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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Broadband integrated polarization splitter and rotator using subwavelength grating claddings

We present a broadband integrated photonic polarization splitter and rotator (PSR) using adiabatically tapered coupled waveguides with subwavelength grating (SWG) claddings. The PSR adiabatically rotates and splits the fundamental transverse-magnetic (TM 0 ) input to the fundamental transverse-electric (TE 0 ) mode in the coupler waveguide, while passing the TE 0 input through the same waveguide. The SWGs work as an anisotropic metamaterial and facilitate modal conversions, making the PSR efficient and broadband. We rigorously present our design approaches in each section and show the SWG effect by comparing with and without the SWG claddings. The coupling coefficients in each segment explicitly show a stronger coupling effect when the SWGs are included, confirmed by the coupled-mode theory simulations. The full numerical simulation shows that the SWG-PSR operates at 1500–1750 nm (≈250 nm) wavelengths with an extinction ratio larger than 20 dB, confirmed by the experiment for the 1490–1590 nm range. The insertion losses are below 1.3 dB. Since our PSR is designed based on adiabatical mode evolution, the proposed PSR is expected to be tolerant to fabrication variations and should be broadly applicable to polarization management in photonic integrated circuits.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Preliminary conceptual design of fast neutron spectrum nuclear thermal rocket cores using monolithic uranium nitride fuel

This paper presents a few nuclear thermal rocket (NTR) reactor core preliminary conceptual designs based on the use of monolithic uranium nitride (UN) fuel plates or pins clad with tungsten (W), alloyed or unalloyed as needed. High-assay low-enriched uranium (HALEU) as well as high-enriched uranium (HEU) are considered. Nominal core thermal powers are between 300 and 1100 MW corresponding to thrusts between 66 kN and 240 kN. The estimated thrust-to-reactor-weight ratios (T/WRx) achievable with the HEU plate and pin configurations are between 3.8 and 10.9. The T/WRx achievable with the HALEU plate and pin configurations are lower, 1.9–4.0, than those achievable with HEU. It is noteworthy that even a relatively modest increase in fuel enrichment above the HALEU limit enables a significant increase in T/WRx. It must be emphasized that these results do not yet account for uncertainties and that the appropriateness of the assumed fuel thermal design limits for normal operation (i.e., peak centerline temperature of 3100 K) will need to be confirmed. Finally, just like for the other UN-based fuels currently considered for NTR applications (i.e., CERMET and CERCER), the proposed UN fuel systems need significant testing in prototypic conditions to confirm the appropriate performance for safe NTR applications. However, the monolithic fuel systems may have, overall, a higher technology readiness level (TRL), and, consequently, shorter fuel development and validation time may result.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Demonstration of NEAMS Multiphysics Tools for Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 and FY20 expanded to additional fast reactor applications including lead cooled fast reactors (LFR) and sodium cooled fast reactors (SFR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, and (2) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR). First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study which began in FY19 and continued in FY20. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice. Along with the application on SFR and LFR, the capability of the tools has also been matured to deal with different reactor types and designs. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, in FY19 a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress for PROTEUS-SN and completed with MCNP. The MCNP power distributions were transferred to Nek5000 to perform thermal hydraulic calculations of the control and fuel assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Technical Considerations on MURR Control Blade Design Change and Testing using a New Metal Matrix Composite

The University of Missouri Research Reactor (MURR) is one of six research reactors, including a critical facility, that are pursuing conversion as part of a collaboration with the U.S. Department of Energy National Nuclear Security Administration Material Management and Minimization Office of Reactor Conversion and Uranium Supply, under the U.S. High Performance Research Reactors (USHPRR) conversion project. Five of the six USHPRR are planned to convert from highly enriched uranium (HEU) fuel using a low-enriched uranium (LEU) high assay monolithic alloy of uranium-10 wt% molybdenum (U-10Mo). As part of the conversion safety analysis, it is necessary to demonstrate the safety performance of the proposed core fueled with LEU as compared to the current HEU cores. The MURR reactor is planning to switch to a new control blade design that uses a metal matrix composite of boron carbide (B 4 C) and aluminum as the absorber in place of Boral®. Since MURR is expected to adopt the new metal matrix composite control blade design prior to conversion, the impact of the new blade design on the neutronics characteristics of the MURR cores for conversion are analyzed in this work through updates to incorporate the changes to the blade design in conversion models as they directly impact the LEU conversion safety analysis. The quantitative comparison shows that the neutronics and thermal hydraulic behavior of one metal matrix composite blade replacing a Boral blade is comparable for the two example MURR LEU and HEU cores states considered. Geometrical changes in the metal matrix composite blade design, combined with a 4% increase in areal boron density, showed local heating effects up to 20% higher than the Boral design. As expected, the metal matrix composite showed slightly lower heat depositions and absorber region temperatures for the LEU cases compared to HEU. Although this analysis was comparative for a single blade, maximum control blade temperatures for both Boral, metal matrix composite, and HEU/LEU remained below 100 °C, though additional analysis at a core level could differ. A qualitative irradiation behavior assessment concludes that the mechanisms that may drive swelling and blistering in the current Boral design are eased by the adoption of the metal matrix composite design. The work concludes that the two blade designs are essentially equivalent with regards to neutronics, thermal hydraulics, and expected material behavior under irradiation. However, due to the geometrical changes to the blades including redesigned and thinner cladding, new testing and increased surveillance for distortion and swelling are recommended to confirm the performance of the metal matrix composite control blade design. Where testing is completed prior to conversion, the only anticipated impacts on conversion to LEU U-10Mo fuel would be the need for models and safety analysis incorporating the metal matrix composite control blades.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Pool Boiling CHF Experiment for Cr-alloy-coated Accident Tolerant Fuel Cladding

The severe accident at the Fukushima nuclear power plants forced the nuclear communities to revise and reinforce their safety technology, especially accident tolerance. The main problem of Zr-based alloys is hydrogen generation owing to steam oxidation in a very high temperature environment. Currently, several types of Accident Tolerant Fuel (ATF) claddings that eliminate the steam oxidation problem have been presented as alternatives. Although the ATFs have shown appropriate performances in terms of material engineering, thorough tests including thermal hydraulic concerns are still required to ensure the design compatibilities. In this study, we conducted pool boiling critical heat flux (CHF) experiments using Cr-alloy-coated ATF claddings to verify basic thermal hydraulic performances. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2021

Flow boiling transient critical heat flux tests with stainless steel and FeCrAl: Transient correlation implementation, model calibration, and sensitivity analysis

In this study, steady-state and transient internal-flow critical heat flux (CHF) experiments were carried out under two flow conditions, at atmospheric pressure, on 316 stainless steel (316-SS) and iron-chromium-aluminum (FeCrAl) tubes. Slow power transients at a low mass flow (300 kg/m 2 - s) with very low subcooling (X e =-0.0054) generated premature CHFs, which were prevented by faster power transients and high mass flow (1000 kg/m 2 - s). The measured transient CHFs increased linearly with increasing power transients, compared with the steady-state CHF. Yet, the wall superheat at the CHFs decreased with faster power transients. Transient CHF correlations highlighting heterogeneous spontaneous nucleation were calibrated to the measured CHFs and compared with other existing correlations. Transient CHF multipliers were acquired from pool and flow boiling empirical CHF correlations that were generated. The multipliers were applied to the RELAP5-3D nuclear system code to analyze the discrepancies between the measured data and the predicted CHF and post-CHF behavior, which improved peak cladding temperature predictions by 24.6%. A variance-based global sensitivity study perturbing the experimental uncertainties and cladding material thermal properties showed the diminishing influence of flow boiling heat transfer with increasing power transients, highlighting the significance of the volumetric heat capacity for the cladding integrity during transients. Transient CHF correlations were applied to the most limiting design basis accident: a hot-zero-power reactivity-initiated accident on a generic pressurized water reactor RELAP5-3D model.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Phased Array Ultrasonic Examination of Reactor Coolant System (Carbon Steel-to-CASS) Dissimilar Metal Weld Mockup Specimen

In the summer of 2009, Pacific Northwest National Laboratory (PNNL) staff traveled to the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina, to conduct phased-array ultrasonic testing on a large bore, reactor coolant pump nozzle-to-safe-end mockup. This mockup was fabricated by FlawTech, Inc. and the configuration originated from the Port St. Lucie nuclear power plant. These plants are Combustion Engineering-designed reactors. This mockup consists of a carbon steel elbow with stainless steel cladding joined to a cast austenitic stainless steel (CASS) safe-end with a dissimilar metal weld and is owned by Florida Power & Light. The objective of this study, and the data acquisition exercise held at the EPRI NDE Center, were focused on evaluating the capabilities of advanced, low-frequency phased-array ultrasonic testing (PA-UT) examination techniques for detection and characterization of implanted circumferential flaws and machined reflectors in a thick-section CASS dissimilar metal weld component. This work was limited to PA-UT assessments using 500 kHz and 800 kHz probes on circumferential flaws only, and evaluated detection and characterization of these flaws and machined reflectors from the CASS safe-end side only. All data were obtained using spatially encoded, manual scanning techniques. The effects of such factors as line-scan versus raster-scan examination approaches were evaluated, and PA-UT detection and characterization performance as a function of inspection frequency/wavelength, were also assessed. A comparative assessment of the data is provided, using length-sizing root-mean-square-error and position/localization results (flaw start/stop information) as the key criteria for flaw characterization performance. In addition, flaw signal-to-noise ratio was identified as the key criterion for detection performance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Steam Oxidation, Burst and Critical Heat Flux Testing of FeCrAl Cladding in the Severe Accident Test Station

Since August 2011, Oak Ridge National Laboratory (ORNL) has been evaluating the steam oxidation resistance of a wide range of materials to identify a potential accident tolerant fuel (ATF) cladding [1-14]. The primary goal of an ATF cladding is to enhance safety margins in light water reactors (LWR) during beyond design basis accident scenarios (such as the nuclear accident at Fukushima Daiichi in March 2011) without requiring changes to the reactor design or fuel enrichment, or significantly affecting reactor economics during normal operation.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Water-based Ria Testing in TREAT: Commissioning and Early Results

We have completed a series of reactivity-initiated accident commissioning tests with the static water capsule in the Transient Reactor Test Facility and completed some visual and non-destructive post-irradiation examinations on the fuel rods. The test campaign included a calibration test followed by five tests in the Static Environment Rodlet Transient Test Apparatus capsule. The conditions varied from room temperature and pressure up to 200°C and 2.5 MPa, with energy depositions varying between ~500–1100 J/gUO2. The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Overview of the Water-Based RIA Testing in TREAT

A series of reactivity-initiated accident (RIA) commissioning tests have been completed with the static water capsule in the Transient Reactor Test (TREAT) Facility. The test campaign included a calibration test followed by five tests in the Static Environment Rodlet Transient Test Apparatus (SERTTA) capsule. The conditions varied from room temperature and pressure up to 200°C and 2.5 MPa with energy depositions varying between ~500-1100 J/gUO2. The series of tests allowed for a number of instrumentation qualifications and demonstrations including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. This paper documents the design of the capsule and highlight the results from the commissioning tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Overview of the Water-Based RIA Testing in TREAT

We have completed a series of reactivity-initiated accident commissioning tests with the static water capsule in the Transient Reactor Test Facility. The test campaign included a calibration test followed by five tests in the Static Environment Rodlet Transient Test Apparatus capsule. The conditions varied from room temperature and pressure up to 200°C and 2.5 MPa, with energy depositions varying between ~500–1100 J/gUO2. The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. This paper documents the design of the capsule and highlights the results from the commissioning tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A progress update on selected AREVA NP advanced BWR methodologies

This article presents an overview of select advances in AREVA NP methods concerning 3 topics: Analytical fuel rod model, Transient and accident simulation, and ATWSi post dryout treatment. Concerning the first topic, the analytical fuel rod model relies heavily on first-principles closed-form solutions instead of ever finer nodalization and higher dimensions of standard numerical methods such as finite differences or finite elements. This analytical fuel rod model takes full account of the complications that made numerical analysis necessary, e.g. variation of pellet thermal conductivity with temperature and burnup, and the radial dependence of fission power deposition in pellets. The model also calculates thermo-mechanical parameters such as clad stress and strain, fission gas release, pellet sintering and swelling, and pellet-clad gap size and thermal resistance. The model is suited for innovative fuel designs such as doped pellets. Concerning the second topic, the BWR evaluation model, AURORA-B, is an advanced transient and accident simulator including SRELAP5, a two fluid thermal hydraulic code comprehensively reviewed by international safety authorities for BWR and PWR applications; RODEX4, a best estimate fuel performance code used to evaluate the local thermal-mechanical behavior of fuel rods during postulated transients and accidents; and MB2-K, a neutron kinetics extension of the steady state core simulator MICROBURN-B2. The analytical domains will include transients as well as control rod drop and loss of coolant accidents. Extension of RODEX4 have been made to accommodate analysis of chromium-doped fuel. Concerning the third issue,The new AREVA post-dryout models can accurately predict dryout/re-wet followed by an ultimate failure to re-wet under oscillatory ATWSi (Anticipated Transient Without SCRAM with core instability) conditions. These models, while not a boiling curve based method, can qualitatively reproduce the characteristics of the boiling curve.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

limiting rods selection in large break LOCA analysis with the RELAP5-3D code

The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10 CFR 50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently more fuel rods need to be analyzed under LOCA conditions to maintain the safety margin, compared to the current practice in which only one hot rod needs to be analyzed. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of the limiting rods under LOCA conditions. In this work, the best-estimate plus uncertainty analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. With the new analysis method, the limiting transient case and the limiting rods can be easily identified in response to the proposed new rulemaking. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Processing of Nanostructured Ferritic Alloys for Advanced Reactor Core Materials

The Fuel Cycle Research and Development Program is investigating methods of burning minor actinides in a transmutation fuel. To achieve this goal, the fast reactor core materials (cladding and duct) must be able to withstand very high doses (>200 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). One of the most promising materials for this demanding applications are the class of nanostructured ferritic alloys. These materials are ferritic alloys with high chrome (9-14Cr) with a high density (>10{sup 23}/m{sup 3}), fine distribution of nanosized (∼2 nm) oxide particles and a fine grain size (<0.5 micron). This fine microstructure provides an alloy with high strength at high temperatures and excellent radiation tolerance (e.g. reduced void swelling and ductility retention at low temperatures) but also increases the difficulty of producing engineering parts (e.g. thin walled tubes) from these advanced materials. Thus, extensive research is underway to produce tubes using techniques such as high temperature hydrostatic extrusion, intermediate temperature plug drawing and low temperature pilger processing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Transient Mechanical Test Development for Accident Tolerant Cladding Candidates to Simulate PCMI-like Conditions

This study investigates the transient mechanical response of accident tolerant fuel (ATF) cladding candidates to provide information on potential PCMI-induced deformation and failure of candidate cladding materials. The mechanical testing is applied t to Gen-I FeCrAl and triaxially-braided SiC fiber SiC matrix composite materials under loading conditions like the PCMI during a postulated design basis reactivity insertion accident. To achieve the specific loading conditions, a modified burst test instrument with accurate mechanical pulse control has been designed and constructed. The mechanical response of FeCrAl, SiC-SiC composite tubes were compared with the pre-hydrided zirconium alloy cladding. The specimens subjected to mechanical pulses were investigated in a strain-controlled manner deformation.

Cinbiz, Nedim↗
DOE OSTI2015

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status of INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

The Feasibility and the Benefits of the Advanced Nuclear Fuel Pellet Designs with Radially Varying Fuel Zoning and Burnable Poison Concentration (Final Summary Report)

As part of the work supported by the US Department of Energy (DOE) Office of Nuclear Energy (NE) Gateway for Accelerated Innovation in Nuclear (GAIN) FY 2021 Voucher, Exelon Generation, now Constellation, and Oak Ridge National Laboratory (ORNL) entered into a cooperative research and development agreement (CRADA) to evaluate and assess the feasibility and impact of various conceptual advanced nuclear fuel pellet designs (ANFPDs). The objective of this project was to perform modeling and simulation and analyses using the advanced modeling and simulation capabilities of VERA/BISON, developed by DOE, to determine the viability and benefits of numerous advanced nuclear fuel pellet design concepts in terms of fuel cycle costs, operational safety, and margin improvement. Whereas the detailed coupled neutronic and thermal hydraulic analyses performed using VERA provided in-depth knowledge in terms of fuel cycle performance, the detailed VERA results were used in subsequent fuel performance analyses using BISON. These subsequent analyses focused on several key fuel performance criteria, such as peak fuel centerline temperature (FCT), fission gas release (FGR), gap closure, plenum pressure, and cladding hoop stress. These key fuel performance criteria were analyzed for some of the conceptual fuel designs and were compared with the results obtained for UO 2 fuel. The results provided detailed information to enable better understanding of the performance of the fuel types analyzed. Understanding the advantage of loading these conceptual fuel designs into the core is important not only to Constellation but also to the entire light-water reactor (LWR) fleet in the United States.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A simple structure of all circular-air-holes photonic crystal fiber for achieving high birefringence and low confinement loss

We propose a simple structure of photonic crystal fibers (PCFs) with high birefringence and low confinement loss based on one rectangular centric ring of smaller circular air holes (CAHs) in the fiber core, and three rings of larger CAHs in the fiber cladding. This simple geometry (using all CAHs with two different air hole sizes) is capable of achieving a flexible control of the birefringence, B = 5.501 × 10{sup −3}, and ultra-low confinement loss, 7.30 × 10{sup −5 }dB/km, at an excitation wavelength of λ = 1550 nm. The birefringence value is ∼5.0 times greater than that obtained for conventional CAH PCF. This simple structure has the added advantage from the view point of easy fabrication, robustness, and cost. A full-vector finite element method combined with anisotropic perfectly matched layers was used to analyze the various fiber structures. We have analyzed four cases of CAH PCFs, focusing on the core asymmetry design as opposed to the conventional approach of CAHs or elliptical air holes on the cladding and core. The robustness against manufacturing inaccuracies of the proposed structure has also been further investigated in this work.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗