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DOE OSTI2017

Evaluation of Wilks’ One-sided Non-parametric Formula Against Analytical Parametric Methods

This paper will be presented to the "Sensitivity and Uncertainty Methods" subgroup of the "Mathematics and Computation Division" of the ANS Winter Symposium. Direct comparisons are made via simulation between the Wilks' non-parametric and Gaussian (parametric) methods which are applied in the SASQUATCH and SINDA-SAMPLE thermal-hydraulic codes, respectively, for the Advanced Test Reactor safety basis. The level of relative conservatism is quantified for the Wilks method and shown to be greater than the parametric method for an equivalent probability and confidence level.

95/95↗

DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Operational Experience and Development of a Reliability Model for the ATR Demineralizer System (Slides)

The Advanced Test Reactor (ATR) is a light water reactor with aluminum clad driver fuel. Strict limits are implemented on pH, conductivity, and filterable solids to assure the performance of the driver fuel limit corrosion of the primary coolant system (PCS) pressure boundary components. A bypass demineralizer system is used to maintain PCS coolant within allowable bands for each of the aforementioned variables. Unlike many other reactors, the ATR does not have a filtration system. As such, failure of ion exchange resin in the bypass demineralizer can defeat the purpose of the system and result in high filterable solids in the PCS. Power operations with high filterable solids is prohibited by a technical safety requirement. This paper discusses recent failures of the ATR bypass demineralizer system. The testing and analysis done to deduce the cause of the failures is covered, as well as short term mitigative steps that were taken. To prevent the likelihood of future issues, historical data from the ATR and industry is used to develop a reliability model of the ATR bypass demineralizer system. Evidence of radiolytic decomposition of ion exchange resin was found in the cause investigation. As such, special attention is given to ion exchange resin failure due to high radiation environments. The reliability model is used to develop the ATR testing and maintenance program. Specifically, limits on resin service life are discussed.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2019

Core-to-Specimen Energy Coupling Results of the First Modern Fueled Experiments in TREAT

For this study, a series of experiments was conducted in the Transient Reactor Test (TREAT) Facility following a decades-long operational hiatus. A series of physics tests were then performed emphasizing relevant transients for Light Water Reactor (LWR) fuels research. A series of fueled experiments were then conducted using LWR-type sub-length specimens in inert gas capsules with the primary objective to quantify the ratio of nuclear heat deposited in the core to that deposited in the specimens. This type of data can be difficult to measure but is highly important in the design and interpretation of transient tests. Five fueled capsules were irradiated and energy coupling factors (which express the ratio of nuclear heat deposited per mass in the test specimen to that deposited in the reactor core in units of J/gUO 2 MJ TREAT ) were observed ranging from 2.0 to 2.2 over a wide range of transient energies with good agreement between measurement methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Recent 238 Pu Production Activities at Idaho National Laboratory

The Plutonium-238 ( 238 Pu) Production program at Idaho National Laboratory (INL) is actively qualifying irradiation targets containing 237 Np for the Advanced Test Reactor (ATR) to produce 238 Pu for future National Aeronautics and Space Administration missions. INL qualified and loaded seven targets in the ATR’s south flux trap for cycle 169A, which occurred in Spring 2021. The irradiation qualification program has expanded to additional ATR irradiation positions after two baseline production targets in three positions validated significant production of 238 Pu. The validation model was followed by the PFS-1 experimental test in the ATR Critical Facility that verified 238 Pu production cross sections. This paper outlines the progress and status of the 238 Pu production program at INL. The qualification effort, safety analysis, hardware status, and future activities for qualification of an updated target design for use in the ATR are discussed.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2019

Architecture Improvements and Retesting of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

Nuclear fuel experiments conducted at the Transient Reactor Test Facility (TREAT) will be analyzed with the MAMMOTH reactor physics application, currently under development at Idaho National Laboratory. MAMMOTH natively couples with the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. The implemented methods allow for simulating the irradiation of fuel in a nuclear reactor from the beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing using the same finite element mesh throughout the analysis retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. This report focuses on three significant improvements to the MAMMOTH micro-depletion capability. First, the depletion architecture within MAMMOTH is modified to improve the flexibility and speed of burnup calculations. The results of various benchmark tests demonstrate that the refactoring of the previously implemented capability does not introduce any regression. The second improvement involves the expansion of the data functionality. MAMMOTH can now read the decay data files used by the Oak Ridge Isotope Generation (ORIGEN) code, thus enabling access to data of more than 1,600 isotopes. The third improvement is the capability to perform depletion at a constant power level which was verified via a comparison to ORIGEN.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

Implementation of Depletion Architecture in the MAMMOTH Reactor Physics Application

Nuclear fuel experiments conducted at the Transient Reactor Test Facility (TREAT) will be analyzed with the MAMMOTH reactor physics application, currently under development at Idaho National Laboratory. MAMMOTH natively couples with the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. The methods currently implemented allow for simulating the irradiation of fuel from the beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing using the same finite element mesh throughout the analysis thereby retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. This work focuses on two significant improvements to the MAMMOTH micro-depletion capability. First, the depletion architecture within MAMMOTH is modified to improve the flexibility and speed of burnup calculations. The results of various benchmark tests demonstrate the accuracy of the modified implementation. The second improvement involves the expansion of the data functionality. MAMMOTH can now read the decay data files used by the Oak Ridge Isotope Generation (ORIGEN) code, thus enabling access to data of more than 1,600 isotopes. Comparison of ORIGEN and MAMMOTH for a selected benchmark problem was performed. Both codes yield very similar results, with differences of 0.03% that can be attributed to differences in precision between the two codes. Future work will focus on comparisons between MAMMOTH and ORIGEN for more complex problems.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Pu-238 Production Calculator and Estimation Tool

A MS Excel tool was created to help track Pu-238 production, Np-237 feedstock use, estimate production scenarios, and coordinate planning for shipments. As-run estimates were used as a basis for beginning and end of life isotopic concentrations. Multiple target types were created to demonstrate a proof of concept while actual results were not yet available The initial goals of this production tool were: • Track unirradiated, currently irradiated, and irradiated targets at the Advanced Test Reactor (ATR) • Estimate production yields from irradiated targets in various positions for each cycle • Provide past, current, and future year quantities of Np-237 and Pu-238 at ATR, amounts shipped to ATR and amounts shipped from ATR • Estimate production from current and future target designs based on data from numerical modeling (values will be provided by Idaho National Laboratory (INL)) • After proof of principle, additional work is planned that will further enhance the tool with: o Incorporate the High Flux Isotope Reactor (HFIR) reactor at Oak Ridge National Laboratory (ORNL) for irradiation o Incorporate irradiations and target shipments from ORNL to INL and back to track Np-237 transfers across the Department of Energy (DOE) Complex

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2015

Study of guided wave transmission through complex junction in sodium cooled reactor

Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presented in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Applicability of HBU Fuel Cask R and D Project to Al Clad Fuel Storage - 18558

Aluminum clad spent nuclear fuel (SNF) not destined to be reprocessed at Savannah River's H Canyon will eventually be moved from existing wet and dry storage facilities into standardized dry storage systems (e.g., Universal Canisters), ideally suited for disposal. With the current lack of disposal capacity in the U.S., placement of this SNF into dry storage requires plans for demonstrating to the regulator the safety of both the extended storage and the subsequent transportation to a future disposal facility of this SNF. For commercial SNF, concerns related to extended dry storage and subsequent transportation have been, and continue to be, addressed through several demonstration projects including, but not limited to, one for the lower burnup SNF at Idaho National Laboratory (INL) and one for the higher burnup SNF at Dominion's North Anna Power Station. Lessons learned from the more recent Orano led higher burnup demonstration project (HBU Demo Project) provides relevant information for advancing the concept of a 'smart cask' system for dry interim storage, which will allow for collection of data to support the safety demonstration for extended dry storage and subsequent transportation of the aluminum clad SNF. This 'smart cask' would eliminate the need for multiple demonstration projects that may have to be performed for different types of aluminum clad SNF designs and/or for potentially age distinguishing parameters, such as burnup or initial enrichment. This 'smart cask' would be developed and deployed by leveraging the HBU Demo Project's experience associated with modifying and deploying a TN-32B bolted carbon steel vertical storage cask system to monitor key parameters related to the criticality, structural, confinement, and thermal safety of the SNF and the canister system containing this SNF. This paper will focus on how the extensive design, licensing, and fabrication experience accumulated by the HBU Demo Project's cask system can be leveraged to support the design, licensing, and fabrication of a 'smart cask' system intended to support/confirm the interim dry storage, potential extended storage, and subsequent transportation of aluminum clad SNF, such as the Advanced Test Reactor (ATR) SNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

TEM Characterization of U-Mo Irradiated with High-Energy Xe Ions

U-Mo alloys (with 7-10 wt.% Mo) with a low U-235 enrichment (< 20%) have been developed to be used in high-performance research and test reactors, replacing the current highly enriched uranium (HEU) fuels. A number of in-pile irradiation tests of U-Mo fuels show that drastic microstructural changes occurred during reactor irradiation, including the formation of U-Mo-Al interaction products and large gas-filled pores. These microstructural changes might lead to breakaway swelling and deterioration of plate mechanical integrity. Hence, it is critical to understand the underlying mechanisms of these microstructural changes in order to improve the irradiation behavior of UMo fuels. Heavy ion irradiation was employed in this study to investigate separate effects of the irradiation behavior of U-Mo fuels. 1.7 mm diameter disks punched out from UMo/ Al dispersion fuel mini-plates were irradiated with 84 MeV Xe ions at the ATLAS accelerator at Argonne National Laboratory (ANL). This particular ion source was selected for two reasons: (1) to simulate the irradiation response of the fuel to fission fragment bombardments, and (2) to study fission gas behavior, as fission gas bubble formation and growth is the primary cause of U-Mo fuel swelling. The ion-irradiated samples have been characterized with scanning electron microscopy (SEM). This study focuses on the characterization of the samples using transmission electron microscopy (TEM). With the high magnification details, microstructures in U-Mo along the ion penetration depth were revealed. The U-7Mo/Al dispersion fuel sample irradiated with 84 MeV Xe ions to a dose of 2.9 x 10{sup 17} ions/cm{sup 3} was characterized using TEM. The TEM results exhibit that the sample microstructure varies along the ion penetration depth. Near the sample surface (zone A) is a layer of nano-crystalline U-Mo-Al interdiffusion product. After passing through zone A, the Xe ions resided in zone B, where gas bubbles formed in U-Mo. TEM images of this region displayed that large gas bubbles formed on grain boundaries and high-density nano-bubbles lay inside grains. The random distribution of intragranular bubbles indicates that the irradiation conditions of this sample cannot lead to gas bubble superlattice formation. Outside of the Xe penetration depth (zone C), significant amount of α-uranium precipitates were identified. The fact that no α-uranium precipitates were found in zone B suggests that the dose rate in this irradiation is sufficient to facilitate a phase reversal by means of a displacement spike mechanism. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Design of a Loss of Coolant Blowdown Capsule for Remote Assembly with High Burnup Fuel

The Transient Water Irradiation System is an enhanced capability capsule type irradiation vehicle designed to support fuel safety research for light water reactor specimens in the Transient Reactor Test Facility and is designed to simulate loss of coolant and reactivity-initiated accidents. The capsule was designed, deployed, and commissioned with fresh fuel specimens to validate instrumentation and prepare for previously irradiated high burnup specimens. The irradiation system features an extensive in-situ instrumentation package to detect phenomena typical to light water reactor fuels. To accommodate the assembly with high burnup specimens inside the Hot Fuel Examination Facility, and to ensure instrumentation integrity is maintained throughout assembly, the design was updated to support remote handling. The updated design features a hinge mechanism which allows for remote pre-irradiated specimen loading and assembly while protecting sensitive instrumentation by relocating during loading. Fixtures and equipment have been developed to handle the experiment capsule and components in cell, and to support remote assembly. The experiment module also includes radiation shielding and contamination control to support operations after removal from the hot cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Consulting on HTGR Design and Analysis Methods - Lommers

The Advanced Reactor Technologies – Gas Cooled Reactor (ART-GCR) program has undertaken an effort to develop point designs of a gas-cooled, graphite moderated irradiation test reactor (TR) and a larger demonstration reactor (DR). The Department of Energy (DOE) is specifying requirements and performance metrics for these two systems. Both designs pose challenges, and their development will benefit from interaction with a potential vendor. The work defined in this subcontract supports the effort to generate point designs for gas-cooled test and demonstration reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Calculation of Neutron Damage to Sodium Coolant Inlet Plenum

The coolant inlet plenum is a life limiting component of a Sodium-cooled Fast Reactor (SFR). The inlet plenum is a fixed reactor component which receives sodium from the cold side of the sodium pool redan and directs it into the inlet nozzles of the fuel assemblies. Neutron leakage from the reactor core causes irradiation damage to the steel components of the inlet plenum leading to a reduction in strength. The life of the reactor is essentially tied to the acceptable strength degradation of the inlet plenum. Therefore, typical SFR designs employ a neutron reflector/shield as an integral feature of the fuel assembly. The role of the axial reflector/shield is to: 1) direct fast neutrons back to the core, and 2) attenuate the neutron flux to minimize neutron interaction with the inlet plenum. The longer the reflector and the greater the share of steel within the reflector design, the greater the attenuation. The axial reflector is heated by gamma rays from the core, and must have sufficient coolant penetrations to ensure adequate cool-ability. Also, the sodium coolant must pass through the reflector before reaching the fuel rods above it. The present work investigates varying reflector geometries in support of the Versatile Test Reactor (VTR), a SFR research reactor planned to restart fast reactor irradiation capabilities in the U.S. DOE laboratory complex. This work will demonstrate that the neutron damage to the inlet plenum is relatively insensitive to the number and arrangements of coolant channel penetrations through the reflector.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2023

ATF-2 Sensitivity M&C 2023 (Slides)

Irradiation experiments 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 are modeled with a range of possible irradiation targets. First, neutron flux maps with experiments in those positions are compared to those of a water-filled configuration. Then, several safety and programmatic parameters for a generic accident-tolerant fuel test train are calculated. It is shown that these parameters can exhibit considerable sensitivity to the contents of the H and A experiment positions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Neutron characterization of UN/U-Si accident tolerant fuel prior to irradiation

Neutron imaging and tomographic reconstructions provide non-destructive three-dimensional characterization of nuclear fuel pellets that are typically about a cubic centimeter in volume with energy-resolved neutron imaging at a pulsed neutron source offering unique characterization not possible at a reactor source. Recent measurements at the Los Alamos Neutrons Science Center (LANSCE) studied a series of uranium nitride / uranium silicide ceramic composite pellets. This paper provides examples of spatially resolved measurements of cracks, isotopic enrichment and density with a (current) spatial resolution of approximately 75 microns. Phase and texture measurements were also performed but are not discussed in this paper. The UN/U-Si systems are under consideration as candidate systems for accident tolerant fuel. The examples are taken from measurements on a total of 9 pellets of differing UN/U-Si formulations (2 enrichment levels), U{sub 3}Si{sub 5} reference formulations (two enrichment levels) and pellets with known flaws. The {sup 235}U enrichments ranged from 0.2 to 8.8 wt.%. Pellets of the same composition are scheduled for irradiation in the Advanced Test Reactor this year. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Precipitation of Epsilon Particles in Metal Doped Ceria Irradiated with Energetic Ions

Metallic particles precipitated from fission products in oxide fuel matrix (e.g., UO{sub 2}) have a great impact on the thermal and mechanical properties of the fuel. A burn-up of 40 MWd/kg U results in the conversion of 4% of the uranium to approximately 3% fission products and 1% transuranium elements. Epsilon phase is the major metallic phase of the precipitates with a hexagonal-closed packed (HCP) crystal structure. It can exist in a variety of metallic forms in UO{sub 2}, but typically, it consists of five metals: Mo, Pd, Rh, Ru and Tc. Epsilon particles can form over a wide range of stoichiometries in the five metal system. Simulated compositions with Re as a surrogate for Tc have been investigated as a waste form for {sup 99}Tc, an isotope with a long half-life of 2.13 10{sup 5} years and a high solubility under oxidizing conditions in addition to Mo in the corrosion process. There is a pressing need to understand the factors influencing the formation and growth of the metallic phase in fuel. Efforts are currently underway to perform careful neutron irradiation of nuclear fuel under controlled conditions in a test reactor, recover metallic phases that may be present, and characterize the resulting structure and chemistry. Due to the long duration, high cost and radiological concerns associated with neutron irradiation and sample characterizations, ion irradiation studies of a surrogate material with a full control of the processing parameters can emulate different stages of the particle precipitation, which can add a tremendous value to reactor-based studies. This presentation reports our recent ion irradiation results on the metallic particle precipitation under various irradiation and thermal annealing conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Hybrid Zircex Hydrochlorinator

This project was driven by the desire to create easier and faster ways to load Advanced Test Reactor (ATR) Fuel Elements into the Hybrid Zircex hydrochlorinator. The primary requirements for this project were to minimize the potential release of radioactive contamination, minimize maintenance operations to replace or remove the lid seal, and allow for remote usage. Three preliminary designs were made. These designs allow ATR fuel elements to be loaded into the hydrochlorinator more efficiently while fulfilling the requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗