ISSN:0550-3043    

DOI:10.13182/NT70-A28721

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

The first of this series contained in this issue ofNuclear Applications and Technologyis on the topic of fuel cladding models. The integrated fuel-element performance model must consider irradiation effects on cladding properties and the development of stress analysis and failure analysis techniques for the cladding. The papers presented are intended as a survey of some of the important ideas involving these aspects of fuel element behavior. The papers do not as a body pretend to represent all the work which is going on in the field, but they do serve as an excellent starting point for an understanding of the present state-of-the-art.

ISSN:0550-3043    

DOI:10.13182/NT70-A28722

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

Fast-neutran irradiations in the EBR-II have been completed an biaxial stress rupture, creep, and tensile specimens of AISI 304 and 316 stainless steel. Postirradiation test results show that irradiations in the 480 to 650°C range to fluences of 1×1022n/cm2(E>0.1 MeV) substantially reduce the time-dependent rupture life and ductility of these materials. Tensile ductility is also severely reduced.Bulk-density measurements and electron-microscopy examinations on specimens of annealed 304 from EBR-II core components and mechanical property specimens have been made for fluence levels to 7×1022n/cm2and at temperatures in the 360 to 470°C range. Both the bulk-density measurements and microscopy examinations correlate well and indicate that volume changes of 4% can be expected under these conditions. The temperature and fluence dependency for annealed 304 stainless steel has been determined and can be expressed as:The mechanisms responsible for the observed degradation of mechanical properties and metal swelling are being studied. Some observatians are presented. However, as yet, no adequate nucleatian and growth model has been determined to enable an acceptable extrapolatian of these data-to-goal fluence levels to be achieved in Liquid Metal Fast Breeder Reactor core companents or fuel-pin cladding.

ISSN:0550-3043    

DOI:10.13182/NT70-A28723

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

A model has been developed for the evolution of voids and dislocation loops during fast neutron irradiation of austenitic stainless steel. The model is based on a thermodynamic approach that calculates void nucleation and growth rates in terms of the supersaturation of vacancies and interstitials. It is recognized that the steady-state point-defect concentrations decrease with fluence as the result of the creation of additional sinks (voids and loops).The ability to monitor both the microstructural development and the steady-state concentrations of defects allows discussion of the in-pile mechanical properties. The yield strength of austenitic stainless steel is expected to increase rapidly during irradiation at 400°C due to the effectiveness of voids and dislocation loops as obstacles to dislocation motion. Irradiation at 600°C is predicted to result in a slowly increasing yield strength.In-reactor creep behavior is discussed in terms of a climb-controlled model for a dispersion strengthened system. Radiation-enhanced climb is expected to predominate at lower temperatures and stresses over the thermal climb component. Discussion of the possible effects of neutron flux and fluence on the in-pile steady-state creep rate is also included.

ISSN:0550-3043    

DOI:10.13182/NT70-A28724

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

Short-term tensile data have been generated using a new experimental technique which allows the true total axial strain rate to be maintained constant all the way to fracture. Tensile data at 650°C (1200°F) for irradiated and unirradiated specimens of AISI 304 and 316 stainless steel are presented and compared.A new relationship between low-cycle fatigue and short-term tensile behavior is discussed and applied to data for irradiated and unirradiated material. The effectiveness of this approach is shown to be excellent. This method should allow the low-cycle fatigue behavior for irradiated material to be estimated with acceptable accuracy.Hold times at peak strain have a noticeable effect on fatigue behavior as evidenced in tests at 650°C. These effects are most pronounced when hold periods are employed in only the tension portion of the cycle. An interesting correlation of hold-time data is presented, based on a logarithmic plot of time to fracture vs the length of the hold period. Another important correlation involves a relationship which identifies a method for estimating hold-time effects from a knowledge of the effect of strain rate on low-cycle fatigue behavior.

ISSN:0550-3043    

DOI:10.13182/NT70-A28725

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

The stress and strain distributions produced in nuclear-fuel-element cladding by the expansion of cracked pellets have been calculated both analytically and by numerical methods. As the radial (and transverse) pellet cracks open, the tendency for the cladding to stretch preferentially over them is reduced by frictional sliding at the pelletclad interface. The frictional forces opposing sliding are intensified by a high coolant pressure (which holds the can onto the fuel) while the ability of the clad to resist the frictional forces, without being locally deformed, depends on its strength. The coefficient of friction, the angle between adjacent radial pellet cracks, and the creep properties of the clad have, in theory, strong effects upon the tendency for clad strain to be concentrated over opening pellet cracks; confirmation of the correctness of these deductions has been obtained from laboratory experiments in which cladding has been stretched by cracked pellets on an expanding mandrel.The numerical analysis has enabled a detailed study of the strain-concentrating processes to be made, revealing that swelling of the pellet during a period at reduced-heat rating increases its diameter so that when high rating operation is resumed and the pellet expands, the cladding is stretched by an amount that depends on the magnitude of the prior swelling. During the expansion of the fuel pellet, the radial cracks in it open up and preferentially strain the adjacent cladding so that the clad strain due to fuel swelling, like that due to thermal expansion of the fuel, tends to be concentrated in arcs of cladding adjacent to pellet cracks. This process is repetitive, occurs whatever the magnitude of the coolant pressure, and is accentuated by the presence of a circumferential temperature gradient in the cladding.

ISSN:0550-3043    

DOI:10.13182/NT70-A28726

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

The various additions to the CYGRO fuel-rodanalysis technique in order to calculate ratchetting effects are described. These effects include fuel cracking, clad collapse, friction between fuel and clad, clad anisotropy, and effects of neutron flux on clad creep. By reasonable choice of parameters, good agreement can be obtained with tests on axial elongations of non-freestanding fuel rods. There is a pronounced sensitivity of these predictions to the value of creep enhancement as a result of neutron flux. Predictions of diameter changes are believed to be inherently less accurate because of the masking effects of ridging, wrinkling, and clad collapse.

ISSN:0550-3043    

DOI:10.13182/NT70-A28727

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

The computer program CRASH allows the calculation of triaxial stress and strain states in the sheath of cylindrical fuel elements when creep or plasticity occur in the cladding material. Any creep or plasticity law may be used in the program and any type of external stresses and strains assumed. A short description of the program is given in the first part of this paper. Its second part deals with some applications: examples of plastic thermal ratchetting under cyclic conditions and creep of cladding due to contact pressure are presented. These few examples show that the CRASH program is a useful and flexible tool for the cladding design of fuel elements. Moreover, its calculation time is sufficiently small as to allow intensive parametric studies.

ISSN:0550-3043    

DOI:10.13182/NT70-A28728

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

A three-dimensional calculational method has been developed for use in studying xenon oscillations and their control. This method is based on nodal analysis and a modified one-group theory and includes all major feedback mechanisms which are important in xenon transients. Comparisons have been made between the results obtained with this method and with more sophisticated methods and good agreement has been found in all cases. Significant reductions in computing times are obtainable with this method.

ISSN:0550-3043    

DOI:10.13182/NT70-A28729

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor

摘要:

The neutronic behavior of fusion reactor blankets is discussed, and transport-theory calculations are presented for two blanket designs. The areas investigated are (1) tritium breeding, (2) nuclear heating, and (3) neutron irradiation effects within the vacuum wall of the blanket, i.e., neutron-induced (a) atom displacements and (b) helium and hydrogen production. The two blanket designs considered consist of niobium as the vacuum wall and structural material, lithium or lithium in combination with lithium-beryllium fluoride (called“flibe”) as the coolant, and graphite as the neutron moderator and reflector. The results indicate that the tritium breeding potential of both designs is promising. The results also show that the tritium-breeding and nuclear heating characteristics of the lithium-flibe blanket are inferior to those of the lithium blanket. The calculated atom displacement rates and production rates of helium and hydrogen within the vacuum wall are essentially the same for both blanket designs.

ISSN:0550-3043    

DOI:10.13182/NT70-A28730

出版商:Taylor&Francis    

年代:1970

数据来源: Taylor