Nobuchika Kawasaki

Stress Mitigation Design of Tubesheets With Consideration of Thermal Stress Inducement Mechanism

Adoption of double-wall-straight tube steam generators made of Mod.9Cr-1Mo steel is planned for next generation fast breeder reactors in Japan. One of the major concerns relevant to the SG is structural integrity of tubesheets. In the reactor transient operation, thermal stress is induced by the temperature distribution in tubesheet and the magnitude of it depends on configurations of tubesheet. Stress generation mechanism of tubesheets was revealed through Finite Element analysis. Semi-spherical tubesheet models were investigated for the first survey of the thermal stress mechanism. As calculated results, semi-spherical tubesheet model gave the extensive peak stress around the outermost hole. Recognized thermal stress mechanism of semi-spherical tubesheet is as follows. (1) Dominant thermal stress is hoop stress caused by temperature difference between the perforated region and surrounding region. (2) Thermal stress is insensitive to size of specific portion, although is dominated by interaction mechanism between perforated and surrounded regions. (3) Stress concentration around hole’s edge generates peak stress. (4) Amplitude of peak stress depends on the tubesheet penetration angle and stress concentration becomes high near the outermost hole. Based on the above stress generation mechanism, authors proposed a stress mitigated tubesheet. It is center flatted spherical tubesheet (FST) as improved configuration. Calculated peak stress of FST was smaller than that of semi-spherical tubesheet. Further investigation revealed the detailed stress generation mechanism of FST during thermal transient. There were two different comparable thermal peak stress mechanisms in FST. Both location and magnitude of maximum peak stress depend on sodium temperature histories at thermal transient. One depends on the range (ΔT) of sodium temperature change. This type of peak stress was radial stress caused by the structural discontinuity, and it was located at the outermost hole. The other depends on the rate (dT/dt) of sodium temperature change. This type of peak stress was hoop stress caused by interaction between perforated region and surrounding region, and it was located at the one inner layer hole from outermost layer holes.Copyright

CREEP-FATIGUE STRENGTH EVALUATION OF PERFORATED PLATE AT ELEVATED TEMPERATURE USING STRESS REDISTRIBUTION LOCUS METHOD

This study reports the experimental results carried out at the elevated temperature of 550°C on fatigue strength of the perforated plate. Stress Redistribution Locus (abbreviated as SRL hereafter) Method is applied to predict fatigue life for the specimens having stress concentration. The specimens made of SUS304 stainless steel have through holes with different number and different diameter, accordingly leading to the different stress concentration condition. The inelastic local strain is estimated by the SRL method or the other previous Neuber’s rule, and compared to the experimental results on the crack initiation life at the edge of the hole using the concentrated local strain obtained by these methods. The obtained result is that the SRL method is best used with the onset of failure or crack initiation.Copyright

Parametric Design Study About Seismic Isolation System for Fast Reactor JSFR

Japanese seismic conditions are getting severer and natural frequencies of components are getting lower due to the enlargements of components’ size, therefore response accelerations and buckling margins of reactor vessels were parametrically surveyed with attention to thicknesses, diameters, and isolation frequencies for reviewing necessary isolation specification.For the first, Japanese seismic condition and present specification of JSFR isolation system are introduced in this paper. RV installed floor responses were calculated based on this seismic condition and the relationship of natural frequencies, initiated stresses, and buckling margins against vessel thicknesses and diameters were shown with trend.Expansion characteristic of isolation system was evaluated by parametric acceleration response analyses. Comparing the response of isolation system with 8Hz vertical natural frequency with other natural frequency’s isolation, response ratios against natural frequencies were calculated.Japanese seismic design condition may become severer than present one, and a natural frequency of main component may decrease. However based on the buckling margin with present plant specifications and the expansion characteristic of isolation system, the advanced isolation system with 8Hz vertical natural frequency was selected as the isolation system of JSFR at still present occasion.Copyright

Comparison of Creep-Fatigue Evaluation Methods With Notched Specimens Made of Mod.9Cr-1Mo Steel

In a component design at elevated temperature, creep-fatigue is one of the most important failure modes, and assessment of creep-fatigue life in structural discontinuity is important issue to evaluate structural integrity of the components. Therefore a lot of creep-fatigue life evaluation methods were proposed until now. To compare and assess these evaluation methods, a series of creep-fatigue tests was carried out with notched specimens. All the specimens were made of Mod.9Cr-1Mo steel, which it is a candidate material for a primary and secondary heat transport system components of JSFR (Japan Sodium-cooled Fast Reactor). Mechanical creep-fatigue tests and thermal creep-fatigue tests were performed by using conventional uni-axial push-pull fatigue test machine and thermal gradient generating system with an induction heating coil. Stress concentration levels were adjusted by varying the diameters of notch roots in the both tests. In the test, creep-fatigue lives, crack initiation and propagation processes were observed by digital micro-scope and replica method. Besides those, a series of elastic Finite Element Analysis (FEA) were carried out to predict the number of cycles to failure by several creep-fatigue life evaluation methods. Then these predictions were compared with test results. Several types of evaluation methods which are stress redistribution locus (SRL) method, simple elastic follow-up method and the methods described in JSME FR (Fast Reactor) code were applied. The applicability and conservativeness of these methods were discussed. It was appeared that SRL method gave rational prediction of creep-fatigue life with conservativeness when the factor of κ = 1.6 was applied for all the conditions tested in this study. Comparison of SRL method and simple elastic follow-up method indicated that SRL method applied factor of κ = 1.6 gave the smallest creep-fatigue life in practicable stress level. JSME FR code gave an evaluation 70∼100 times conservative lives comparing with the test results.Copyright

Proposals of Guidelines for High Temperature Structural Design of Fast Reactor Vessels

Main loadings of reactor vessels in fast reactor plants are thermal stresses induced by fluid temperature change at transient operation. Structures respond to them with elastic plastic creep deformation under high temperature conditions. It can induce incremental deformation and creep fatigue crack at critical portions around the sodium surface, thermal stratification layer and core support structures. Those phenomena are so complex that design evaluation becomes sometimes too conservative. In order to achieve precise high temperature design for realizing compact reactor vessels of fast reactor plants, such guidelines are proposed as for thermal load modeling, structural analysis and strength evaluation. This paper gives the brief summary of these guidelines. GUIDELINES FOR THERMAL LOAD MODELING: One of main difficulties of thermal load modeling is their inducement mechanism by interaction between thermal hydraulic and structural mechanics. Design evaluation requires envelope load conditions with considering scatter of design parameters. Proposed guidelines enable precise load modeling by grasping sensitivities of thermal stress to design parameters including thermal hydraulic ones. GUIDELINES FOR INELASTIC DESIGN ANALYSIS: Guidelines are proposed to apply inelastic analysis methods for design of reactor vessels. There are so many influence parameters in inelastic analysis that conservative and unique solutions are hardly found. To overcome such difficulties, mechanism and main parameters of inelastic behaviors of reactor vessels were clarified. Guidelines give conservative results within the same mechanism as expected reactor vessels. HIGH TEMPERATURE STRENGTH EVALUATION METHOD: Incremental deformation and creep fatigue strength evaluation methods were proposed. Accumulated strain is limited within no influence of fatigue and creep-fatigue strength. Taking design conditions of reactor vessels into account, creep fatigue evaluation considers strain concentration and an intermediate stress hold effect on creep-fatigue strength. Influences of thermal aging were also confirmed.Copyright

Spectra Thermal Fatigue Tests Under Frequency Controlled Fluid Temperature Variation: Superposed Sinusoidal Temperature Fluctuation Tests

To clarify frequency-dependent attenuation effects of fluid temperature fluctuation on fatigue strength, thermal fatigue strength tests subjected to superposed sinusoidal temperature fluctuations were performed by the SPECTRA test facility. Fluid temperature waves were generated by superposition of sinusoidal waves, where frequencies were 0.05, 0.2, and 0.5Hz. Two types of superposed waves were selected for the tests, dual and triple ones. The dual one was obtained by superposing two sinusoidal waves whose temperature ranges and frequencies are respectively 200 centigrade and 0.05Hz and 60 centigrade and 0.5Hz at the inlet of test piece. The triple one was the superposition of three sinusoidal waves whose temperature ranges and frequencies are respectively, 150 centigrade and 0.2Hz, 75 centigrade and 0.05Hz and 50 centigrade and 0.5Hz at the inlet of test piece. The longest periods were 20 seconds for both types of waves and it is the fundamental cycle for the thermal fatigue tests. For the dual case, 73,810 cycles fatigue test was performed while for the triple one 116,640 cycles were performed. After these fatigue tests, cylindrical test pieces were cut away from the test loop, and cracks were observed on the inner surface of the test pieces. For the dual wave case, crack initiation occurred from 400 to 600mm position from the inlet of test piece. For the triple wave case, crack initiation occurred from 400 to 600mm position from the inlet of test piece. The corresponded fluid temperature range to crack initiation is from 205 to 220 centigrade for the dual one and from 195 to 215 centigrade for the triple one. Fatigue lives at crack initiation positions were evaluated based on the test conditions. Adopting power spectrum density functions and frequency transfer functions, fatigue lives were predicted within a factor 3 as predicted for single sinusoidal temperature waves in the other tests. To confirm advantages of these functions, fatigue life estimations were compared with those obtained without using these functions. Based on the compared results, these functions are necessary to predict accurate fatigue lives.Copyright

SPECTRA Thermal Fatigue Tests Under Frequency Controlled Fluid Temperature Variation: Strength Tests

Thermal fatigue strength tests subjected to sinusoidal fluid temperature waves were performed by the SPECTRA test facility, where frequencies were 0.05, 0.2, and 0.5Hz. Cracks were observed on the inner surface of cylindrical test pieces after testing. 0.05Hz’s wave caused a greater number of and deeper cracks than 0.5Hz’s wave under the same fluid temperature range and the same fatigue cycles. The crack initiation region of the 0.05Hz’s wave was larger than for the 0.5Hz’s wave. Estimated fatigue failure cycles based on the frequency transfer functions were compared with test results. Frequency-dependency in failure cycles was observed through these test results, and frequency transfer functions could estimate this dependency. The test results supported the fatigue damage evaluation method with frequency transfer functions.Copyright

SPECTRA Thermal Fatigue Tests Under Frequency Controlled Fluid Temperature Variation: Transient Temperature Measurement Tests

The coolant leakage by thermal striping phenomenon should be prevented at nuclear power plants and a lot of efforts are made to develop its evaluation methods. The frequency transfer function method can be used to obtain temperature and stress response to fluid temperature history using transfer function models; therefore it is considered as an excellent evaluation method. To measure temperature response of structures to fluid temperature variations and to confirm their frequency characteristics, transient temperature measurement tests were performed by JAEA. In the transient temperature measurement tests, three different frequencies of sinusoidal fluid temperature waves (0.05, 0.2, 0.5Hz) were controlled using frequency controlled thermal fatigue test equipment (SPECTRA) and temperature responses at inner and outer piping surfaces were measured along the test sections. Frequency effects on temperature attenuation during transfer process from fluid to structures were confirmed and the effective heat transfer function in frequency transfer function method was verified by transient temperature measurement test results.Copyright

Comparison and Assessment of the Creep-Fatigue Evaluation Methods With Notched Specimen Made of Mod.9Cr-1Mo Steel

In components design at elevated temperature, creep-fatigue is one of the most important failure modes, and assessment of creep-fatigue life in structural discontinuities is an important issue in evaluating the integrity of components. Therefore, a lot of creep-fatigue life evaluation methods were proposed until now. To compare and assess the evaluation methods, a series of creep-fatigue test was carried out with notched specimens. All the specimens were made of Mod.9Cr-1Mo steel, which is a candidate material for primary and secondary heat transport system components of the Japan sodium-cooled fast reactor (JSFR). Mechanical creep-fatigue tests and thermal creep-fatigue test were performed by using a conventional uni-axial push–pull fatigue test machine and a thermal gradient generating system with an induction heating. The stress concentration levels were adjusted by varying the notch radius in the each test. The creep-fatigue lives, crack initiation, and propagation processes were monitored by a digital microscope and the replica method. A series of finite element analysis (FEA) was carried out to predict the number of cycles to failure by the several creep-fatigue life evaluation methods. Then, these predictions were compared with the test results. Several types of evaluation methods such are stress redistribution locus (SRL) method, simple elastic follow-up method and the methods described in the design and constriction code for fast reactor (FR) published by the Japan Society of Mechanical Engineers (JSME FRs code) were applied. Through the comparisons, it was appeared that SRL method gave rational conservative prediction of the creep-fatigue life when the factor of κ = 1.6 was applied for all conditions tested in this study. A comparison of SRL method and simple elastic follow-up method indicated that SRL method applied factor of κ = 1.6 gave the smallest creep-fatigue life in practicable stress range level. The JSME FRs code gave an evaluation 70–100 times conservative lives comparing with the test results.

Verification of the Estimation Methods of Strain Range in Notched Specimens Made of Mod.9Cr-1Mo Steel

Several methods of estimating strain range at a structural discontinuity have been developed in order to assess component reliability. In a component design at elevated temperature, estimation of strain range is required to evaluate the fatigue and creep-fatigue damage. Therefore, estimation of strain range is one of the most important issues when evaluating the integrity of a component during its lifetimes. To verify the methods of estimating strain range for discontinuous structures, low cycle fatigue tests were carried out with notched specimens. All the specimens were made of Mod.9Cr-1Mo steel, because it is a candidate material for a primary and secondary heat transport system components of Japan Sodium-cooled Fast Reactor (JSFR). Displacement control fatigue tests and thermal fatigue tests were performed by ordinary uniaxial push–pull test machine and equipment generating the thermal gradient in the notched plate by induction heating. Several notch radii were employed to vary the stress concentration level in both kinds of tests. Crack initiation and propagation process during the tests were observed by a digital microscope and the replica method to define the failure cycles. Elastic and inelastic finite element analyses were also performed to estimate strain range for predicting fatigue life. Then, these predictions were compared with the test results. Several methods such as stress redistribution locus (SRL) method, simple elastic follow-up (SEF) method, Neubers law, and the procedures employed by elevated temperature design codes were applied. Through these comparisons, the applicability and conservativeness of these strain range estimation methods, which is the basis of the fatigue and creep-fatigue life prediction, are discussed.