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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

On the influence of the power plant operational history on the inventory and the uncertainties of radionuclides relevant for the final disposal of PWR spent fuel - 15149

A mandatory requisite for official granting safe final disposal approval of SNF is the comprehensive specification and declaration of the nuclear inventory in SNF by the NPP-operators. The whole radiochemical characterization of each spent fuel assembly hides many difficulties such as very high radiation burden, destruction method, the inhomogeneous burn-up of fuel assembly, very high cost. Thus, non-destructive characterization is usually applied on industrial scale. In this case only several nuclides can be measured, and the rest is calculated using simulations and typical irradiation history. Like any other German rad-waste acceptance requirements all these declarations will have to be checked and verified by an independent inspector. In this verification process not only values of radionuclide (RN) activities are important, but also their uncertainties. The simple simulation using burn-up calculations based on typical and generic reactor operational parameters do not answer the question about possible uncertainties observed in real reactor operations. At the same time, the details of irradiation history are often unknown, which complicates the assessment of declared RN inventories. Therefore, the development of dedicated generic proof tools, which can be applied to different types of waste, often of unknown origin and history is worth a serious consideration. Here, we have compiled a set of burn-up calculations considering the operational history of 339 published or anonymized real PWR fuel assemblies. This gives us the possibility to calculate the realistic variation or spectrum of RN inventories. These histories were used as a basis for 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's), i.e. initial enrichment (IE), specific power (SP), irradiation time (IT), downtime (DT), boric acid concentration (BA), moderator and fuel density (MD and FD) and temperatures (MT and MT) used as input parameters for the simulation models. Then, burn-up calculation codes provide the anticipated results for the entire nuclear inventory such as characteristic activity values and the corresponding minimum and maximum limits as a function of the fuel burn-up. These results are being validated with the experimental data from various publications and the online database - SFCOMPO and subsequently transferred and summarized into our own database. Ultimately, this database will provide the basis of proof tools for the operator independent verification of the declared nuclear inventories for the quality and safety assessment of a radioactive waste compound and bounding values for certain RN as a function of burn-up. The RN inventory can hence be verified without taking credit of the whole of the operational history of the SNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

A Design and Fabrication Options Trade Study for Metallic Fuel without Internal Sodium Bonding

Nuclear fuels using alloys of uranium, or metallic fuels, have many beneficial properties. The classical metallic fuel design uses a loose fitting cylindrical “slug” of fuel placed inside stainless-steel cladding tubes where the gap is filled with sodium. This sodium bond is liquid at operating temperature and conducts heat from the slug to the cladding, especially in early life before fuel swells into contact with the cladding. Despite the benefits of sodium bonding, there is a desire to develop metallic fuel technologies without it chiefly to reduce chemical reaction hazards in spent fuel storage from sodium fast reactors operating on once-through fuel cycles. Elimination of the sodium bond may also help unlock potential benefits for fuel fabrication, reactor neutronics, and compatibility with other types of reactors. Creating a sodium-free metallic fuel revolves around the problem of manufacturing fuel slug geometries which are in close contact with the cladding at beginning of life to facilitate heat transport while alleviating fuel-cladding chemical interactions (FCCI) at this interface and providing enough free volume to accommodate fuel swelling. Accelerating development and qualification of this fuel system will require careful selection of design and manufacturing options. To this end, a design trade-off study was performed to evaluate candidate options. Several design and manufacturing options were assessed, weighted, scored, and ranked with respect to fabrication, normal reactor operation, off-normal scenarios, and back-end considerations. This effort was performed both for “baseline” needs, which represented a once-through fuel cycle at temperatures and burnups known to be viable for sodium-bonded metallic fuel, and for “enhanced” needs to represent opportunities for closed fuel cycles and/or more aggressive temperatures/burnups. The outcomes of this study prioritized a baseline technology using U-Zr alloy with additives to mitigate FCCI, produced in annular slug geometry by continuous casting, clad in austenitic stainless-steel alloy, and followed by a final step to swage the cladding down to close the gap. This study prioritized an enhanced fuel technology using U-Mo alloy, also produced by continuous casting into an annular geometry, followed by coating/plating with an FCCI barrier on the slug, again with a final step to swage the cladding diameter down using oxide dispersion strengthened steel. It was noted that development of the enhanced fuel technology would entail more risk, thus U-Zr alloy was put forth as a backup to U-Mo if challenges are encountered with FCCI barriers, and advanced ferritic/martensitic steels are put forth as a backup to oxide dispersion strengthened steels if swaging and welding are found unworkable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

ACRRF High-Bay Dose Calculations using MCNP (Part A)

Analytical tools and models have been developed as a starting point for directly assessing dose in the Annular Core Research Reactor Facility (ACRRF) due to reactor operation. Key results include peak dose along the Central Cavity (CC) centerline (beamline) at the cavity level, dose throughout the High-Bay (HB), and dose on the facility roof for partially-shielded reactor operation where the 4” insert is removed from the CC Shield Plug (SP). Model results in the beamline are benchmarked against measured doses from passive dosimetry evaluations. Personnel total (neutron and gamma) dose in the ACRRF HB is calculated using Monte Carlo N-Particle (MCNP). Various CC and SP configurations are analyzed, including unshielded (no SP) and partially shielded (SP installed but 4” insert removed). Novel application of Variance Reduction (VR) techniques, namely the Surface Source Write (SSW) and Surface Source Read (SSR) capabilities in MCNP, enable impressive resolution (in a Monte Carlo modeling sense) of dose throughout much the facility. The VR techniques reduce stochastic error for challenging tallies, with more advanced techniques explored in the companion to this report (Part B) [1]. Supplementary studies (including a verification analysis) and pedagogic evaluations in Part B involve neutron spectra, angular distributions, and the dose impact of facility characteristics. With the SP 4” insert removed and the Lead-Boron (44”) Bucket (LB–44) in the reactor cavity, Total Effective Dose (TED) within the CC beamline is ≈140 rem per 300 MJ of reactor yield (or 3900 rem per hour at 100% Steady-State (SS) power). With no SP (unshielded) and a Free-Field (FF) cavity, TED within the beamline is ≈610 rem per 300 MJ (or 17000 rem per hour at 100% SS power). Due to the predicted collimation of radiation by the reactor pool (and partial SP, if present), beamline dose is much greater than the scattered radiation field surrounding the cavity and reactor tank. Comparisons are made to beamline dosimetry measurements to validate the model. Model predictions agree reasonably well (⪅10%) with measured quantities of neutron fluence, gamma fluence, and spectral metrics. Away from the beamline, comparisons made to previous dose measurements in the HB agreement within an order of magnitude.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Molecular dynamics study of interstitial He clusters in nickel

This study presents a molecular dynamics analysis focusing on the behavior of interstitial helium (He) clusters in nickel (Ni), examining their formation, stability, and migration energetics. Consistent with previous research, we found that the binding energies of interstitial helium within a helium cluster are positive and increase with the cluster size, indicating a preference for helium atoms to cluster together. However, our findings also reveal that while the formation energy increases monotonically with cluster size, the increase in binding energy is non-monotonic. Importantly, small He clusters were observed to be thermally unstable at reactor operational temperatures (approximately 600 K), with the He 2 cluster exhibiting instability even at room temperature. With a binding energy of 0.49 eV for a He 4 cluster, we hypothesize that for helium bubbles to form via homogeneous nucleation (i.e., through trap mutation) at reactor operating temperatures, the helium concentration must be high enough to facilitate the formation of helium clusters of at least size 4 or larger. As expected, interstitial helium and small helium clusters are highly mobile. This mobility was observed not only at room temperature but also at temperatures as low as approximately 200 K. Furthermore, the mean squared displacement method has been utilized to determine the migration barriers and the corresponding prefactors for clusters ranging from He 1 to He 6

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Corrosion of Containment Alloys in Molten Salt Reactors and the Prospect of Online Monitoring

The aim of this review is to communicate some essential knowledge of the underlying mechanism of the corrosion of structural containment alloys during molten salt reactor operation in the context of prospective online monitoring in future MSR installations. The formation of metal halide species and the progression of their concentration in the molten salt do reflect containment corrosion, tracing the depletion of alloying metals at the alloy salt interface will assure safe conditions during reactor operation. Even though the progress of alloying metal halides concentrations in the molten salt do strongly understate actual corrosion rates, their prospective 1 st order kinetics followed by near-linearly increase is attributed to homogeneous matrix corrosion. The service life of the structural containment alloy is derived from homogeneous matrix corrosion and near-surface void formation but less so from intergranular cracking (IGC) and pitting corrosion. Online monitoring of corrosion species is of particular interest for molten chloride systems since besides the expected formation of chromium chloride species CrCl 2 and CrCl 3 , other metal chloride species such as FeCl 2 , FeCl 3 , MoCl 2 , MnCl 2 and NiCl 2 will form, depending on the selected structural alloy. The metal chloride concentrations should follow, after an incubation period of about 10,000 hours, a linear projection with a positive slope and a steady increase of <1 ppm per day. During the incubation period metal concentration show 1 st order kinetics and increasing linearly with time. Ideally, a linear increase reflects homogeneous matrix corrosion, while a sharp increase in the metal chloride concentration could set a warning flag for potential material failure within the projected service life, e.g. as result of intergranular cracking or pitting corrosion. Continuous monitoring of metal chloride concentrations can therefore provide direct information about the mechanism of the ongoing corrosion scenario and offer valuable information for a timely warning of prospective material failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

PCI analysis of a commercial PWR using BISON fuel performance code

Through the early 2000’s, nuclear fuel cladding failure resulting from pellet-cladding interaction in commercial pressurized water reactor fuel was managed successfully by incorporating power ramp restrictions. Within the past decade, however, cladding failures have again begun to appear, often correlated to missing pellet surface defects, and further aggravated by modern reactor operating strategies demanding higher fuel duty operations. In response, renewed efforts in fuel behavior modeling and simulation are being pursued to improve our understanding of the conditions that can lead to missing pellet surface related cladding failure. This paper describes a methodology used to model pellet-cladding interaction related failures observed in commercial pressurized water reactors using the BISON fuel performance code developed at Idaho National Laboratory. In order to accomplish this goal, a three-step process is followed: 1) A failure-stress threshold is developed by using BISON code to model ramp test with known pellet-clad interaction related failures, 2) BISON analysis is applied to a commercial pressurized water reactor cycle startup with one rod confirmed to contain a missing pellet surface defect, and 3) an missing pellet surface sensitivity study of a cycle startup of the same commercial pressurized water reactor which contained a failed fuel rod of un-identified type. The analysis results described in this paper serve to illustrate the BISON code capabilities for modeling pressurized water reactor fuel subjected to higher duty power events.

Nuclear Science & Technology↗
DOE OSTI2020

Criticality Safety and Fuel Performance Considerations for Enrichment Above 5 Weight Percent in the Uranium Dioxide Fuel Cycle

The U.S. Nuclear Regulatory Commission (NRC) is preparing for anticipated licensing applications requesting fuel enrichments of UO 2 ceramic fuels enriched in excess of 5 weight percent (w/o) 235 U for commercial light-water reactors (LWRs). PNNL has been tasked with providing technical assistance to the NRC related to the review and approval of requests for enrichment increases above the current 5 w/o 235 U enrichment limit. This report will provide the agency with technical assistance to enhance the staff’s knowledge base and to identify developmental needs required to support the licensing of greater than 5 w/o UO 2 ceramic fuels. This report provides a background on increasing enrichment of UO 2 LWR fuel beyond 5 w/o 235 U. The primary concern of increasing enrichment is the impacts to criticality safety. This report will focus on criticality safety as it relates to enrichment, conversion of uranium hexafluoride to UO 2 , fuel assembly fabrication, transportation of fresh/spent fuel assemblies, reactor operations, and storage of fresh/spent fuel assemblies. The report provides a general summary of each of the steps in the UO 2 fuel cycle starting from transportation of enriched UF 6 through dry cask storage of spent fuel assemblies. This report also discusses other impacts of increased enrichment during reactor operation and the surrounding fuel cycle activities. Current needs regarding both data and analytical codes to support the increase above 5 w/o 235 U are identified and discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of accident tolerant control rod for light water reactors

For an accident tolerant control rod (ATCR) designed to improve LWR (light water cooled reactor)core safety, the following inherent characteristics are required: sufficient neutron absorption, high-temperature stability and high miscibility with fuel materials to avoid uncontrollable recriticality. In the present work an ATCR is being developed to improve the safety of light water reactors under all conditions including severe accidents. Promising neutron-absorbing materials for the ATCR include rare-earth sesquioxides (RE{sub 2}O{sub 3}), hafnium dioxide (HfO{sub 2}) and mixtures of these materials. A reactivity worth analysis of these materials in representative light water reactors revealed that a conventional control rod can be replaced by an ATCR without significant impact on the reactor operation conditions. A high-temperature compatibility of the candidate ATCR materials and iron, which is a main component of control rod cladding, was examined experimentally, and no significant reaction occurred upon heating to 1600 Celsius degrees for 1 hour. That is, the ATCR is expected to have no significant damage prior to extensive fuel pin failure in severe accidents, thus avoiding uncontrollable recriticality after reactor shutdown even in case of injection of non-borated or insufficiently borated water for emergency cooling of the reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

In-Line Method for Calculating Peak Neutron Absorption by Fission Products Following a Change in Reactor Power

When the power level in an operating reactor is reduced the concentration of xenon-135 and samarium-149 temporarily increases as the nuclides (and their immediate precursors) equilibrate to the new power level. During this period, the increased inventory of these fission products causes an insertion of negative reactivity that can last for tens of hours. Understanding and predicting this transient reactivity effect is important in reactor design and operation. Explicit depletion calculations are often used to predict the reactivity behavior following a power change. However, these calculations require many time-steps to fully resolve the reactivity trend, and are correspondingly expensive. In 2010 an in-line method for estimating the peak xenon mass in a core following shutdown was described. However, there is no guarantee that the peak negative reactivity condition for a core will occur at peak xenon mass. Therefore, in this summary, we extend this original work using a new method that can predict the peak neutron absorption rate in the core due to {sup 135}Xe and {sup 149}Sm following any reduction in power. The proposed method can also be used to calculate the peak positive reactivity transient due to an increase in reactor power, but this scenario is of less interest in most reactor design applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Oxidation Resistance in 1200°C Steam of a FeCrAl Alloy Fabricated by Three Metallurgical Processes

Abstract FeCrAl alloys are a leading candidate material for accident tolerant fuel cladding due to their good performance in both normal light-water reactor operating conditions as well as their resilience to high-temperature accident scenarios. For commercial-scale production, new fabrication techniques need to be investigated. In this study, the effects of fabrication methods on the high-temperature steam oxidation performance of C26M (Fe12Cr6Al2Mo in wt.%) were investigated. Three variants of C26M were manufactured: wrought (cast and forged) (WC26M), powder metallurgy hot isostatic pressing (PMC26M), and laser powder bed fusion additive manufacturing (AMC26M). All three variants were exposed to steam at 1200°C for 2 h. Results showed no significant variation in mass change between the variants after steam exposure. All three variants effectively formed stable protective alumina films with ~0.6–1.3 µ m thickness. This study suggests FeCrAl alloys have excellent resilience to high-temperature steam in nuclear reactor accident scenarios regardless of the fabrication method.

Materials Science↗
DOE OSTI2020

High-resolution neutron time-of-flight measurements for light water at the Spallation Neutron Source (SNS), Oak Ridge National Laboratory

Series of light water inelastic neutron scattering experiments have been made at the Oak Ridge National Laboratory (ORNL), Spallation Neutron Source (SNS) covering temperatures ranging from 295 K to 600 K and pressures of 1 bar and 150 bar. The temperatures and pressures ranges correspond to that of pressurized light water reactors. The inelastic scattering measurements will help the development of light water thermal scattering kernels, also known as S (α,β) thermal scattering law (TSL), in a consistent fashion given the amount and the quality of the measured data. Light water thermal scattering evaluations available in existing nuclear data libraries have certain limitations and pitfalls. This paper introduces the state of the art of the light water thermal scattering cross-section data not only for room temperature but as well as for reactor operating temperatures, i.e. 550 - 600 K. During the past few years there has been a renewed interest in re-investigating the existing TSL models and utilize the recent experimental data or perform molecular dynamics simulations. It should be pointed out that no single TSL evaluation is based entirely on experimental data and one has to rely on TSL models or a combination of both. New TOF measurement of light water at the SNS, with a detailed description of the experimental setup, measurement conditions, and the associated foreseen results is presented in this paper. The analysis of the experimental data would help in validating the existing approach based on old experimental data or based on molecular dynamic simulations using classical water models, knowledge of which is very important to generate TSL libraries at reactor operating conditions.

Leal, Luiz C.↗
DOE OSTI2025

Automated Resonance Fitting for Nuclear Data Evaluation

Global and national efforts to deliver high-quality nuclear data to users have a wide-ranging impact, affecting applications in national security, reactor operations, basic science, medicine, and more. Cross section evaluation is a major part of this effort, combining theory and experimentation to produce recommended values and uncertainties for reaction probabilities. Resonance region evaluation is a specialized type of nuclear data evaluation that can require significant manual effort and months of time from expert scientists. In this article, non-convex non-linear optimization methods are combined with concepts of inferential statistics to infer a resonance model from experimental data in an automated manner that is not dependent on prior evaluation(s). This methodology aims to enhance the workflow of a resonance evaluator by minimizing time, effort, and the potential for bias from prior assumptions, while enhancing reproducibility and documentation, thereby addressing well-known challenges in the field.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Test results of an ITER relevant FPGA when irradiated with neutrons

The data acquisition and control instrumentation cubicles room of the ITER tokamak will be irradiated with neutrons during the fusion reactor operation. A Virtex-6 FPGA from Xilinx (XC6VLX365T-1FFG1156C) is used on the ATCA-IO-PROCESSOR board, included in the ITER Catalog of I and C products - Fast Controllers. The Virtex-6 is a re-programmable logic device where the configuration is stored in Static RAM (SRAM), functional data stored in dedicated Block RAM (BRAM) and functional state logic in Flip-Flops. Single Event Upsets (SEU) due to the ionizing radiation of neutrons causes soft errors, unintended changes (bit-flips) to the values stored in state elements of the FPGA. The SEU monitoring and soft errors repairing, when possible, were explored in this work. An FPGA built-in Soft Error Mitigation (SEM) controller detects and corrects soft errors in the FPGA configuration memory. Novel SEU sensors with Error Correction Code (ECC) detect and repair the BRAM memories. Proper management of SEU can increase reliability and availability of control instrumentation hardware for nuclear applications. The results of the tests performed using the SEM controller and the BRAM SEU sensors are presented for a Virtex-6 FPGA (XC6VLX240T-1FFG1156C) when irradiated with neutrons from the Portuguese Research Reactor (RPI), a 1 MW nuclear fission reactor operated by IST in the neighborhood of Lisbon. Results show that the proposed SEU mitigation technique is able to repair the majority of the detected SEU errors in the configuration and BRAM memories. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Monte Carlo simulations of Cu/Ni–Si–Mn co-precipitation in duplex stainless steels

First-Passage Kinetic Monte Carlo (FPKMC) simulations of species migration in duplex stainless steels were performed in order to establish relationships between alloy content, segregation behavior, and the formation of Ni-Si-Mn rich particles in cast duplex stainless steels during thermal aging. The Ni-Si-Mn-rich second phase forms after extended aging at reactor operating temperatures and degrades alloy properties. Simulations of Ni-Si-Mn cluster formation were validated through comparison with experimental results obtained through Atom Probe Tomography (APT) on similar alloy compositions, identifying several trends. First, Cu promotes the formation of Ni-Si-Mn clusters, but only when Ni or Mn prefer segregation to the surface of Cu particles; without the segregation of these species, the critical composition for the clusters to form was not achieved. Second, the width of precipitate-denuded zones near ?/d interfaces increases with decreasing Cu content. This finding was in strong agreement with APT results, further validating our model. Finally, our model predicts that Ni-Si-Mn cluster formation will be the most extensive when the Si:Mn ratio is approximately 1:1 and the least extensive when one of these key elements is less concentrated (Mn = 0.5 at.% or Si = 0.5 at.%). The implications of these results on how to improve alloy properties are discussed.

Cu co-precipitation, Atom probe tomography (APT), ↗
DOE OSTI2021

Isotope effects on transport in LHD

Abstract Isotope effects are one of the most important issues for predicting future reactor operations. Large helical device (LHD) is the presently working largest stellarator/helical device using super conducting helical coils. In LHD, deuterium experiments started in 2017. Extensive studies regarding isotope effects on transport have been carried out. In this paper, the results of isotope effect studies in LHD are reported. The systematic studies were performed adjusting operational parameters and nondimensional parameters. In L mode like normal confinement plasma, where internal and edge transport barriers are not formed, the scaling of global energy confinement time ( τ E ) with operational parameters shows positive mass dependence ( M 0.27 ; where M is effective ion mass) in electron cyclotron heating plasma and no mass dependence ( M 0.0 ) in neutral beam injection heating plasma. The non-negative ion mass dependence is anti-gyro-Bohm scaling. The role of the turbulence in isotope effects was also found by turbulence measurements and gyrokinetic simulation. Better accessibility to electron and ion internal transport barrier (ITB) plasma is found in deuterium (D) plasma than in hydrogen (H). Gyro kinetic non-linear simulation shows reduced ion heat flux due to the larger generation of zonal flow in deuterium plasma. Peaked carbon density profile plays a prominent role in reducing ion energy transport in ITB plasma. This is evident only in plasma with deuterium ions. New findings on the mixing and non-mixing states of D and H particle transports are reported. In the mixing state, ion particle diffusivities are higher than electron particle diffusivities and D and H ion density profiles are almost identical. In the non-mixing state, ion particle diffusivity is much lower than electron diffusivity. Deuterium and hydrogen ion profiles are clearly different. Different turbulence structures were found in the mixing and non-mixing states suggesting different turbulence modes play a role.

Physics↗
DOE OSTI2017

Risks to Ecological Receptors Posed by Contaminants of Potential Concern in the Lower Three Runs Cooling Ponds and Canals

The upper portion of Lower Three Runs includes several ponds, reservoirs, and canals that were formerly used as a cooling system for nuclear production reactors. This area was divided into nine exposure areas (EAs) for the assessment of environmental contamination resulting from past reactor operations and other industrial processes. A tiered screening process identified several contaminants of potential concern including aluminum, cyanide, lead, manganese, mercury, DDD, DDE, and DDT. Risks posed by these contaminants to ecological receptors (river otter, belted kingfisher, raccoon, and blue heron) were assessed using contaminant exposure models that estimated contaminant intake resulting from ingestion of food, water, and sediment/ soil and compared these intakes with toxicity reference values (TRVs). The contaminant exposure models showed that the TRVs were not exceeded in the otter model, exceeded by aluminum in EA 7 (Pond 2 and associated canals) in the raccoon model, and exceeded by mercury in EAs 2, 3 (Pond B), 6 (Par Pond), and 8 (Ponds 4 and 5 and Canal to Pond C) in both the kingfisher and blue heron models. Hazard quotients (total exposure dose divided by the TRV) were 2.8 for aluminum and 1.7- 3.6 for mercury. The primary route of exposure for aluminum was the ingestion of soil, and the primary route of exposure for mercury was the ingestion of mercury contaminated fish. Elevated levels of mercury in fish were at least partly the result of the aerial deposition of mercury onto Lower Three Runs and its watershed. The atmospheric deposition of mercury creates pervasive contamination in fish throughout the Savannah River basin. Another possible source of mercury was the discharge of mercury contaminated Savannah River water into the Lower Three Runs cooling ponds and canals during previous years of reactor operation. This contamination originated from industries located upstream of the SRS. The aluminum exceedance for the raccoon was likely the result of naturally high aluminum levels in SRS soils rather than SRS operations. Aluminum exceedances have previously been observed in relatively undisturbed background locations as well as areas affected by SRS operations. Aluminum exceedances are more likely with the raccoon than the other receptors because it consumes more soil as a result of its feeding habits. Sensitivity analysis showed that model uncertainty can be reduced by adequate sampling of key variables (e.g., fish and sediments). Although sediment samples were collected from all EAs, fish samples were not collected from three EAs and some analytes (pesticides and cyanide) were not measured in fish. Water-to-fish concentration ratios were used to estimate contaminant levels in fish when direct measurements from fish were unavailable; however, such estimates are potentially less accurate than direct measurements.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

The Effect of Localized Pin Flux on Short-Term Gamma Intensities of Spent Fuel Assemblies

Nuclear reactor operation for electricity generation involves irradiation of a fuel assembly for an extended period of time until a certain burnup is achieved. To harness the full nuclear energy potential and flatten the flux around the core, fuel assemblies are shuffled to one or more different locations within the core. Post-irradiation practice typically involves storing the spent fuel assembly in a spent fuel pool. While in the spent fuel pool, aspects of the fuel assembly continue to evolve, particularly the isotopic concentration as nuclides decay. Numerous nondestructive assay (NDA) safeguards techniques exist and have been proposed for performing verification on spent fuel assemblies to ensure they have not been tampered with. Adigun et al. recently proposed measuring both the neutron and gamma signatures from spent fuel from discharge to final disposal, however, the research in only investigated the neutron part of this approach. In the conclusion, the authors caveated the need for the gamma results to make more certain conclusions about a spent fuel assembly, as a combination of the neutron and gamma NDA measurements will limit the range of possible initial enrichment and burnup combinations. This paper focuses on the production of the gamma source. It is important to understand the driving mechanisms behind the emissions signatures at shutdown and after long cooling time periods, to better determine the correlation between the measured passive gamma emission spectra and the gamma emissions per second of the assembly. The objective is to demonstrate how the effect of localized flux within a reactor affects the gamma emissions per second (or gamma intensity) present in an assembly at shutdown. This study examined the photons emitted from the nuclides present within the fuel assembly, and not the photons escaping from the fuel assembly. The latter point of transporting the photons present within the spent fuel assemblies to an appropriate detector will be addressed in a future paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

In-situ coating of silicon-rich films on tokamak plasma-facing components with real-time Si material injection

Abstract Experiments have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from 3.9– 7.5 × 10 19 m −3 and input powers ranging from 5.5 to 9 MW. The small Si pellets were delivered with the impurity granule injector at frequencies ranging from 4 to 16 Hz corresponding to mass flow rates of 5–19 mg s −1 (1– 4.2 × 10 20 Si s −1 ) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite samples were exposed to the scrape-off layer and private flux region plasmas through the divertor material evaluation system to evaluate the Si deposition on the divertor targets. The Si II emission at the sample correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings containing silicon oxides, of which SiO 2 is the dominant component. No evidence of SiC was found, which is attributed to low divertor surface temperatures. The in-situ and ex-situ analysis found that Si-rich coatings of at least 0.4–1.2 nm thickness have been deposited at 0.4–0.7 nm s −1 . The technique is estimated to coat a surface area of at least 0.94 m 2 on the outer divertor. These results demonstrate the potential of using real-time material injection to form Si-enriched layers on divertor PFCs during reactor operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗