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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

The decommissioning imperative - motivating progress in us used fuel management - 15185

When the Yucca Mountain project was terminated in 2010 there was approximately 3500 metric tons of used fuel stored at Independent Spent Fuel Storage Installations (ISFSIs) at which there was no operating reactor (10 shutdown plant sites and the GE Morris facility). By 2015, four additional sites with five reactors will have shut down, causing the amount of used fuel 'stranded' at sites where there is no operating reactor to nearly double to approximately 6900 metric tons which will be stored at a total of 15 sites. At each of these sites, once the reactors are decommissioned, the only thing preventing the land from being returned to communities for other uses is the remaining presence of the used fuel. The cost of maintaining 15 stand-alone ISFSIs is born by electric ratepayers and taxpayers. Going forward, the costs of maintaining stand-alone ISFSIs is certain to grow even if the number of stand-alone ISFSIs does not grow beyond the already announced shutdown of the Oyster Creek reactor in 2019 (which would increase the amount of stranded used fuel by another 10% and create a 16. stand-alone ISFSI). One reason for this cost escalation will be the need for ISFSI owners to renew storage licenses, and implement aging management programs to support these renewals, due to the continued delay in the federal program for removing the used fuel. Storage licenses at 10 of 16 stand-alone ISFSIs will expire by 2021, which is the earliest projected date at which the Department of Energy (DOE) has indicated it could have a pilot interim storage facility available - and nearly 3 decades before DOE projects a repository will be available. The burden and cost of continuing to renew storage licenses - and manage aging storage facilities - at 16 stand-alone sites instead of moving stranded used fuel to a single consolidated storage location or repository will likely be significant. Industry resources are increasingly being channeled into dry storage aging management activities as public and regulator interest in potential long-term degradation mechanism continues to grow. Whether the ultimate destination for used fuel is Yucca Mountain or some other location, the current stalemate preventing progress in the federal program to move it must be broken - and the increasing prominence of the decommissioned reactor quandary may very well be a pivotal factor in breaking that stalemate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Immediate Dismantling of a Large Fleet of LWR NPPs: Consequences for Spent Fuel and Waste Management - 16027

The experience feedback shows that the complete dismantling of one Light Water Reactor (LWR) unit is an activity now globally well under control (reactor pressure vessel and its internals removal and cutting included). However some specific industrial difficulties may arise in view of the dismantling of numerous LWR units simultaneously. Indeed more than 300 LWR units are currently operated worldwide, which have been commissioned these last 50 years. So in some countries, many units may be permanently shut down over a period of few years in the next decades. Such a situation addresses the issue of the overall management of large quantities of spent nuclear fuel (SNF) and decommissioning waste for the concerned units. For example, in France, the legislative and regulatory framework for the nuclear facilities favors their dismantling 'as soon as possible' after their permanent shutdown which implies as well to limit the duration of the transition period from operation to decommissioning. Furthermore 58 LWR units have been commissioned between 1977 and 1999 in France - on average more than 2 units per year. In this context, the operator (Electricite de France - EDF) plans to remove quickly the SNF then to perform dismantling actions immediately after the permanent shutdown of the LWR units. One issue is to remove the SNF from all the relevant units, even if this removal is simultaneous in many units (permanently shut down or still under operation). Similar issue has to be taken into account regarding the management of the radioactive waste (RW) produced by the dismantling and clean-up actions, notably the RW that cannot be disposed of in a near surface repository. One method followed by the French technical support organization (Institut de Radioprotection et de Surete Nucleaire - IRSN) to analyze these issues is the use of estimates of flows of SNF and RW. These estimates are based notably on a phasing-scenario and a planning template defined for the decommissioning of the units of one nuclear power plant (NPP) and coupled to an overall schedule for phase out all the units of the fleet. The estimates relative to the next decades can be compared to the current experience feedback of flows of SNF and RW for units under operation, in order to identify risks when facing decommissioning. The risks highlighted are driven by key parameters (as duration of the main dismantling actions) of the estimates which can be adapted to minimize their impact. On this basis, it is possible to identify the key-factors to dismantle each unit of NPPs and phase out the fleet regarding SNF and RW management. It is noteworthy that this work needs to be done in any case upstream the studies and the implementation of dismantling actions. Finally it can be underlined that another issue is the human resources (staff, skills and knowledge) necessary to perform all the decommissioning actions, but this aspect is not addressed in the present paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A WM Facility for the Processing of Radioactive Materials Arising from the Decommissioning Activities of the Belgian Nuclear Power Plant of Doel - 16042

Doel Nuclear Power Plant is located in the northern part of Belgium in the harbour of Antwerp on the Scheldt River. On this 80-hectare site Electrabel, the Belgian utility, operates four PWRs with a total generation capacity of 2911 MWe. Those nuclear power units were commissioned in 1975 (Doel 1 and 2 - twin units), in 1982 (Doel 3) and in 1985 (Doel 4). In light of the Belgian governments decision to definitively shutdown the oldest units of the site (Doel 1 and 2 - 433 MWe each) in 2015 at the end of their 40-year design lifetime, Electrabel launched a D and D program in partnership with Tractebel Engineering in July 2012. During the D and D preparation phase prior to the final shutdown, studies performed by Tractebel Engineering showed the added benefit of a purpose-built Waste Management Facility (WMF) specifically equipped to process contaminated and activated components generated by the in-situ dismantling activities. The WMF is a key infrastructure contributing to a large extent to the Waste Minimization Policy within the global D and D program. The present paper aims at presenting its strategic, logistic, infra-structural, and design principles. At the end of 2014, the Belgian government and Electrabel entered into discussions to envisage a 10-year lifetime extension of Doel 1 and 2 units. Therefore, Electrabel decided to put the D and D program on hold. Nevertheless, the preparatory studies for the future D and D activities on Doel site are set on solid assumptions and pave the way towards a safe, cost-effective and waste minimizing decommissioning process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Transitioning From Operations to Decommissioning - 16049

With the recent unplanned shutdown of a number of U.S. Nuclear Power Plants, a benchmarking effort was conducted in late 2104 to capture the lessons learned from these recently shutdown sites. The specific plants benchmarked for this effort included San Onofre Nuclear Generating Station (SONGS), Kewaunee, Zion, and Crystal River Nuclear Power Stations. The benchmark results found for the most part, site personnel had little or no decommissioning experience and in fact, were not aware of the challenges facing a plant in transitioning from operations to decommissioning. This lack of experience and knowledge resulted in a steep learning curve, a loss of productivity, and significant challenges by plant personnel prepared the site for decommissioning. The results of the benchmark identified a number of areas and challenges facing an operating plant during this period. All sites visited identified four key areas where significant cost savings and planning effectiveness could have been realized if they had been better prepared. The areas included transition planning, staffing, stakeholder involvement, and cost estimating. It was recognized if these areas are aggressively managed early in the process, it will have a significant positive impact on the effectiveness and cost of decommissioning and can better set the stage for a successful project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Overview of the U.S. Department of Energy, Office of Nuclear Energy's Nuclear Fuel Storage and Transportation Planning Project

The U.S. Department Energy, Office of Nuclear Energy (DOE-NE) is laying the groundwork for implementing interim storage, including associated transportation, per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High- Level Radioactive Waste, and to develop a foundation for a new nuclear waste management organization. The organization within DOE-NE that is responsible for this planning is the Office of Nuclear Fuels Storage and Transportation (NFST). The technical work to support this planning effort is being conducted by DOE's national laboratories. DOE's overarching goal is to develop options for decision-makers on the design of an integrated waste management system. The goal is to develop an integrated plan to (1) implement interim storage; (2) improve the overall integration of storage as a planned part of the waste management system; (3) prepare for the large-scale transportation of SNF and high-level waste (HLW), with an initial focus on removing SNF from the shutdown reactor sites; and (4) develop the foundational information, resources, and capabilities needed to support future implementation decisions and actions. Near-term objectives are to (1) develop and maintain an integrated plan to accomplish the Administration's Strategy goals, (2) improve integration of storage as a planned part of the waste management system, (including evaluating standardization of dry cask storage systems), (3) develop and evaluate design options for an integrated waste management system, (4) develop and apply systems analyses to provide quantitative estimates of system impacts of utility actions' impacts on the system and inform future decisions, (5) prepare for the large-scale transportation of SNF and HLW, (with an initial focus on removing SNF from the shutdown reactor sites), and (6) establish and maintain a unified, integrated SNF database and analysis system to characterize the input to the waste management system. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Alternate Packaging Options for Remote-Handled Transuranic Waste Disposal at WIPP - 19093

Sandia National Laboratories/New Mexico (Sandia) began a remote-handled (RH) transuranic (TRU) repackaging campaign two weeks before the Waste Isolation Pilot Plant (WIPP) shutdown in February 2014. Since a contract with Nuclear Waste Partnerships (NWP), the WIPP Management and Operation (M and O) contractor, was already in place for certified visual examination (VE) and dose-to-curie (DTC) support, repackaging continued and was completed in October 2015. Twenty-five 55-gallon drums of RH TRU waste were packaged, awaiting certification and shipment to WIPP. However, in February 2017 at the National TRU Program (NTP) Corporate Board meeting, Sandia learned that RH TRU waste normally transported in removable lid canisters (RLCs) and 72-B casks was not being accepted for disposal when WIPP re-opened after the shutdown. One option, shielded container assemblies (SCAs), was available and approved for RH TRU waste disposal at the WIPP. Sandia made the decision to pursue the use of SCAs. The use of SCAs allow RH TRU waste to be shipped as contact-handled (CH) in HalfPACTs, emplaced on the floor of the WIPP instead of in boreholes in the walls, if the dose rate is less than 200 millirem/hour on contact. The RH TRU waste remains on the WIPP inventory as RH. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Migration to a fully-integrated control room - 349

The nuclear industry is faced with a pressing need to reduce operating costs in order to remain competitive with other forms of electric generation. While current industry initiatives are focused on incremental gains by streamlining work processes and reducing worker burdens, very few, if any, involve significant reductions in the base workload of a nuclear plant. Migration to a fully-integrated control room affords just such an opportunity. This report describes an NPP digital instrumentation and control (I and C) system design, along with its corresponding human systems interface (HSI) design, which achieves high plant availability, while selectively applying digital equipment redundancy and integrating traditional I and C system partitions, to minimize the digital equipment needed, and thereby minimize initial installation costs and ongoing operations and maintenance (O and M) costs. An additional key component of digital equipment minimization is optimization of the digital interface to conventional plant sensors and controlled plant devices (e.g., pumps, valves), which requires no disturbance to plant cables and results in eliminating thousands of electro-mechanical relay control components. The design is rooted in the highly integrated digital technology applied to the most recent advanced light water reactors under United States (U.S.) Nuclear Regulatory Commission (NRC) design certification review and approval. While this design employs technology from new plants, that technology has been adapted to the modernization of current operating NPPs. Hence, the highly integrated digital I and C and HSI design is referred to as the compact control room digital modernization (CDM). Along with minimal digital equipment, the corresponding HSI design employs compact operator workstations with visual display units (VDU) and soft controls for both the main control room (MCR) and remote shutdown room (RSR), to immediately improve human performance and ultimately facilitate operating staff reduction. Some advanced light water reactors have demonstrated this staff reduction through full scope simulator testing of normal and abnormal plant conditions, with U.S. licensed operating crews from currently-operating U.S. plants. Digital equipment minimization, compact operator workstations, and optimized plant component interfaces that require no disturbance to field cables, also facilitate implementation of the CDM over multiple phases, most of which can be conducted during normal refueling shutdown windows. This phased implementation method includes a novel transition of the I and C, MCR and RSR with an operations centered focus. This paper presents the CDM design, along with qualitative cost/benefit comparisons to currently operating plants with conventional analog I and C and HSI technology; this generation of I and C/HSI is referred to as a conventional control room (CCR). A CCR also includes NPPs that have been upgraded with digital systems, but retain the current partitions between plant I and C systems (i.e., no I and C integration) and retain the current plant system to control board configuration of current NPP control rooms. Even though these comparisons are qualitative, a compelling case is presented for the CDM. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2017

Integration of the ITER Tokamak C-model with the Building Model

The pellet injector system is designed to accommodate the requirement of continuous plasma refueling during ITER Tokamak operations. Through pellet flight tubes/injection lines, the pellets can directly reach the interior of the tokamak vacuum vessel. The pellet injector system is located at a port cell outside of the Tokamak bio-shield. The plasma radiation source in the tokamak vacuum vessel, together with the highly activated surrounding water in port cells, present major radiation challenges for the pellet injector system during ITER operation and after shutdown. These sources produce a radiation field encompassing the entire Tokamak complex. The ITER neutronics reference model C-model is an ITER radiation transport model. It represents a 40 deg. regular torus sector out to the bio-shield, which consists of 38 individual systems. The Monte Carlo N-Particle (MCNP) input of the C-model consists of more than a hundred thousand cell definitions and surface definitions, and over three hundred material compositions. The latest tokamak complex MCNP model is significantly more detailed than any previous releases. It consists of 36,863 cell definitions, 57,105 surface definitions, and a total of 21 different types of materials. The C-model and tokamak building model share the bio-shield as the boundary between them. In general, the radiation fields in the building due to the plasma source during tokamak operation are determined by generating the radiation source in the inner faces of the bio-shield. An inherent limitation of this approach is that it can only approximate the directions in which streaming paths are produced through the bio-shield plugs. To better estimate the shutdown dose rate for the pellet injector system due to the plasma source, we have decided to integrate the tokamak model (C-model) with the complex building model. Both MCNP models are provided by the ITER International Organization (IO) to the Oak Ridge National Laboratory (ORNL) Radiation Transport (RT) group.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Examining Two-Phase Flow with Optical Sensors

In lieu of the events that occurred at Fukushima in 2011, passive safety shutdowns for nuclear reactors has become a focus area for recent research. At Argonne National Laboratory this research has been simulated using air as a means of cooling down a experimental reactor core. This operates on the principals of natural convection and can be further studied in the report devised by Lisowski et al. in 2014. The air reactor cavity cooling system (RCCS) at Argonne National Laboratory is currently being transitioned into a water cooled system for passive safety reactor shutdown systems. In this system eight water pipes (1.5'' schedule 160 stainless steel) will transfer heat by means of natural convection. However, the primary network pipe will be 4 inch schedule 40 stainless steel pipe. The adiabatic chimney portion of this piping segment will be under analysis for void fraction. Based on the internal temperature of the water pipes, the characteristic regimes of two-phase flow that could exist inside of the pipe are an area of interest for experimental analysis. Determining which phase the water exists is instrumental to understanding the state of the processes inside of the water cooling pipe. Knowledge of these states allows for analysis of the efficiency of the RCCS as it shuts down. The state of two-phase flow can be measured by the analysis of the fluids void fraction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Recovery from loss of reductant feed to a U/Pu separation contactor, magnox reprocessing plant, Sellafield - 14217

Following a revealed sustained loss of reductant feed to the Primary Separation 4 (PS4) U/Pu separation contactor in the Magnox Reprocessing Plant in 2014 the solvent extraction cycles were safely shutdown. However, initial mass balance calculations indicated that over four times the typical mass of Pu could be present in PS4. Due to the loss of reductant feed, much of this Pu was expected to be in the solvent phase of Tributyl Phosphate and Odourless Kerosene (TBP/OK) where it would be difficult to observe with the neutron monitors located under PS4. Samples were taken of the solvent in PS4 which confirmed the presence of Pu in solvent; normally almost all of the Pu in PS4 is in the aqueous phase. The volume ratio of solvent to aqueous in the PS4 contactor is high. This meant that simply re-dosing PS4 with reductant could not be supported as a recovery option due to the potential to realise high Pu concentrations in the aqueous phase. This paper details the recovery options which were considered and the ultimate decision to recover PS4 using Gadolinium (Gd) poisoning and Gd-dosed reductant feeds. The recovery included: - The need to consider the impact of Gd poisoning on monitoring of the recovery with the PS4 neutron monitors; - Close support from criticality safety personnel on the plant at all times during the recovery; - Production of operator aids and short fit for purpose safety documentation to assist with the recovery. The recovery of PS4 proceeded safely and smoothly. There were no losses of Pu to raffinate streams despite the prolonged shutdown. The plant was then safely rundown to flush out residual Pu and Gd. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Advanced wind turbine control development using field test analysis for generator overspeed mitigation

Abstract Turbulent and gusty wind conditions can cause generator overspeed peaks to exceed a threshold that then lead to wind turbine shutdowns, which then decrease the energy production of the wind turbines. We derive so‐called “gust measures” that predict when generator overspeed peaks may occur. These gust measures are then used to develop advanced controllers to mitigate generator overspeed peaks so that wind turbines can operate more robustly in difficult wind conditions without exceeding generator overspeed thresholds that would lead to turbine shutdown events. The advanced controllers are demonstrated in nonlinear aeroelastic simulations using the open‐source wind turbine simulation tool OpenFAST. To increase the realism of the simulations, they are run using field‐replicated wind conditions and a wind turbine model based on data from an experimental field campaign on a downscaled demonstrator of a novel extreme‐scale, two‐bladed, downwind rotor design.

17 WIND ENERGY↗
DOE OSTI2020

Preliminary design of control rods in the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

Recent studies on Molten Salt Reactors (MSRs) showed that the excess reactivity at the beginning of the operation is large for many fueling strategies and must be compensated by a reactivity control system. The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). Three different initial fissile loadings are considered: 233 U, reactor-grade Pu, and transuranic (TRU) elements as a startup fuel. We applied six different absorbing materials to investigate the main operational and safety parameters in the SD-TMSR: natural B 4 C, enriched B 4 C with 90% 10 B, HfB 2 , HfH 1.62 , Eu 2 O 3 , and Gd 2 O 3 . The present work focuses on control rod design, integral and differential control rod worth, shutdown margin, and shadowing effects at steady-state. We employed the SERPENT-2 Monte-Carlo code to calculate the reactivity worth and analyze the performance of the reactivity control system. We showed that 233 U and reactor-grade Pu startup cores maintain adequate shutdown margin with all considered absorbers. Finally, this report proposes a design of control rod clusters that compensate the excess reactivity of the SD-TMSR loaded with different initial fissile material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Scalable modular dynamic molten salt reactor system model with decay heat

Nodal dynamic models allow rapid reduced-order simulations of complex systems. This paper presents a publicly available system dynamic model of a molten salt reactor system with a topology of the Molten Salt Reactor Experiment (MSRE), in MATLAB/Simulink, building upon a model previously published and validated with MSRE data. It adds scaling of nominal power, dynamic representation of decay heat generation, a generic decay heat removal system, improved documentation, modularity, and test cases. One hour of system transient runs in less than 10 min on a laptop. This model is intended for first order engineering calculations, developing insights into system transient behavior, equipment sizing, parameter sensitivity studies. The model contains an easy to use toolkit which can be customized and expanded to describe any molten salt reactor system. Finally, scenarios involving off-normal transients are presented. Relation between shutdown reactivity insertions, decay heat removal, core recriticality, and safe shutdown conditions are discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Thermal and electric multidomain dynamic model for integration of power grid distribution with behind-the-meter devices

As renewable energy sources like solar and wind power become more integrated into the grid, coordinated control of behind-the-meter devices is crucial for enhancing grid flexibility and reliability and for meeting cost targets, with standardized models being developed to support this transition. The increasing flexibility and uncertainty of integrated renewable energy grids, along with interactions between various subsystems, make traditional steady-state modeling insufficient to capture transient and dynamic behaviors. Current models (e.g., composite load and battery equivalent models) focus on thermodynamic or electrical characteristics but overlook critical electromechanical interactions. This limits the ability to share performance information for grid services and hampers fast dynamic simulations. In addition, motor stalling is usually triggered by a fault event and attributed to the characteristics of the mechanical torque of the motor, resulting in absorption of a large amount of reactive power during the stalling period. Further, this significant withdrawal of reactive power will deteriorate the dynamic voltage stability of power grids and cause delayed voltage recovery. Therefore, an in-depth modeling of the thermodynamics or mechanical torque is essential to study the impacts of the realistic torque characteristics of those behind-the-meter devices on power system voltage stability. This study developed a dynamic multidomain model for building HVAC systems, such as air-source heat pumps, to simulate their thermal and electrical responses to grid transients. The model can accurately predict power metrics with a mean absolute percentage error of 10%, by validating against with power system computer-aided design performance data. Case studies demonstrate the model capability of capturing the transient response to sudden voltage changes, rapid load fluctuations, and system shutdowns respectively. During a sudden voltage drop (30% for 0.1s), a fully loaded heat pump’s motor speed dropped, continued declining, and shut down after 3.6s, with severe power oscillations and a torque spike. A partially loaded unit experienced temporary oscillations but stabilized. Under higher building loads, compressor speed increased from 64% to 100%, with power and torque rising before stabilizing. In safety-triggered shutdowns, power decreased after minor fluctuations, and torque briefly spiked before dropping to zero.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2020

Implementation and verification of PyNE R2S with DAG-OpenMC

The mesh-based Rigorous-Two-Step (R2S) method has been widely used in the accurate estimation of the shutdown dose rate (SDR) of fusion systems. Several mesh-based R2S code has been implemented based on Monte Carlo particle transport code MCNP5 or DAG-MCNP5 and inventory calculation code such as FISPACT-II, ALARA or ACAB. Recently, the OpenMC, a community-developed Monte Carlo particle transport code, implemented the photon particle transport capability, which enables OpenMC to be used in R2S. In this paper, PyNE R2S is extended with the support of DAG-OpenMC, i.e., OpenMC with DAGMC. Modifications that allow PyNE to read the flux result in the state point file of OpenMC has been implemented to PyNE R2S workflow. A custom source routine of OpenMC that reads the photon source file from PyNE R2S and sampling photon particle in OpenMC photon transport has been implemented in OpenMC. With these modifications, PyNE R2S is now capable of performing R2S calculations with both DAG-MCNP5 and DAG-OpenMC. The ITER FNG dose rate benchmark problem has been used to validate the code. The FNG neutron source has been modified for neutron transport with DAG-OpenMC. The shutdown dose rate of 19 cooling times has been calculated and compared with the experimental data and the computational results of other R2S codes. The results of PyNE R2S with DAG-OpenMC show satisfactory agreement with the experimental data and other computational results. Therefore, we consider PyNE R2S with DAG-OpenMC is a reliable code to calculate the SDR of fusion systems.

fusion↗
DOE OSTI2025

Consequence analyses of sabotage-induced radiological releases in sodium-cooled fast microreactors

Analysis of three sodium-cooled fast microreactors (SFMs) with thermal powers of 10, 30, and 50 MWt showed that smaller reactors result in lower radiological consequences during a postulated sabotage-induced event because of their reduced core inventory. All SFMs used U-10Zr metal fuel enriched to 15 wt% high-assay low-enriched uranium and operated until their respective effective multiplication factor (k eff ) reduced to less than 1 or until the end of their operational lifespan. Sabotage scenarios were simulated at this point, when the fuel inventory within the core contains the highest-level of radioactivity. Radionuclide core inventories were calculated using the SCALE code at shutdown and 3 days post-shutdown. Dose consequence analyses were performed for three sabotage scenarios using the RASCAL tool. As microreactor developers plan for minimal on-site or complete off-site emergency response, it remains essential to evaluate their physical protection needs and potential hazards, including assessing postulated sabotage-induced events that could become more relevant. SFM licensees should identify a credible worst-case, major accident, estimate release source terms, and perform dose consequence analyses to evaluate site-specific physical protection measures. In conclusion, this recommendation supports a risk-informed, performance-based approach, aligning with applicable regulatory requirements, i.e., 10 CFR Parts 100 and 53 rulemaking in the United States.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Cells Respond to Distinct Nanoparticle Properties with Multiple Strategies As Revealed by Single-Cell RNA-Seq

The impact of distinct nanoparticle (NP) properties on cellular response and ultimately human health is unclear. This gap is partially due to experimental difficulties in achieving uniform NP loads in the studied cells, creating heterogeneous populations with some cells “overloaded” while other cells are loaded with few or no NPs. Yet gene expression studies have been conducted in the population as a whole, identifying generic responses, while missing unique responses due to signal averaging across many cells, each carrying different loads. Here we applied single-cell RNA-Seq to alveolar epithelial cells carrying defined loads of aminated or carboxylated quantum dots (QDs), showing higher or lower toxicity, respectively. Interestingly, cells carrying lower loads responded with multiple strategies, mostly with upregulated processes, which were nonetheless coherent and unique to each QD type. In contrast, cells carrying higher loads responded more uniformly, with mostly downregulated processes that were shared across QD types. Strategies unique to aminated QDs showed strong upregulation of stress responses, coupled in some cases with regulation of cell cycle, protein synthesis and organelle activities. In contrast, strategies unique to carboxylated QDs showed upregulation of DNA repair and RNA activities, and decreased regulation of cell division, coupled in some cases with upregulation of stress responses and ATP related functions. Together, our studies suggest scenarios where higher NP loads lock cells into uniform responses, mostly shutdown of cellular processes, whereas lower loads allow for unique responses to each NP type that are more diversified, proactive defenses or repairs of the NP insults.

Environmental Molecular Sciences Laboratory↗