[en] Primary coolant flow monitoring margin for one of the plants Westinghouse services was approaching 0% due to Steam Generator tube plugging and design changes. The approved methodology requires calibrating the primary coolant flow to the calorimetrically measured flow. The calorimetric flow measurement has a high uncertainty and has errors due to process variations. The path identified to gain margin was to develop a new method to reduce the uncertainty of the reference flow measurement. The new method identified for the uncertainty reduction was to determine the reference flow based on diverse, independent indications of flow. Utilizing the variance weighted averaging technique, the method produces a more accurate best estimate reference flow. Our team used three alternate indications available in the plant and the simulation of the flow loop as the diverse indications of flow. The benefit to the plant was a 60% reduction of the uncertainty. Introduction: Standard monitoring method: based on reactor coolant pump pressure differentials, periodically calibrated to the calorimetrically measured flow. - RCP involvement: Difficulty with RCP DP based flow measurement: it needs to be calibrated to a reference flow since the conditions of performance testing, if it is done at all, differ from the operating conditions. - Calorimetric involvement: Difficulties with calorimetric flow measurement: high uncertainty due to process noise and the process variation based biasing. - Plant Problem: Cycle independent calibration, applied at various plants, would not be adequate at the plant without RSG due to high uncertainty and low monitored flow. Increased resistance could put plant operability at risk. - Identification of the problem: Identified reference flow uncertainty as dominant margin reduction - Proposed Solution: Use Diverse Methods to reduce uncertainty and get more accurate best estimate flow, calibrate pump DP data to best estimate flow to allow plant to not have to change procedures except for removing the periodic calibration to calorimetric flow. - Implementation of Solution: Simulation of true phenomenon based on true physical modeling as method one - Implementation of Solution: Utilizing DP signals for Core and Steam Generator and calibrating based on simulation. - Implementation of Solution: Utilizing Design Basis method for flow measurement - Data combination: Instrument results must be poolable to utilize data and confirm each instruments validity - Data combination: Utilizing variance weighting of poolable instruments to reduce uncertainty and obtain best estimate flow - Final calibration: Calibrate Pump DP result to best estimate flow, possibly gain margin. - Final uncertainty calculation: combine Pump DP uncertainty with variance weighting uncertainty to gain margin. (authors)

[en] Aim: provenance determination of unknown nuclear material: - demonstrated for spent nuclear fuel; - information sought for unknown: fuel type, reactor type where fuel was irradiated, final burnup; Using an isotopic finger-printing method: - U, Pu or Pu isotopics or fission products; - simulations of fuel evolution during irradiation, using ORIGEN; - multivariate statistical tools. Fuel considered: simulated commercial spent fuel for a range of burnups: - PWR UO₂ 3.1% and 3.5% ²³⁵U, - PWR thermal MOX, - BWR UO₂ 3.2% ²³⁵U, - CANDU-N natural U, - CANDU-S UO₂ 3.2% ²³⁵U, - fast Reactor MOX; simulated commercial spent fuel for a range of burnups: - PWR UO₂ 3.1% and 3.5% ²³⁵U, - PWR thermal MOX, - BWR UO₂ 3.2% ²³⁵U, - CANDU-N natural U, - CANDU-S UO₂ 3.2% ²³⁵U, - fast Reactor MOX; 'unknown' spent fuel: - PWR 1: UO₂ 3.1% ²³⁵U (26 GWd/t), - PWR 2: UO₂ 3.1% ²³⁵U (32 GWd/t). Procedures: U, Pu or Pu isotopic compositions or fission products: - isotopic composition of unknown spent fuel, - simulated for commercial spent fuel from a range of nuclear power reactors → comparison of compositions through factor analysis → unknown has the provenance of the commercial spent fuel with which it exhibits the most similar composition. In conclusion: different reactor-fuel types well resolved; fuel and reactor type accurately predicted; burnup predicted to within 5% of declared; different reactor-fuel types. (authors)

[en] Since the very beginning of its brief history, nuclear energy was doomed to public attention - because of its first application. For 50 years of existence it failed to become one of traditional energy technologies, which the society would assess on the basis of its actual advantages (such as energy efficiency, resource availability and environmental acceptability). Nuclear weapons and crisis of confidence resulting from severe accidents have both formed the attitude to nuclear. This paper considers the basic antinuclear arguments, such as proliferation, waste and severe accidents. The current status of relations between nuclear energy and the public is still close (not only in Russia, but also in almost all European countries) to this state of politicization of nuclear and constant irrational fear radiation causes among people. Nevertheless, the positive trend in the attitude towards nuclear energy is obvious, both in Russia and in the world. In 2006, the long-expected 'new nuclear energy policy' (with returned budgetary financing of the new nuclear build) was announced in Russia at the highest governmental level. After that the worldwide recognition of the need to develop nuclear energy was only growing. The scale of global energy development is so large that all sources capable of making a contribution will find their demand. In the same time, public opinion in the world inseparably connects the issue of energy security with measures to combat climate changes. The '2 deg. C problem', if solvable at all, could be addressed only by simultaneous implementation of all possible emission reduction measures (including carbon-free energy technologies) on an unprecedented scale. Emission-free nuclear energy can actually become a system capable of sustainable and prompt development. This paper considers the issues, which could hamper nuclear development and negatively impact the public attitude towards nuclear. (authors)

[en] Over the years, electronic technology has played a very important role in the area of plant maintenance and operation. Enterprise Asset Management (EAM) and Enterprise Operation Management (EOM) have been developed and enhanced to improve plant personnel efficiency, productivity and safety. At the same time, CAD technology has been commonly used in plant maintenance and operation processes mostly in the area of design. In nuclear power plants, 2D CAD drawings (P and ID, electrical, mechanical, etc.) have long been used to represent not only the future plant configuration but also the 'as built' plant configuration. Over the last few years the opportunity to combine the two technologies (EAM/EOM and CAD) has emerge. The question regarding the benefits of displaying and/or updating EAM/EOM data such as clearances/Tagout/lockout, alignment checklist, radiation areas, scaffolding, chemical risks,..in a graphical format using 2D or 3D CAD drawings has often been asked. This is why EDF R and D and Ventyx have collaborated in developing a prototype that allows plant personnel to graphically display, create, and modify alignment procedures and clearances using an EAM (Asset Suite) and EOM (eSOMS) software. This paper describes some of the methodology as well as the tools used to develop such a prototype. (authors)

[en] Nuclear power plants require a number of electro-mechanical devices, for example, Control Rod Drive Mechanisms (CRDM's) to control the raising and lowering of control rods and Reactor Coolant Pumps (RCP's) to circulate the primary coolant. There are potential benefits in locating electro-mechanical components in areas of the plant with high ambient temperatures. One such benefit is the reduced need to make penetrations in pressure vessels leading to simplified plant design and improved inherent safety. The feature that limits the ambient temperature at which most electrical machines may operate is the material used for the electrical insulation of the machine windings. Conventional electrical machines generally use polymer-based insulation that limits the ambient temperature they can operate in to below 200 degrees Celsius. This means that when a conventional electrical machine is required to operate in a hot area it must be actively cooled necessitating additional systems. This paper presents data gathered during investigations undertaken by Rolls-Royce into the design of high temperature electrical machines. The research was undertaken at Rolls-Royce's University Technology Centre in Advanced Electrical Machines and Drives at Sheffield University. Rolls- Royce has also been investigating high temperature wire and encapsulants and latterly techniques to provide high temperature insulation to terminations. Rolls-Royce used the experience gained from these tests to produce a high temperature electrical linear actuator at sizes representative of those used in reactor systems. This machine was tested successfully at temperatures equivalent to those found inside the reactor vessel of a pressurised water reactor through a full series of operations that replicated in service duty. The paper will conclude by discussing the impact of the findings and potential electro-mechanical designs that may utilise such high temperature technologies. (authors)

[en] On June 3, Spanish Nuclear Safety regulatory body, CSN, declared unanimously 'Santa Maria de Garona' nuclear power plant, a BWR reactor, as entirely compliant with all the safety requirements to extend its operation for the 2009-2019 period. Nevertheless, on July 3, Spanish government allowed Garona Nuclear Power Plant for 4 years more, two years more than the design life, 40 years. From June 3 until July 3 there were several public demonstrations in support of Garona continuity. Every day a lot of debates, interviews and articles, regarding Garona issue took place in TV channels, radio stations and newspapers. This paper shows all the activities that the owner, workers, unions and non-profit associations carried out to show the good state of the facility, also includes a discussion regarding the data showed by the government to shut-down the facility eight years before lifespan. And finally, what to do in those cases is discussed in this paper. (authors)

[en] Since 1957, Tecnatom is giving engineering specialized services to the Electric Power Generation Plants. These activities also extend to other industrial areas like aerospace sector. TECNATOM it's made up of highly qualified specialists with an extensive experience and dedicated spirit and a pool of young people who interact with them to ensure the continuity of the company as a great team. As an important part of our activities, the 'Spares Engineering', complements the reliability and guarantee of a committed company. Nowadays, because of different reasons, there is a growing problematic in the collection of safety equipment and its spares in the NPPs. Tecnatom has set several actions in motion directed to improve the availability of these equipment and spares, by means of life extension of equipment installed, by means of the development of equipment and spares alternatives to the originals, or by means of services in relation with the spare parts. The actions developed until now have provided satisfactory results in technical and economical terms for the Plants. We have developments at present in course, that we hope they will increase this contribution to solve the problems of obsolescence in NPP's. (authors)

[en] The Probabilistic Safety Analysis (PSA) is a systematic and comprehensive methodology to evaluate risks associated with a complex engineered technological entity. Risk in a PSA is defined as a feasible detrimental outcome of an initiator. Those initiators can be 'classical' transient as the loss of main feedwater, loss of the secondary heat sink, etc.. or accident (LOCA - Loss Of Coolant Accident, SGTR - Steam Generator Tube Rupture, LOOP - Loss Of Offsite Power, etc..) In a PSA, risk is characterized by two quantities: the magnitude (severity) of the possible adverse consequence, the likelihood (probability) of occurrence of each consequence. Consequences are expressed numerically (for this purpose: the core damage) and their likelihoods of occurrence are expressed as probabilities or frequencies (i.e., the number of occurrences or the probability of occurrence per unit time). The total risk is the expected loss: the sum of the products of the consequences multiplied by their probabilities. This lead to the parameter CDF: The Core Damage Frequency, which is expressed by unit of time. The main advantage of this risk calculation is to have a global, integrated, overview of the plants and their systems. This allows to have an objective and quantitative point of view on the importance of the equipments, human action, or common cause failures that can challenge the plant's safety. A total PSA model is divided in three levels: Level one, which consider the core damage; Level two, which consider the robustness of the containment; Level three, which consider the impact of the radiological release on the public. For the purpose of the risk matrix, a level one PSA is needed. The scope of a PSA model is important to have a good characterization of the plant's risk. The matrix makes more sense if you have a full scope level one model, containing, furthermore the internal events, the fire and flooding, but also seismic event (if relevant). Asymmetries are also classical in the PSA models. Those asymmetries are characterized by the fact that, for instance, a LOCA is considered to occur only on a certain primary loop. So, to take that weakness into account, we have decided to group the different redundant equipments of a function as one. This leads to some conservatism. For instance, if you have 3 LPSI pumps, you group those and consider as one. Uncertainties are also a threat for the PSA models. You must be sure that the model you use do not have 'cliff edge' effects. A complete sensitivity study can be enough to demonstrate that the uncertainties are bounded and understood. By use of the PSA models, we can give an importance factor to each equipment. This qualitative approach of quantitative data is expressed by two quantities: The Fussel Vessely (FV) and the Risk Increased Factor (RIF). The combination of those 2 factors can lead to the ranking of the equipment in three categories: Highly Safety Significant (HSS), Medium Safety Significant (MSS), Low Safety Significant (LSS). This ranking can be done globally, for all sequences, or initiator by initiator. The latter one is used to construct our matrix. It permits a larger comprehension of the phenomena following an initiator. And this is the goal of our risk matrix. A frequency of occurrence is assigned to each initiators, depending on generic data, or plant specific data for the more frequent ones. As an initiator occurs, you have a conditional probability to go to core damage. We call it the severity. We will assign a set of colour, representing the importance of the severity and the frequency of occurrence. This permits us to have a clear overview of what could happen to the plant. One of the main advantages of this matrix is to link, with a risk approach, equipments that are not linked together within the technical specifications of the plant. As example, in the NPP of GDF-Suez, we have a second level for the safeguards systems. In the design phase, it was to provide a protection against external hazards such as plane crash. Those systems can also operate as backup of the first level safeguard systems. This second level is not linked with the first level in our technical specifications. The risk matrix put them at the same level. (authors)

[en] The 250 MWe (140 MWe since 1993) Phenix sodium cooled fast reactor was shut down on March 6, 2009. Before decommissioning, a final set of tests were performed during the May 2009 - January 2010 period covering core physics, fuel behaviour and thermal-hydraulics areas. Detailed analysis of the tests results is ongoing. It will be used for the extension of the validation of ERANOS and DARWIN codes for core physics, TRIO-U and CATHARE for Thermal-hydraulics and GERMINAL for fuel behaviour. In addition, the program included two tests related to the comprehension of the four negative reactivity transients (AURN in French acronym) experienced during the reactor operation in '89 and '90 and not yet fully explained. This was also a great opportunity to involve young engineers in the different processes like the design of the tests, their carrying out, and the analysis of the results. The standard instrumentation of the reactor was completed by specifically designed devises. (authors)

[en] One of the most challenging areas in the operation of nuclear power plants (NPP) is related to the management of plant design modifications. Plant modifications can be made to improve reliability, facilitate operation, improve safety or get better results. In any of these situations, plant modifications imply many different activities that have to be done in a coordinated manner. NUREG-0711 (Human Factors Engineering Program Review Model) shows a global approach to manage most of these activities. Although this approach is mainly focused on the design and construction of new plants, it can also be applied to plant modification management. Successful global management will require performing every activity in a specific order, taking advantage of the output coming from some tasks as input for others and finalizing every task when necessary. This will provide the best results in terms of quality, time required for implementation, safe and reliable operation and maintenance, and cost. Tecnatom is involved in most of the activities related to the operational areas and has applied a global approach to get advantages in terms of quality and cost, which is outlined in this paper. As an example of this approach, the Vandellos NPP experience is shown in this presentation. Vandellos NPP carried out an important design modification that consists of replacing an old essential service water system with a new one. This was a three-year project that implied the construction of new reservoirs, new buildings, the implementation of new equipment, and new panels in the main control room. This paper shows the way in which all of these activities were performed. (authors)