Nuclear analysis was carried out for the Heliotron-H reactor. Due to the helical geometry of the system, 14 MeV neutron streaming inherent in the reactor exists; some of source neutrons generated in the plasma region can see the coil shield surface in a direct line-of-sight manner and stream through blanket regions without collisions.
With an appropriate coordinate transformation from the helical geometry, analysis on the neutron streaming was made possible by a method coupling the three-dimensional Monte Carlo code and two-dimensional discrete-ordinates transport code.
For each radiation response parameter in the helical magnet, the streaming neutrons contribute 4050%. This fact strongly suggests the need of taking account of the contribution from the streaming neutrons to provide an adequate shielding for the helical magnet.

View full abstract

An accident diagnostic system DISKET has been developed to identify the cause and the type of an abnormal transient of a nuclear power plant. The system is based on the knowledge engineering (KE) and consists of an inference engine IERIAS and a knowledge base. The main features of DISKET are the following :
(1) Time-varying characteristics of transients can be treated.
(2) Knowledge base can be divided into several knowledge units to handle a lot of rules effectively.
(3) Programming language UTILISP, which is a dialect of LISP, is used to manipulate symbolic data effectively.
For the verification of DISKET, performance tests have been conducted for several types of accidents. The knowledge base used in the tests was generated from the data of various types of transients produced by a PWR plant simulator. The results of verification studies showed a good applicability of DISKET to reactor accident diagnosis.

View full abstract

The applicability of the thermal response of an electrically heated simulated rod mostly used in loss-of-coolant-accident (LOCA) experiments to that of a nuclear fuel rod is a concern for the safety evaluation of a reactor. The present analysis describes the characteristics of the thermal response for both electrically heated and nuclear fuel rods during typical reflood conditions for a PWR-LOCA. A model describing the radial temperature field in the rod is developed based on the scheme in HETRAP code by Malang and incorporated into a reflood analysis code, REFLA for that purpose.
The calculations applied to the existing reflood tests gave good agreement with experiments, showing the validity of the present model. The analysis has shown that the nuclear fuel rod tends to give a lower clad temperature and a sooner quench time than the electrically heated rod in a typical reflood condition, due to the smaller gap heat transfer and smaller heat capacity of the cladding. The thermal response of nuclear fuel rods was found to be well simulated by the use of an electrically heated rod with Zircaloy cladding and a clad-to-fuel gap.

View full abstract

Magnetic susceptibility was measured on the samples of U_1-xGd_xO₂ prepared by a coprecipitation method from 1.7 K to room temperature. The behavior of magnetic susceptibility is contributed almost from trivalent gadolinium ion because of its very large spin magnetic moment of ⁸S. For x<40%, the magnetic susceptibility-temperature curve obeyed Curie-Weiss law and values of magnetic susceptibility increase linearly with increasing gadolinium concentrations. For the higher concentrations of gadolinium the magnetic susceptibility-temperature curve has a peak at 34 K, showing a magnetic field dependence of magnetic susceptibility at liquid helium temperature which would be expected for the case of a weak magnetic anisotropy in antiferromagnets.

View full abstract

Isotope effects in the lithium amalgam formation were studied by using the mercury cathode and LiOH solutions. The electrolyses were carried out at different conditions of both applied voltages 4-10 V, and the concentrations of initially charged LiOH, mol/dm3. Higher efficiency of electrolytic amalgam formation was observed at higher LiOH concentrations and higher applied voltages. At these conditions, however, significant amount of solid amalgam was produced in the mercury phase. From the isotopic analyses on the samples taken during the electrolyses, it was found that the isotopic equilibrium was attained between the aqueous and the liquid amalgam phases. The isotopic equilibrium constant (isotope separation factor) was determined as 1.056 (average value for all the experiments) at 20°C.

View full abstract

The methods of estimation of the exchange capacity of mixed bed demineralizer used in nuclear power plants were studied for the purpose of protection against contamination of condensed water with leaked sea water flowing as a coolant in the heat exchanger. It was identified by numerical calculations and experimental works that the performance of mixed bed could be represented by the model for a system of single sort of ion exchange resin and single solute.
The intraparticle diffusivities and ion exchange equilibrium constants for the monitoring minicolumns can be obtained experimentally with the breakthrough curves for different packed height and the equilibrium constant for 1 m packed height can be known by extrapolation. The constants for 1 m packed height are substituted to the theoretical approximate formula which has been already recognized as a solution for the transient behavior of the adsorption system of a single solute, then the breakthrough time for actually working column of 1 m packed height can be calculated. For the rough estimation of degree of degradation for the deteriorated resin, a simple prediction method was presented in which the relations between the initial concentrations of the effluent from the minicolumns and the packed heights were applied.

View full abstract

Transient thermal-hydraulic behaviors of the JRR-3 which is an open-pool type research reactor has been analyzed with the THYDE-P1 code. The focal point is the thermal-hydraulic behaviors related to the core flow reversal during the transition from forced circulation downflow to natural circulation upflow. In the case of a loss-of-coolant accident (LOCA), for example, the core flow reversal is expected to occur just after the water pool isolation from the primary cooling loop with a leak. The core flow reversal should cause a sudden increase in fuel temperature and a steep decrease in the departure-from-nucleate-boiling ratio (DNBR) and the phenomenon is, therefore, very important especially for safety design and evaluation of research reactors.
Major purposes of the present work are to clarify physical phenomena during the transient and to identify important parameters affecting the peak fuel temperature and the minimum DNBR. The results calculated with THYDE-Pl assuming the sequences of events of the loss-of-offsite power and LOCA help us to understand the phenomena both qualitatively and quantitatively, with respect to the safety design and evaluation.

View full abstract

In summary, shield optimization calculations have been carried out for the choke coil shield of the GAMMA R. We could obtained the shield distribution in which an expensive material, tungsten, is effectively used to protect the choke coil. This optimal distribution could not have been obtained by varying the homogenized composition parametrically. Such an optimized distribution can decrease the amount of tungsten by 40% compared to the original homogeneous shield, while retaining the same dpa rate in the SC coil.
These results provide the basic information for further study of the choke coil design for the GAMMA R.

View full abstract

Liquid potassium is considered a promising candidate for adopting as heat transfer medium in nuclear fusion and liquid metal fast breeder reactors, and for this reason a number of studies⁽¹⁾ have been reported on its transport properties such as thermal conductivity and kinematic viscosity. Relatively little attention, however, has so far been given to the radiative properties of this material, despite the indispensable nature of such knowledge for theoretically evaluating radiation heat transfer between surfaces bounding the medium.

View full abstract

A major radiation source in boiling water reactors (BWR) for personnel exposure during shutdown maintenance has been identified to be activated corrosion products, mainly ⁶⁰Co, deposited on the primary system piping wall. Consequentry, information on Co transport mechanisms is vital to reduce this buildup in the primary system. It is generally considered that Co ions, which are corrosion products, are adsorbed on oxides deposited on the fuel rod surface, activated by neutrons, and then released as radioactive ions. As cobalt ferrite (CoFe₂O₄) is a thermodynamically probable compound under BWR conditions⁽¹⁾, ⁶⁰Co is considered to be released from it. Cobalt oxide (CoO) is also formed on the fuel rod surface when the concentration of Co ions in the coolant exceeds its solubility limit. Water pH might play an important role in reactions of adsorption and release because protons involve in these processes. However, effects of water pH on adsorption and release at high temperature have not been experimentally clarified yet.
In this study, release rates of Co ions from CoFe₂O₄ and CoO particles were measured at 285°C and various pH values. Then 9 mechanism for their release was oronosed.

View full abstract