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These are the slides accompanying the summary "ATF-2C Physics Safety and Scoping Analysis" in the Transactions of the American Nuclear Society.
This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171A ATF which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains fluence rate values corresponding to the particular elevations (relative to the 80-ft. core elevation) where the measurements were taken. The data in this report consists of (1) a table of the ATR power history and distribution, (2) a listing of the calculated thermal and fast neutron fluence rates, and (3) plots of calculated both the thermal and fast neutron fluence rates. The fluence rates reported are for the average power levels given in the table of power history and distribution. During the irradiation of Cycle 152A, the RML performed additional flux measurements to assess core reconfiguration issues. Measurements using cadmium covered as well as bare neutron monitors in the Southwest, Southeast, and H positions of the Center lobe were performed to determine the division in neutron energy between epithermal and thermal neutrons. This is defined as the resonance correction defined in this report. It is used to quantify both thermal and fast neutron flux. For more information please refer to letter, “RADIATION MEASUREMENTS LABORATORY MEASUREMENTS OF IN CORE ATR PHYSICS TESTING DURING CYCLE 152A” dated January 24, 2013, to Casey Stengel from C. C. Jensen. The measured values documented in RML Procedure, “ACMM-3600, Flux Monitoring,” will be used indefinitely until further measurements supersede the current values.
The following report documents the conceptual design for the ATF-2 Ramp experiment. The experiment aims to fill the in-pile irradiation testing gap to conduct integral ramp testing, which was created by the closure of R2, Osiris, and Halden test reactors. The concept involves the simultaneous ramping of three fuel pins using a power axial locator mechanism in the Loop-2A testing facility in the center flux trap of the advanced test reactor (ATR). The three fuel pins are ramped in individual coolant channels containing a prototypic PWR environment. Rod failure will be detected using a fuel rod elongation sensor attached to the fuel pin upper end cap. An axial stack of concentric hafnium and zirconium shrouds will be used to shape the flux around the test pins to create different power levels in each pin and to ensure the peak power location of each pin remains in the center of the test rod. Monte Carlo simulations are used to demonstrate the viability of this design concept.
Flow boiling heat transfer tests were conducted to evaluate the Critical Heat Flux (CHF) and Heat Transfer Coefficient (HTC) of conventional and accident tolerant fuel (ATF) cladding materials, i.e., bare Zircaloy-4 (Zr4), and Zircaloy-4 coated with Chromium using physical vapor deposition (PVD) (Zr4-Cr-PVD) and cold spray process (Zr4-Cr-CS), respectively. The tests were performed on a single heater rod with a uniform heat flux profile, at the atmospheric pressure, inlet temperature (24 °C) and mass flow rate (750 kg/m 2 s). HTC’s were increased by 5.2% and 3.3% on Zr4-Cr-CS compared to bare Zr4 and Zr4-Cr-PVD. Improved HTCwas attributed to the increased void fraction and high roughness (Ra = 532 nm) of the Zr4-Cr-CS. In detail, the micro-cavities, which act as bubble seeds, were entirely distributed on the Zr4-Cr-CS, and these micro-cavities generated smaller and faster bubbles, thus the void fraction increased by 10.9% compared to bare Zr4. HTC is improved by these smaller and faster bubbles which can remove heat from the heater rod surface more efficiently. To identify the exact location of CHF, the surface temperature profile was measured using advanced fiber-optic sensors which have high temporal/spatial resolution (distance between point to point: ~2.5 mm, frequency ~100 Hz). CHF occurred at 80–95% along the heated length and showed an 11.6% reduction on Zr4-Cr-CS compared to bare Zr4. We hypothesize the CHF reduction was caused by the lower wettability of the Zr4-Cr-CS which reduces the liquid supply to the surface, and vigorous bubble accumulation owing to high void fraction near the outlet, which results in early formation of the vapor film. During the post-CHF quenching phase, the rougher Zr4-Cr-CS showed larger cooling rates compared to the bare Zr4 and Zr4-Cr-PVD, preventing of oxidation by chromium layer. In conclusion, we demonstrate that the high roughness on the Zr4-Cr-CS increased the number of micro-cavities on the surface, resulting in 5.2% increase in HTC and improvement in the quenching heat transfer performance whereas CHF was 11.6% reduced compared to bare Zr4. Overall statistically there is little effect in the coating with regards to CHF and slight improvement in HTC.
Chromium-coated zirconium alloys are one of the promising candidates for accident-tolerant fuel cladding (ATF) tubes for light water reactors (LWRs). In this study, the high temperature oxidation and degradation of two types of Cr coatings (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on prototype rod segment samples filled with ZrO 2 pellets and tightly sealed with welded end caps. Isothermal tests at 1100 and 1200 °C were terminated by quenching with water; transient tests were performed up to 1500–1600 °C until complete coating failure. The positive effect of both types of Cr coatings was observed in all tests. Pre-damaged specimens showed only locally increased oxidation of the scratched zone, but no negative effects on the adjacent Cr coating. Finally, the behavior of the two types of coatings is compared and the degradation mechanisms are discussed based on hydrogen release data, extensive metallographic post-test examinations, and the current state of global research.
Over the years, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) (2015-2018, 2020) and Consortium for Advanced Simulation of Light Water Reactors (CASL) (2019) programs have provided support for development of Accident Tolerant Fuel (ATF) material models in the BISON fuel performance code. Since the beginning, the goal has been to utilize a multiscale modeling approach to gain a physical understanding of the fuel concepts of interest and to develop mechanistic models in the absence of a large amount of experimental data. This work builds upon that of previous years. In particular we present newly updated fission gas release models for both gas behavior in Cr 2 O 3 -doped UO 2 and U 3 Si 2 fuels, and a new creep model for U 3 Si 2 . The validation exercises completed last year are revisited with the latest models and the results updated. A brief summary of recent modeling activities for FeCrAl cladding is also provided.
A detailed study of the effects of gravitational radiation on the relative positions of objects was carried out with the hope that astrometric detection of gravitational radiation might be possible. The results are discouraging. It would appear that narrow-field instruments in the Astrometric Telescope Facility (ATF) class are still several orders of magnitude less accurate than would be required for this very delicate kind of measurement. However, the situation changes considerably when wide-field instruments are considered. An instrument such as POINTS ought to be able to detect gravitational radiation at this level with ease. The list of bright quasars has been augmented. Of particular interest is the finding of quasars near two open clusters and one planetary nebula. These would be useful in determining absolute parallaxes of these objects.
Test process, milestones and inputs are unknowns to first-time users of the ATF. The User Test Planning Guide aids in establishing expectations for both NASA and non-NASA facility customers. The potential audience for this guide includes both internal and commercial spaceflight hardware/software developers. It is intended to assist their test engineering personnel in test planning and execution. Material covered includes a roadmap of the test process, roles and responsibilities of facility and user, major milestones, facility capabilities, and inputs required by the facility. Samples of deliverables, test article interfaces, and inputs necessary to define test scope, cost, and schedule are included as an appendix to the guide.
Presentation on the RIA Testing Program for ATFs at TREAT for the 11th Annual EPRI/DOE/INL Joint Workshop on Accident Tolerant Fuel
These are the slides accompanying the summary "Optimizing the ATF-2Ramp Power Profile" (PRS/CON-24-04674).
This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171B which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order.
Abstract not provided.
Irradiation experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor are typically assumed to have little effect on one another. This assumption does not hold true for certain experiments in close proximity. To evaluate the impacts on safety and programmatic parameters of experiments in the center flux trap, the contents of the adjacent H and inner-A positions were modeled with a range of possible irradiation targets. First, neutron flux maps with experiments in those positions were compared against a baseline configuration. Next, several safety and programmatic parameters for a generic accident-tolerant fuel test train were calculated. It was shown that these parameters can exhibit considerable sensitivity to the contents of the H and A experiment positions.
Racks in the New Fuel Vault (NFV) and Spent Fuel Pool (SFP) and accompanying fuel handling procedures are currently designed with a maximum enrichment of 5.0 wt. %. Publicly available source documents were used to develop baseline models of fresh fuel storage configurations.
Boiling water reactors can experience anticipated operational occurrences, where an expected deviation from normal operating conditions is witnessed for short time scales. One example of such transients arises from cladding dryout, where stable boiling becomes unstable, resulting in cladding heat up before re-wetting occurs. To understand the impact of such transients on fuel cladding lifetime and performance, cyclic dryout conditions were imposed upon fresh fuel cladding in the absence of irradiation. Pressurized cladding tube segments were exposed to rapid temperature increases to intermediate temperatures below a cladding rupture threshold and deformation was monitored in situ via 3-D digital image correlation. The impact of cycle frequency on cladding deformation was assessed. A clear understanding of elastic and plastic strain responses were developed for Zr alloy deformation under temperature transients, and these results can be used to support material model development and validate fuel performance codes.
Presentation for a meeting on accident tolerant fuels. Brings together information from already cleared reports.
This report documents post-irradiation examination (PIE) activities performed in FY 2025 at Oak Ridge National Laboratory on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~40 GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide thermometry was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target.