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At least 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

High Fidelity Forward Model Development for Nuclear Reactor Spent Fuel Technical Nuclear Forensics

The fidelity of the forward model within a spent fuel forensic analysis system was improved by using two unique methodologies. The first consisted of developing a system to create accurate one-group neutron cross-section libraries for any user specified reactor system. In such, a detailed model is developed using the depletion code MONTEBURNS. During MONTEBURNS execution, cross-section libraries are generated at every user specified burnup step in time. These libraries could be developed for many reactor systems, then housed in a database and used for analyzing unknown fuel samples. The forensic analysis system for spent fuel resulted in higher accuracy at predicting the initial uranium isotopic compositions and burnup from spent fuel samples. Using this method, the error in results was reduced from the order of 1-6% down to less than 1% when recovering a fuel sample's burnup and initial uranium isotopic composition. The second method consisted of implementing 2D/3D reactor depletion codes as the forward model within the system's framework. This method would allow the usage of potentially recoverable geometric information from an unknown sample. No predetermined cross-section library is required for the system using this method, therefore potentially reducing model error associated with the neutron flux spectrum. The accuracy of the recovered initial uranium isotopic compositions and burnup from spent fuel samples was also improved using this method, even more so than the first. For MTR reactors, the error using this method was significantly reduced and was driven to below 0.5%. However, additional research may be required to determine the ideal fission yield and recoverable energy per fission for cases where significant amounts of 239 PU are bred and burned throughout the life of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Design change management in regulation of nuclear fleets: World nuclear association's working groups on Cooperation in Reactor Design Evaluation and Licensing (CORDEL)

The 60 year life of a reactor means that a plant will undergo change during its life. To ensure continuing safety, changes must be made with a full understanding of the design intent. With this aim, regulators require that each operating organisation should have a formally designated entity responsible for complete design knowledge in regard to plant safety. INSAG-19 calls such an entity 'Design Authority'. This requirement is difficult to achieve, especially as the number of countries and utilities operating plants increases. Some of these operating organisations will be new, and some will be small. For Gen III plants sold on a turnkey basis, it is even more challenging for the operating company to develop and retain the full knowledge needed for this role. CORDEL's Task Force entitled 'Design Change Management' is investigating options for effective design change management with the aim to support design standardization throughout a fleet's lifetime by means of enhanced international cooperation within industry and regulators. This paper starts with considering the causes of design change and identifies reasons for the increased beneficial involvement of the plant's original vendor in the design change process. A key central theme running through the paper is the definition of responsibilities for design change. Various existing mechanisms of vendor-operator interfaces over design change and how they are managed in different organisational and regulatory environments around the world are considered, with the functionality of Owners Groups and Design Authority being central. The roles played in the design change process by vendors, utilities, regulators, owners' groups and other organisations such as WANO are considered The aerospace industry approach to Design Authority has been assessed to consider what lessons might be learned. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Iso-nuclear tungsten dielectronic recombination rates for use in magnetically-confined fusion plasmas

Under the auspices of the IAEA Atomic and Molecular Data Center and the Korean Atomic Energy Research Institute, our assembled group of authors has reviewed the current state of dielectronic recombination (DR) rate coefficients for various ion stages of tungsten (W). Subsequent recommendations were based upon available experimental data, first-principle calculations carried out in support of this paper and from available recombination data within existing atomic databases. If a recommendation was possible, data were compiled, evaluated and fitted to a functional form with associated uncertainty information retained, where available. This paper also considers the variation of the W fractional abundance due to the underlying atomic data when employing different data sets.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Effect of vanadium oxide addition on thermomechanical behaviors of borosilicate glasses: Toward development of high crack resistant glasses for nuclear waste disposal

The effect of of V2O5 additions (up to 8 mol%) on the physical properties (e.g., density, glass transition temperature and refractive indices) and mechanical behaviors (e.g., hardness, fracture toughness, elastic modulus) of a soda-lime borosilicate glass with a composition of 60.2SiO(2)-16.0B(2)O(3)-12.6Na(2)O-3.8Al(2)O(3)-5.7CaO-1.7ZrO(2) (mol%) have been investigated systematically. It was found that introducing V2O5 decreases the glass transition temperature and hardness while increases the refractive index and crack initiation load. Composition-property relationships, especially on mechanical properties, were examined in order to provide insights for the development of new glass materials with outstanding mechanical properties. The results show that V2O5 can be a glass component for engineering and insuring high crack resistant glasses and glasses with long-term stability. Additionally, elastic properties as a function of temperature were studied using Brillouin light scattering and temperature-independent elastic moduli behaviors were observed.

Lu, Xiaonan↗
DOE OSTI2015

Effect of specimen size and grain orientation on the mechanical and physical properties of NBG-18 nuclear graphite

Here we present a comparison of the measured baseline mechanical and physical properties of with grain (WG) and against grain (AG) non-ASTM size NBG-18 graphite. The objectives of the experiments were twofold: (1) assess the variation in properties with grain orientation; (2) establish a correlation between specimen tensile strength and size. The tensile strength of the smallest sized (4 mm diameter) specimens were about 5% higher than the standard specimens (12 mm diameter) but still within one standard deviation of the ASTM specimen size indicating no significant dependence of strength on specimen size. The thermal expansion coefficient and elastic constants did not show significant dependence on specimen size. Lastly, experimental data indicated that the variation of thermal expansion coefficient and elastic constants were still within 5% between the different grain orientations, confirming the isotropic nature of NBG-18 graphite in physical properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗