Heterogeneous cores for improved safety performance: A case study: The supercritical water fast reactor

Light water reactor (LWR) technology is nowadays the most successful commercial application of fission reactors for the production of electricity. However, in the next few years, nuclear industry will have to face new and demanding challenges: the need for sustainable and cheap sources of energy, the need for public acceptance, the need for even higher safety standards, the need to minimize the waste production are only a few examples. It is for these very reasons that a few next generation nuclear reactor concepts were selected for extensive research and development; super critical water reactors are among them. The use of a supercritical coolant would allow for higher thermal efficiencies and a more compact plant design, since steam generators, or steam separators and driers would not be needed, hence achieving a better economy. Moreover, because of the high heat capacity of supercritical water, relatively less coolant would be needed to refrigerate the reactor, therefore the feasibility to design a water cooled fast reactor: the supercritical water fast reactor (SCFR). This system presents unique features combining well-known fast and light water reactor characteristics in one design (e.g. a tendency to a positive void reactivity coefficient together with loss of coolant accident – LOCAs as a design basis accident). The core is in fact loaded with highly enriched MOX fuel (average plutonium content of 23%), and presents a peculiar and significant geometrical and material heterogeneity (use of radial and axial blankets, solid moderator layers, 12 different enrichment zones). The safety analysis of this very complex core layout, together with the optimization of the void reactivity effect through core design, is the main objective of this work.     

The need for innovative nuclear reactor and fuel cycle systems: Strategy for development and future prospects

The paper reviews the major nuclear power challenges in safety, waste and non-proliferation areas that are hindering large-scale participation of nuclear power in the world energy market. It reviews also the ongoing R&D in reactor and fuel cycle areas and discusses the need for international cooperation in innovative R&D activities to address the above challenges. The main question is how to address these challenges and at the same time improve economic attractiveness of nuclear power in the global deregulated electricity markets.     

The inherently-safe power reactor DYONISOS : Dynamic nuclear inherently-safe reactor operating with spheres

A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydrodynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h.     

Cermet fuel in a light water reactor: a possible way to improve safety. Part I. Fabrication and characterization

The concept of cermet (ceramic-metallic) fuel for LWRs is considered. Cermet fuel utilization allows one to reduce the fuel operating temperature and to increase reactor safety under DBAs (Design Basis Accident). A general consideration of the cermet fuel application as well as design features and fabrication techniques of cermet fuel pins are presented in the paper.     

Cermet fuel in a light water reactor: a possible way to improve safety. Part II. Testing the fuel

This paper summarizes what is done for the experimental testing of cermet fuel with various matrix materials. Low neutron absorption, high heat conductivity, good corrosion resistance in water, low chemical interaction with cladding (zirconium alloy) and UO₂ in normal and accident conditions, technological ability — are the requirements of the matrix material [1]. Suitability of the proposed solutions to the cermet fuel design with respect to these requirements was proven through a series of experiments simulating fuel operating and accidental conditions.     

Nuclear safety of spent fuel pool storage at nuclear power plants with VVER-1000: physical principles and structural solutions

The physical principles and structural solutions concerning nuclear safety in handling nuclear fuel at nuclear power plants with VVER-1000 are examined. It is shown that a structure of storage racks in the holding pools based on the spacing of hexahedral boron steel pipes, the structure acting as a combined neutron trap, will increase the pool capacity as compared with the loosely packed pools and will make the handling of nuclear fuel nuclear safe under normal operating conditions and during accidents. In the future, this construction will make it possible to use fuel assemblies with high uranium content and fuel enrichment above 5%.     

The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis

In September 1988, the United States Nuclear Regulatory Commission issued a revised emergency core cooling system rule for light water reactors that allows, as an option, the use of best estimate plus uncertainty methods in safety analysis. To support the 1988 licensing revision, the United States Nuclear Regulatory Commission and its contractors developed the code scaling, applicability and uncertainty evaluation methodology to demonstrate the feasibility of the best estimate plus uncertainty approach. The phenomena identification and ranking table (PIRT) process, Step 3 in the code scaling, applicability and uncertainty methodology, was originally formulated to support the best estimate plus uncertainty licensing option. Through further development and application, the PIRT process has shown additional utility as a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. The generic PIRT process, including typical and common illustrations from prior applications that promoted further development of the process, are described. Analysis of the results of the prior applications is also described. The analysis results provide information that can help guide future applications of the process in a graded approach based on phenomena relative importance.     

The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the B-500SKDI reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology.     

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application

The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper (Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.     

The autoassociative neural network in signal analysis: III. Enhancing the reliability of a NN with application to a BWR

This is the third and last of a series of papers trying to unveil the opaqueness of neural networks structure through a geometrical approach [Marseguerra M., Zoia, A., 2005a. The autoassociative neural network in signal analysis: I. The data dimensionality reduction and its geometric interpretation. Ann. Nucl. Energy 32, 1191–1206, Marseguerra, M., Zoia, A., 2005b. The autoassociative neural network in signal analysis: II. Application to on-line monitoring of a simulated BWR component. Ann. Nucl. Energy 32, 1207–1223]. Artificial neural networks (NN) provide a powerful tool in the operation of complex systems, such as nuclear power plants, in that they are suitable to determine the relationship between measured variables and control parameters on the basis of input-output examples. However, their major drawback is the fact that they always provide an output to the user, regardless of the appropriateness of the input. In this paper, we propose to adopt an autoassociative neural network (AANN) to work in cooperation with the NN to first assess the well-posedness of the desired neural model and to successively establish the appropriateness of the input data. The neural algorithm has been applied to a nuclear problem: the estimation of the reactivity forcing function parameters from the values of the measured neutron flux in a BWR reactor (provided by a reduced-order literature model). In this example, the AANN was able to suggest through geometrical considerations how to decompose the dataset in order to obtain a successful training for the NN and thereafter to validate the input data, thus enhancing the reliability of the NN model output.     

Constitutional studies on carbide nuclear fuels for fast breeder reactors: III. The uranium-barium-carbon ternary system at 1400°C,with observations on the uranium-strontium-carbon system

The ternary system U-Ba-C has been examined at 1400°C and the solid-state compatibility lines established. No compound formation was found to occur and solubility effects were found to be minimal. A tentative examination of compositions in the U-Sr-C system indicates that it is of a similar form to that of the U-Ba-C system.