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Considerations for Nuclear Thermal Propulsion Reactor Technology Readiness Level Advancement

Nuclear thermal propulsion (NTP) is a candidate in-space propulsion technology for crewed missions to Mars. An NTP engine relies on the use of a reactor as a heat exchanger to directly heat a hydrogen propellant which is expanded through a nozzle for high specific impulse (> 850s), high thrust (>15 klbf) propulsion. NTP reactor and engine technologies are currently being developed to enable a future NTP engine prototype demonstration. Technology readiness level (TRL) definitions are a useful tool to plan necessary technology advancement tasks as they provide guidance on expected development activities and necessary level of fidelity of test article or testing conditions for demonstration tasks. However, technology readiness definitions for an NTP reactor and related demonstration tasks have not yet been defined. This paper identifies relevant TRL definitions by the National Aeronautics and Space Administration (NASA) and Department of Energy (DOE). Based on these definitions, preliminary recommendations on testing conditions and important parameters to consider for NTP reactor technology development planning are provided for TRL 1 - 6. As a part of this effort, functional characteristics and important performance parameters for the reactor are identified, as well as a summary of relevant literature which was considered when assessing development tasks capable of meeting TRL definition criteria.

nuclear thermal propulsion↗

DOE OSTI2012

Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could actually help extend the life of a fuel pin in a “long pin” reactor design to a higher peak fuel burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Supercell Depletion Studies for Prismatic High Temperature Reactors

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challenges exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time-dependent measures of performance including uranium usage, TRU inventory, and radiotoxicity to evaluate the complex impacts of transition from the current uranium-fueled LWR system, and other more realistic impacts that may not be intuited from the time-independent steady-state conditions of the end-state fuel cycle. These analyses were performed using the Verifiable Fuel Cycle Simulation Model VISION. Compared with static calculations, dynamic results paint a different picture of option space and the urgency of starting a FR fleet. For example, in a static analysis, there is a sharp increase in uranium utilization as CR exceeds 1.0 (burner versus breeder). However, in dynamic analyses that examine uranium use over the next 1 to 2 centuries, behavior as CR crosses the 1.0 threshold is smooth, and other parameters such as the time required outside of reactors to recycle fuel become important. Overall, we find that there is no unambiguously superior value of TRU CR; preferences depend on the relative importance of different fuel cycle system objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Developing a Comprehensive Software Suite for Advanced Reactor Performance and Safety Analysis

This paper provides an introduction to the reactor analysis capabilities of the nuclear power reactor simulation tools that are being developed as part of the U.S. Department of Energy s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit. The NEAMS Toolkit is an integrated suite of multi-physics simulation tools that leverage high-performance computing to reduce uncertainty in the prediction of performance and safety of advanced reactor and fuel designs. The Toolkit effort is comprised of two major components, the Fuels Product Line (FPL), which provides tools for fuel performance analysis, and the Reactor Product Line (RPL), which provides tools for reactor performance and safety analysis. This paper provides an overview of the NEAMS RPL development effort.

Multi-physics↗

DOE OSTI2014

Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Scanning tunneling microscope assembly, reactor, and system

An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Target-fueled nuclear reactor for medical isotope production

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Seismic attenuation system for a nuclear reactor

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Seismic attenuation system for a nuclear reactor

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Mobile heat pipe cooled fast reactor system

A mobile heat pipe cooled fast nuclear reactor may be configured for transportation to remote locations and may be able to provide 0.5 to 2 megawatts of power. The mobile heat pipe cooled fast reactor may contain a plurality of heat pipes that are proximate to a plurality of fuel pins inside the reactor. The plurality of heat pipes may extend out of the reactor. The reactor may be configured to be placed in a standard shipping container, and may further be configured to be contained within a cask and attached to a skid for easier transportation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Radioactivity Calculator for Research and Test Reactor Neutron-Activated Materials

The Idaho National Laboratory (INL) Nuclear Science User Facility (NSUF) supports nuclear energy-relevant research into the effects of radiation on nuclear fuels and structural materials. The NSUF provides access, at no cost to the researcher, to the Advanced Test Reactor (ATR), the High-Flux Isotope Reactor (HFIR), and other university and national laboratory research and test reactors for radiation effects experiments. Post-irradiation examination of materials poses some difficulties when studying neutron-irradiated materials due to the radioactivity levels of the materials. Inside a test reactor, under high neutron flux, stable materials can activate (capture a neutron) and become unstable or radioactive. These newly created radioactive materials decay, based on their respective half-lives, and emit energetic radiation. Nuclides with short half-lives are able to decay to sufficiently low radiation levels prior to the sample being removed from the reactor. However, nuclides with longer half-lives present a challenge since they must be handled with extreme caution and stored until their radioactivity levels decrease. If immediate examination is required, the examination may take place in a hot cell, which can dramatically increase the cost of the project. The NSUF has access to a variety of hot cells and characterization capabilities at national laboratories, universities, and industrial sites across the United States. In order to estimate the post-irradiation radioactivity levels of materials early in the experiment cycle, we created a radioactivity calculator, hosted on the NSUF web site (nsuf.inl.gov), which will consider three stages of neutron activations and decays for sample materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Fast Modular Reactor Nuclear Design Parameters of Fuel Cycle and Power Distributions

This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 ºC. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutronics calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Evaluation of Oak Ridge National Laboratory Health Physics Research Reactor Operation Data for Critical Benchmark Creation [Abstract]

The Oak Ridge National Laboratory (ORNL) Health Physics Research Reactor (HPRR) was a research reactor designed and built at ORNL in 1961. The critical assembly was using a highly enriched uranium and molybdenum alloy as the fuel, and it could be operated in steady-state or burst modes. The reactor was used for about 25 years to produce a lot of publications related to dosimetry, radiobiology and radiation detectors testing before its decommissioning in 1987. In recent years, the idea of using legacy operation data from the to create a valuable critical accident alarm system shielding benchmark arose. Such a benchmark has been submitted to the International Criticality Safety Benchmark Experiment Project (ICSBEP) Technical Review Group for a potential inclusion in the 2022 version of the handbook. Another way to use the valuable data from the operation of the HPRR is to evaluate the feasibility of the creation of a subcritical or prompt supercritical benchmark for inclusion in the ICSBEP or the International Reactor Physics Experiments Evaluation Project (IRPhEP) handbooks. To initiate a burst, the HPRR had to be operated in a slightly subcritical state for a few minutes. Then, the insertion of the burst control rod would greatly increase the reactivity of the system and start the burst. No critical configuration of the HPRR critical assembly could be located. The only information available concerns stable subcritical and prompt supercritical states, found in a burst experiments’ logbook. In the recovered logbook pages, information about 8 different bursts is available. The information includes the rods positions before and during a burst, the recorded subcritical reactor period and reactivity, and the burst fission yield derived from the temperature elevation sulfur pellet irradiation analysis. By using the HPRR logbook information and the as-built drawings of the critical assembly, a highly detailed model of the HPRR was created with SCALE 6.2.4/KENO-VI. Eight KENO-VI models were created to replicate the sub-critical assembly configurations described in the eight bursts from the recovered logbook pages. KENO-VI calculates k eff and it can be linked to a reactivity value in cents by using the delayed neutron fraction B eff , also calculated by KENO-VI. KENO-VI can also be used to model the prompt super-critical configurations of the HPRR and to assess the similarity with the burst measurements by comparing the calculated k eff and the measured fission yields between each burst. Unfortunately, high uncertainty exist and the obtained discrepancies between experiments and calculation results are high, compromising the creation of a valuable critical benchmark from HPRR operation data. The reasons of the discrepancies and potential ways to solve them are explored.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Kinetics and Reactor Modeling of VFA Ketonization for Sustainable Aviation Fuel Production

Ketonization of volatile fatty acids (VFAs) produced via arrested methanogenesis of wet waste represents the only unit operation of wet waste upgrading to sustainable (net-zero or negative life cycle CO2 emissions) aviation fuel not currently in industrial practice. Ketone product yields of close to 100% have been obtained during gas-phase reactions of VFAs over oxide catalysts at the laboratory scale, but design of an industrial ketonization reactor requires understanding of the impacts of reactant and product partial pressures, deactivation, and heat and mass transport phenomena on observed ketonization rates. Our work leverages rigorous kinetic analysis of ketonization processes to inform reactor scale-up efforts through packed-bed reactor modeling. We first present results of a kinetic study of ketonization of a model VFA, hexanoic acid, over an industrial ZrO2 catalyst performed in a packed-bed microreactor in conditions free of significant heat or mass transfer gradients. Major findings of the analysis include: (i) hexanoic acid saturates all catalyst active sites at relatively low partial pressure (~10 kPa) and (ii) ketonization products 6-undecanone, water, and CO2 inhibit reaction rates. Kinetic data are used to fit a rate expression quantifying the functional dependence of ketonization rate on partial pressures of VFA reactants and ketone, water, and CO2 products. A packed-bed reactor model describing vapor-phase hexanoic acid ketonization over ZrO2 extrudate catalysts is developed based on the kinetic model. The effects of (i) bed- and pellet-scale mass- and heat-transfer limitations and (ii) axial pressure drop guide our development of recommendations for optimal sizing, temperature, and influent flow composition of an industrial-scale ketonization reactor. The quantitative understanding of VFA ketonization developed in this study represents an advance toward derisking the VFA ketonization step of wet waste upgrading to sustainable aviation fuel.

biojet↗

DOE OSTI2012

Experience in Remote Demolition of the Activated Biological Shielding of the Multi Purpose Research Reactor (MZFR) on the German Karlsruhe Site - 12208

In 2009, WAK Decommissioning and Waste Management GmbH (WAK) became owner and operator of the waste treatment facilities of Karlsruhe Institute of Technology (KIT) as well as of the prototype reactors, the Compact Sodium-Cooled Fast Reactor (KNK) and Multi-Purpose Reactor (MZFR), both being in an advanced stage of dismantling. Together with the dismantling and decontamination activities of the former WAK reprocessing facility since 1990, the envisaged demolishing of the R and D reactor FR2 and a hot cell facility, all governmentally funded nuclear decommissioning projects on the Karlsruhe site are concentrated under the WAK management. The small space typical of prototype research reactors represented a challenge also during the last phase of activated dismantling, dismantling of the activated biological shield of the MZFR. Successful demolition of the biological shield required detailed planning and extensive testing in the years before. In view of the limited space and the ambient dose rate that was too high for manual work, it was required to find a tool carrier system to take up and control various demolition and dismantling tools in a remote manner. The strategy formulated in the concept of dismantling the biological shield by means of a modified electro-hydraulic demolition excavator in an adaptable working scaffolding turned out to be feasible. The following boundary conditions were essential: - Remote exchange of the dismantling and removal tools in smallest space. - Positioning of various supply facilities on the working platform. - Avoiding of interfering edges. - Optimization of mass flow (removal of the dismantled mass from the working area). - Maintenance in the surroundings of the dismantling area (in the controlled area). - Testing and qualification of the facilities and training of the staff. Both the dismantling technique chosen and the proceeding selected proved to be successful. Using various designs of universal cutters developed on the basis of wall saws, both the activated steel liner and the inner reinforcing layer were cut remotely in one process. This allowed for the efficient execution of the following remote concrete removal steps using mining techniques. The electro-hydraulic demolition excavator that was purchased and then modified turned out to be an ideal tool carrier system with rapid-exchange coupling. Due to the high availability, no major delays occurred. This also was a result of the consistently implemented maintenance and repair concept. With the excavator installed in a modifiable scaffolding suspended from a rotating carrier ring, all dismantling areas could be reached and treated in spite of the small space. Thanks to an optimum organization of work-flows, routine change of dismantling work, and maintenance or repair, the iterative radiological measurement campaigns could be integrated in the whole activity without the dismantling work being disturbed significantly. The ventilation system with pressure grading and pre-filtration units ensured a low contamination level in the dismantling area. It was also possible to manage the dust formed by the milling of concrete surfaces. As it was possible to further cut metal parts and crushed concrete later on, residue flows were optimized. The planned overall period for testing, dismantling the bio-shield and removing the equipment was 36 months. The final duration was 39 months. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

A Frozen Soil Barrier to Control Groundwater Inflow into Damaged Reactors at the Fukushima Daiichi Nuclear Power Station - 16613

Reactor buildings and support facilities at the Fukushima Daiichi Nuclear Power Station (F1 Site) were damaged by a March 2011 tsunami. In response, the Tokyo Electric Power Company (TEPCO) is implementing a number of countermeasures to limit the releases and impacts of contaminated water to the surrounding environment. The diverse countermeasures work together in an integrated manner to provide different types, and several levels, of environmental protection. In general, the strategy represents an example of a 'defense in depth' concept that is used for nuclear facilities around the world. One of the key countermeasures is a frozen soil barrier encircling the damaged reactor facilities. The frozen barrier is intended to limit the flow of water into the area, reduce the amount of water entering damaged reactors, and reduce the resulting volume of contaminated water that requires treatment and storage. The frozen soil barrier was designed and installed by a team from TEPCO and Kajima Corporation. A group of scientists and engineers from US Department of Energy National Laboratories provided independent evaluation of the frozen barrier design and operational plans along with technical recommendations to the TEPCO team. The frozen soil barrier design extends to about 30 m depth; the bottom of the barrier is in a low permeability interval, and the total barrier length around the reactors is just over 1.5 km. The barrier required 1927 total boreholes: 1568 for freeze pipes plus 359 for temperature monitoring arrays. Drilling of all of the boreholes was completed November 9, 2015. Construction of the refrigeration plant is complete and all of the above ground piping/manifolds are in the final stages of construction. The DOE laboratory independent assessment of the frozen soil barrier concluded that the technical characteristics of a frozen barrier are relatively well suited to Fukushima-specific hydrogeologic conditions and the need for reducing the inflow of water into damaged reactors at the F1 Site. The scale of the Fukushima barrier is bounded by industry experience and the equipment and infrastructure proposed for the ground freezing is well understood. The on-site pilot test at Fukushima indicated predictable ground freezing and supported the full scale design parameters. These factors increase the confidence in the frozen soil barrier project underway at Fukushima. TEPCO is currently working with the Japanese Nuclear Regulatory Authority (NRA), providing operational plans/strategies for the frozen soil barrier and modeling results of the projected performance. Full scale frozen soil barrier operations are to begin after authorization from NRA. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Oxide Growth and Validation for Aluminum Research Reactor Fuel - 19422

Aluminum fuel cladding will undergo general corrosion with resulting formation of adherent oxide films during in-reactor and post-discharge exposures under various water conditions and temperatures. These oxides include chemically-bound water that can pose challenges for extended dry storage. This paper describes laboratory experiments performed to produce hydroxide (oxide) films on aluminum 1100, 6061 and 5052 series alloy substrates immersed in water at controlled temperatures. The results are compared to the oxide film present on a spent fuel assembly irradiated in research reactor. This work is part of a broad investigation to address knowledge gaps and technical data needs for dry storage of aluminum-clad spent nuclear fuel (SNF). To adequately assess the risks posed by chemically-bound water on SNF in dry storage and strategies for removing chemically bound water, it is necessary to understand the expected composition and loading of oxide films on SNF of different service histories and to investigate the response of such oxides to irradiation and potential drying methods. These efforts require both the growth of films under controlled conditions and the production of non-radioactive surrogates for testing. The current work comprises experimental investigation of oxide growth via (1) laboratory experiments to produce hydroxide films on aluminum alloy substrates immersed in water at controlled temperatures and (2) characterization and comparison of laboratory-grown oxides to oxides on service-exposed aluminum alloys. The important factors affecting oxide growth and characteristics typical of oxide films formed by aqueous corrosion under conditions similar to in-reactor are reviewed. Aluminum oxide films were grown under a variety of controlled temperature conditions and durations to serve as surrogates for oxides found on aluminum-clad spent nuclear fuel due to aqueous corrosion. The chemically-bound water found in oxides formed by aqueous corrosion poses challenges for extended dry storage, requiring investigation of how the oxide loading and characteristics vary with exposure history and to provide the initial condition for investigation of drying methods. Characteristics including chemical composition, oxide morphology, and film thickness were measured to assess whether the surrogate films were representative of in-service-produced oxides on SNF. Tests were conducted in liquid water at atmospheric pressure in the temperature range 50-100 deg. C, with and without heat flux through the corroding surface, to replicate in-service conditions, and at 185 deg. C inside pressure vessels to replicate pre-filming procedures. Three different alloys commonly used in aluminum clad research reactor fuel were considered: Al-1100, Al-6061, and Al-5052. Specimens were characterized by weighing before and after exposure to measure oxide mass loading, X-ray diffraction to identify oxide composition, and scanning electron microscopy to examine the oxide morphology and film thickness. The oxide type and the morphological characteristics of the film on a fuel element from the Uruguay RU1 research reactor were determined and compared to the laboratory surrogates. The results show some differences compared to the 50 deg. C laboratory specimen that are rationalized with the service storage history of the element. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗