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At least 613 records · Page 34
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DOE OSTI2023

A Fast Reactor Irradiation Experiment Design in the ATR

Modern modeling techniques were used to investigate a proposed method for fast neutron irradiations in an existing thermal-spectrum reactor, the Advanced Test Reactor (ATR). This method builds upon pre-existing ideas, where fast flux is increased by surrounding the specimens with fissionable “booster fuel” but diverges from historical approaches by using an already developed fuel element design used in the Belgian Reactor 2 (BR2) as the booster fuel while leveraging modern 3-D modeling and simulation techniques. Design evaluations and neutronics simulations were performed to evaluate the performance of a BR2 fuel element irradiated in an ATR flux trap with test pins in the central channel of the BR2 fuel element. These efforts have yielded promising results. Adding a BR2 fuel element in the northeast (NE) flux trap of ATR was predicted to result in a 150% increase to the incident fast neutron flux with a fast (>0.1 MeV) to thermal (<0.625 eV) neutron flux ratio ranging from approximately 50 to 150, dependent on the material used for thermal neutron filtering and volume of moderator within the central channel of the BR2 element. The predicted annual fast neutron fluence (>0.1 MeV) ranges from 7.9 × 1021 to 9.1 × 1021 n/cm2. Given the relatively large fast to thermal neutron flux ratio, the calculated radial power profiles within 4.3 mm outer diameter U10Zr fueled specimens irradiated within the BR2 booster fuel element are adequate representations of those within fast neutron reactors. The predicted radial power profiles are not flat, but they are more prototypic than those seen in advanced fuels tests which began in ATR in 2003. Another distinct advantage of this experiment design is that full-scale test pins can be irradiated to augment the ongoing series of reduced scale advanced fuels tests. The proposed experiment design irradiated within a BR2 fuel element in a flux trap of ATR offers an improved alternative to the current testing of advanced reactor fuels in ATR. Selected results from this design evaluation are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and the process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Measured Thermal and Fast Neutron Fluence Rates for ATF-1 Holders During ATR Cycle 157D

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 157D 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 measurements of the fluence rates corresponding to the particular elevations relative to the 80-ft. core elevation. The data in this report consist of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, and (3) plots of 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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

Destructive post-irradiation examination was performed on AGR-1 fuel Compact 4-1-1, which was irradiated to a final compact-average burnup of 19.4% FIMA (fissions per initial metal atom) and a time-average, volume-average temperature of 1072°C. The analysis of this compact focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy, electron microscopy, and elemental analysis. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed TRISO or failed SiC layers (as indicated by very low uranium inventory in all of the leach solutions). The total fractions of the predicted compact inventories of fission products Ce-144, Cs-134, Cs-137, and Sr-90 that were present in the compact outside of the SiC layers were <2×10 -6 , based on DLBL data. The Ag-110m fraction in the compact outside the SiC layers was 3.3×10 -2 , indicating appreciable release of silver through the intact coatings and subsequent retention in the OPyC layers or matrix. The Eu-154 fraction was 2.4×10 -4 , which is equivalent to the inventory in one average particle, and indicates a small but measurable level of release from the intact coatings. Gamma counting of 61 individual particles indicated no particles with anomalously low fission product retention. The average ratio of measured inventory to calculated inventory was close to a value of 1.0 for several fission product isotopes (Ce-144, Cs-134, and Cs-137), indicating good retention and reasonably good agreement with the predicted inventories. Measured-to-calculated (M/C) activity ratios for fission products Eu-154, Eu-155, Ru-106, Sb-125, and Zr-95 were significantly less than 1.0. However, as no significant release of these fission products from compacts was noted during previous analysis of the AGR-1 capsule components, the low M/C ratios are most likely an indication of a bias in the inventories predicted by physics simulations of the AGR-1 experiment. The distribution of Ag-110m M/C ratios was centered on a value of 1.02 and was fairly broad (standard deviation of 0.18, with values as high as 1.42 and as low as 0.68). Based on all data gathered to date, it is believed that silver retention in the particles was on average relatively high, but that the broad distribution in values among the particles represents significant variation in the inventory of Ag-110m generated in the particles. Ceramographic analysis of particle cross-sections revealed many of the characteristic microstructures often observed in irradiated AGR-1 particles from other fuel compacts. Palladium-rich fission product clusters were observed in the IPyC and SiC layers near the IPyC-SiC interface of three Compact 4-1-1 particle cross-sections. In spite of the presence of fission product clusters in the SiC layer, no significant corrosion or degradation of the layer was observed in any of the particles examined.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

AGR-1, AGR-2 and AGR-3/4 Dimensional Change Data Analysis

A series of Advanced Gas Reactor (AGR) experiments have been completed in the Advanced Test Reactor at Idaho National Laboratory in support of qualification and development of tristructural isotropic fuel. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder. These capsules are instrumented with thermocouples embedded in the graphite, enabling temperature control. The fuel compacts are composed of fuel particles surrounded by a graphitic A3 matrix material. Dimensional change in AGR fuel compacts is vital because the swelling or shrinkage affects the size of the gas gaps that are used to control temperatures. Analysis of dimensional change in the AGR fuel compacts is needed to establish the variables directly relating to compact shrinkage. The variables initially identified for consideration were matrix density, compact density, fuel packing fraction, uranium loading, fuel particle diameter, cumulative fast neutron fluence, and volume average time average fuel temperature. In addition to the data available from the AGR experiments, the analysis included specimens formed from the same A3 matrix material used in Advanced Graphite Creep (AGC) experiments, which provide graphite creep data during irradiation for design and licensing purposes. The primary purpose of including the AGC specimens was to encompass dimensional behavior at zero packing fraction, zero uranium loading, and zero particle diameter. All possible combinations of first-order variable regressions were considered in the analysis. The study focused on identifying the best regression models for percent change in diameter, length, and volume. Bootstrap analysis was used to ensure the resulting regression models were robust and well-performing. The variables identified as very significant in predicting change in one or more dimensions (diameter, length, and volume) are volume average time average temperature, fast fluence, compact density, packing fraction and fuel particle diameter. Due to the presence of confounding effects between several variables, interpretation of these results is equivocal; the use of multiple statistical tests provides additional confidence in the conclusion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Progress on Pu-238 Production at Idaho National Laboratory From March 2021 to February 2022

Idaho National Laboratory has completed irradiation of Np-237 targets in the Advanced Test Reactor’s (ATR) South Flux Trap (SFT) and I-7 positions. INL also progressed qualification of new ATR Gen 1 Np-237 targets for the North East Flux Trap (NEFT), inner A, and H positions. This paper gives an overview of operational and technical activities from March 2021 to February 2022.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Progress on Pu-238 Production at INL From March 2021 to February 2022

Idaho National Laboratory has completed irradiation of Np-237 targets in the Advanced Test Reactor’s (ATR) South Flux Trap (SFT) and I-7 positions. INL also progressed qualification of new ATR Gen 1 Np-237 targets for the North East Flux Trap (NEFT), inner A, and H positions. This slide show is based upon artifact INL/CON-22-65729-Rev000 and gives an overview of operational and technical activities from March 2021 to February 2022.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

Here, the PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination measurements provided data on release of these fission products from fuel compacts and fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Moreover, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the PARFUME predictions and the modeled diffusivity of strontium in UCO.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Analysis of Neutron Fluence Monitors for The USU ATR NSUF Project

The USU (Utah State University) Advanced Test Reactor Nation Science User Facility (ATR NSUF) experiments to test the performance of hafnium-aluminum composites were conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) during cycles 149A through 151A from April 14, 2011, to February 13, 2012. Four sets of two neutron fluence monitor capsules wee irradiated in four different combinations of reactor cycles. The assemblies were irradiated in B2 position of the ATR. Neutron fluence monitors were located at 8 positions in the assemblies. These monitors were previously fabricated at Pacific Northwest National Laboratory (PNNL). After irradiation, the monitors were sent to PNNL for analysis. The measured activation rates were used to adjust the neutron spectra at each location in the assemblies to determine the best fit to the neutron fluence spectra. Radiation damage (dpa) calculations were performed at all locations using the adjusted neutron spectra. The report was revised after receipt of corrected irradiation history for Cycle 149A. A plot of the fast neutron flux as a function of height in the B2 position was also added to the report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171A

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171A SR which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171B.SR

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171B SR which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171A IR TMIST 3B

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171A 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171A MP-2 (Rev. 1)

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171A MP-2 which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171A MP-2

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171A MP-2 which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Measured thermal and fast neutron fluence rates for ATF-1 holders during ATR cycle 164B. Revision 1

This report contains the thermal (2200 m/s) and fast (E> lMeV) neutron fluence rate data for the ATF-1 holders located in core for ATR Cycle 162B 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 measurements of the fluence rates corresponding to the particular elevations relative to the 80-ft. core elevation. The data in this report consist of ( 1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, and (3) plots of both the thermal and fast neutron fluence rates.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Measured Thermal and Fast Neutron Fluence Rates ATR Cycles 171A AFC (Rev. 1)

This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 171A AFC 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗