On November 9,2001,after the pipe rupture incident of the Residual Heat Removal system (RHR), plant personnel found the leak from the bottom of the Reactor Pressure Vessel (RPV). Afterwards, with underwater visual inspection, plant personnel found an axial cracking on one of the stub tube's weldments. In order to join the Nickel base material (Alloy 600) stub tube, a similar weld material (Alloy 182) is deposited to the low-alloy metal (LAS) vessel. From the examination of a boat sample it was found that the cracking in the Alloy 182 weld metal was due to interdendric (intergranular) stress corrosion cracking, which had progressed into the Alloy 600. Residual and applied stress during an in-service analysis explained that the location could have high tensile stress (330MPa and over). In order to repair this cracking, a replacement method was applied. The stub tube and weld joint including the crack area was completely removed, and a new stub tube consisting of high corrosion resisted material was installed and welded. Remote automatic equipment was applied during the replacement process because of the high radiation environment. After inspecting the rest of the 88 stub tube's joints, there were no indications of any further problems. For higher reliability, the application of laser-peening technique is being examined.

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Past and ongoing STPNOC Owner investments in plant information technology (such as database query applications and other client workstation tools) have made it possible for plant staff to utilize information contained in the CWMS to quickly link equipment failure modes to related preventative maintenance (PM) activities. The FW systems STPNOC maintain comprise 3,769 TPNSs which in turn represent about 16,300 failure modes. Risk-informed asset management (RIAM) of FW PM activities requires the failure modes to be modeled in the STPNOC BOP (availability) model. In this paper we present initial development of a STPNOC process for supporting PM optimization, applying the STPNOC BOP and STPNOC RIAM alpha-level software applications. In this work presented, PM activities are evaluated (for profitability and nuclear safety) in the STPNOC RIAM alpha-level application. PMs are optimized for profitability and nuclear safety results are used to ensure STPNOC maintains or improves upon high levels of nuclear safety. In the BOP availability application the level of detail of the FW system is enhanced to support plant decision-making at the component failure mode and human error mode level of indenture for the FW system. The enhanced FW model and modeling techniques are presented. Results of case studies in FW system PM optimization using STPNOC data are presented.

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Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron irradiation, stress and corrosion by high temperature water. It is, therefore, essential to perform in-pile SCC tests, which are material tests under the conditions simulating those of actual LWR operation, in order to clarify the precise mechanism of the phenomenon, though mainly out-of-pile SCC tests for irradiated materials have been carried out in this research field. There are, however, many difficulties to perform in-pile SCC tests. Performing in-pile SCC tests, essential key techniques must be developed. Hence as a part of development of the key techniques for in-pile SCC tests, we have embarked on development of the test technique which enables us to obtain the information concerning the effect of such parameters as applied stress level, water chemistry, irradiation conditions, etc. on the crack initiation behavior. Although it is difficult to detect the crack initiation in in-pile SCC tests, the crack initiation can be evaluated by the detection of specimen rupture if the cross section area of the specimen is small enough. Therefore, we adopted the uniaxial constant loading (UCL) test with small tensile specimens. This paper will describe the current status of the development of several techniques for in-pile SCC initiation tests in JMTR and the results of the performance tests of the designed testing unit using the out-of-pile loop facility.

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In spite of the control of the water chemistry of SG secondary feed-water in PWR-SG, SG TSP BEC holes, which are the flow path of secondary water, are often clogged. In the past, the trending of BEC hole blockage rate has conducted by evaluating ECT original signals and visual inspections. However, the ECT original signals of deposits are diversified, it becomes difficult to analyze them with the existing evaluation method using the ECT original signals. In this regard, we have developed the secondary side visual inspection system, which enables the high-accuracy evaluation of BEC hole blockage rate, and new ECT signal evaluation method.

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As previously reported at ICONE 6 in New Orleans (1996), the use of various innovative maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems. Further cost refinement of the station maintenance strategy is being evaluated via the applicability of statistical analysis of historical failure data. Since the statistical evaluation was initiated in 1999 significant progress has been made in demonstrating the viability of stochastic methods in Candu maintenance. Some of the relevant results were presented at ICONE 10 in Washington DC (2002). Success with the graphical displays and the relatively easy to implement stochastic computer programs was sufficient to move the program along to the next significant phase. This next phase consists of investigating the validity of using subjective elicitation techniques to obtain component lifetime distributions. This technique provides access to the elusive failure statistics, the lack of which is often referred to in the literature as the principle impediment preventing the use of stochastic methods in large industry. At the same time the technique allows very valuable information to be captured from the fast retiring "baby boom generation". Initial indications have been quite positive.

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The subject of this work is to show a new method based on well verified CFD code for the particular purpose of moisture reduction in BWR's for low leakage loading. A complete 3D-steady state is generated for a vessel region starting with the pump entrance, going through the lower head, core upper mixing chamber and the stand pipes. The flow parameters at the stand pipes outlets are then used as a input for a empirical model for analysis of the cyclones and dryers performance. The empirical model is based on appropriate experimental data base. The parameters at the exit of the dryers are then cross section averaged in order to obtain the final moisture at the vessel exit. Comparing two specific BWR fuel cycle states for low leakage loading we find out that a reduction of about 40% of the peripheral mass flow leads to : (a) Effective reduction the moisture content in the fresh steam ; (b) The averaged void over the core is slightly increased over the first half of the fuel cycle and remains almost unchanged over the end of the fuel cycle which is positive for the fuel utilization ; (c) The reduced mass flow amount over the periphery is pushed through the remaining central part of the core which increases the safety margins for the most loaded bundle. The method is already in use for performance optimization of existing plants and for design of a new plants.

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The paper presents general approach, used methods and form of documentation of the results have been applied within the shutdown and low power PSA (SPSA) study for Mochovce NPP, Unit 1,Slovakia. The SPSA project was realized by VUJE Trnava Inc., Slovakia in 2001-2002 years. Level 1 SPSA study for Mochovce NPP Unit 1 covers internal events as well as internal (fires, floods and heavy load drop) and external (air craft crash, extreme meteorological conditions, seismic event and influence of surrounding industry) hazards. Mochovce NPP consists of two operating units equipped by VVER 440/V213 reactors safety upgraded before construction finishing and operation start. 87 safety measures based on VVER 440 operational experience and international mission insights were implemented to enhance its' operational and nuclear safety. The SPSA relates to full power PSA (FPSA) as a continuation of the effort to create harmonized level 1 PSA model for all operational modes of the plant with the goal to use it for further purposes as follows : ・Real Time Risk Monitor, ・Maintenance Optimalization, ・Technical Specifications Optimalization, ・Living PSA, etc.

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In order to perform research activity for aging countermeasure of nuclear power plant effectively, Plant Life Engineering Center (PLEC) was established in Japan Power Engineering and Inspection Corporation (JAPEIC) in April 2000 sponsored by Ministry of Economy, Trade and Industry (METI, called as MITI at that time). Results of technical survey for research and development for aging phenomena have been summarized in a table (Research Map) categorizing them into "inspection and monitoring", "evaluation method for aging" and "preventive maintenances and refurbishment". Necessary research themes have been extracted from the Research Map consulting to experts of the specified research area. Medium and long-term research perspective (Research Perspective) has been established which contains prioritized research themes and outlined specification of each theme. Several new research themes proposed by various organizations and selected by PLEC as effective for the regulation activities of METI are identified and proposed to be funded by METI every year. There are about ten on-going research programs funded by METI. Their progress and performance are evaluated annually to improve their efficiency including their alteration, abolition and integration. This cycle of research is going to be attained successfully. Technology Advisory committee composed of members from various field of nuclear power including prefectural and municipal governments supervises the PLEC activity to concentrate national wide potentials and to secure transparency, openness and neutrality. This paper also provides an outline of the aging related research projects currently conducted by JAPEIC under the auspices of METI.

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In Nuclear Power Plants, the reliability of all the safety systems is very critical from the safety viewpoint and it is very essential that the required reliability requirements be met while satisfying the design constraints. From practical experience, it is found that the reliability of complex systems such as Safety Rod Drive Mechanism is of the order of (10)^<-4> with an uncertainty factor of 10. To demonstrate the reliability of such systems is prohibitive in terms of cost and time as the number of tests needed is very large. The purpose of this paper is to develop a Bayesian reliability demonstrating testing procedure for exponentially distributed failure times with gamma prior distribution on the failure rate which can be easily and effectivel used to demonstrate component/subsystem/system reliability conformance to stated requirements. The important questions addressed in this paper are : With zero failures, how long one should perform the tests and how many components are required to conclude with a given degree of confidence, that the component under test, meets the reliability requirement. The procedure is explained with an example. This procedure can also be extended to demonstrate with more number of failures. The approach presented is applicable for deriving test plans for demonstrating component failure rates of nuclear power plants, as the failure data for similar components are becoming available in existing plants elsewhere. The advantages of this procedure are the criterion upon which the procedure is based is simple and pertinent, the fitting of the prior distribution is an integral part of the procedure and is based on the use of information regarding two percentiles of this distribution and finally, the procedure is straightforward and easy to apply in practice.

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A complete in-service inspection (ISI) program for nuclear research reactors requires several steps that have to be performed in sequence. These steps were described in many logic flow charts (LFC's). These logic flow charts might be separately used as a guide for future work in case of developing expert system software for research reactors. These charts included a general LFC describing the steps required to complete the ISI program. Details of this general LFC describing the steps required to complete the ISI program. Details of this general LFC are given in many modules. The next step of the LFC is to define the class of the component to be inspected. For each component class there are related modules that should be followed (examination requirements modules, examination methods modules, and flaw acceptance/rejection modules). If the component is accepted according to the requirements found in American Society of Mechanical Engineers (ASME) CODE section XI, the record and report requirement modules will be used to complete the ISI program. In the present work the inspection requirements for welded parts in the Second Egyptian Testing Research Reactor (ETRR-2) are considered. ETRR-2 is a pool- type reactor with an open water surface and variable core arrangement. Its core power is 22 MW_<th>, cooled and moderated by light water and with beryllium reflectors. It contains platetype fuel elements (MTR type, 19.7% enriched uranium) with aluminum clad. This reactor consists of 57 systems and around 200 subsystems, these systems contain many mechanical components such as; tanks, piper, valves, pumps, heat exchangers, cooling tower, air compressors and suppors.

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The complete understanding of intergranular stress corrosion cracking of Inconel 600 in high-temperature water is, despite more than 30 years of research, still somewhat in the future. Especially the early phase of the development of cracks seems to be beyond the present state-of-the-art explanations. An effort was therefore made by the authors to construct a computational model of the crack growth kinetics at the grain-size scale. The main idea is to divide continuum (e. g., polycrystalline aggregate) into a set of sub-continua (grains). Random grain structure is modelled using Voronoi-Dirichlet tessellation. Each grain is assumed to be a monocrystal with random orientation of crystal lattice. Elastic behaviour of grains is assumed to be anisotropic. Crystal plasticity is used to describe (small to moderate) plastic deformation of monocrystal grains, caused mainly by the strains along the "incompatible" grain boundaries and at triple points. Finite element method is used to obtain numerical solutions of strain and stress fields. The analysis is currently limited to two-dimensional models. The paper focuses on the dependence of crack tip loading (J-integral) on the random orientation of neighboring grains. The limited number of calculations indicate that the incompatibility strains, which develop along the boundaries of randomly oriented grains, influence the local stress fields (J-integrals) at crack tips significantly.

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Temper bead welding technique is one of the most important repair welding methods for large structures for which post weld heat treatment (PWHT) is difficult to perform. In this study the optimum condition of weld thermal cycles to carry out temper bead weld repair for pressure vessel steel SQV2A are investigated taking interest in the improvement of the characteristics of coarse grained heat affected zone (CGHAZ) re-heated to the temperature between 670℃ (Ac1 temperature) and 837℃ (Ac3 temperature). Thermal/mechanical simulator Gleeble 1500 is employed to give specimens repeated thermal cycles. Improvability of the Charpy absorbed energy and the hardness for the thermal cycles is discussed. The absorbed energy of the CGHAZ improved to about 70% of that of as-received steel and the hardness of the CGHAZ decreased to lower than 350HV1 when the peak temperature of the succeeding thermal cycle reached to the temperature from 600℃ to Ac1. The absorbed energy of the CGHAZ degraded to about 25% of that of as-received steel and the hardness of the CGHAZ was still higher than 350HV1 when the peak temperature of the succeeding thermal cycle reached to the temperature from Ac1 to Ac3. In order to improve the absorbed energy to about 70% of that of as-received steel but to decrease the hardness to lower than 350HV1,additional one more thermal cycle ; that is, triple thermal cycle is required. The absorbed energy of the CGHAZ improved to more than 70% of that of as-received steel but the hardness of the CGHAZ was still higher than 350HV1 when the peak temperature of the succeeding thermal cycle was higher than Ac3. The absorbed energy improved to almost the same as that of as-received steel but the hardness was still higher than 350HV1 even after the following thermal cycle, the peak temperature of which is less than Ac1. To reduce the hardness to lower than 350HV1,fourth thermal cycles are required. The recommended temper bead thermal cycles were suggested that in the second thermal cycle the peak temperature Tp2 should be selected as lower than Ac1 but near Ac1. In this case triple thermal cycles temper bead process are enough to improve the characteristics of CGHAZ. When the other peak temperatures (that is, higher than Ac1) in the succeeding thermal cycle are applied to CGHAZ, quadruple or more thermal cycle temper bead process should be applied.

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Intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) internals materials is a well understood phenomenon. The degree of stress corrosion cracking (SCC) varies from plant to plant depending on weld heat affected zone residual stresses and the operating water chemistry history of each plant. Since late 80's plants have paid a significant attention to reactor water chemistry control and have seen the beneficial effect of reduced incidents of SCC as a result of good water chemistry control. In addition to water chemistry control by lowering the reactor water impurity ionic chloride, sulfate, copper and maintaining feedwater oxygen and iron at the desired levels, plants have also attempted to lower the electrochemical corrosion potential (ECP) of internal surfaces by practicing hydrogen water chemistry (HWC) to reduce the propensity of materials to undergo SCC. Majority of the BWRs worldwide have employed HWC to mitigate IGSCC of reactor internals. Since recently, many BWRs have also employed a new technology called NobleChem^<TM> and low HWC which improves the effectiveness of hydrogen to lower the ECP to HWC specification value of -230 mV(SHE) or lower. This paper highlights some of the shortcomings of the SCC mitigation practices that need further attention to derive the full benefit of HWC and other mitigative actions against IGSCC. The topics that will be discussed in this paper will include, 1) operating BWRs at low feedwater hydrogen levels, 2) effect of hydrogen interruptions and the importance of high hydrogen availability, 3) dissolved oxygen monitoring as a means of controlling SCC, 4) ECP measurement at the end of long sample lines and its impact on SCC mitigation decisions, and 5) using main steam line dose rate as a criterion to determine SCC mitigation effectiveness.

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The process used by French utility, concerning the Reactor Pressure Vessel assessment, applied on 58 PWR NPPs 3 loops and 4 loops Reactor, involves the verification of the integrity of the component by finite element mechanical studies, in all conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and considering all of parameters. This approach, is based on mechanical safety studies, to demonstrate the absence of risk of failure by brittle fracture. For these mechanical studies two major input data are necessary : 1-the fluence distribution and the fluence values and RTNDT during the lifetime in operation for each NPPs, 2-the thermal-hydraulic and mechanical evaluation and temperature distribution values in the downcomer. The main results must show significant margins. The major tasks and expertises engaged by EDF were : -more precise assessment of the fluence and neutron flux calculations, -better knowledge of the vessel material properties, including the effect of radiation, -the NDE inspection program based on the inspection of the vessel wall, with a special NDE tool to inspect the area in subcladding zone

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