A feasibility study of the particle interaction method for the flow regimes with the chemical reaction; (Report under the contract between JNC and Toshiba Corporation)

Shirakawa, Noriyuki*; *; *; *

JNC TJ9440 2000-008, 47 Pages, 2000/03

JNC-TJ9440-2000-008.pdf:1.96MB

The numerical thermohydraulic analysis of a LMFR component should involve its whole boundaly in order to evaluate the effect of chemical reaction within it. Therefore, it becomes difficult mainly due to computing time to adopt microscopic approach for the chemical reaction directly. Thus, the thermohydraulic code is required to model the chemically reactive fluid dynamics with constitutive correlations. The reaction rate denpends on the binary contact areas between components such as continuous liquids, droplets, solid particles, and bubbles. The contact areas change sharply according to the interface state between components. Since no experiments to study the jet flow with sodium-water chemical reaction have been done, the goal of this study is to obtain the knowledge of flow regimes and contact areas by analyzing the fluid dynamics of multi-pahse and reactive components mechanistically with the particle interaction method. For the first stage of the study, the applicability of this method to the nalysis of a liquid jet into the other liquid pool was investigated. Based on the literatures, we investigated the jet flow mechanisms and analyzed the experiment of a water jet into a gasoline pool. We also analyzed SWAT3/Run19 test, the jet flow in a rod bundle, to study the applicability of the method to a complicated boundary without a chemical reaction model. The calculated fluid dynamics was in good agreement with the experiment. Furthermore, we studied and formulated the paths of phase change and chemical reaction, and conceptually designed the adopting the heat-transfer-limited phase change model and the synthesizd reaction model with a water-hydrogen conversion ratio.

The development and application of overheating failure model of FBR steam generator tubes

Hamada, Hirotsugu; *; *; *; Hiroi, Hiroshi*

PNC TN9410 98-029, 122 Pages, 1998/05

PNC-TN9410-98-029.pdf:14.03MB

The following items have been studies to evaluate overheating failure of FBR generator heat transfer tubes: (1)To establish a structural integrity analysis method. The strength standard values for 2.25Cr-1Mo steel was established taking account of time dependent effect to overheating failure mechanism based on high temperature (700 - 1200C) creep data and was validated by tube rupture simulation test data. (2)To improve and validate blow down analytical method. The analytical result by use of BLOOPH, the FBR blow down code, was compared with that by use of RELAP-5, the general purpose thermo-hydraulic code, and a good agreement was obtained. (3)To quantitatively validate the entire overheating analysis model by sodium water reaction data Sodium-water reaction tests of SWAT-3 and LLTR were analyzed using above mentioned analytical method. The ductile fracture occurred earlier than the creep fracture in the analysis and the comparison of tube failure times with the experiments showed sufficient conservativeness. Based on the above studies, the analytical method was applied to PFR superheater leak event and the Monju steam generator accidental analysis. The followings were quantatitively shown through the analysis: (1)The most important cause that multi-tube failure occurred in the 1987 PFR superheater-2 leak is that the superheater did not equip a fast steam dump system at the time of the leak event. (2)Overheating failure will not occur under any operational conditions of Monju in both steady state and transient phases such as water/steam blow-down. (3)Although safety margin becomes small when the water/steam flow rate becomes small during the blow-down, the modification of the plant such as hastening blow-down by equipping more relief valves will drastically improve the safety margin.

Overheating failure analysis of steam generator tubes II; Overheating failure analysis of U.K.PFR superheater

Hamada, Hirotsugu; Tanabe, Hiromi

PNC TN9410 96-027, 41 Pages, 1995/12

PNC-TN9410-96-027.pdf:1.02MB

If a sodium-water reaction jet was formed due to water leakage in an FBR steam generator(SG), neighboring tubes would suffer from overheating. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such a severe overheating condition. So far, an analytical model using the structural integrity analysis code, FINAS, has been prepared and validated by the explosive torch overheating test data. This report presents the results on the overheating failure analysis of the under-sodium leak in the PFR superheater(SH), 1987. In the SH with slow steam dump system in 1987, neighboring overheated tubes are failed about 3 seconds after the SH isolation, which is shown both by the leak in the PFR and its analysis. For the SH in which a fast steam dump system was installed after the leak of 1987, the analysis shows no tube failure due to the fast steam depression and cooling effect inside. These results indicate that the FINAS model adequately predicts the overheating failure and the specific SH design and operation possibly result in further growth of the leak. It is concluded that steam blow effect is extremely important and the analysis model presented here is useful for the overheating failure evaluation of the SGs.

None

PNC TN1410 95-087, 89 Pages, 1995/10

PNC-TN1410-95-087.pdf:7.31MB

None

Neural Network Predictive and Anticipatory Control Algorithms for a Neural Adaptive Control System

Ugolini; Yoshikawa, Shinji; Ozawa, Kenji

PNC TN9410 95-210, 11 Pages, 1995/09

PNC-TN9410-95-210.pdf:0.47MB

The proper control of the outlet steam temperature of the evaporator is of major importance for improving the overall performance of the balance of plant of a nuclear power reactor. This report presents a predictive and an anticipatory control algorithms based on the artificial neural network (ANN) technique. The two control algorithms are embedded on a model reference adaptive control system based on the ANN technique, defined as MRAC. It has already been illustrated that nonlinear dynamical systems such as the evaporator of a nuclear power plant can be controlled by an MRAC system. However, little attention has been devoted on exploiting the forecasting potential of the ANN technique for enhancing the accuracy and improving the efficacy of the control action of the MRAC system. The improved MRAC system has been tested to simulate the behavior of a fast breeder reactor (FBR) evaporator and to control its outlet steam temperature. The simulation results indicate that the performance of the MRAC system substantially improves when the predictive and the anticipatory control algorithms are activated.

Ultra-high temperature tensile properties on Mod.9Cr-1M0, 2.25Cr-1Mo and SUS321 steel(I)

; Yoshida, Eiichi;

PNC TN9410 94-262, 120 Pages, 1994/09

PNC-TN9410-94-262.pdf:6.07MB

This study clarified the tensie properties of Mod.9Cr-1Mo, 2.25Cr-1Mo and SUS321 steels at ultra-high temperature up to 1,200C which will be used in analysys and evaluation of the tube burst in steam generators of fast breeder reaetors. (1)Tensile strength of Mod.9Cr-1Mo, 2.25Cr-1Mo and SUS321 steels at 1,200C were 2.5, 2,and 2.5kg/mm, respectively. (2)The difference for tensile strength and 0.2% yeild strength between specimen heat rate and heat holding time could not be found in the present. (3)The temperatures of the tube burst at the maximum internal pressure of 150kg/cm corresponding to the practical use condition were expected to be approximately 960C for Mod.9Cr-1M0, 860C for 2.25Cr-1Mo and 1040C for SUS321, respectively. These tests result will be reflected to evaluation of tube burst by sodium water reaction.

Preliminary design for reconstruction of SWAT-3 facility

*; *; *; *; *; *; *

PNC TJ9164 94-006, 133 Pages, 1994/03

PNC-TJ9164-94-006.pdf:3.4MB

This report gives an applicability of SWAT-3 facility and contents of the reconstruction in order to confirm a DBL (Design Basis Leak) for the demonstration reactor SG. (1).Test Cndition and test case. Evaluation of the wall temperature for adjacent heat transfer tubes under the sodium-water reaction event was performed. (a)As the effect of tube rupture due to overheating, failure of upper part of the helical coil was severer than one of the lower part. (b)The wall temperature depends on the water side condition. (c)Reference test condition, whici is water leak rale about 1 kg/s, failure of upper part of the helical coil and 30% partial load, was selected. A total of ten test cases were decided. (2).System and Components Design. (a)Large leak sodium-water reaction analyses including water injection rate analysis and quasi-steady pressure analysis were performed. The maximum water leak rate of 1 DEG was 7.2 kg/s and the water leak rate at the quasi-steady state was 3.1 kg/s. The maximum pressure was 18.1kg/cma at the piping between the reaction vessel and IHX, the pressure was within the design condition of SWAT-3 facility. (b)Based on the results of the large leak sodium-water reaction analyses, a reaction vessel, water heaters and a dump tank were designed and their design specification were clarified. The reaction vessel was a scale of one third of the demonstration reactor SG and it was designed to be able to conduct the water injection test twice with one test unit. (c)A system and piping diagram, and many kinds of list (Piping list, Valve list, instrumentlist) were made up. (3).Reconstruction scope and arrangement plan. The reconstruction scope and a layout for the components and piping were clarfied and the arrange ment plans were made up. (4)Reconstruction period. The recoastruction period and man power for the design, fabrication, inspection and installation were studied and the reconstruction schedule was made up.

Wastage characteristics of high-chrome steel heat transfer tube; Intermediate leak wastage tests

Shimoyama, Kazuhito; Hamada, Hirotsugu; Tanabe, Hiromi; Usami, Masayuki

PNC TN9410 93-212, 134 Pages, 1993/09

PNC-TN9410-93-212.pdf:5.99MB

A one-through unit type steam generator (SG) having the Mod.9Cr-1MO Steel for its heat transfer tube is considered to be promising for the development of large FBR SGs. Wastage data of the tube material was already obtained for the micro-/small leak region as formerly reported. Therefore, intermediate leak wastage tests were conducted in the range from 10 g/s to around 200 g/s by using the SWAT-1 test facility and the test results are summarized as follows: (1)The wastage resistivity of the Mod.9Cr-1Mo steel is between that of 2.25Cr-1Mo steel and austenitic stainless steel; namely, the Mod.9Cr-1Mo steel has about half the of wastage rate of the 2.25Cr-1Mo steel. An experimental wastage formula in the intermediate leak region was derived from the test data. (2)Almost all of the wastage profile of target tubes was toroidal type and it became about half the cross section area of the 2.25Cr-1Mo steel. An experimental formula on initial leak diameters versus equivalent secondary failure diameters was derived in the intermediate leak region. These test results would be applied to failure propagation analysis code LBAP which is to be used for the design of a one-through unit type SG.

Mockup test apparatus for the inspection system of steam generator tubes; Design and Manufacturing

; ; ; ;

PNC TN9410 92-254, 76 Pages, 1992/07

PNC-TN9410-92-254.pdf:2.02MB

A verification test of the inspection system of Monju steam generator(SG) tubes will be performed in near future. Mockup Test Apparatus for the inspection system of SG tubes was manufactured and installed at Mechatronics Application Reserch Facility (MARF) in OEC. The test apparatus has the same specification, which is prepared for verification test, as Monju plant; for instance, which are dimension and material of tubes, and workability for the inspection equipment. About one hundred and forty SG tubes are radially arranged in tube sheets in Monju SG, however, three tubes, inner, center and outer one, are sellected in this test apparatus for testing of inspection system, It was verified that the test apparatus was manufactured with the same accuracy and dimension as Monju. System verification test is planned using this test apparatus.

Evaluation of strength of Mod.9Cr-1Mo weldments; 1st Report: Evaluation of fatigue strength

; ;

PNC TN9410 92-148, 65 Pages, 1992/02

PNC-TN9410-92-148.pdf:1.51MB

Mod.9Cr-1Mo steel is the material whose future utilization is expected as an advanced material of the steam generater of the Fast Breeder Reactors. A procedure for evaluation of weldment is being developed as one of the main concerns for the utilization of this material. The purpose of this report is to propose a fatigue strength evaluation method of Mod.9Cr-1Mo weldment which incorporates the effect of the heat affected zone which forms the softest portion of the weldment on the strength of the weldment. The mechanisms of strain concentration in the weldments of Mod.9Cr-1Mo steel was analysed and fatigue fracture was evaluated in terms of strain concentration. It was shown that the maximum strain concentration occured in the heat affeted zone at the begining of cyclic strain loading but that as a result of cyclic softning which was evident with the base metal but negligible with the heat affected zone the maximun strain concentration moved on to the base metal after half-life under the assumption that the hardness of the base metal and the heat affected zone coincide with each other at about half-life. Furthermore, based on this result, fatigue damage was evaluated. It was shown that the accumulated fatigue damage was not the maximum in the heat affected zone and that fatigue failure did not occur in the heat affected zone as far as the present analyses concerns. It was also made clear that fatigue failure occur in the base metal due to the strain concentration at strain ranges higher than elastic reagion and that it occur in the weld metal due to its inferior fatigue strength compared with the base metal. As a result, it was clarified that the fatigue strength of Mod.9Cr-1Mo weldment is reasonably evaluated by FE-analysis, which observation is able to be applied to the analysis of structures with weldments to be evaluated.