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Monitoring of fuel element failure in a nuclear reactor is an important process to ensure safe operation of the nuclear reactor. It is feasible to detect fuel element rupture by analysing the radionuclides of fission products and activation products in the primary coolant. To identify the radionuclides in the primary coolant, the most convenient method to measure their activity concentrations is γ-ray spectrometry. A γ-ray spectral analysis method for real-time detection of fuel element failure is proposed in this paper. The method aims to implement a Gaussian smoothing processing for spectral data, extract peaks of spectral data through mathematical morphological transformation, determine left and right peak boundaries with a gradient descent algorithm and finally estimate net peak area with an image method. The method can rapidly recognize a designated key radionuclide from an obtained spectrum and extract peak information to meet the demands of real-time monitoring of fuel element failure in pressurized water reactors.

An incremental approach to introducing thorium to the conventional pressure-tube heavy-water reactor natural uranium fuel cycle is investigated. The approach involves the replacement of the centre fuel element of the bundle with an element of thorium dioxide. Increasing the operating margin of a key safety parameter, the coolant void reactivity, is a prime motivating factor. The analyses showed that the simple use of a single pin of thorium is unlikely to be economically advantageous due to a large burnup penalty and increased fuel costs. However, a slight reduction in the void reactivity is observed, and this approach does allow the exploitation of the energy potential available in thorium as an alternative nuclear fuel resource through the development of a U-233 resource. This bundle concept may also be advantageous from a fuel disposal point of view, as the fuel requires less time in storage before emplacement in a deep geological repository.

O Reator Multipropósito Brasileiro, conhecido como RMB, visa a produção de radioisótopos para aplicação na saúde, indústria e meio ambiente, irradiação e teste de materiais, realização de pesquisas científicas em física nuclear e aplicação de feixe de nêutrons em apoio a pesquisas científicas e tecnológicas desenvolvidas em várias áreas do conhecimento. O combustível para esse reator é do tipo MTR (Materials Testing Reactor) e será produzido pelo Centro de Combustível Nuclear do IPEN, que já vem desenvolvendo e fabricado rotineiramente esse tipo de combustível e tem como desafio aumentar a produtividade das instalações para atender à atual e futura demanda. Objetivou-se neste trabalho estudar a influência da folga entre o briquete e a cavidade da moldura no aparecimento de trincas no núcleo da placa combustível. Para isso, realizou-se uma análise do processo de produção das placas a serem usadas no Reator Multipropósito Brasileiro e estudou-se diferentes folgas definiu-se uma folga padronizada. Foram laminadas e radiografadas placas combustíveis montadas com diferentes folgas entre a moldura e o briquete, identificando-se a ocorrência de trincas no núcleo. As informações obtidas permitiram definir as dimensões e tolerâncias adequadas dos briquetes e das cavidades das molduras que garantem a produção de placas combustíveis dentro dos padrões aceitáveis, melhorando a produtividade do Centro de Combustível Nuclear. Além da folga entre o briquete e a cavidade da moldura, a qual era considerada a principal causa para o trincamento do núcleo, outras causas também foram investigadas. Tanto a folga como também a qualidade do acabamento superficial da matriz de compactação do briquete mostraram-se ser a causa para o trincamento do núcleo durante a laminação das placas combustíveis. The Brazilian Multi-Purpose Reactor, known as RMB, aims to produce radioisotopes for use in health, industry and the environment, irradiation and material testing, conducting scientific research in nuclear physics and applying a neutron beam in support of scientific and technological research developed in various areas of knowledge. The fuel for this reactor is of the MTR (Materials Testing Reactor) type and will be produced by the Nuclear Fuel Center of IPEN, which has already developed and routinely manufactured this type of fuel and has the challenge of increasing the productivity of the installations to meet the current and future demand. The objective of this work was to study the influence of the gap between the briquette and the frame cavity on the appearance of cracks in the fuel plate core. For this, an analysis of the production process of the fuel plates to be used in the Brazilian Multi-Purpose Reactor was carried out and different clearances were studied, a standardized clearance was defined. Fuel plates assembled with different gaps between the frame and the briquette were laminated and radiographed, identifying the occurrence of cracks in the core. The information obtained allowed to define the appropriate dimensions and tolerances of the briquettes and cavities of the frames that guarantee the production of fuel plates within acceptable standards, improving the productivity of the Nuclear Fuel Center. In addition to the gap between the briquette and the frame cavity, which was considered the main cause for cracking the core, other causes were also investigated. Both the clearance and the quality of the surface finish of the briquette compaction matrix proved to be the cause for the cracking of the core during the rolling of the fuel plates.

While developing advanced nuclear energy systems, one needs to carefully consider the performance of the fuel in a particular reactor. If a metallic fuel is used in a sodium fast reactor (SFR), many phenomena can affect the performance of the fuel, including the one that this chapter discusses, namely, fuel–cladding chemical interaction (FCCI). During irradiation, a metallic fuel swells and eventually contacts the cladding, at which time FCCI can occur. During this process, interdiffusion occurs between the fuel, cladding, and fission products, which can result in the formation of interaction zones on the inner surface of the cladding that can become brittle or may contain relatively low-melting phases. The result of this process can be cracking and failure of the cladding. Detailed information on FCCI in irradiated metallic SFR fuel is minimal in the literature. The metallic fuels with the most data include U–Zr and U–Pu–Zr alloys. The bulk of the irradiation data that is available was generated from the destructive examination of individual fuel elements irradiated in the Experimental Breeder Reactor-II over the course of a 30-year time frame. Three examination techniques were employed to characterize FCCI in these fuel elements: optical metallography, electron probe microanalysis, and scanning electron microscopy. The results of these examinations have led to a reasonable understanding of FCCI and how it affects fuel performance. The fuel elements that were characterized included U–Zr or U–Pu–Zr alloy fuels and HT-9, D9, or Type 316 stainless steel cladding. Out-of-pile testing has also been conducted using annealed diffusion couples comprised of alloy and cladding constituents to improve the understanding of fuel–cladding interactions. Furthermore, to investigate liquefaction in FCCI zones, actual irradiated fuel segments or complete fuel pins have been heated to high temperatures and then characterized to ascertain the development of liquid phases at the fuel–cladding interface. Finally, transient tests have been performed on irradiated fuel pins to investigate microstructural development at the fuel–cladding interface. This chapter describes the data that is available with respect to out-of-pile and in-pile testing that can be employed to develop an understanding of FCCI in irradiated metallic nuclear fuels.

This paper presents a preliminary neutronic conceptual design for a heavy water research reactor core substituting low-enriched uranium fuel for natural uranium dioxide fuel; focusing on reducing the amount of plutonium in the spent fuel. In this study, two core models are proposed and investigated. These result in a smaller core configuration and consequently less plutonium inventory at the end of the cycle. Parameters such as multiplication factors, fuel burn-up, and radial power peaking factor are calculated by the MCNPX2.6 code. The results show that in converted reactor cores the mass of plutonium is decreased about 85% comparing to the main core without substantial changes in the core and fuel dimension.