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[en] A new tendency in nuclear reactor conceptual design is to include safety criteria through accident analysis to address the selection of the engineering solutions and the value of the main design parameters. In this work, the concept of design map is used to correlate reactor safety performance with the design parameters. The effect of different design parameters on characteristic safety variables, referred to as 'observable variables', extracted from reactor evolution during accidents, is analyzed, and the concept of 'safety design maps' is introduced. The sensitivity of these 'observables variables' regarding changes in design parameters is visualized. Several safety design maps are built from the performance of an integral type reactor during loss of heat sink (LOHS) and main steam line break sequences without SCRAM to show the technique potentiality. Maximum reactor pressure vessel (RPV) pressure and minimum departure from nucleate boiling ratio are chosen as 'observable variables' and their sensitivity to geometry-related parameters and reactivity coefficient is studied. Multiple-parameter single design maps and combined design maps for both accidental sequences are built as examples. The results show the usefulness of this technique to balance and optimize reactor design through an early engineering step. A computer code (HUARPE) has been developed in order to simulate these transients. The cooling circuit, the steam dome, the pressure vessel structures and core models are considered
[en] Highlights: • Conjugating: • ALARA. • BEPU. • Extended safety margin. • Independent assessment. - Abstract: The paper aims at fixing bases for possible strengthening of current Nuclear Reactor Safety (NRS) and safety analysis: this is done by combining the logical frameworks connected with the terms As-Low-As-Reasonably-Achievable (ALARA), Best-Estimate-Plus-Uncertainty (BEPU), Extended-Safety-Margin (E-SM) and Independent-Assessment (IA). ALARA is an early principle in Nuclear Reactor Safety: designers and operators must do their best to minimize doses to the humans. BEPU is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margins constitute an established concept in NRS: a few dozen SM values must be calculated in current safety analyses and demonstrated to be acceptable. The concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit, thus forming the E-SM. Here ‘the environment’ includes the personnel in charge of activities connected with the NPP. The E-SM implies the formulation of some ten-thousands safety margins values, which shall correspond to a similar number of monitored variables. IA is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design. The implementation of the idea in the paper brings to an additional safety barrier for existing and future nuclear reactors which may reduce the probability of core melt, presumably at an affordable cost for the industry.
[en] Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited from PSA is indeed the systematic way which PSA follows. This is more important especially for the new reactors.
[en] In 1989 Framatome and Siemens, the two most experienced European nuclear power plant suppliers, decided to join the efforts for the development of a new reactor type for the next generation in their equally owned subsidiary nuclear power international (NPI). In 1992 Electricite de France and the major German utilities operating nuclear power plants merged their own development programs with that of nuclear power international and initiated the European pressurized water reactor (EPR) project. In order to reach the two major targets of the project, the licensability in both countries, France and Germany, and the competitiveness of nuclear energy with other alternative energy sources, the design basis which had differently developed in the two countries needed to be harmonized. In parallel, the licensing authorities of both countries extended their existing cooperation in the field of a safety survey of existing nuclear power plants to the definition of safety criteria for the next generation of nuclear power plants. Through this cooperation the licensability of EPR in France and Germany will be assured. Continuously performed cost analysis show in addition that also the second target of the project, the competitiveness with alternative primary energy sources, can be achieved. Thanks to the fruitful cooperation between all parties involved, satisfactory results have been achieved not by a simple superposition of existing design features but through a careful evaluation and combination of the best available alternatives. At the end of 1997 the basic design results were compiled in a final report. Subsequently an optimization phase was launched that further improves the competitiveness of the power generation costs. (orig.)
[en] Highlights: • SALTO peer review service is designed for reviewing of ageing management and NPPs’ preparedness LTO. • It has been established as an effective tool to review the compliance with IAEA safety standards. • The important issues for safe LTO are being identified by SALTO missions. • Analysis of those issues is provided in the paper. • This peer review service is strongly recommended for NPPs prior to entering LTO period. - Abstract: This paper presents main IAEA activities for safe long term operation (LTO) which includes establishment of IAEA Safety Standards and other LTO related documents, fostering information exchange and establishing databases and provision of SALTO (Safety Aspects of Long Term Operation) peer review service. This paper provides insights into IAEA SALTO peer review service objectives, scope and methodology. The SALTO peer review service was designed to assist nuclear power plant (NPP) operators in adopting a proper approach to LTO of their plants and in implementing complete and appropriate activities to ensure that plant safety will be maintained during the LTO period. The SALTO peer review service can also support regulators in establishing or improving regulatory and licensing strategies for LTO of NPPs. Issues derived from 19 SALTO missions and 2 LTO modules of OSART (Operational Safety Review Team) missions conducted during the period of 2005 to March 2014 are also analyzed in this paper
[en] In this study, we analyzed human and organizational deficiencies in the trip events of Korean nuclear power plants. K-HPES items were used in human error analysis, and the organizational factors by Jacobs and Haber were used for organizational deficiency analysis. We proposed the use of CCDP as a risk measure to consider risk information in prioritizing K-HPES items and organizational factors. Until now, the risk significance of events has not been considered in human error and organizational deficiency analysis. Considering the risk significance of events in the process of analysis is necessary for effective enhancement of nuclear power plant safety by focusing on causes of human error and organizational deficiencies that are associated with significant risk
[en] A problem of design-based cycles counting for fatigue assessments of structures without initial cracks is considered in terms of requirements of strength analysis regulations for nuclear engineering components according to crack nucleation criterion. It is shown that a higher conservatism of these requirements for design stage analysis does not permit to apply the widely distributed rain-flow method. An alternative algorithm of cycles counting is presented. It satisfies the regulations requirements and is easy to realize in design practice. (orig.)
[en] An approach to evaluate the influence of phenomenological uncertainty in reliability assessments is proposed. A simple scheme is developed to update initial estimates of the proneness to failure as the time during which the system operates increases. Conclusions are drawn with respect to the significance of very high reliability estimates. (orig.)
[en] In less than 10 years, the first commercial pressurized water reactor (PWR) plant in Korea will reach its official design life. As part of safety activities, developed countries have already implemented periodic safety review (PSR) or equivalent programs to check and improve the safety of operating nuclear power plants (NPP) during their plant life. At the end of 1999, it was decided by the Korean Atomic Energy Safety Committee to adopt the PSR program and to apply it to Korean operating NPP. Since Kori Unit 1 started the review for the first tentative application of PSR as a model case in May 2000, it is now progressing well. Management of aging is one of the major factors to be considered in PSR and life extension of a nuclear plant. This paper is intended to introduce the regulatory aspect and strategy of Korean PSR. The background and scope of basic PSR guidelines are described, and a summary of technical criteria for aging management, which shows a regulatory direction for PSR, is also presented
[en] The loading data for the design and fatigue analyses of the primary system components are derived from the plant operating transients as specified in the Operational Manual, the accident analyses and from the external event requirements. The methods used, applicable codes and standards, and QA procedures produce a design concept that contains significant safety margins. Measurements conducted in pre-Konvoi plants and during commissioning of the Konvoi plants deliver additional verifications of the analyses. (orig.)