An elevator group control system with floor-attribute controlmethod and system optimization using genetic algorithms

A new concept is proposed for an elevator group control system which can change control settings according to individual floor utilization situations. The floor-attribute-based control method uses a combination of floor-attribute-based evaluation and car-attribute-based evaluation. Computer simulations show it can improve multiple preferential door settings at the same time. Additionally, an on-line parameter tuning method using genetic algorithms is proposed. The floor attribute control method needs several dozen control parameters to be tuned according to each building's traffic, and the proposed method can do it. The elitist-preserving strategy is an effective way to keep control stability. A diversity-recovering operation, which initializes population, excluding the elitist individual, when traffic flow data have changed, improves the following ability for transitions of utilized conditions     

Irradiation behavior of HTGR coated particle fuel at abnormally high temperature

Irradiation behavior of high temperature gas-cooled reactor (HTGR) coated particles under temperature transient conditions was investigated in accordance with a design-base accident scenario for HTTR, a 30 MWth HTGR under construction at JAERI. One of the scenarios predicts that the fuel temperature of the block-type fuel elements rises to abnormally high temperature by blocking a coolant channel with some foreign substance. For simulating this scenario the fuel compacts incorporating the coated particles were irradiated at normal temperature in three capsules, followed by temperature transient up to a maximum of approximately 2000°C. The post-irradiation examinations, including surface inspection, metrology, ceramography and a measurement of coated particle failure were applied to the fuel compacts to investigate the thermal-transient effect on the fuel integrity. Integrity of the fuel compact was also assessed by an estimation of tangential stress introduced into the compact by the temperature transient.     

Studies on a protamine (galline) from fowl sperm. 4. Degradation of galline by trypsin-like protease of fowl sperm heads

Galline, a protamine of domestic fowl, was obtained by two preparation procedures from the semen of a strain of White Plymouth Rock and submitted to fractionation by column chromatography on Bio-Gel CM-30. In the first procedure the specimen prepared from sperm heads was purified by the use of distilled water and dilute acetic acid and fractionated into almost eight fractions (G-I-G-VIII) in the same way as the specimen from a strain of New Hampshire (1,2). No difference could be found between galline specimens from the two different strains based on the amino acid and terminal analyses of each fraction. The specimen of galline from sperm heads purified with 1% citric acid (the second procedure) was composed of only one component, which was isolated as a single peak. The smaller fractions, G-I-G-VII, were found to be derived from G-VIII by the action of trypsin-like protease contained in the extract of sperm heads with 1% citric acid. This enzyme seems to originate in the acrosome of fowl spermatozoa. Consequently, it is concluded that intact galline is composed of only one molecular species and its total amino acid sequence is represented by the completed formula of G-VIII as shown in the preceding paper (4).     

Enhanced conduction-corrected modified effective temperature as the outdoor thermal environment evaluation index upon the human body

This paper aims to clarify the handling technique of the solar radiation in an element of the thermal environment evaluation indices and to add expansions and improvements to conduction-corrected modified effective temperature ETF (Kurazumi et al., 2009) that can quantify the comprehensive effect on sensational and physiological sense and the effect of individual meteorological elements on the same evaluation axis applicable to an outdoor environment. Mean radiant temperature and radiant heat transfer coefficient of the outdoor space was defined. Enhanced conduction-corrected modified effective temperature ETFe that is ETF including short-wave solar radiation in outdoor space was defined. This sensational and physiological climatic environment index can make temperature convert each effect of difference in posture; air velocity; long-wave radiation in the outdoor space; short-wave solar radiation; contact surface temperature and humidity into individual meteorological elements. The addition of each temperature-converted factor is possible and quantifying the composite effect on sensational and physiological sense in the outdoor spaces as well as the discrete effect of each meteorological element is possible on the same evaluation axis. Consequently, it is possible to make the climate modification effects due to tree shade and areas of water that improve the urban thermal environment quantitatively explicit.     

Performance of ZrC-Coated Particle Fuel in Irradiation and Postirradiation Heating Tests

The ZrC-coated UO₂ particle is a promising fuel for high-temperature gas-cooled reactors. Particle fuels with multiple layers of pyrolytic carbon and ZrC have been irratiation-tested to a maximum fast-neutron fluence exceeding 2 × 10²⁵/m² ( E < 29 fJ). In-reactor fission-gas release measured at a burnup of 1.5 at.% was minimal. The failure fraction by postirradiation examination was null for all samples. The ZrC-coated particles at 4 at.% burnup were postirradiation heated to 2400°C/min without failure until after 6000 s at the maximum temperature. It was found that the ZrC layer could sustain a large strain at such high temperatures. The behavior is in strong contrast with that of SiC of standard Triso coating, which is brittle to very high temperatures.     

Evaluation of recriticality behavior in the material-relocation phase for Japan sodium-cooled fast reactor

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency has selected the advanced loop-type fast reactor, so-called Japan sodium-cooled fast reactor (JSFR). Through the evaluation of event progressions during hypothetical core-disruptive accident (CDA) under the design extension condition, a CDA scenario for JSFR has been evaluated. It has already been demonstrated that in-vessel retention (IVR) against CDA could be achieved by taking adequate design measures under best estimate conditions.

The whole sequence of CDA scenario for JSFR was categorized into four phases according to the progress of core-disruption status. In the third phase, so-called material-relocation phase, the accident events would progress in the subcritical state. However, if the uncertainties about the molten state of core remnant and their discharge behavior outward from core are conservatively superposed, the disrupted core may lead up to recriticality.

In the present study, the factors leading to recriticality in the material-relocation phase were investigated using the phenomenological diagrams, and the recriticality behaviors were evaluated through parametric analyses using SIMMER-III/IV codes. The results of parametric analyses suggested that a significant mechanical energy leading to the boundary failure of reactor vessel would not be released even assuming recriticality due to the uncertainties about molten state and discharge behavior. Through the present evaluation of the hypothetical recriticality event, the CDA scenario for JSFR could obtain further robustness from the viewpoint of achieving IVR.     

A scenario of core disruptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency (JAEA) has selected the advanced loop-type fast reactor, so-called JSFR. The safety design requirements of JSFR for Design Extension Condition (DEC) are the prevention of severe accidents and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, in particular, the In-Vessel Retention (IVR) against postulated Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the in-vessel cooling of degraded core materials are evaluated so as to achieve IVR. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulations. This is an unprecedented approach to the construction of a CDA scenario, and is an effective method to objectively investigate the factors leading to IVR failure and the design measures against them. It was concluded that mechanical/thermal failures of the reactor vessel due to power-excursion/thermal-load could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.     

Single molecular conductance measurements of molecular junction of Au/1,4-phenylenediamine/Au

It was reported recently that the diamine-terminated molecules show two sets of single molecular conductance peaks in the conductance histogram. Although we found another set of conductance value of 1,4-diaminobutane in a lower current range, it was difficult to determine the conductivity definitely because the compound has different conformations with different gauche contents within its molecular chain. To make it easier to determine and analyze a single molecular conductance we measured here the conductance of 1,4-phenylenediamine, whose conformation cannot be changed in terms of the gauche contents. As a result, new sets of conductance other than those reported recently [L. Venkataraman, J.E. Klare, C. Nuckolls, M.S. Hybertsen, M.L. Steigerwald, Nature 442 (2006) 904] were observed.     

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.