Article

Relaxation of the requirements on loop height and heat transfer area of a passive heat removal system in integral SMR using a self-powered booster

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Abstract

In this study, compactness improvement capacity of a self-powered efficiency booster is analyzed for natural circulation systems used in integral small modular reactors (SMR). Based on thermoelectric principle, the booster converts heat energy into electricity directly, which is used to produce additional driving force for natural circulation flow and heat transfer. Thus, flow and heat transfer are enhanced and such systems do not rely on gravity solely. Consequently, requirement on loop height and heat transfer area can be relaxed. Numerical simulations and experimental tests have been carried out. The results indicate that, using the booster, loop height is no longer a determining factor for natural circulation systems, and lower height can be used for the same performance. Furthermore, if the desirable heat transfer rate was kept, the heat exchanger area can be reduced by a factor of 25%–78%. With these results, compactness improvement capacity of the booster is proved.

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The basic concept of an innovative advanced marine reactor with a passive safety system, MRX (Marine Reactor X) has been established for the primary application to ship propulsion. The design goals of the reactor system, to be lightweight and compact, and to be enhanced in safety and reliability, are achieved with adoption of new technologies. The MRX is of the integral-type PWR with 100 MW of thermal output. Adoption of a water-filled containment makes the MRX extremely lightweight and compact. The total weight and volume of MRX is about 1600 tons and 1210 m3, which is half that of the first Japanese nuclear ship, ‘Mutsu’, reactor, although the reactor power of MRX is three times greater than that of the ‘Mutsu’ reactor. Numbers of active components in the reactor system are greatly reduced, compared with loop type PWRs, by adopting an integral type reactor and the passive system. Safety was evaluated by both experiments and analyses. Core damage occurrence frequency estimated by probability safety analysis (PSA) is of two orders smaller than those of existing PWRs. Feasibility study on economics is conducted by comparing the total operation costs of a nuclear container ship installing the MRXs with a diesel engine ship. The nuclear ship has the advantage for greater speed and larger amounts of cargo carried.
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The transients and setpoint simulation/system-integrated modular reactor (TASS/SMR) code has been used to identify the safety margin of a 65-MWt advanced integral reactor and to evaluate its design performance. Although, the code has been verified by using simplified and analytical problems as well as a reliable system code, its validation has not been fully established. This paper deals with a validation of the TASS/SMR code by using two kinds of separate effect tests related to heat transfer at a helically coiled steam generator. The heat transfer experiments were performed by using a full-scale prototype of the steam generator cassette of the advanced integral reactor and a scaled-down steam generator cassette. Analytical results show that the TASS/SMR code predicts the thermal hydraulic parameters, including the system pressure and fluid temperature at the primary and secondary sides of the steam generator cassette, and the heat transfer rate through the steam generator cassette well. The validation results in this study show that the TASS/SMR code is applicable for heat transfer calculations related to the helically coiled steam generator of the advanced integral reactor.
Article
The development of the HTGR type reactor is reviewed. Design and operating characteristics are outlined. Safety aspects and fuel cycle considerations are described.
Article
Salient features of the International Reactor Innovative and Secure (IRIS) are presented here. IRIS, an integral, modular, medium size (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. Described here are the features of the integral design which includes steam generators, pumps and pressurizer inside the vessel, together with the core, control rods, and neutron reflector/shield. A brief summary is provided of the IRIS approach to extended maintenance over a 48-month schedule. The unique IRIS safety-by-design approach is discussed, which, by eliminating accidents, at the design stage, or decreasing their consequences/probabilities when outright elimination is not possible, provides a very powerful first level of defense in depth. The safety-by-design allows a significant reduction and simplification of the passive safety systems, which are presented here, together with an assessment of the IRIS response to transients and postulated accidents.
Article
Although a major focus of nuclear power reactor development efforts in industrialised countries is on large evolutionary units and design modifications that take maximum advantage of successful proven features and components, consideration is also given to utilisation of passive safety systems and inherent safety features. The various advanced reactor designs: evolutionary, large water-cooled reactor designs; evolutionary, medium size water-cooled reactor designs; concepts requiring substantial development; gas-cooled reactor concepts; and liquid metal-cooled fast reactors, incorporate a wide variety of passive safety features for initiation of safety systems, for residual heat removal and for containment heat removal. Organizations have established testing programmes to confirm their performance. A number of IAEA activities have lead to the conclusion that the use of passive safety features can be a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions. However, care should be taken to evaluate possible new failure mechanisms, and both passive and active systems should be assessed from the standpoint of reliability and economics. Key technical issues include: the quantification of reliability over a wide range of conditions: economics, speed of action, plant ageing, demonstration of technical feasibility, in-service testing, ease of maintenance and minimization of personnel radiation exposure. Many member states conduct substantial work on the design, modelling, development and reliability assessments of passive safety systems. Continued information exchange can benefit the involved member states, and the IAEA is providing a forum for review of programmes, project directions, and the results achieved.
Article
A methodology has been developed to evaluate the reliability of passive systems characterised by a moving fluid and whose operation is based on thermal–hydraulic (T-H) principles. The methodology includes:•identification and quantification of the sources of uncertainties and determination of the important variables;•propagation of the uncertainties through T-H models and assessment of T-H passive system unreliability;•introduction of passive system unreliability in the accident sequence analysis.Each step of the methodology is described and commented and a diagram of the methodology is presented. An example of passive system is presented with the aim to illustrate the possibilities of the methodology. This example is the Residual Passive heat Removal system on the Primary circuit (RP2), an innovating system supposed to be implemented on a 900 MWe Pressurized Water Reactor.
Article
The Westinghouse AP1000 Program is aimed at making available a nuclear power plant that is economical in the US deregulated electrical power industry in the near-term. The AP1000 is a two-loop 1000 MWe pressurizer water reactor (PWR). It is an uprated version of the AP600. Passive safety systems are used to provide significant and measurable improvements in plant simplification, safety, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 35 years of operating PWR experience. The AP1000 received Final Design Approval from the United States Nuclear Regulatory Commission in September 2004; the AP1000 has also received Design Certification by the USNRC in December 2005. The AP1000 and its predecessor AP600 are the only nuclear reactor designs using passive safety technology licensed anywhere in the world. The safety performance of AP1000 has been verified by extensive testing, safety analysis and probabilistic safety assessment. AP1000 safety margins are large and the potential for accident scenarios that could jeopardize public safety is extremely low.Simplicity is a key technical concept behind the AP1000. It makes the AP1000 easier and less expensive to build, operate, and maintain. Simplification also provides a hedge against regulatory driven operations and maintenance costs by eliminating equipment subject to regulation. The AP1000's greatly simplified design complies with NRC regulatory and safety requirements and the EPRI advanced light water reactor (ALWR) utility requirements document.Plans are being developed for implementation of the AP1000 plant. Key factors in this planning are the economics of AP1000 in the de-regulated US electricity market, and the associated business model for licensing, constructing and operating these new plants.
Article
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. Includes bibliographical references. Passive cooling systems sometimes use natural circulation, and they are not dependent on emergency AC power or offsite power, which can make designs simpler through the reduction of emergency power supplying infrastructure. The passive system approach can lead to substantial simplification of the system as well as overall economic benefits, and passive systems are believed to be less vulnerable to accidents by component failures and human errors compared to active systems. The viewpoint that passive system design is more reliable and more economical than active system design has become generally accepted. However, passive systems have characteristics of a high level of uncertainty and low driving force for purposes of heat removal phenomena. These characteristics of passive systems can result in increasing system unreliability and may raise potential remedial costs during a system's lifetime. This study presents a comprehensive comparison of reliability and cost taking into account uncertainties and introduces the concept of flexibility using the example of active and passive residual heat removal systems in a PWR. The results show that the active system can have, for this particular application, greater reliability than the passive system. Because the passive system is economically optimized, its heat removal capacity is much smaller than that of the active system. Thus, functional failure probability of the passive system has a greater impact on overall system reliability than the active system. Moreover, considering the implications of flexibility upon remedial costs, the active system may more economical than the passive system because the active system has flexible design features for purposes of increasing heat removal capacity. by Jiyong Oh. Ph.D.
Decay heat generation in fission reactors
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Passive safety systems and natural circulation in water cooled nuclear power plants
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Advances in small modular reactor technology developments, A Supplement to: IAEA Advanced Reactors Information System (ARIS)
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An introduction to nuclear waste immobilisation
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Decay heat generation in fission reactors Chap. 8. Probabilistic, Possibilistic, and Deterministic Safety Analysis, Nuclear Applications
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