Characterization of simulated burnup fuel by nanoindentation

A simulated burnup UO₂ based fuel (150 GWd/t) was prepared by solid-state reactions. The phase equilibria of the simulated fuel were evaluated by XRD and SEM/EDX analysis. Nanoindentation tests were performed for the simulated fuel at room temperature in air. The modulus and hardness of the matrix phase and oxide precipitates that exit in the simulated fuel were directly evaluated by the nanoindentation.     

High enrichment of N isotope by ion exchange for nitride fuel development

Two kinds of nitrogen-isotope enrichment experiments were conducted for the purpose of developing a process to obtain highly enriched ¹⁵N by ion-exchange chromatography. As the first stage experiments, high enrichment of ¹⁴N was carried out by displacement chromatography using columns packed with cation-exchange resin and feed ammonium ion of which ¹⁵N abundance is natural namely 0.36%. In the second stage, high enrichment of ¹⁵N was studied starting from ca. 80% enriched ¹⁵N feed material; after 30 m migration 99.67% ¹⁵N was obtained. The height equivalent to a theoretical plate (HETP) of ion-exchange separation is usually very small, but the results clearly indicate that the HETP is enlarged when the high enrichment zone of ¹⁴N or ¹⁵N is developed at the front or the rear band boundary, respectively. The information obtained is very important and should be useful for designing when the plant for the high enrichment isotope separation is designed.     

Nitrogen isotope enrichment for nitride fuel by using hybrid chemical exchange process

In order to develop a feasible process for production of enriched nitrogen, chromatography operation was carried out to study the nitrogen isotope separation by using a high porous cation ion exchange resin(SQS-6). The ammonium ion adsorption band initially charged in the resin bed using 0.5 mol·dm⁻³ NH₄OH solution and was eluted in a reverse breakthrough manner. The temperature of column was kept constant at 308K by temperature-controlled water through a jacket. A constant length of adsorbed bands were maintained throughout the elution. The enrichment of ¹⁴N at the front boundary region and ¹⁵N at the rear boundary region were confirmed. The attention was placed on enrichment high purity of ¹⁴N (99.99%). An enrichment factor of ¹⁴N was determined to be 0.026±0.001 and HETP was evaluated 0.18 cm. HETP of front boundary region is obviously lager than that of the rear part.     

Measurement of 2-dimensional local instantaneous liquid film thickness around simulated nuclear fuel rod by ultrasonic transmission technique

An accurate knowledge of hydro-dynamic behavior of a liquid film flow on nuclear fuel rods is indispensable for analysis of the CHF under postulated loss-of-coolant-accidents in boiling water reactors. This work is concerned with a new development of ultrasonic transmission technique for film flow measurements. The technique adopted a rotating reflector, capable of measuring time-dependent spatial distribution of liquid film thickness around a simulated nuclear fuel rod. The scanning time is currently 4 ms for reconstruction of one image of the circumferential film thickness distribution.     

Single-phase subchannel mixing in a simulated nuclear fuel assembly

Single-phase subchannel mixing data were obtained from a 25 rod square array by measuring precisely subchannel exit temperatures over a range of test conditions. A least-squares type statistic operating on exit enthalpy differences was developed to ascertain, in conjunction with the COBRA-II subchannel computer analysis, an optimum value for the coefficient of Rowe and Angles's single-phase mixing correlation β. When the same analytical procedures were applied to data taken under conditions of subcooled nucleate boiling, an approximately linear correlation of β with average exit quality was found. The values obtained were for conditions of natural, or unaided mixing, and for design purposes should be considered as a dependable lower limit. Commercial rod-spacer designs intended to increase mixing will, of course, show substantially higher values of β.     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Experimental data on burn-out in a simulated BWR fuel bundle

Critical heat flux (CHF) experiments have been carried out on a 16-rod test section having the typical geometry of boiling water reactor (BWR) fuel elements and in particular a 366 cm length. Heat fluxes were uniform, both axially and radially. The tests were carried out for the CNEN Plutonium Program on CISE's 8 MW IETI-3 facility, at 71 kg/cm² abs, mass velocities of 12–200 g/cm² s and inlet sub-cooling of 15–180°C. Each corner rod was instrumented with four separate thermocouples to detect nnd locate the initiation of CHF, while the other rods were instrumented with four-junction thermopiles.     

Physics and safety studies of a compact ADS design loaded with nitride fuel

We have performed transient analysis of a nitride fueled and heavy liquid metal cooled accelerator driven system, in which the pitch-to-diameter ratio of pin lattice was set at 1.26. Unprotected loss-of-flow, unprotected transient-over-power and protected loss-of-heat-sink transients were simulated in a geometrical model of the suggested ADS design, using safety parameters obtained with the MCNP/MCNPX code.The simulations indicated that the suggested ADS design could survive the full set of transients, thanks to the introduction of the austenitic 15/15Ti stainless steel as fuel cladding material.Thus, the viability of an ADS design with small pin pitch and concomitant high proton source efficiency could be confirmed.     

Variation in the oxygen potential of a mixed-oxide fuel with simulated burnup

The oxygen potentials of solid solutions of UO₂, PuO₂, and the oxides of selected fission-product elements simulating stages in the burnup of a mixed-oxide fuel to 10 atom per cent have been measured at temperatures from 900 to 1100°C. At a given temperature and deviation from stoichiometry, the oxygen potential increases linearly with simulated burnup.     

Wet-chemical dissolution of TRISO-coated simulated high-temperature-reactor fuel particles

Chemical etching with different mixtures of acidic solutions has been investigated to disintegrate the two outermost coatings from tri-structural isotropic coated particles containing zirconia kernels, which are used in simulated particles instead of uranium dioxide. A scanning electron microscope (SEM) was used to study the morphology of the particles after the first etching step as well as at different stages of the second etching step. SEM examination shows that the outer carbon layer can be readily removed with a CrO₃–HNO₃/H₂SO₄ solution. This finding was verified by energy dispersive spectroscopy (EDS) analysis. Etching of the silicon carbide layer in a hydrofluoric–nitric solution yielded partial removal of the coating and localized attack of the underlying coating layers. The SEM results provide evidence that the etching of the silicon carbide layer is strongly influenced by its microstructure.     

Safety properties of bn-1200 with mixed oxide and nitride fuel: a comparison

Along with mixed oxide fuel, the possibility of using in BN-1200 dense nitride fuel, making it possible to attain higher technical-economic performance, is also studied. However, safety analysis will determine the choice of fuel type. In this connection, it is important to perform a comparative analysis of the inherent safety properties for variants of the BN-1200 core with mixed uranium-plutonium and nitride fuel for the most serious unanticipated loss-of-power accident with failure of all emergency protection organs of the reactor simultaneously. A two-dimensional version of the COREMELT computer code was used in the calculations. The computational analysis showed that the inherent safety of BN-1200 is much greater with nitride than with mixed uranium-plutonium fuel.     

BWR fuel reloads design using a Tabu search technique

We have developed a system to design optimized boiling water reactor fuel reloads. This system is based on the Tabu Search technique along with the heuristic rules of Control Cell Core and Low Leakage. These heuristic rules are a common practice in fuel management to maximize fuel assembly utilization and minimize core vessel damage, respectively. The system uses the 3-D simulator code CM-PRESTO and it has as objective function to maximize the cycle length while satisfying the operational thermal limits and cold shutdown constraints. In the system tabu search ideas such as random dynamic tabu tenure, and frequency-based memory are used. To test this system an optimized boiling water reactor cycle was designed and compared against an actual operating cycle. Numerical experiments show an improved energy cycle compared with the loading patterns generated by engineer expertise and genetic algorithms.     

Nondestructive measurement of the coating thickness for simulated TRISO-coated fuel particles by using phase contrast X-ray radiography

The TRISO-coated fuel particle for a HTGR (high temperature gas-cooled reactor) is composed of a nuclear fuel kernel and outer coating layers. The coating layers consist of a buffer PyC (pyrolytic carbon) layer, an inner PyC (I-PyC) layer, a SiC layer, and an outer PyC (O-PyC) layer. X-ray radiography is one of the nondestructive alternatives to measure a coating thickness without generating a radioactive waste. Phase contrast X-ray radiography technology is more powerful for acquiring a radiograph with clear boundaries, when compared with a conventional X-ray radiography. The contrast can be enhanced for weakly absorbing materials in a phase contrast X-ray radiograph by detecting an intensity variation due to the variation of a phase of the X-rays in the boundary between two objects. Phase contrast X-ray radiograph was acquired from a simulated TRISO-coated fuel particle with a micro-focus X-ray imaging system. The coating thickness was nondestructively measured from the phase contrast X-ray image for the fuel particle.     

Behavior of metallic precipitate in an irradiated simulated fuel

A simulated fuel specimen which was irradiated at the HANARO research reactor up to 3300 MWd/tU of a burn-up at the condition of 36 kW/m of a maximum linear power was studied by a shielded EPMA (Electron Probe Micro-Analyzer). In order to obtain an accurate analysis results, chemical and EPMA analyses were also performed on un-irradiated fresh simulated fuel, the results of which were compared with those of the irradiated simulated fuel. This study concentrated on the metallic precipitates of the irradiated simulated fuel specimen which contained lots of fission products. Among the several properties of the metallic precipitate, its size and composition were investigated. A large metallic inclusion was also observed in the irradiated simulated fuel, from which X-ray photographs were taken to analyze its properties.     

A theoretical procedure for detection of simulated cracks in a pipe by the direct current–potential drop technique

Based on the direct current–potential drop (dc–pd) technique, an efficient theoretical detection procedure is developed to identify the existence of simulated cracks in a pipe. By this procedure, the electric potential on a ‘pseudo’ perfect pipe needs to be calculated in advance by finite element method. The proposed defect influence factor, which is defined as the ratio of the electric potential of the defective pipe divided by that of the ‘pseudo’ perfect one, is then employed to reveal the effect of cracks on the electric potential. By depicting the contours of the defect influence factor with sufficient resolution, not only the position, but also the shape and length size of cracks in the pipe can be identified accurately by the detection criterion devised in this work. The types of detectable through-wall cracks include circumferential crack, inclined crack, and multiple cracks. Good detection results show the merits of the procedure developed for the identification of the simulated cracks as described above in the pipe structure.     

Thermal expansion of TRU nitride solid solutions as fuel materials for transmutation of minor actinides

The lattice thermal expansion of the transuranium nitride solid solutions was measured to investigate the composition dependence. The single-phase solid solution samples of (Np_0.55Am_0.45)N, (Pu_0.59Am_0.41)N, (Np_0.21Pu_0.52Am_0.22Cm_0.05)N and (Pu_0.21Am_0.18Zr_0.61)N were prepared by carbothermic nitridation of the respective transuranium dioxides and nitridation of Zr metal through hydride. The lattice parameters were measured by the high temperature X-ray diffraction method from room temperature up to 1478 K. The linear thermal expansion of each sample was determined as a function of temperature. The average thermal expansion coefficients over the temperature range of 293-1273 K for the solid solution samples were 10.1, 11.5, 10.8 and 8.8 × 10⁻⁶ K⁻¹, respectively. Comparison of these values with those for the constituent nitrides showed that the average thermal expansion coefficients of the solid solution samples could be approximated by the linear mixture rule within the error of 2-3%.