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Tanaka, Yoshihiko; Kasahara, Naoto
JNC TN9400 2003-037, 95 Pages, 2003/05
JNC-TN9400-2003-037.pdf:4.11MB
The advanced loop-type reactor system, one of the promising concepts in the Feasibility study of the FBR Cycle, adopts many innovative ideas to meet the challenging requirements for safety and economy. As a results, it seems that the structures of the reactor system would be subjected to severer loads than the predecessors. 0ne of the countermeasures to them is the design by inelastic analysis. In the past, many studies showed that structural design by inelastic analysis is much more reasonable than one by conservative elastic analysis. However, inelastic analysis has hardly been adopted in nuclear design so far. One of the reasons is that inelastic analysis has loading history effect, that is, the analysis result would differ depending on the order of loads. It seems to be difficult to find the general solution for the loading history effect. Consequently, inelastic analysis output from the four deferent thermal load histories which consists of the thermal load cycle including the severest cold shock ("C")and the one including the severest hot shock ("H") were compared with each other. From this comparison, it was revealed that the thermal load history with evenly distributed "H"s among "C" s tend to give the most conservative damage estimation derived from inelastic analysis output. Therefore, such thermal load history pattern is proposed for the structural design by inelastic analysis.
Ando, Masanori
JNC TN9400 2003-033, 53 Pages, 2003/04
JNC-TN9400-2003-033.pdf:1.84MB
Reactor vessel of fast reactor plant contains high temperature liquid sodium in its inside and its upper end is supported by a low temperature structure made of concrete. Therefore, a sharp temperature gradient will arise at the vessel wall near the sodium surface. For this reason, generated thermal stress around this part might be larger than yield stress of that materia1. Then it is necessary to use inelastic analysis for predicting detailed behavior of it. However, there are no complete constitutive equations for inelastic analysis until now. Therefore inelastic analysis results will depend on the selected constitutive equation, and it is not easy to apply to a design. In this research, author analyzed about mentioned part by using some kinds of constitutive equations, and compared their results each other to research about dependence of analysis results on constitutive equations. The results obtained in this research are as follows. (1)Results from inelastic analyses under monotonic load depend on constitutive equations, however they tend to be on the unique stress relaxation curve. (2)When both stress values in results from elastic-plastic analysis and elastic-creep analysis are the same, their inelastic strain would have also similar distributions. (3)The inelastic analysis results under cyclic loads are not on the stress relaxation curve. One of this reason, rachet strain might change the distribution of inelastic strain. (4)On the cyclic analysis case (start-up-normal operation-trip), analysis results using monotonic stress-strain curve predicted larger rachet strain than that of the results using cyclic stress-strain curve.
Asayama, Tai
JNC TN9400 2003-036, 66 Pages, 2003/03
JNC-TN9400-2003-036.pdf:0.93MB
For the commercialization of fast breeder reactors, (System Based Code), a completely new scheme of a code on structural integrity, is being developed. 0ne of the distinguished features of the System Based Code is that it is able to determine a reasonable total margin on a structural system, by allowing the exchanges of margins between various technical items. Detailed estimation of failure probability of a given combination of technical items and its comparison with a target value is one way to achieve this. However, simpler and easier methods that allow margin exchange without detailed calculation of failure probability are desirable in design. The authors have developed a simplified method such as a (design factor method) from this viewpoint. This report describes a (Vector Method),which has been newly developed. Following points are reported: 1)The Vector Method allows margin exchange evaluation on an (equi-qualitly assurance plane) using vector calculation. Evaluation is easy and sufficient accuracy is achieved. The equi-quality assurance plane is obtained by a projection of an (equi-failure probability surface in a n-dimensional space, which is calculated beforehand for typical combinations of design variables. 2)The Vector Method is considered to give the (Quality Assurance Index Method) a Probabilistic interpretation. 3)An algebraic method was proposed for the calculation of failure probabilities, which is necessary to obtain a equi-failure probability surface. This method calculates failure probabilities without using numerical methods such as Monte Carlo simulation or numerical integration. Under limited conditions, this method is quite effective compared to numerical methods. 4)An illustration of the procedure of margin exchange evaluation is given. It may be possible to use this method to optimize ISI plans; even it is not fully implemented in the System Based Code.
Shimakawa, T.*; Nakamura, Kyotada*
JNC TJ9400 2002-007, 54 Pages, 2003/03
JNC-TJ9400-2002-007.pdf:1.45MB
None
Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai; Sagayama, Yutaka*; Dozaki, Koji*; Shibamoto, Hiroshi*; Tanaka, Yoshihiko*
JNC TY9400 2002-025, 889 Pages, 2003/01
JNC-TY9400-2002-025.pdf:26.72MB
None
Morishita, Masaki; Kitamura, Seiji; *; *; *
JNC TY9400 2002-024, 914 Pages, 2003/01
JNC-TY9400-2002-024.pdf:37.28MB
None
Yoo, Yeon-Sik
JNC TN9400 2002-079, 32 Pages, 2003/01
JNC-TN9400-2002-079.pdf:0.79MB
This study is concerned with crack opening displacements (CODs) of cylinders with a circumferential through-crack which is subjected to tension and in-plane bending loads. Most studies about crack opening behavior have performed on membrane and global bending stresses. Moreover, they cannot be valid for large-scale structures. For simplicity on evaluation for structural integrity, crack opening displacement has been often calculated by plate or pipe model considering almost stresses as a membrane component. However, it is important to investigate ones close to real crack opening behaviors under stress states for reliability on evaluation. The results must be directly related to evaluate leakage detection in reactor vessel and the primary piping system of FBR structures. From that purpose, a series of FEM analyses were performed, and hence the characteristics of COD under an in-plane bending stress were compared with those under a membrane stress. In addition, the plate model was indicated to be unreasonable for application on large-scale pipes by comparing the plate model with the pipe model. The results of this study are expected to be valid for leakage evaluation of high temperature structures especially.
*; *; *; *; Tsukimori, Kazuyuki
JNC TY9400 2002-012, 101 Pages, 2002/08
JNC-TY9400-2002-012.pdf:5.83MB
None
Kasahara, Naoto; Furuhashi, Ichiro*; *; Ando, Masanori; *
JNC TN9400 2002-047, 107 Pages, 2002/08
Temperature fluctuation from incomplete fluid mixing induces fatigue damages on structures of nuclear components, which should be prevented. For rational evaluation of fatigue crack initiation against this phenomenon, the authors have developed a frequency response function of thermal stress to fluid temperature. Since an actual failure mode has crack initiation, propagation and penetration processes, Fracture mechanics approach such as repair based on crack propagation characteristics and crack arrest evaluation are effective to prevent failure. This study proposed a frequency response function of stress intensity factors to fluid temperature fluctuations in order to evaluate thermal fatigue based on crack propagations. Stress intensity factor decreases according to crack propagation under high cycle fluctuation. On the other han, it increases under low cycle one and membrane constraint conditions.
Tanaka, Yoshihiko; kasahara, Naoto
JNC TN9400 2002-038, 95 Pages, 2002/06
JNC-TN9400-2002-038.pdf:2.56MB
PARTS, Program for Arbitrary Real Time Simulation is being developed: it is cxpected to make great contribution to fast reactor components' design by enabling integration of thermal hydraulic and structural analysis. At this moment, the Green function method is mainly used as a stress analysis method for PARTS. The Green function is a description of the relationship between input and response of a system. Strictly, the response depends only on the natures of the input and the system. However if a function precisely simulating their natures is established based on the response to elemental inputs (pulse wave, step wave, etc.,), it becomes possible to find approximate responses to random and/or complicated inputs. This procedure is called "Green function method". This method is applicable to structural design of the fast reactors. Green function method finds thermal transient stress arising in structures in the form of convolute integration corresponding to coolant fluids' step-changes of temperature. It is expected to calculate faster than Finite Elemental Method (FEM) that solves innumerable balance equations of stress and strain at every time step. In order to apply the Green function method to actual plant design in near future, it is necessary to prove that the method gives appropriate results even under the conditions assumed in plant design works. The authors have successfully developed the Green function method which had been applicable only to a cylinder contacting with sole fluid under constant thermal transfer rate into the one being applicable to a cylinder with primary and secondary fluids under step-changing thermal transfer rates. In this report, applicability of Grecn function method to structural design of a components of complicated shape cxposed to thermal transients of the two independent coolant systems under changing heat transfer rate. As an example, the internal components of the intermediate heat exchanger (IHX) of the advanced loop ...
Kawasaki, Nobuchika;
JNC TN9400 2002-017, 89 Pages, 2002/05
JNC-TN9400-2002-017.pdf:1.66MB
Both reliability and safety have to be further improved for the successful commercialization of FBRs. At the same time, construction and operation costs need to be reduced. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of load setting, fabrication, inspection, maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Crack initiation probability and crack penetration probability are considered as candidate indexes. In material field, probabilistic material distributions are requested to calculate those index probabilities. Therefore, using 316FR base metal material strength as a sample, a scope of necessary distributions, a method of distribution decision, and an evaluation method of decided distributions are considered. The scope of distribution is decided by a comparison with strength evaluation methods. Those are yield stress, ultimate stress, creep rupture equation, creep strain equation, monotonic stress-strain equation, cyclic stress-strain equation, fatigue failure equation, creep rupture elongation, and so on. The method of distribution decision is fixed during this sample trial, and by this procedure, above scoped characters were approximated into normal distributions, lognormal distributions or Weibull distributions. Decided distributions are evaluated into three ranks from the view points of data range, the number of data, and the coefficient of correlations.
; *
JNC TN9400 2002-014, 48 Pages, 2002/04
JNC-TN9400-2002-014.pdf:1.23MB
For the commercialization of fast breeder reactors, economy must be drastically improved so that they ccan be competitive to future generation light water reactors. Reliability and safety must also be increased. For this purpose, the authors have been developing the FBR System Based Code since fiscal year 2000. One of the core concepts of the System Based Code is margin exchange and methodologies that realize it were strongly needed. This report proposes a methodology for margin exchange that is based on failure probability calculation. The effectiveness of the methodology is demonstrated by a worked example, which quantitatively shows a possibility of margin exchange between two designs of FBR reactor vessels, one made of forged rings with no ISI, and the other a welded vessel with ISI implemented. Failure probabilities for both designs were calculated considering thermal fatigue crack initiation and propagation. By setting an appropriate design life, calculated cumulatcd failure probabilities for both designs became equivalent, which means margin exchange hold for those designs. Based on this methodology, a simpler method for margin exchange, systematize design factors, was proposed for the implementation in the System Based Code. This methodology does not need the detailed calculation of filure probabilities. Current safety factors for fatigue assessment for example, which is 2 for strain range and 20 for numbers of cycles to failure, are replaced by systematized design factors, which consists of multiple values of factors corresponding to target reliabilities and arrays of technical options that are provided in the partial code of the System Based Code. This report showed some example values determined based on the above worked example. Those results showed that the two methodologies proposed in this report are promising to be implemented hl the System Based Code. Finally, items to be developed regarding two methodologies were clarified.
; Hosogai, Hiromi*; Furuhashi, Ichiro*; kasahara, Naoto
JNC TN9400 2001-121, 44 Pages, 2002/02
JNC-TN9400-2001-121.pdf:1.16MB
PARTS, Program for Arbitrary Real Time Simulation is being developed: it is expected to make great contribution to fast reactor components' designing work by enabling integration of thermal hydraulic and structural analysis. Since PARTS is a tool to perform the integrated thermal hydraulic-structural analysis under various conditions, it needs to calculate rapidly. At the point, the Green function method seems to be the most Promising stress analysis procedure for PARTS. The Green function method figures out thermal transient stress arising in structures in the form of convolute integration corresponding to fluids' step temperature changes. It is expected to calculate faster than Finite Elemental Method. Hitherto, the Green function method has been used to describe the response to sole thermal fluid with a constant heat transfer coefficient. In this report, the Green function method is extended to cope with a cylinder touching two thermal fluids with variable heat transfer coefficients (inside and outside surfaces contacting with primary and secondary coolants respectively) and is confirmed to be sufficiently applicable to such condition.
Yoo, Yeon-Sik
JNC TN9400 2001-120, 27 Pages, 2002/01
These days, this interest on LBB(Leak before Break) design becomes to be rising in the viewpoint of the cost reduction and structural inter-grity for the commercialization of FBR plants, LBB design enables pla-nts to be shut down safely before occuring unstable fracture by dete- cting the leak rates even if a crack initiates and penetrates a wall thickness. It is necessary to assess crack growth and penetration be- havior considering in-service conditions under operation temperature, leak retes considering detector capability and unstable fracture quan-titatively for LBB assessment. The governing service of FBR can be identified thermal expansion stress andthermal transient stress be- cause the operation temperature is higher than that of LWR and the li-quid metals contain higher heat transfer coefficient than water. On that reason, the use of 12Cr type ferrite steel in the primary cooling system is investigated for reducing stress in the design of large- scale FBR. This stud
kasahara, Naoto; *
JNC TN9400 2001-118, 69 Pages, 2002/01
JNC-TN9400-2001-118.pdf:1.58MB
Temperature fluctuation from incomplete fluid mixing induces fatigue damages on structures of nuclear components, which should be prevented. For rational analyses of this phenomenon, the authors have developed a frequency response function of thermal stress induced by fluid temperature fluctuations on the fixed spatial boundary. 0n the other hand, actual components have other stress generation modes from multi-dimensional spatial fluctuations of fluid temperature such as traveling of thermal stratification layers and hot/cold spots. This study has extended the frequency response function to be applied to spatial fluctuations problems of fluid temperature. Different characteristics of spatial fluctuation problems from fixed ones are their stress dependencies on both frequencies and traveling distances. When the distance is long and the frequency is very low, thermal stress is not attenuated. At actual components, stresses usually attenuate because the distances are short and frequencies are high. A proposed frequency response function rationally evaluates thermal stress induced by spatial traveling according to their traveling distances and frequencies.
kasahara, Naoto; *
JNC TN9400 2001-085, 68 Pages, 2001/09
JNC-TN9400-2001-085.pdf:1.41MB
Temperature fluctuation from incomplete fluid mixing induces fatigue damages on structures of nuclear components, which should be prevented. For rational analyses of this phenomenon, the authors have developed a frequency response function of thermal stress induced by one-dimensional temperature gradient across wall thickness. 0n the other hand, it is pointed out that existence of other stress modes from multi-dimensional structure with complex constraint conditions. This study has extended the frequency response method for adoption to multi-dimensional problems by introducing constraint efficiency factors. Applicability of this function was validated for multi-dimensional problems such as thermal stratification problems and hot/cold spot ones.
Kawasaki, Nobuchika;
JNC TN9400 2001-090, 102 Pages, 2001/07
JNC-TN9400-2001-090.pdf:2.56MB
Both reliability and safety have to be further improved for the successfull commercialization of FBRs. At the same time, construction and operation costs need to be reduced to a same level of future LWRs. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of fabrication, installation, plant system design, safety design, operation and maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Failure probability is considered as a candidate index. Therefore we decided to make a model calculation using failure probability and judge its applicability. We first investigated other probabilistic standards like ASME Code Case N-578. A probabilistic approach in the structural integrity evaluation was created based on these results, and also an evaluation flow was proposed. According to this flow a model calculation of creep-fatigue damage was performed. This trial calculation was for a vessel in a sodium-cooled FBR. As the result of this model calculation, a crack initiation probability and a crack penetration probability were found to be effective indices. Last we discussed merits of this System Based Code, which are presented in this report. Furthermore, this report presents future development tasks.
Furuhashi, Ichiro*; kasahara, Naoto
JNC TN9400 2001-089, 24 Pages, 2001/07
JNC-TN9400-2001-089.pdf:0.63MB
The present analytical investigations of thermal expansion stresses at hot-leg piping of advanced sodium-cooled loop-type reactors clarified their mechanisms and sensitivities to design parameters. The major mechanisms of thermal stresses identified in this study are summarized as: (1)The elbow has a large flexibility and acts like the hinge in the pipelines. (2)Vertical displacement of the IHX with the optimized support position can compensate thermal expansion displacement of the piping system in the vertical direction. (3)This piping system absorbs thermal expansion loads in the horizontal direction by bending transformation of the inner and outer vertical pipes. The major design parameters and their sensitivities are evaluated as follows: (1)Flexibi1ity of the elbow(high): The current design conditions are in an insensitive range. (2)Stiffness of the Y-piece (Intermediate): Stress of the Y-piece is sensitive and other portions are insensitive. A design with the rigid Y-piece is feasible. (3)Stiffness of the IHX nozzle (Insensitive): A design with the rigid nozzle is feasible. (4)Relative displacement in the vertical direction (Sensitive): The current support position is optimized and deviation within 
10mm is feasible. (5)Radius of the elbow (Insensitive): The current design with the short elbow is feasible. Adoption of l2Cr steel makes the current design of the hot leg piping system feasible. The sensitivity diagrams obtained from this study can predict thermal expansion stresses with other values of the design parameters as well.
Ando, Masanori
JNC TN9400 2001-088, 57 Pages, 2001/07
JNC-TN9400-2001-088.pdf:1.42MB
A reactor vessel of fast reactor plants contains high temperature liquid sodium and its upper end is supported from a concrete structure of low temperature. Therefore, a sharp temperature gradient will arise at the vessel wall near the sodium surface and a large thermal stress will be generated around this part. From the economical viewpoint, a simply designed reactor vessel is desirable, but this usually tends to make the stress larger. For this reason, design evaluation methods that can permit large thermal stress are necessary. The purposes of this research are to grasp influence factors and the sensitivities of the thermal stress to those factors, and to identify important points of the structural design of the reactor vessels. The results obtained from this research are as follows. (1)Sn (range of primary+secondary stress intensity) is sensitive to the thermal boundary conditions in the reactor vessels. By adjusting the heat transfer between cover gas and reactor vessel, Sn decreased. (2)By considering variation of the sodium surface level during starting up, the sign of the thermal stress near the sodium surface is reversed, and Sn is thereby increased. (3)By changing thickness of vessel to 30mm from 50mm, Sn is decreased by 15%. These results can be utilized for the pursuit of improved design of the reactor vessels.