Technical information engineering system /

"United States Atomic Energy Commission, Contract No. AT (11-1)-171."

Nuclear science and technology for ceramists : proceedings of a symposium, April 7-12, 1966.

Issued May 26, 1967.

Nuclear science; a bibliography of selected unclassified literature.

Mode of access: Internet.

Engineering materials list : cumulative drawing number Index through supplement 24 /

AEC Report No. TID-4101 REV. 1.

Nuclear technology : a selected list of references /

April 1955.

Radiological surveillance study at the Haddam Neck PWR nuclear power station

Mode of access: Internet.

A three region steam drum model for a nuclear power plant simulator (BRENDA) /

"Date published: August 1981."

Nuclear reprocessing and proliferation : alternative approaches and their implications for the Federal budget /

Mode of access: Internet.

Effects of nuclear radiation on men and materials.

Mode of access: Internet.

Reactor handbook.

Includes bibliographies.

High stress monitoring of prestressing tendons in nuclear concrete vessels using fibre-optic sensors

Maintaining the structural health of prestressed concrete nuclear containments is a key element in ensuring nuclear reactors are capable of meeting their safety requirements. This paper discusses the attachment, fabrication and characterisation of optical fibre strain sensors suitable for the prestress monitoring of irradiated steel prestressing tendons. The all-metal fabrication and welding process allowed the instrumented strand to simultaneously monitor and apply stresses up to 1300 MPa (80% of steel's ultimate tensile strength). There were no adverse effects to the strand's mechanical properties or integrity. After sensor relaxation through cyclic stress treatment, strain transfer between the optical fibre sensors and the strand remained at 69%. The fibre strain sensors could also withstand the non-axial forces induced as the strand was deflected around a 4.5 m bend radius. Further development of this technology has the potential to augment current prestress monitoring practices, allowing distributed measurements of short- and long-term prestress losses in nuclear prestressed-concrete vessels. © 2014 Elsevier B.V.

Evaluation of Two Ultrasonic Systems for Analysis of Porosity in Ceramic

The Ultrasound Laboratory of the Nuclear Engineering Institute (LABUS / IEN) has developed an ultrasonic technique to measure porosity in nuclear fuel pellets (UO2). By difficulties related to the handling of UO2 pellets, Alumina (Al2O3) pellets have been used in preliminary tests, until a methodology for tests with pellets of UO2 could be defined. In a previous work, in which a contact ultrasonic technique was used, good results were obtained to measure the porosity of Alumina pellets. In the current studies, it was found that the frequency spectrum of an ultrasonic pulse is very sensitive to the porosity of the medium in which it propagates. In order to define the most appropriate experimental apparatus for using immersion technique in future tests, two ultrasonic systems, available in LABUS, which permit to work with the ultrasonic pulse in the frequency domain were evaluated . One system was the Explorer II (Matec INSTRUMENTS) and the other the ultrasonic pulse generator Epoch 4 Plus (Panametrics) coupled with an oscilloscope TDS 3032B (Tektronix). For this evaluation, several frequency spectra were obtained with the two equipment, by the passage of the ultrasonic wave in the same pellet of Alumina. This procedure was performed on four different days, on each day 12 ultrasonic signals were acquired, one signal every 10 minutes, with each apparatus. The results were compared and analyzed as regard the repeatability of the frequency spectra obtained.

CHARACTERIZATION OF RADIATION DAMAGE TO A NOVEL PHOTONIC CRYSTAL SENSOR

New methods of nuclear fuel and cladding characterization must be developed and implemented to enhance the safety and reliability of nuclear power plants. One class of such advanced methods is aimed at the characterization of fuel performance by performing minimally intrusive in-core, real time measurements on nuclear fuel on the nanometer scale. Nuclear power plants depend on instrumentation and control systems for monitoring, control and protection. Traditionally, methods for fuel characterization under irradiation are performed using a “cook and look” method. These methods are very expensive and labor-intensive since they require removal, inspection and return of irradiated samples for each measurement. Such fuel cladding inspection methods investigate oxide layer thickness, wear, dimensional changes, ovality, nuclear fuel growth and nuclear fuel defect identification. These methods are also not suitable for all commercial nuclear power applications as they are not always available to the operator when needed. Additionally, such techniques often provide limited data and may exacerbate the phenomena being investigated. This thesis investigates a novel, nanostructured sensor based on a photonic crystal design that is implemented in a nuclear reactor environment. The aim of this work is to produce an in-situ radiation-tolerant sensor capable of measuring the deformation of a nuclear material during nuclear reactor operations. The sensor was fabricated on the surface of nuclear reactor materials (specifically, steel and zirconium based alloys). Charged-particle and mixed-field irradiations were both performed on a newly-developed “pelletron” beamline at Idaho State University's Research and Innovation in Science and Engineering (RISE) complex and at the University of Maryland's 250 kW Training Reactor (MUTR). The sensors were irradiated to 6 different fluences (ranging from 1 to 100 dpa), followed by intensive characterization using focused ion beam (FIB), transmission electron microscopy (TEM) and scanning electron microscopy (SEM) to investigate the physical deformation and microstructural changes between different fluence levels, to provide high-resolution information regarding the material performance. Computer modeling (SRIM/TRIM) was employed to simulate damage to the sensor as well as to provide significant information concerning the penetration depth of the ions into the material.

A new age of fuel performance code criteria studied through advanced atomistic simulation techniques

A fundamental step in understanding the effects of irradiation on metallic uranium and uranium dioxide ceramic fuels, or any material, must start with the nature of radiation damage on the atomic level. The atomic damage displacement results in a multitude of defects that influence the fuel performance. Nuclear reactions are coupled, in that changing one variable will alter others through feedback. In the field of fuel performance modeling, these difficulties are addressed through the use of empirical models rather than models based on first principles. Empirical models can be used as a predictive code through the careful manipulation of input variables for the limited circumstances that are closely tied to the data used to create the model. While empirical models are efficient and give acceptable results, these results are only applicable within the range of the existing data. This narrow window prevents modeling changes in operating conditions that would invalidate the model as the new operating conditions would not be within the calibration data set. This work is part of a larger effort to correct for this modeling deficiency. Uranium dioxide and metallic uranium fuels are analyzed through a kinetic Monte Carlo code (kMC) as part of an overall effort to generate a stochastic and predictive fuel code. The kMC investigations include sensitivity analysis of point defect concentrations, thermal gradients implemented through a temperature variation mesh-grid, and migration energy values. In this work, fission damage is primarily represented through defects on the oxygen anion sublattice. Results were also compared between the various models. Past studies of kMC point defect migration have not adequately addressed non-standard migration events such as clustering and dissociation of vacancies. As such, the General Utility Lattice Program (GULP) code was utilized to generate new migration energies so that additional non-migration events could be included into kMC code in the future for more comprehensive studies. Defect energies were calculated to generate barrier heights for single vacancy migration, clustering and dissociation of two vacancies, and vacancy migration while under the influence of both an additional oxygen and uranium vacancy.