Methodology for the Determination of a Set of Safety Factors That are Consistent for Design Code and Fitness-for-Service Code: Case Study for Fast Breeder Reactors

2009 ◽  
Author(s):  
Tai Asayama
Keyword(s):  
Failure Modes ◽  
Fatigue Crack Initiation ◽  
Seismic Load ◽  
Safety Factors ◽  
Design Code ◽  
Secondary Stress ◽  
Primary Stress ◽  
Creep Fatigue ◽  
Fast Breeder Reactors ◽  
Breeder Reactors

This paper proposes a method for determining a set of safety factors taking account of multiple failure modes and their interactions. The purpose of the work is to materialize the System Based Code concept for the applications to fast breeder reactors. Current structural codes prevent failure by limiting primary stress, excessive strain, buckling and fatigue damage. However, the relationships between safety factors in these criterion and failure modes are not necessarily clear. For example, safety factors in the limitation of primary stress are considered to cover not only ductile fracture but also fracture due to crack like defects. When the System Based Code concepts, one of the most important of which is designing to target reliability, it is essential to determine safety factors so that they explicitly correspond to particular failure modes. This paper deals with the most important two failure modes to be prevented in fast breeder reactors (FBRs), that is, primary stress due to seismic load and secondary stress due to creep-fatigue. Safety factors that are consistent for design code and fitness-for-service code are derived by the following steps: 1) Formulations of continuous evaluation of reliability are derived for both fracture by primary stress and creep-fatigue crack initiation and propagation due to secondary stress, with their interaction taken into account, 2) Reliability is calculated for various combinations of loading conditions, 3) Safety factors corresponding to various levels of target reliabilities are investigated and compared with currently used ones. The safety factors thus determined not only have firm physical basis but also contribute to enlarge design windows for fast breeder reactor components. Items to be further investigated for the methodology to be implemented in current code are also discussed.

Study on Simplified Prediction Method of Thermal Ratchet Deformation Based on Parallel Bar Model

2014 ◽  
Author(s):  
Toshiaki Kokufuda ◽  
Naoto Kasahara
Keyword(s):  
Elevated Temperature ◽  
Inelastic Deformation ◽  
Evaluation Method ◽  
Failure Modes ◽  
Prediction Method ◽  
Inelastic Strain ◽  
Primary Stress ◽  
Redistribution Mechanism ◽  
Creep Fatigue ◽  
Calculation Results

For elevated temperature structures such as fast breeder reactor components, inelastic deformation is likely to occur because of reduction of yield stress and occurrence of creep deformation. The typical failure modes for elevated temperature structures are excessive deformation caused by the accumulation of inelastic deformation and creep fatigue caused by inelastic strain concentration at structure discontinuities. In order to prevent such failures, it is necessary to evaluate inelastic deformation adequately. Thermal ratchet deformation, namely the progressive plastic deformation induced by cyclic thermal stress with uniform primary stress, has some possibility resulting in excessive deformation. ASME boiler & pressure vessel code provides elastic evaluation methods for thermal ratchet. However, these methods are so focused on preventing thermal ratchet deformation, that it could be too conservative under some conditions. Therefore, a simplified elastic evaluation method to quantify thermal ratchet deformation is desired. In this paper, the simplified prediction method for thermal ratchet deformation is proposed using parallel bar model, which represents stress redistribution mechanism of cylindrical vessels. The solution of thermal ratchet deformation of parallel bar model was derived and compared with FEM calculation results of cylindrical vessels. This theoretical solution is proposed as a prediction method for thermal ratchet deformation of cylindrical vessels. The applicable area of the proposed prediction method is the cylindrical vessel under linear and parabolic temperature distribution through the wall thickness.

Radiation Damage Studies with the HVEM

1972 ◽  
Vol 30 ◽  
pp. 680-681
Author(s):  
J. J. Laidler ◽  
B. Mastel
Keyword(s):  
Radiation Damage ◽  
Stainless Steels ◽  
Volume Expansion ◽  
Austenitic Stainless Steels ◽  
Test Facility ◽  
Fast Breeder Reactors ◽  
Breeder Reactors ◽  
Under Construction ◽  
Development Laboratory ◽  
Insight Into

One of the major materials problems encountered in the development of fast breeder reactors for commercial power generation is the phenomenon of swelling in core structural components and fuel cladding. This volume expansion, which is due to the retention of lattice vacancies by agglomeration into large polyhedral clusters (voids), may amount to ten percent or greater at goal fluences in some austenitic stainless steels. From a design standpoint, this is an undesirable situation, and it is necessary to obtain experimental confirmation that such excessive volume expansion will not occur in materials selected for core applications in the Fast Flux Test Facility, the prototypic LMFBR now under construction at the Hanford Engineering Development Laboratory (HEDL). The HEDL JEM-1000 1 MeV electron microscope is being used to provide an insight into trends of radiation damage accumulation in stainless steels, since it is possible to produce atom displacements at an accelerated rate with 1 MeV electrons, while the specimen is under continuous observation.

Structural Integrity for Nuclear Power Plant Against Excessive Load on Design Extension Condition

2013 ◽  
Author(s):  
Daigo Watanabe ◽  
Kiminobu Hojo
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Factors ◽  
Design Code ◽  
Acceptance Criteria ◽  
Integrity Assessment ◽  
Current Design ◽  
Factor Design ◽  
Excessive Load ◽  
Load And Resistance Factor

This paper introduces an example of structural integrity evaluation for Light Water Reactor (LWR) against excessive loads on the Design Extension Condition (DEC). In order to assess the design acceptance level of DEC, three acceptance criteria which are the stress basis limit of the current design code, the strain basis limit of the current design code and the strain basis limit by using Load and Resistance Factor Design (LRFD) method were applied. As a result the allowable stress was increased by changing the acceptance criteria from the stress basis limit to the strain basis limit. It is shown that the practical margin of the LWR’s components still keeps even on DEC by introducing an appropriate criterion for integrity assessment and safety factors.

Appraisal of Allowable Loads by Simplified Rules

1986 ◽  
Vol 108 (2) ◽  
pp. 131-137
Author(s):  
D. Moulin
Keyword(s):  
Experimental Investigation ◽  
Plastic Region ◽  
Thin Structures ◽  
Geometric Imperfections ◽  
Buckling Behavior ◽  
Simplified Method ◽  
Post Buckling ◽  
Initial Geometric Imperfections ◽  
Fast Breeder Reactors ◽  
Breeder Reactors

This paper presents a simplified method to analyze the buckling of thin structures like those of Liquid Metal Fast Breeder Reactors (LMFBR). The method is very similar to those used for the buckling of beams and columns with initial geometric imperfections, buckling in the plastic region. Special attention is paid to the strain hardening of material involved and to possible unstable post-buckling behavior. The analytical method uses elastic calculations and diagrams that account for various initial geometric defects. An application of the method is given. A comparison is made with an experimental investigation concerning a representative LMFBR component.

Comparison of Failure Modes of Piping Systems With Wall Thinning Subjected to In-Plane, Out-of-Plane, and Mixed Mode Bending Under Seismic Load: An Experimental Approach

2010 ◽  
Vol 132 (3) ◽  
Author(s):  
Izumi Nakamura ◽  
Akihito Otani ◽  
Masaki Shiratori
Keyword(s):  
Dynamic Behavior ◽  
Failure Modes ◽  
Seismic Load ◽  
Piping System ◽  
Shake Table ◽  
Shake Table Tests ◽  
Wall Thinning ◽  
Piping Systems ◽  
Out Of Plane ◽  
Plane Bending

Pressurized piping systems used for an extended period may develop degradations such as wall thinning or cracks due to aging. It is important to estimate the effects of degradation on the dynamic behavior and to ascertain the failure modes and remaining strength of the piping systems with degradation through experiments and analyses to ensure the seismic safety of degraded piping systems under destructive seismic events. In order to investigate the influence of degradation on the dynamic behavior and failure modes of piping systems with local wall thinning, shake table tests using 3D piping system models were conducted. About 50% full circumferential wall thinning at elbows was considered in the test. Three types of models were used in the shake table tests. The difference of the models was the applied bending direction to the thinned-wall elbow. The bending direction considered in the tests was either of the in-plane bending, out-of-plane bending, or mixed bending of the in-plane and out-of-plane. These models were excited under the same input acceleration until failure occurred. Through these tests, the vibration characteristic and failure modes of the piping models with wall thinning under seismic load were obtained. The test results showed that the out-of-plane bending is not significant for a sound elbow, but should be considered for a thinned-wall elbow, because the life of the piping models with wall thinning subjected to out-of-plane bending may reduce significantly.

Progress in Developing Constitutive Equations for Inelastic Design Analysis

1983 ◽  
Vol 105 (3) ◽  
pp. 273-276 ◽  
Author(s):  
C. E. Pugh
Keyword(s):  
Constitutive Equations ◽  
Inelastic Strain ◽  
Short Term ◽  
Inelastic Material ◽  
Fast Breeder Reactors ◽  
Breeder Reactors ◽  
Term Creep ◽  
Inelastic Design ◽  
Creep Deformations

A summary is given of the constitutive equations that have been developed for use in design assessments of elevated temperature components of liquid metal fast breeder reactors. The discussion addresses representations of short-term (plastic) and long-term (creep) inelastic material responses. Attention is given to improved representations of the interactions between plastic and creep deformations. Most of the discussion is in terms of constitutive equations that make use of the concept of separating the total strain into elastic, plastic, and creep portions. Additionally, some discussion is given of progress being made toward establishing design equations based on unified measures of inelastic strain that do not distinguish different strain portions.

Stress Analysis and Structure Optimization for the Opening Flat Cover of Pressure Vessels

2012 ◽  
Vol 538-541 ◽  
pp. 3253-3258 ◽  
Author(s):  
Jun Jian Xiao
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Stress Concentration ◽  
Stress Analysis ◽  
Structure Optimization ◽  
Pressure Vessels ◽  
Secondary Stress ◽  
Element Analysis ◽  
Primary Stress ◽  
Flat Cover

According to the results of finite element analysis (FEA), when the diameter of opening of the flat cover is no more than 0.5D (d≤0.5D), there is obvious stress concentration at the edge of opening, but only existed within the region of 2d. Increasing the thickness of flat covers could not relieve the stress concentration at the edge of opening. It is recommended that reinforcing element being installed within the region of 2d should be used. When the diameter of openings is larger than 0.5D (d>0.5D), conical or round angle transitions could be employed at connecting location, with which the edge stress decreased remarkably. However, the primary stress plus the secondary stress would be valued by 3[σ].

Constitutional studies on carbide nuclear fuels for fast breeder reactors

1980 ◽  
Vol 89 (1) ◽  
pp. 13-26 ◽  
Author(s):  
M. Peatfield ◽  
N.H. Brett ◽  
H.R. Haines ◽  
P.E. Potter
Keyword(s):  
Nuclear Fuels ◽  
Fast Breeder Reactors ◽  
Breeder Reactors
Sign in / Sign up

Export Citation Format

Share Document