Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany.In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented.Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact.A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.     

Boiling water reactor development and its mechanical-structural requirements and problems

A brief review of the state of development of boiling water reactors is given, pointing out differences in the boiling water reactor technology of GE and AEG. Design problems of large components of the boiling water reactor primary circuit are outlined and the safety of steel reactor pressure vessels is discussed giving special consideration to problems of stress analysis and fracture mechanics. The characteristics of the design of concrete structures considering extreme accident conditions are described. Finally, mechanical problems of the BWR fuel element cannings are briefly discussed.     

Boiling water reactor fuel heatup during a loss of coolant while refueling

The design of nuclear power plants includes provisions for heat removal from the reactor core in the event there is a loss of reactor coolant while shut down. Boiloff from decay heat can lead to inventory reduction and fuel heatup if no coolant makeup is available. Certain decay heat removal system failures in boiling water reactors can drain the upper vessel and downcomer. This leaves the water inside the core shroud at the same level as the top of the jet pumps. This becomes the starting point from which further inventory reduction is possible through boiloff. This study investigated the core thermal response following such a scenario. A simple model of the core was used for analysis of this sequence. The goal of the analysis was to determine the time at which the water in the core would boil down and fuel heat up to a specified temperature (1256 K). It is this interval during which the operator can take action that will mitigate the transient.     

Theoretical study of flashing and water hammer in a supercritical water cycle during pressure drop

During a loss of coolant accident (LOCA) the pressure of the coolant can drop significantly in the vicinity of the leak. It will be shown that unlike in pressurized water reactors (PWRs) where this pressure drop can cause only sudden vaporization - also called flashing - in supercritical water cooled reactors (SCWRs) it can cause sudden condensation (condensation-induced water hammer), too. The reason is that from supercritical state the system can go to metastable liquid as well as to metastable vapour state after LOCA. Relaxation from metastable fluid states is a fast process, followed by a local positive or negative pressure-jump, which might increase the damage around the leak. Conservative estimation will be given for the magnitude of these pressure jumps caused by the flashing or water hammer by assuming various initial pressure losses. In our calculations, three different equations of state are used: the simple van der Waals EoS; the Redlich-Kwong as an empirical development; and the more sophisticated non-cubic Deiters equation of state. These equations are able to describe metastable states qualitatively but with different accuracy. These calculations can help us to map the local immediate effect of any sudden pressure drop and therefore it can help to design better safety protocols.     

Boiling experiments for the validation of dryout models used in reactor safety

The coolability of fragmented corium is a major issue in reactor safety. Since the long-term coolability of such particle beds is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the pressure field inside the debris has a strong effect on the cooling potential in multi-dimensional cases as expected in severe accidents in light water reactors (LWR). Therefore, the determination of the pressure field for two-phase flows in porous media is one central point of interest.In this context simulation models and in particular dryout models were developed for reactor safety analyses which have to be validated by reliable experimental data. Therefore, basic experimental investigations have been carried out with inductively heated steel balls of 6 or 3 mm diameter to provide a database for the validation and modification of the friction laws included in these dryout models.The performed boiling and dryout experiments show clearly that models without the explicit consideration of the interfacial drag cannot predict the pressure distribution inside a boiling particle bed, not even qualitatively. Against it, models with an explicit consideration of the interfacial drag can describe the distribution of pressure inside a boiling particle bed.     

Study on the stability behaviour of a natural circulation pressure tube type boiling water reactor

The stability behaviour of a natural circulation pressure tube type boiling water reactor (BWR) has been investigated analytically. The analytical model considers homogeneous two-phase flow, a point kinetics model for the neutron dynamics and a lumped heat transfer model for the fuel dynamics. The results indicate that both Type I and Type II density-wave instabilities can occur in the reactor in both in-phase and out-of-phase mode of oscillations in the boiling channels of the reactor. The delayed neutrons were found to have strong influence on the stability of Type I and Type II density-wave instabilities. Also, the stability of the reactor is found to increase with increase in negative void reactivity coefficient unlike that observed previously in vessel type BWRs. Decay ratio map was predicted considering the effects of channel power, channel inlet subcooling, feed water temperature and channel exit quality, which are useful for the design of the reactor.     

Cross-section modelling effects on pressurised water reactor main steam line break analyses

Because of the strong asymmetric overcooling effects occurring during a PWR main steam line break (MSLB) event, an accurate analysis of this transient requires the use of 3-D kinetics methods. An assessment has been made of the relative performance of the two kinetics solvers currently employed at PSI for such analyses, viz. CORETRAN and SIMULATE-3 K. For the purpose, the simulation of a hypothetical MSLB in a real operated PWR MOX cycle has been considered, employing consistent 3-D core models with specified thermal-hydraulic boundary conditions at the lower and upper plenums. Although the employed cross-section library is in both codes based on the same set of homogenised 2-group cross-sections prepared with CASMO-4, significant differences are shown to occur due to the smaller moderator reactivity coefficient calculated in CORETRAN. It is found that this stems largely from differences in the cross-section formalism, i.e. the manner in which feedback dependencies are modelled and interpolated for the cross-section sets.In particular, the CORETRAN cross-section formalism induces an inadequate treatment of coupled feedback effects, principally between boron density and moderator temperature, which renders the MSLB dynamics predictions quite sensitive to the methodology employed during the cross-section preparation. As such, transient-specific cross-section libraries need to be produced for reliable MSLB analysis in this case. The cross-section model for SIMULATE-3 K, on the other hand, is shown to be adequate for accurately capturing the coupled reactivity effects occurring during an MSLB. In this case, the sensitivity of the results to other sources of uncertainties becomes more apparent, e.g. to those related to the neutron data and/or the thermal-hydraulic boundary conditions. Considering that many other state-of-the-art advanced kinetics solvers have cross-section formalisms similar to that of CORETRAN, effects of the type currently investigated need to be taken into account while developing methodologies for assessing neutronics-related uncertainties in best-estimate transient analysis.     

Experimental thermohydraulic stability map of a Freon-12 boiling water reactor facility with high exit friction

The natural-circulation characteristics and the density-wave stability characteristics of the natural-circulation Freon-12 facility DESIRE for a specific configuration have been determined by systematically performing experiments in the whole operating range. A large amount of data has been gathered to be used for future benchmarking of computer codes for the calculation of boiling water reactor (BWR) stability. Contrary to expectations, it was found that, for low subcooling values, the stability of the facility improves as the power is increased when keeping the subcooling number constant. This result can serve as a challenging benchmark for models and codes, since it is not in line with the experience from the majority of analytical and numerical models.     

事故条件下模块化小堆与同厂址大堆的辐射影响对比分析

本文针对海南小堆的实际厂址环境特征,根据机组初步的二级PSA源项,利用更实际的CALPUFF烟团模式开展事故条件下小堆和大堆对场外公众的辐射影响分析,比较不同事故下对周边居民和工作人员的受照特征。按照针对小堆的剂量准则,确定各种天气条件下满足该准则的距离,有助于更深入地认识小堆的事故特征及应急计划区划分等问题,为相关工程实践和应急监管工作提供参考。     

The physics of rewetting in water reactor emergency core cooling

Surface rewetting is essential for the re-establishment of normal and safe temperature levels following dryout in rod clusters or boiler tubes, or following postulated loss-of-coolant accidents in water reactors. Rewetting experiments have been performed with tubes and rods with a wide range of materials and experimental conditions (surface temperatures 300–800°C, constant water flows 0.1–30 g s⁻¹). The physical processes involved in the rewetting of high temperature surfaces have been shown to be identical for both falling water films and bottom flooding. The variation of rewetting velocity with mass flow has been determined, and shown to be independent of hydraulic diameter over the range 0.2–6 mm of practical interest. Data have also been obtained on the mass ‘carryover’ fraction. Theoretical solutions for the rewetting velocities have been obtained by analysis of thermal conduction in the surface. At low mass flows, effectively one-dimensional (axial) conduction cools the surface ahead of the rewetting front, and gives agreement with experiment. At higher mass flows the rewetting velocity is substantially independent of surface thickness and conductivity. The present data and the available world data for rewetting are shown to be in agreement with the theory.     

Phase Doppler anemometer: commissioning tests for measurement of water aerosol sizes and velocities in flashing jets

A state-of-the-art phase Doppler Anemometer (PDA) has been commissioned at AECL Research, Whiteshell Laboratories to undertake the measurement of size and velocity of water droplets generated in flashing jets. Experimental data on size and velocity distribution of water aerosols in flashing jets are required to support licensing of current multiple-unit and single-unit Canada deuteruim uranium stations. This paper presents the methodology involved in choosing the magnitudes of the various operating parameters of the PDA such as laser power and sensitivity of photomultiplier tubes in obtaining the experimental data. The various calibration and validation procedures used are also discussed. Size and velocity distributions in a typical flashing jet are presented.     

The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor

The reprocessing actinide materials extracted from spent fuel for use in mixed oxide fuels is a key component in maximizing the spent fuel repository utility. While fast spectrum reactor technologies are being considered in order to close the fuel cycle, and transmute these actinides, there is potential to utilize existing pressurized heavy water reactors such as the CANDU^®1 design to meet these goals. The use of current thermal reactors as an intermediary step which can burn actinide based fuels can significantly reduce the fast reactor infrastructure needed. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a typical CANDU nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel lattice was created and the two neutron group properties were transferred to RFSP in order to create a 3 dimensional time average full core model. The model was created with typical CANDU limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU reactors as an intermediary step in burning reprocessed spent fuel and reducing actinide burdens within the end repository. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle using existing and proven reactor technologies.     

The future 700 MWe pressurized heavy water reactor

The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor.