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Liquid-Metal-Bonded Gap for Light Water Reactor Fuel Rods
Principal Investigator: D. Olander

Liquid metal, which consists of 1/3 weight portion each of Pb, Sn, and Bi, was proposed to be the bonding substance in place of helium in the light-water reactor fuel-cladding gap to eliminate a large DT over the gap before closure occurs. However, experimental tests revealed that by simply loading fuel pellets through liquid metal at atmospheric pressure, unfilled regions formed in the gap due to surface tension effect. Results from HEATING 7.3 heat transfer code indicated that this could potentially lead to local fuel overheating. A technique was developed to completely eliminate the unfilled spots for both the dished PWR and non-dished BWR pellets by using the right combination of evacuation and pressurization. The technique was performed on a zircaloy cladding (~ 30 cm long) and UO2 pellets. To visually examine the bond integrity, the cladding was torn apart and the liquid metal completely filling the gap was observed, as shown in the Figure.

In addition, numerical calculations of the delay in fission gas release due to the initial reduction in UO2 operating temperature were studied, and the results showed that the onset of the release depended strongly upon the saturation value used and was very case-specific. Calculated differences between time-to-saturation with liquid metal and He in the gap were as high as ~ 1 year, and as low as 1-2 days.

Future plans include full-scale fabrication mockup of a 4-m long fuel rod, exploring ultra sonic testing to non-destructively examine the liquid metal bond integrity, and performing instrumented-rod irradiation test/PIE.

Properties of UZrH fuel
Principal Investigator: D. Olander

A novel fuel design has been proposed in which the usual UO2 fuel is replaced by a hydride of uranium and zirconium. The advantage of the new design is the placement of the neutron moderator (hydrogen) directly in the fuel. This feature results in a more compact core for the same power and a more rapid shutoff of reactor transients. The research examines the materials properties of the hydride fuel under power reactor conditions. Among the topics investigated are the compatibility of the fuel with Zircaloy cladding and the reaction of steam with the hydride, a situation that occurs in the event of a failed cladding.

Design of an on-line process for boron removal from LWR coolant water
Principal Investigator: D. Olander

Figure 1 – Pentaborate Diester: This figure shows the structure of a polyborate anion bound with a water soluble polymer. In this case, a five-boron polyanion with a minus three overall charge bridges two functional groups.

Natural boric acid, about twenty percent strongly-absorbing boron-10, is added to PWR coolant to provide necessary beginning of cycle reactivity control. As burnup increases, solid ion-exchange resin bed remove borate anions, maintaining the proper reactivity control value for the coolant. Lithium hydroxide balances the pH, but the resulting solution has a high ionic strength, leading to accelerated stress corrosion cracking of steam generator tubes. PWR’s using progressively higher enrichments require a greater value of chemical reactivity control at the beginning of the fuel cycle, and thus need more lithium hydroxide to balance pH. By using boric acid enriched in boron-10, less pH correction is necessary, and the resulting lower ionic strength slows stress corrosion cracking.

Using enriched boric acid (EBA) necessitates the development of an efficient boron recovery process to offset the high initial cost of EBA. The process of selectively binding solvated metals to water-soluble polymers and subsequently separating them from the solvent by passing the solution through an ultrafiltration membrane is collectively termed polymer filtration. Polymer filtration is an effective technique for removing actinides from aqueous solutions, and if successfully applied to boron, it could make use of EBA much more economical.

Boron presents a challenge through its complicated and quite unique chemistry. Borate anions self-polymerize at high concentrations, forming molecules with up to five boron atoms per molecule. In addition, multiple polyanionic species can exist in equilibrium. The water-soluble polymers themselves behave as most polymers do, in that their binding characteristics are highly sensitive to solution conditions such as pH and temperature, and also that they behave chemically more as an average of many species rather than as one discrete species. An effective recovery process must be operable with both a high efficiency and high recovery, and must be robust enough to survive in a process stream loaded with corrosion products and failed fuel constituents.

Our current research involves understanding the chemical behavior of the polymer/boron/water system. By employing a number of developed chemical speciation computer codes, fed with physiochemical data determined by experimentalists within the project, we can determine how a theoretical process might behave chemically under a variety of conditions relevant to the process design. When the chemistry is sufficiently well known, we will transition into the actual process design, with the end goal being to present deployable technology for use with existing reactors.

Preventing corrosion of spent fuel by water in the Yucca Mountain Repository
Principal Investigator: D. Olander

Corrosion of spent fuel exposed to the water dripping through the drifts in the repository is the main source of fission product and actinide element release to the far-field rocks and potentially to the water table. Yucca mountain water is corrosive to UO2 because it is exposed to air and because it contains dissolved carbonates. The technique studied involves placing large quantities of nonradioactive UO2 between the downflowing water and the spent fuel. In this way, the dissolution reactions will take place primarily with the nonradioactive UO2, and the water reaching the spent fuel will have been stripped of its reactive components (O2 and carbonate ions). In the laboratory, synthetic Yucca mountain water drips through a small column containing UO2 pieces and the composition, pH, eH and dissolved uranium) of the outlet water is measured.

Fusion Reactor Materials
Principal Investigator: B. Wirth, Wirth Research Group

The development of fusion as a viable energy source depends on ensuring structural materials integrity. Structural materials in fusion reactors will operate in harsh radiation conditions including high displacement rates from 14 MeV neutrons with accompanying high levels of hydrogen and helium production and will experience severe degradation of mechanical properties. The development of structural materials for use in such a hostile environment is predicated on understanding the underlying physical mechanisms responsible for microstructural evolution along with corresponding dimensional instabilities and mechanical property changes. This research aims to apply predictive, physically based multiscale modeling to improve understanding of the underlying mechanisms of material changes in the fusion environment, with the ultimate objective to aid development of advanced materials. The multiscale modeling methodology involves a hierarchical approach, integrating ab initio electronic structure calculations, molecular dynamics (MD) simulations and kinetic Monte Carlo (KMC) over the relevant length and time scales to model the fates of defects and solutes (including hydrogen and helium) and thus, predict microstructural evolution in ferritic/martensitic-based alloys. Further, the prediction of irradiated microstructures is linked to quantitative predictions of mechanical properties through the use of molecular dynamics simulations of the interaction between moving dislocations and the radiation obstacles. An important component in this research is integration with experiments, both to validate the modeling and to assist in experimental interpretation, including experimental studies to understand the diffusion, trapping and fate of helium in ferritic-martensitic steels, the thermal stability and interactions of nanometer sized oxide dispersion precipitates in these steels and the radiation hardening and embrittlement.

Embrittlement of Reactor Pressure Vessel Steels
Principal Investigator: B. Wirth, Wirth Research Group

The objective of this work is to answer the outstanding questions regarding the irradiation hardening and embrittlement of reactor pressure vessel steels, which threaten to limit the operating lifetime of nuclear power plants worldwide. It is well established that RPV steel hardening and embrittlement results from the formation of sub-nanometer defect cluster-solute complexes and a high number density of ultra-fine, nanometer sized copper-manganese-nickel precipitates.Yet, a number of outstanding questions and controversies remain that impact predictions of hardening and embrittlement at longer irradiation exposure times, to be experienced during lifetime extension. Significant controversy exists regarding the actual composition of the nanoprecipitates and, the identity of the defect cluster-solute complexes has largely rested on interpretations of annealing signatures and simulation results. This research combines computational multiscale modeling with state of the art experimental characterization using positron annihilation spectroscopy and small angle neutron scattering, to determine the atomic characterand composition of the nano-precipitates and sub-nanometer vacancy-solute clusters and resolve the outstanding questions related to the lifetime extension of light water reactors.

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