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Sumino, Kozo; Aoyama, Takafumi; Emoto, Takehiko
PNC TN9410 96-233, 27 Pages, 1996/08
It is important to accurately evaluate Corrosion Products (CPs) behavior in a FBR plant to reduce the personnel exposure due to CP deposition. A series of measurements of the gamma-ray dose rate distribution has been carried out in the Experimental Fast Reactor JOYO to characterize the CP behavior in the primary coolant system. The Plastic Scintillation Fiber (PSF), which is a position sensitive radiation detector that can detect the radiation dose rate at the relevant position in the fiber, was introduced to upgrade the gamma-ray distribution measurement in JOYO. In order to apply the PSF for the JOYO environment, the specification of the fiber was modified to obtain a wide range of sensitivity from about 0.01mSv/h up to 10mSv/h. This range covers the gamma-ray dose rate in the JOYO primary coolant system. In higher gradients of dose rate distribution than that of the PSF's position resolution, the measured data were unfolded by using the response matrix with a successive approximation method to reproduce the narrow distribution. As a result of the study on PSF, the continuous gamma-ray dose rate distribution was able to be measured by PSF in a few minutes, whereas point data at 1 m intervals were laboriously obtained by a set of Thermo-luminescence Dosimeters (TLDs). It was confirmed that the measurement of CP behavior upgraded significantly by using a detailed gamma-ray dose rate distribution.
PNC TN1100 93-008, 26 Pages, 1992/12
FBR development activities in Japan have been performed by the government in cooperion with private enterprises. The prototype reactor "MONJU" is now undergoing functnal testing, and the first demonstration plant is in the conceptual design stage. R for commercial plants has been conducted for several years. Commercial plants are quired to be superior to LWRs with regard to economy, safety, and reliability. Accoingly, the Power Reactor & Nuclear Fuel Development Corporation and the Japan Atomicower Company set up specific R&D goals for commercialization, identified plant conces, and planned the necessary R&D activities. In order to make the concepts a realit both government and private enterprises must play a roll in developing and demonstring FBR technologies through construction and operation of prototype and demonstrati plants. In addition, they must perform FBR optimization activities, such as an enhced safety core and a passive decay heat removal system etc., according to lo
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PNC TN9410 91-169, 87 Pages, 1991/08
We have been developing a inspection tool in order to apply for ISl of 32B sized elbow piping on MONJU primary cooling circuit. Experiments and examination were carried out on design work to decide the design parameters. We obtained the results as follows, (1)The design parameters of un ultrasonic probe, such as the ultrasonic element's diameter, frequency and refraction angle, were chosen as the most suitable for fraw detection. (2)It was confirmed that the most stable scanning of the scanner was achieved by the control of 3 driving wheels. (3)The scanner could be carried by 2 persons, it's weight was 38kg. (4)The handling mechanism for thermal insulator consist's of three divided piecies and lt's driving source is pneumatic. The conclusions of this study were as follows, (1)An equipments of the volumetric examination for large scale elbow piping was designed, and it could be use at 80
C without supplying and collecting apparatus for couplant. (2)This system must be contribute the reduction of human radiation dose.
; ; ; ; Ozawa, Kenji; ; Terunuma, Seiichi
PNC TN9410 91-187, 41 Pages, 1991/07
A reactor vessel in JOYO provides three induction type level meters which is defined in the safety protection system. They have two kinds of measuring range and display the sodium level below to the discharge nozzle of the primary cooling system. One is from 350mm about the normal sodium level to 1,600mm below it and other two sets are from 350mm above to 350mm below it. This report describes a special monitoring method of sodium level in the occurrence of the abnormal sodium level incident which reaches it more than 1600㎜ below the normal sodium level in the reactor vessel. The special monitoring method uses the discharge sodium pressure of the primary auxiliary cooling pump. A discharge sodium pipe from the primary auxiliary cooling pump is located in the bottom of the reactor vessel and it's discharge pressure is correlated with the reactor vessel sodium level which works back pressure to the pump. Therefore, it was assumed that abnormal sodium level which reaches it more than 1600mm below the normal sodium level can be monitored using this discharge sodium pressure. A verification test was conducted to measure the correlation of the discharge sodium pressure and the reactor vessel sodium level. Main results obtained from this test were as follows. (1)Validity of this special level monitoring method was confirmed in the sodium level range from normal to 3,390㎜ below it and in case of sodium level changing which is decreased at the rate of 47.5m
/h by this test during the system sodium drain work. (2)A correlation equation is obtained using parameters of discharge sodium pressure, flow and temperature of the primary auxiliary cooling system to gain sodium level of reactor vessel. (3)Parametor chart of the reactor vessel sodium level was made using multi regressive analysis.
*; *; *; *
PNC TN941 83-08, 51 Pages, 1983/03
Experimental Fast Reactor "JOYO" completed its 75MWt power operation with the Mark-I core (breeding core) in Dec., 1981. The Auxiliary Core Cooling System (ACCS) has been operated satisfactorily since the first sodium charge in Jan., 1977. This paper describes the operation history until the end of Mark-I Operation. (1)Accumulated operation time of Primary Auxiliary Cooling System with circulating pump on during outages for annual inspection or others was only 530 hours, while the rest being occupied by the counterflow from main circulating punps. Auxiliary circulating pump experienced the automatical start only when the reactor scrammed and reactor sodium level lowered because of the failure of overflow make-up pump in July, 1981. (2)Secondary auxiliary cooling system had been operated approximately 39,140 hours in full flow rate, meanwhile circulating pump failed 4 times because of power loss.
*; *; *; *; *; *; Imazu, Akira
PNC TN941 77-172, 44 Pages, 1977/10
This paper presents the cooperative work by PNC PNC and CRIEPI on the AE application to a creep-fatigue test of a Type 304 long elbow test assembly, that is, almost the 1/2-scale model of the primary coolant piping component of the proto-type FBR "MONJU". The creep-fatigue test was performed in air at 600C with the displacement-controlled in-plane bending load. AE characteristics, such as ring-down count rates, signal waveforms, peak-amplitude distributios and signal location patterns, were analysed in the process of the stationary creep-fatigue loading and the special loading for evaluating the effectiveness of AET to detect the piping detects.
