Hiroshi Shibamoto

Development of Guidelines for Thermal Load Modeling

Conceptual design studies of Japanese commercialized fast reactors are now carried out as “Feasibility Study on Commercialized FR Cycle Systems”. Aiming for economical improvement, these plants adopt innovative design, which leads to make thermal loads more severe. To certify the design concepts and validate structural integrity, research and development of Fast Reactor Structural Design Standard (FDS) for commercialized fast reactor components is now under way. Among them, a set of guidelines for thermal load modeling is under development to overcome above thermal problems. System thermal transient loads sometimes become critical for plant design. Prediction of these loads has difficulties because many influence factors exist. So that, two kinds of modeling methods are recommended. One is the multi-linear approximation method to envelop scatter of those factors. Another is a combination method of thermal hydraulic-structure total analysis and the Design of Experiments. This method can grasp relation between influence factors and induced thermal stress without conservative design factors. Furthermore, screening method prior to modeling is provided. In actual plants, high cycle fatigue failure sometimes occurred, due to thermal striping phenomenon. The guidelines also deal with the modeling method of thermal striping loads. For consideration of attenuation mechanism of temperature fluctuation, a frequency response function of thermal stress is utilized. This function enables us to evaluate sensitivities of thermal stress to frequencies of temperature fluctuation, constraint conditions of components etc.Copyright

Development of the Guideline on Inelastic Analysis for Design

The guideline on inelastic analysis for design, one of the key items of Fast Reactor Design Standard (FDS), is being developed. The basic policies of this guideline are as follows: (a) to emphasis conservative analysis output rather than nominal value representing actual behavior, (b) to clarify the applicable area for assurance of conservative results. With such concepts, it would be possible that the guideline provides useful explanations on the manner of analysis and estimation in the form of concrete examples of design as well as general rules (somehow vague). As the first step of the guideline development, the following five issues to be solved were extracted: 1) applicable area, 2) selection of constitutive equation, 3) modeling method of the load history, 4) ratchet strain and creep fatigue damage evaluation methods by inelastic analysis and 5) example design problems to check users’ analysis quality and to complement the general rules. In parallel, inelastic analyses with the promising constitutive equations were applied by way of trial to obtain rough presumption on their effects on structural design of the components. As a result, all inelastic analyses provided smaller cumulative strains and equivalent strain ranges than the existing design method based on elastic analysis, suggesting advantage of introducing them into actual design.Copyright

Extension of Applicable Area of Thermal Stress Charts (Thermal Stress of Plate Subjected to Heat Transfer on Both Surfaces)

Thermal stresses in plate structures subjected to heat transfer on both surfaces are analyzed. Through analytical study, new design charts for temperature and stress are developed. Design charts for plate, where one surface is adiabatic, are conventionally used. By using new ones, the applicable area of design charts can be greatly extended. New temperature charts are normalized by steady-state temperature. New stress charts are normalized by the steady-state stress at a fixed back surface temperature in order to reduce reading errors. Maximum stress charts for step or ramp change of fluid temperature are developed. Stress reduction by transferring step change to ramp change can be read directly from the charts.

The Study on Evaluation Method for Primary Stress without Evaluation Section

This paper describes an evaluation method of primary stress in three dimensional (3-D) structures. In “Design by Analysis” for nuclear components, the stresses in structures are classified into the primary and the secondary stresses. The primary stress in axisymmetric structures can be evaluated by linealization of stress distribution in the specified section, but it is difficult to define the evaluation section in the 3-D structures, and to evaluate the primary stress with the conventional procedure. From this reason, the alternative evaluation method is needed. In this study, the evaluation method of primary stress in 3-D structures with elastic-plastic analysis is proposed utilizing the feature of primary stress that is independent from stress redistribution. The proposed method is verified through application to example problems.

Measurement of Thermal Ratcheting Strain on the Structures by the Laser Speckle Method

Prevention of thermal ratcheting is an important problem for high temperature components of fast breeder reactors that are subjected to cyclic thermal loads. To clarify ratcheting behaviors, structural model tests were planned. Strain measurement is important for understanding the thermal ratcheting phenomenon The conventional measurement by strain gauge is difficult at high temperature. Then, Laser speckle strain measurement system using the dual-beam set-up was developed to apply to high temperature structural model tests. This system was applied to the thermal ratcheting tests, which demonstrated the actual operative conditions of reactor vessels. Through comparison with uniaxial test results obtained by extensometers, the laser speckle method was verified. Measured data of structural model tests were utilized to certify the guidelines of inelastic analysis for design, which provide prediction method of strain in components of fast reactor.Copyright

A Rational Identification of Creep Design Area Using Negligible Creep Curves

For extension of non-creep design area and simplification of design procedures, a rational identification method of creep design area by negligible creep (NC) curves was studied. NC curves of six kinds of austenite stainless and ferrite steels for fast reactors were determined based on domestic material data. NC curves provide the relation between temperature and time that does not induce damageable creep strain under the constant stress 1.5Sm (Sm: design stress intensity). In existing Japanese design guides, non-creep design area is severely restricted by constant upper temperature limit for austenite stainless steel and ferrite steel. In the case of 316FR steel and SUS410J3, which are candidate materials of Japanese commercialized fast reactors and have excellent material property, this limit can be extended by NC curve concept considering the duration of high temperature operation. NC curves under secondary stress considering stress relaxation were also studied. However, rationalization effect was insufficient whereas evaluation process was too complex. Therefore, at the present stage, NC curves at constant stress level 1.5Sm were adopted to identify creep design area. The concept of NC curve was introduced into the interim structural design guide for commercialized fast reactors in Japan to simplify the creep design of fast reactor systems. Utilizing these curves, non-creep design becomes possible for components operated at comparatively lower temperature in normal condition.Copyright

Application of a Classification Method to Obtain Primary Stresses Without Evaluation Sections to Perforated Structures

Conceptual design studies of Japanese commercialized Fast Breeder Reactor (abbreviated to FBR) are carried out. With the aim of reducing the construction cost, number and size of components are reduced. For example, Intermediate Heat Exchangers (abbreviated to IHXs) are integrated with primary pumps; number of main cooling loops is reduced from 3 to 2 and piping is shortened; the reactor vessel is simplified. Accompanying the reduction of cooling loops, Steam Generators (abbreviated to SGs) become larger, and semi-sphere perforated plates in 3-D structures are adopted for large sized steam generator to endure primary stress caused by pressurized steam. In “Design by Analysis” for these nuclear components, the stresses in these structures are classified into primary and secondary stresses. Conventional method of stress classification uses an evaluation section; however, it is difficult to define evaluation sections in 3-D structures. For this reason, an alternative evaluation method without evaluation sections, which can be easily applied to 3-D structures, is necessary. Primary stress is the so-called load controlled stress; this is decided as the stress in equilibrium with external loading and is independent of inelastic behavior of materials. Paying attention to above feature, there is an idea to obtain primary stress from a Re-distribution node (abbreviated to R-node), where stress is constant during stress re-distribution. One of concrete method is GLOSS (Generalized local Stress Strain) method proposed by R. Seshadri et al. [1–3]. To generate artificial stress re-distribution, this method needs to carry out two elastic analyses with different values of material constants. All stresses are different between two calculations except R-node and the stress at this point can be evaluated as the primary stress. This method adopts elastic analysis with special material constants that is determined from post-processing of stress distributions of the first elastic analysis due to the insufficient power of earlier computer. Recent progress of computer technologies reduces computational time and cost of inelastic calculation. Furthermore, it is not difficult for an elastic-plastic calculation to be carried out with classical constitutive equations provided by commercial codes. In this paper, an alternative method is proposed, which uses elastic-plastic analysis for artificial stress re-distribution instead of elastic analysis in GLOSS method. The proposed method is confirmed through the application to the example problem for a thick cylinder and the ability to be applied to actual structures is confirmed through the application to the example problem for a perforated plate.Copyright

Research and Development Issues for Fast Reactor Structural Design Standard (FDS)

For the realization of safe and economical fast reactor (FR) plants, the Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on a research project titled “Feasibility Study on Commercialized FR Cycle Systems.” To certify the design concepts and validate their structural integrity, the research and development of the “Fast Reactor Structural Design Standard (FDS)” is recognized as being an essential theme. FDS considers the general characteristics of FRs and the design requirements for their rationalization. Three main problem areas related to research and development issues were identified by FDS. The first is “refinement of failure criteria,” which takes characteristic design conditions into account. The next is the development of “guidelines for inelastic design analysis” in order to predict the elastic plastic and creep behaviors of high-temperature components. Finally, efforts are being made toward preparing “guidelines for thermal load modeling” for FR component design where thermal loads are dominant. Their research plans and current status are explained.Copyright