experiments, FLECHT.
In addition, the insensitiveness of the core flooding mass flow rate against the decay heat was analytically confirmed.

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Knoop hardness was measured in order to study plastic anisotropy of irradiated fuel cladding tubes. Results are summarized as follows. The lowest Knoop hardness was obtained when the indenter was so oriented that {1010} <l210> prism slip played the dominant role in indentation formation. The dependence of Knoop hardness on neutron fluence implied that the damage caused by neutron irradiation affected the prism slip of α-zirconium more strongly. The considerable amount of plastic anisotropy observed in the unirradiated tubes showed a tendency to decrease with increasing neutron fluence.

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Empirical formulas for the 14 MeV (n, p) and (n, α) cross sections given by Levkovskii were modified separately in three ranges of mass number, in each of which, coefficients modifying Levkovskii's formulas were determined by least-squares fitting to experimental cross sections. The resulting modified formulas yielded cross sections representing markedly smaller chi-square deviations from experimental values, and moreover gathered closer to unity, compared with calculation using Levkovskii's original formulas.

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A pressure control system failure test series was conducted at the Rig of Safety Assessment (ROSA)-III test facility to evaluate the effect of the pressure control system on thermal-hydraulic phenomena during a small break loss-of-coolant accident (LOCA) of a boiling water reactor (BWR). The break was assumed at the recirculation pump suction line. The pressure control system had no effect for breaks greater than 5%. For breaks less than 5%, if the pressure control system was inactive, the core was uncovered temporarily because of the bubble collapse due to pressure rise by the closure of the main steam isolation valve (MSIV), and the fuel rod surface temperature rose high during this period. However, the peak cladding temperature (PCT), which occurred mainly during the later core uncovering by boil-off, was lower in a LOCA with pressure control system failure than in a corresponding LOCA with intact pressure control system. This is because the emergency core cooling systems (ECCSs) were actuated earlier in a LOCA with pressure control system failure due to lower system pressure. The PCT was well below the present safety criteria of 1, 473 K even if the pressure control system and high pressure core spray system (HPCS) (the severest single failure in ECCS) were assumed to be inactive.

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