RCC-MR Code Research Articles

Investigation on buckling of FBR vessels under seismic loadings with fluid structure interactions

The critical thin walled shell structures in the reactor assembly of a pool type fast breeder reactor (FBR) are the main vessel, inner vessel and thermal baffles. On these structures, the seismic events impose major forces by developing high dynamic pressures, thereby causing a concern on structural integrity due to buckling. An integrated analysis for determining realistic forces and critical buckling loads at any instant during the seismic event has been carried out for the vessels of a typical 500 MWe pool type fast breeder reactor. The dynamic forces including pressure distributions generated on the vessel surfaces are extracted from the seismic analysis carried out for the reactor assembly. The seismic forces thus computed from axisymmetric analysis are transmitted appropriately to 3D shell geometries for the buckling analysis. In view of high computational time needed for carrying out buckling analysis at every time increment, the elastoplastic analysis is carried out only at a few critical time steps which are identified based on strain energies that are associated with the shear and compressive stresses developed at the portions of the vessels prone to buckle. The shear buckling of main vessel straight portion and buckling of toroidal portion of inner vessel and thermal baffles are found to be important. The possible randomness of support excitation time histories is accounted for by compressing as well as expanding the nominal time histories by 10%. Buckling strength reduction factors due to the initial geometrical imperfections are adopted from the literature. The inner vessel is found to be the most critical component which buckles under seismic forces induced by a safe shutdown earthquake with a load multiplier of 1.52, which is higher than the minimum factor of safety of 1.3 required as per the design code RCC-MR [RCC-MR: edition, 2002. Design and construction rules for mechanical components for FBR nuclear islands, vol. 1, section I, subsection B. AFCEN, Paris, in press] for service level D conditions.

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part V: Elements of validation

French nuclear codes include flaw assessment procedures: the RSE-M Code “Rules for In-service Inspection of Nuclear Power Plant Components” and the RCC-MR code “Design and Construction Rules for Mechanical Components of FBR Nuclear Islands and High Temperature Applications”. Development of analytical methods has been made for the last 10 years in the framework of a collaboration between CEA, EDF and AREVA-NP, and by R&D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of articles consists of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II–IV provide the compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). This part presents validation of the methods, with details on the process followed for their development and of the evaluation accuracy of the proposed analytical methods.

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part I: General overview

Two French nuclear codes include flaw assessment procedures: the RSE-M code “Rules for In-service Inspection of Nuclear Power Plant Components” and the RCC-MR code “Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications”. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R&D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of five parts: this first one presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II–IV provide compendia for specific components: plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for their development and on evaluation of the accuracy of the proposed analytical methods. This first article of the series presents an overview of the calculation of K I and J in these two codes and describes briefly the defect assessment analyses. Specific details in the Appendix A16 of RCC-MR (LBB procedure and creep analyses) are also introduced in this article.

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part IV: Cracked elbows

Two French nuclear codes include flaw assessment procedures: the RSE-M Code “Rules for In-service Inspection of Nuclear Power Plant Components” and the RCC-MR code “Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications”. Development of analytical methods has been made for the last 10 years through a collaboration between CEA, EDF and AREVA-NP, and through R&D actions involving CEA and IRSN. These activities have led to unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of papers is composed of five parts: the first presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II–IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation of the methods, with details on their accuracy. This paper presents the stress intensity factor and J calculation for cracked elbows. General data applicable for all defect geometries are first presented, and then, compendia for K I and σ ref calculations are provided for the available defect geometries.

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part II: Cracked plates

French nuclear codes include flaw assessment procedures: the RSE-M Code “Rules for In-service Inspection of Nuclear Power Plant Components” and the RCC-MR code “Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications”. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R&D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II–IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for the development and evaluation of the accuracy of the proposed analytical methods. This second article in the series presents all details for the stress intensity factor and J calculations for cracked plates. General data applicable for all defect geometries are first presented, and then, available defect geometries where compendia for K I and σ ref calculation are provided are given.

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part III: Cracked pipes

French nuclear codes include flaw assessment procedures: the RSE-M Code “Rules for In-service Inspection of Nuclear Power Plant Components” and the RCC-MR code “Design and Construction rules for mechanical components of FBR nuclear islands and high-temperature applications”. An important effort of development of these analytical methods has been made for the last 10 years in the frame of collaboration between CEA, EDF and AREVA-NP, and in the frame of R&D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress-intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of the RSE-M and in the 2007 edition of the RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II–IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation, with details on the accuracy of the proposed analytical method. This third part in the series presents details of the stress intensity factor and J calculations for cracked pipes. General data applicable for all defect geometries are first presented, and then, compendia for K I and σ ref calculations are provided for specific cases.

Fatigue crack propagation analysis of surface cracks in pipe elbows

Piping elbows are one of the critical components of the cooling piping systems in power plants. This paper presents the details of fatigue crack propagation analysis based on code procedure and FEM and compares the results with experimental observations. The fatigue crack propagation was evaluated based on the inelastic range J integral, J s, which was obtained by applying the plastic correction factor (from RCC-MR code) to the elastic J-integral values obtained from the code as well as from the elastic finite element analysis. The fatigue crack propagation was also studied directly using the elastic–plastic J-integral value obtained from the finite element analysis. It was observed that the fatigue life calculated based on the modified J s-integral values was closer on the conservative side to the experimental results, than that from J-integral obtained directly from elastic–plastic finite element analysis.

Experimental creep life assessment of tubular structures with geometrical imperfections in welds with reference to fast reactor plant life

Creep damage is a major life limiting factor for components operating at high temperature. For the fast breeder reactor (FBR), the hot sodium pool components and the steam generator are critical in determining the operating temperatures from structural mechanics considerations. Therefore, creep fatigue damage at various critical locations is estimated based on viscoplastic analyses, in compliance with the high temperature design rules of RCC-MR. In addition, welded joints of steam generator tubes with spigots made of modified 9Cr–1Mo and the welds of fuel pins made of 20% cold worked stainless steel to annealed end plugs, which have certain geometrical imperfections in the welds are also critical and the applicability of existing design rules to these components needs to be investigated carefully. Towards assessing the life of such tubular components realistically, accelerated tests were conducted simulating the creep damage to rupture. The generated test data were investigated by applying the existing creep damage assessment procedure recommended in the design code RCC-MR and the results demonstrated that with the recommended multiaxial creep damage assessment rule, the creep crack initiation lives have been predicted accurately. Further, the extrapolated life predicted by applying an appropriate Larson–Miller parameter indicates that the circumferential welds in the steam generator tubes and fuel pins are not life limiting factors in the design.

Thermal fatigue behaviour for a 316 L type steel

This paper deals with initiation and growth of cracks produced by thermal fatigue loadings on 316 L steel, which is a reference material for the first wall of the next fusion reactor ITER. Two types of facilities have been built. As for true components, thermal cycles have been repeatedly applied on the surface of the specimen. The first is mainly concerned with initiation, which is detected with a light microscope. The second allows one to determine the propagation of a single crack. Crack initiation is analyzed using the French RCC-MR code procedure, and the strain-controlled isothermal fatigue curves. To predict crack growth, a model previously proposed by Haigh and Skelton is applied. This is based on determination of effective stress intensity factors, which takes into account both plastic strain and crack closure phenomena. It is shown that estimations obtained with such methodologies are in good agreement with experimental data.

Austenitic stainless steels, status of the properties database and design rule development

In parallel with the new tasks initiated to substantiate the existing database for the reference structural material (type 316LN-IG) of ITER, interim design criteria are being developed to guide subsequent design stages. The French RCC-MR codes for fast breeder reactors, incorporating rules from other ITER partner codes and those needed to meet specific fusion requirements, are used for this purpose. This paper presents the current status of materials data and design rules for type 316LN-IG steel and describes how the irradiation effects are taken into account.

Operating temperatures for an FBR

Structural mechanics aspects related to operating temperatures of a typical pool type 500 MWe fast breeder reactor are discussed. The critical high temperature components are analysed in detail based on elastic, inelastic and viscoplastic deformation theories, and life is predicted in accordance with the rules of design code RCC-MR 87. Analysis indicates that the control plug is the most critical component in the reactor which limits the reactor outlet temperature to 820 K with a temperature rise of 160 K across the core.

Behaviour of first-wall components under thermal fatigue

Physical and thermal/mechanical interaction with plasma is felt to be the most limiting phenomenon for the first wall lifetime of the next generation Tokamak fusion reactors. This paper describes the thermal fatigue test programme on NET/ITER FW components. The lifetime predictions performed according to ASME and RCC-MR codes are discussed on the basis of the numerical evaluation and of the experimental results.

Codes and Standards for Fast Breeder Reactor Components—A European Point of View

In 1984, five European governments (Belgium, France, Italy, the United Kingdom, and the Federal Republic of Germany) agreed to a coordinated program for FBR research, development, and construction. The ultimate outcome must be a common design code. The authors discuss how drawing up such a design code represents a major challenge in the field of international collaboration. It seems likely that the existing RCC-MR code will be used as the starting point for the development of European design rules. Recently, European manufacturers set up a committee with the object of amending and supplementing RCC-MR. The results of this committee are discussed.