The Reactor Pressure Vessel (RPV) is an essential component liable to limit the lifetime duration of PWR plants. The structural integrity assessment of defects in RPV subjected to PTS transients made at an European level generally doesn't take into account the potential beneficial effect of the load history (warm pre-stress WPS). A 3-year European Research & Development programme has been started in January 2002 as part of the Fifth Framework of the European Atomic Energy Community (EURATOM). The SMILE project ('Structural Margin Improvements in aged-embrittled RPV with Load history Effects') is one of a 'cluster' of Fifth Framework projects in the area of plant Life Management. It aims to give sufficient elements to demonstrate, to model and to validate the beneficial WPS effect in a RPV structural integrity assessment. Finally, this project aims to harmonize the different approaches in the European Codes and Standards regarding the inclusion of the WPS effect in the RPV structural integrity assessments. All elements necessary to propose a method to take into account this effect will be gathered or obtained. This will be done through experimentals works, leading to a deep understanding of metallurgical and mechanical phenomena, and through numerical works and development of models. The results will permit a much more precise prediction of possible brittle fracture in a RPV during a severe PTS transient. The present paper describes the aims and objectives of SMILE project, its interactions with NESC (Network for Evaluating Steel Components) and other European projects, and gives details of its various Work-Packages. Finally, the on going progress of the project is described.

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In the case of Boiling Water Reactors, Inter Granular Stress Corrosion Cracking (IGSCC) near the critical circumferential welds of SS 304 core shroud has been reported worldwide. Potential safety concerns were raised by regulatory bodies for the 360 degrees circumferential separation of core shroud following the postulated pipe break in the re-circulation line. Such separation of the shroud might either prevent full insertion of the control rods or open a gap in the shroud large enough to preclude adequate core cooling. Based on Electrical Power Research Institute (EPRI) co-ordinated Vessel and Internals (VIP) project, NRC through its generic letter 94-03 issued guidelines for the assessment of core shroud response for design basis blow down accidents due to re-circulation line break. The in-service inspection carried out at TAPS-BWR has demonstrated that there is no indication of flaw like defects. However, a major safety evaluation programme was initiated to assess the integrity of TAPS-BWR core shroud in Reactor Safety Division, BARC, Trombay. One of the important issues after the initiation of the blow down is determination of acoustic load and the associated fluid-structure interaction response evaluation of the core shroud. The present paper focuses on this problem and the coupled fluid-structure interaction analysis results are reported for the core shroud of TAPS-BWR with an in-house three-dimensional finite element code FLUSHEL. For the safety evaluation of the TAPS-BWR core shroud the performance of code FLUSHEL was evaluated with the analysis of standard benchmark example of HDR-PWR core barrel blow down experimental results. The blow down induced depressurisation wave is traced within the two phase fluid medium of downcomer and lower plenum and is found to be consistent with the experimental results (within an accuracy of 11%) of reported blow down test results on a full scale reactor vessel of PWR design. Modifications were made in code FLUSHEL to account for the blow down induced phase change and the influence of acoustic speed variation in the dispersive fluid medium was appropriately accounted. The estimation of the critical flow for blow down due to LOCA was carried out with systematic review of Burnell's model, Moody's homogeneous equilibrium model and Leung's generalised equilibrium model. The adequacy of Leung's generalised model was established for the prediction of sub-cooled and two-phase blow down induced critical discharge for HDR-PWR and TAPS-BWR problems respectively. After the validation of the code the coupled analysis of core shroud and downcomer annulus fluid for TAPS-BWR was undertaken for the postulated recirculation line break. It has been demonstrated that the acoustic Helmholtz modes of the downcomer annulus and the shroud shell multi-lobe modes of TAPS-BWR are well separated. The transient dynamic response of the core shroud shows that the acoustic load induced stresses are within service level D limits of Section III NB of ASME Boiler and Pressure vessel Code.

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In components which operate at high temperatures, changes in conditions at the beginning and end of operation or during operation result in transient temperature gradients. If these transients are repeated, the differential thermal expansion during each transient may result in thermally induced cyclic stresses. The extent of the resulting fatigue damage depends on the nature and frequency of the transient, the thermal gradient in the component, and the material properties. Components, which are subjected to thermally induced stresses generally, operate within the creep range so that damage due to both fatigue and creep has to be taken into account. In order to select the correct materials for these hostile operating environmental conditions, it is vitally important to understand the behaviour of mechanical properties such as creep-fatigue properties of these materials. This paper reports the results of standard creep-fatigue tests conducted using P122 (HCM12A or 12Cr-1.8W-13.5Cu) high temperature boiler material. P122 is one of the latest developed materials for high temperature environments, which has the potential to be successful in such hostile operating environments. The tests were conducted at temperatures ranging from 550℃ to 700℃ at 50℃ intervals with strain ranges of ±1.5 tp ±3.0% at 0.5% intervals and a strain rate of 4x (10)^<-3> s^<-1> with an application of 10-minute tensile hold time using a closed-loop hydraulic instron material testing machine with a servo hydraulic controller. The results confirm that P122 is comparable to conventional high temperature steels.

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Calandria tubes and liquid injection shutdown system (LISS) tubes in a pressurized heavy water reactor (PHWR) is known to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with LISS tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags to both tubes in the PHWR. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted.

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In case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Reactor, the interaction between fuel and liquid sodium creates a high pressure gas bubble in the core. The violent expansion of this bubble is similar to an explosion; it loads and deform the vessel and the internal structures. The experimental test MARA10 simulates a HCDA in a mock-up schematising simply the reactor block of a Fast Breeder Reactor : the external vessel, the roof as well as the main internal structures of the reactor are represented. The vessel is filled with water, topped with an air blanket. The test is fired using an explosive charge. This paper presents the numerical model used in the simulation of the MARA10 test with the fast dynamics code EUROPLEXUS. A specific HCDA constitutive law has been implemented in this code to simulate this kind of explosion. The external structures are represented by shells. Coupled fluid-structure computations are carried out.

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The MARS test-facility represents a small scale replica of a Fast Breeder Reactor. This mock-up contains all the significant internal structures of a Fast Breeder Reactor block : core, core support structures, diagrid and diagrid support, neutron shielding, baffles, internal vessel, core catcher, core cover plug, pumps, intermediary and emergency heat exchangers... The external geometry of the reactor is also represented precisely : roof slab including openings for the passing of the components, rotating plugs, main vessel of variable thickness...

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In the past, a complicated piping system under complex loading and boundary conditions may have to be simplified or decoupled to several simple models for analysis purpose. The results of those simplified analysis are generally approximate and very conservative. With the current advance of today's computer technology, a very detailed analysis using special features of a piping stress analysis program such as Bechtel Standard Program ME101 is feasible to provide more realistic results of such a complex pipe stress analysis problem. This paper presents the analysis methodology of such a complex pipe stress analysis problem - a buried jacket piping system connected between two buildings. The buried jacket piping system is a double containment piping consists of 3″ process piping and 6″ encasement piping. The 3″ processing piping is to transfer radioactive waste material from one building to another for processing. The 6″ encasement piping provides secondary containment in case of a leakage from the 3″ processing piping. One way weight supports, two way plate type guide supports and three way plate type guide supports with lugs are used between the process piping and encasement piping to support and control the 3″ process pipe movements. The 6″ encasement piping is anchored at the penetrations of the two building walls. The pipe mathematical model includes the process piping and encasement piping coupled together by those supports. Soil springs in vertical, lateral and axial direction of the pipe are developed to simulated the under ground condition. Based on this underground soil condition, the piping system is dynamically analyzed for seismic loads, and is statically analyzed for various temperature conditions, weight effects, seismic anchor movement and building settlement effects. The details of analysis and the results such as the interactions due to seismic motions and thermal expansions, soil interaction and the clearance are discussed and evaluated. The paper further compares the results of this analysis and the results of a simplified analysis based on the decoupled models. The final conclusions address the importance of the design parameters for this buried jacket piping system. A detail analysis and solution of this investigation will be discussed. With the results from this study, this paper further proposes a simple approach to decouple this type of analysis. Tables of summaries to facilitate various results and operating/event scenarios are included in this paper. The final conclusion addresses the importance of the critical design parameters for this buried jacket piping system.

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A water-shower cooling method was developed. In the new method, the surface of one side of a plate is heated by welding torch, and then it is immediately cooled by a water-shower. Temperature around the surface of the plate decreases rapidly when the water-shower begins. But the temperature inside of the plate remains high. During cooling, tensile stress occurs only on the surface of the plate. And tensile plastic strain remains near the surface of the plate. After moderate cooling, the tensile plastic strain generates compressive residual stress. To estimate the efficiency of this method, temperature histories during the welding process and residual stresses in type 316L stainless-steel plates were analyzed by finite element analysis. The analytical results were verified by comparing them with experimental measurements. The results can be summarized as follows. In the case that a plate was welded at heat-input rate of 30 kJ/cm with preheating at 215℃, and then cooled rapidly by water-shower, longitudinal and transverse residual stresses at the surface of the plate were decreased and became compressive; the highest value of this compressive stress was more than 200 MP3. In the case that a plate was welded in a rate of 30 kJ/cm without preheating and was cooled rapidly by water-shower, longitudinal and transverse residual stresses at the surface of the plate were reduced. On the other hand, in the case that a plate was welded and cooled in ambient-air, tensile residual stress occurred at the welded surface. The highest value of the tensile stress was more than 150 MP3. It can be concluded from these results that the new water-shower cooling method can effectively reduce tensile stress in welded joints.

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In pressurized light water reactor fuel assemblies, spacer grids support nuclear fuel rods both laterally and vertically. The fuel rods are supported by spacer grid springs and grid dimples that are located in the grid cell. Supporting the fuel rods mainly relies on the characteristics of the spacer grid springs, which are closely related to the shape to the springs. To improve the characteristics of the springs, a structural optimization method is employed for the shape design of spacer grid springs using commercial codes. Design requirements are defined and a design process is established for the spacer grid springs. The design process includes mathematical optimization as well as a practical design methods. Shapes of the grid springs are optimized to have better performance.

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The stunning rate of events after striking the push button AZ5 in April 26 1986 while that only rises. Boundless even for the super-power world states the complex of scientific, technological, organizational, economical and social problems became in 1991 unique property of Ukraine. It has added to the operational power reactors (now 13) at practical absence of an infrastructure of a closed fuel cycle. At the same time Ukrainian economics "always" will depend on nuclear power engineering. In it are very much positive aspect concerning high technological and scientifically based contents and future non-alternative of the nuclear power industry on a global scale. The errors in an estimation of separate links of such composite model are not killed mutually, but only add. Uncertainty in estimations of natural or public processes will cause to large uncertainty of general forecast. A laborious transaction of the rules production or the legitimated algorithms of the activity realization reach the foreseen controllability. On our view the following logical thesis of such concept should be comprehension that the rules of decommissioning of a nuclear-power plant should provide the controllability with matched activities of a few generation of performers. The impressive achievements of scientific-technological revolution of last decades are accompanied "non-regalement" from the point of view of life on a planet by disastrous effects. The nuclear technologies overtake in this sense with that feature, that the "half-life" periods of these consequences often much more large than the whole written history of mankind. The most distant consequences of the long-term processing with radioactive materials bound on our view with the human factor. If for 30-100 years beforehand it is possible to count destiny of radiological contamination or green meadows but to provide behavior of the people or society, as a whole is high-gravity even per annum forward. Objectivity of laws of history outflows from that each strives to its aims, but as a whole leaves so, as did not want anybody.

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Safety-related cables used at PWR plants in Japan must pass the environmental qualification test in accordance with IEEE Std.323 and 383. IEEE Std.383 describes the method of accelerated aging of cables that corresponds to the aging during in-service period. However, IEEE Std.775,etc. raise questions about the testing process in which (1) conditions for the accelerated thermal aging are estimated by extrapolating the activation energy, which is evaluated by the Arrhenius plot obtained in relatively high temperature range, (2) irradiation at high dose rate (<10kGy/h) is permitted, and (3) irradiation is carried out sequentially after the thermal aging test. Based on these background, we obtained the Arrhenius plot in relatively low temperature range to evaluate the activation energy using cables with insulation material such as ethylene propylene rubber, silicone rubber and polyvinyl chloride. And we experimented variety of simultaneous exposure tests to be exposed irradiation of 10-100Gy/h and thermal condition of 60-120degree, and presented at ICONE9 a part of the evaluation result of data obtained from these tests. After these study, the date of the activation energy at low temperature range was improved, because of getting the more advanced thermal aging data and the data of simultaneous radiation aging late. We also revised "the master curve" based on the method of "superposition of time dependent data" described in IEC 1244-2,and established the method to determine the accelerated aging conditions under various environmental conditions of existent plant according to this method. In addition, we conducted LOCA exposure test after the low acceleration simultaneous aging test simulating the natural aging at the existent plant, and as a result, we identified almost correctly the critical value of in-service period o cables made of each insulation material before the accident assumed in the design, from the result of cable soundness judgment based on the test specified in IEEE Std.383. Consequently, we could determine the control value of "elongation at break" at the nondestructive aging diagnosis of cable that was presented at ICONE7. At present, we have been confirming the long-term soundness of cable by applying our cable condition monitoring method (non-destructive aging diagnosis method, "the master curve") in several PWR plants.

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