A new double-energy double-velocity measurement system for the thermal neutron-induced fission fragments was developed. In this system, the energies of fission fragments were measured by silicon surface barrier detectors (SSB) and the velocities of them by time-of-flight method utilizing thin film detectors as start detectors and SSBs as stop detectors. With this system, the energies and the velocities of the thermal neutron-induced fission fragment of ²³³U were measured. The prompt neutron distribution was determined from the experimental data.

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A knowledge based system has been developed to support designers in control rod programming of BWRs. The programming searches through optimal control rod patterns to realize safe and effective burning of nuclear fuel. Knowledge of experienced designers plays the main role in minimizing the number of calculations by the core performance evaluation code. This code predicts power distribution and thermal margins of the nuclear fuel.
This knowledge is transformed into 'if-then' type rules and subroutines, and is stored in a knowledge base of the knowledge based system, The system consists of working area, an inference engine and the knowledge base. The inference engine can detect those data which have to be regenerated, call those subroutine which control the user's interface and numerical computations, and store competitive sets of data in different parts of the working area.
Using this system, control rod programming of a BWR plant was traced with about 500 rules and 150 subroutines. Both the generation of control rod patterns for the first calculation of the code and the modification of a control rod pattern to reflect the calculation were completed more effectively than in a conventional method.

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The air entrainment mechanisms and rate at the free surface of water were investigated experimentally for a drawing flow into a vertical suction pipe in a cylindrical test vessel. It was found that the flow mechanisms of air entrainment were grouped into three typical types ; (1) the vortex-induced entrainment, (2) the conically depressed surface with entrainment of relatively large bubbles, and (3) the falling flow along the side wall of the suction pipe with entrainment of fine bubbles. In the absence of a forced circulation, the air flow rate of the vortex-induced entrainment was well correlated by the Froude number. The proposed analytical model for the air entrainment rate of the vortex-induced entrainment agreed well with the experimental results in the presence of a forced circulation. The transition mechanisms between three types of entrainment were investigated, and it was found that the entrainment rate has the minimum value in the transition between the vortex-induced entrainment and the falling flow due to the stability of the air core.

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The natural circulation core cooling is important during small break loss-of-coolant accidents and transients involving loss of forced-circulation core cooling. This paper presents results of natural circulation experiments conducted in the ROSA-III (Rig of Safety Assessment No. 3) facility under reduced mass inventory conditions. Three distinct modes of BWR natural circulation were identified in the experiments : the primary circulation mode, the internal circulation mode and the open-loop (core uncovery) mode. One-dimensional quasi-steady analytical model was developed for the internal natural circulation mode to determine the core uncovery criterion. The model predicted well the two-phase mixture level inside the core shroud for quasi-steady state natural circulation at pressures of 7.35 and 2.06 MPa, and scaled core powers below 20%, and was used to predict BWR natural circulation behaviors.

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A study was made to investigate the transgranular fracture mechanism of zirconium in an iodine environment using the data from stress corrosion cracking (SCC) tests of irradiated zirconium-lined tubes and the analyses of the force on dislocations in zirconium liner during SCC testing.
Scanning electron microscope (SEM) and metallurgical observations on incipient cracks indicated that the transgranular fracture, or cleavage facet, might occur near the prism planes {1010} of the hexagonal close-packed (hcp) lattice.
Large failure strain of zirconium-lined tubes suggested that plastic strain might be prerequisite for the cracking of zirconium in an iodine environment. And it also indicated that the cracking might be a kind of corrosion assisted one and not pure SCC. Therefore, an analysis was made to study available slip modes for the cracking. The results showed that a slip of the dislocation with Burgers vector b= 1/3 <1120> on {1010} might be responsible for the microcrack formation of zirconium.

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Thermal wave signals from nickel ion deposits produced on a heated rod surface in boiling water at atmospheric pressure was measured by two kinds of thermal wave methods, photothermal deflection and photoacoustics. Thickness of the deposits was determined from calibration curves for NiO which was their possible chemical state. It was revealed that the methods served as useful tools for in situ thickness measurement in high temperature water.
Nickel ion was deposited in circular patterns (about 1 mmφ) at boiling nucleation sites. The deposit thickness (0.11.1 μm) was proportional to time (020 h) and nickel ion concentration (2050 ppm). This indicated that nickel ion was deposited by the evaporation-dryout mechanism, the same as for iron crud deposition in BWR's. The deposition rate for nickel ion (2.7 × 10^-3) was about 1/6 of that for iron crud. The difference was attributed to dissolution of nickel ion from NiO.

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The migration behavior of radionuclides (⁶⁰Co, ⁸⁵Sr and ¹³⁷Cs) in aquifer has been studied under a steady flow of water by using aquifer model apparatus. In the portion of high concentration ( ?? 10^-3 μCi/g), water flow has slight influence on migration of these radionuclides. On the other hand, in the portion of low concentration (10^-4 μCi/g), ⁶⁰Co and ¹³⁷Cs are particularly influenced by the flow, but the migration of ⁸⁵Sr is relatively small. Such phenomena support that the migration of ⁸⁵Sr is subject to the ion exchange reaction, and the migration of ⁶⁰Co and ¹³⁷Cs may be due to both ion exchange reaction and particulate moving mechanism.
The concentration distribution observed is compared with that of the calculation by predicting equation based on ion exchange reaction. In the distribution of high concentration portion, the observed on each radionuclide is almost same as to the calculated, but in the distribution of low concentration portion, both observed on ⁶⁰Co and ¹³⁷Cs are very different from the calculated. Since migration behavior of ⁶⁰Co and ¹³⁷Cs cannot be described by predicting equation only based on ion exchange reaction, it is, therefore, necessary to predict those migration by the equation consisting of ion exchange reaction and particulate moving mechanism.

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The effects of emergency core coolant (ECC) temperature on the performance of the emergency core cooling system (ECCS) during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) were investigated experimentally using the Rig of Safety Assessment (ROSA)-III integral test facility. The ECC temperature had no direct influence on the ECCS core cooling performance, since ECC became nearly saturated before reaching the core irrespective of the initial temperature, however, had indirect effects by changing the vessel pressure response. The ECCS injection timings and flow rates, and the core inlet flooding behavior were affected. The measured peak cladding temperature (PCT) was not affected by the ECC temperature for both large (200%) and small (5%) break tests.

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An experimental study of upward air-water bubbly two-phase flow in an entry region was performed with various rough wall test tubes. The objective of the work is to clarify the effects of wall roughness and entry length on void profile. The fluid flows in the vertical circular test tube of 25 mm I.D. under nearly atmospheric pressure, at room temperature. The void profile changes from a pattern similar in appearance to the saddle shape which has local void peaks near the wall, into the power law shape whose curve is approximated by a power law formula, with increasing wall roughness and/or entry length. That is, wall roughness and entry length have a similar effect upon void profile. There are two patterns in the power law shape, a pattern with sharp center peak and a pattern with obtuse center peak. As wall roughness and/or entry length increase, the void profile changes from the former pattern to the latter pattern. At enough long entry length (L/D ?? 150), every void profile has almost the same power law shape independent of wall roughness. Some void profiles are asymmetric to the axis.

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In the analysis of the core thermal-hydraulic behavior during the reflood phase of a PWR-LOCA, current safety evaluation codes like WREM code system are usually limited to use in narrow region where the employed empirical correlations are validated. In order to make a safety evaluation code more flexibly applicable, the empirical correlations in TOODEE2 code in the WREM code system was replaced with the core model built in the REFLA code. By changing the multiplication factor for the calculated heat transfer coefficient for the region above the quench front, the predicted clad surface temperatures were compared with those measured in Cylindrical Core Test Facility (CCTF) tests.
It was found that the multiplication factor 0.9 gives always a conservative prediction against CCTF data.

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A facility was reconstructed for producing tritium in 40 TBq per batch. Gaseous tritium was extracted from neutron-irradiated ⁶Li-A1 alloy targets by heating at 700°C under vacuum and collected in uranium. The recovery yield of tritium was about 100% and the isotopic purity of the product was about 95%. Through the production run, no leakage of tritium from the facility was observed.

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