Engineering topics

Bays, Samuel E

Publications and source records attributed to Bays, Samuel E.

16 records

DOE OSTI2024

Streaming in a Nuclear Grade Sandwich Composite for Microreactor Shielding

The NGSC is a new approach to develop a shield structure for microreactors which combines the biological shielding with the reactor pressure vessel. Six layers of SS316 skins and core materials are present in the NGSC, where the core materials are reduce the neutron and gamma dose. Previous work has examined how a simplified NGSC can be optimized for cost, dose, and weight. This work explored the inclusion of SS36 ribs, which helps maintain the structural integrity of the NGSC, affects the transportation of radiation through the NGSC. For B$_4$C layers, the addition of ribs reduces neutron absorption but increase photon absorption. For WB$_4$-cermet layers, the addition of ribs reduces neutron absorption and reduces photon absorption.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

A317_Fast and Thermal Neutron Spectrum Dosimetry Measurements in the Advanced Test Reactor large-B and small-I Positions Following the Sixth Core Internals Change-out

The Advanced Test Reactor (ATR) has a wide variety of irradiation positions that have had experiments that were developed by users from around the world. Most experiment irradiations rely on thoroughly benchmarked numerical models. However, some irradiation positions in ATR are not as well- characterized and are complicated by spectral perturbations from control cylinder orientation. The “small I” and “large-B” irradiation positions are located nearby control cylinders and suffer from these flux perturbations from control cylinder orientations that change during an irradiation cycle to maintain the desired core power distribution. Models predicted that these types of position would exhibit both spectral shifts and amplitude changes in neutron flux, but few measurements have been conducted to benchmark these predictions. Recently, requalification testing was performed to confirm the operational readiness of the ATR following the completion of the Core Internals Change-out (CIC). These tests provided a unique opportunity to validate the analytical methods used to simulate the ATR because nearly all components in the reactor were in a clean as-built state, significantly reducing modelling uncertainties. One subset of the testing included characterization of the fast and thermal neutron flux in the “small-I” and “large-B” positions using silver, cobalt, and nickel neutron dosimetry. In contrast to typical irradiation cycles, the control cylinders were held in position during the post-CIC nuclear testing. The specific activity of these dosimeter wires was measured following two low-power tests with different control cylinder positions.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2024

Mass Optimization of a Multilayered Shield for Transportable Microreactors

The ability to easily transport microreactors is a major selling point for deploying microreactors to remote areas. However, this creates a unique shielding challenge, especially when the microreactor is being shipped after irradiation. A traditional reactor configuration utilizes a separate biological shield and pressure vessel to meet radiological shielding and pressure needs. The limited space available for transportable microreactors for both shielding and pressure vessels requires a revised assessment of separating out the biological shield and pressure vessel. To address these concerns, we examine a nuclear-grade sandwich composite (NGSC) that combines the reactor pressure vessel and biological shielding functions into a single component. Through a series of optimization problems for both transportation and operational use cases, the NGSC is able to minimize dose, minimize the vessel cost, and ensure that weight requirements are met for transportation. Initial results show that using a tungsten-tetraboride cermet in the first two layers of a six-layer NGSC provides adequate shielding for both use cases. These results show promise that an NGSC has enough overlap between operational and transportation cases to help reduce the design space for future analysis and assessment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Preliminary Analysis of a Nuclear Grade Sandwich Composite for Transportable Micro-Reactor Shielding

Transportable microreactors pose a unique shielding challenge compared with conventional reactors. Many vendors are seeking to transport a microreactor to a site, operate it for a number of years, and then transport it to a secondary site. This involves ensuring the reactor has sufficient shielding during operations and post-irradiation transportation to ensure the safety of those around it, all while ensuring it can be transported via traditional means. This work examines the use of a nuclear grade sandwich composite (NGSC) which seeks to combine the reactor pressure vessel and biological shielding into a single unit. Through this work, it was determined that the NGSC was sufficient to adequately reduce neutron and gamma current exiting the structure to warrant further investigation as a potential shielding option for microreactors.

42 ENGINEERING↗

DOE OSTI2024

Utilizing a Virtual Sodium-Cooled Fast Reactor Digital Twin to Aid in Diversion Pathway Analysis for International Safeguards Applications

We report digital twin technology has the potential to improve the effectiveness of international safeguards inspectors by providing a tool which can: first, perform an accurate diversion path analysis, identify their indicators, and required sensors to detect them; and second, monitor facilities in real-time using critical data streams that benefit from this safeguards-by-design approach. Safeguards inspectors are required to visit facilities and verify the nuclear material to ensure no diversion has taken place and detect misuse of the facility; however, this analysis and verification effort is time consuming, and with limited funding it is imperative that time spent at a nuclear facility is focused on key areas. A virtual digital twin of three prototypic sodium fast reactors was developed, where diversion and misuse scenarios were explored to determine how a digital twin could provide inspectors with an understanding of how proliferation may occur and where the most likely areas for proliferation would be. For each of the three reactors, an optimization algorithm was able to find core designs which would be difficult to detect via sensors alone; however, the use of a machine learning adapter provided by the digital twin was able to show general trends in where proliferation as likely to take place.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Startup Physics Testing of Advanced Reactors

This report documents the startup physics testing from initial fuel loading through ascension to full power for past advanced reactor startup physics testing programs. The review includes an assessment of what nuclear physics data was measured, why this data was measured, how was the measurement made, and the agreement with predictive reactor performance calculations of that time. The purpose of this review is to establish historical precedence for test inclusion for future advanced reactors planned for demonstration at the National Reactor Innovation Center (NRIC). The historical review includes reactor designs considered to be significantly different from current light-water reactor designs, or use simplified, inherent, passive, or other innovative means to accomplish their safety functions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Shielding of Transportable Fission Batteries

This is a generic presentation on the basics of radiation shielding as well as potential issues and possible solutions facing shielding fission batteries. It is to be presented at the Lightweight Materials Workshop on Nov 8 & 9 in EROB-159.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Towards an Agent-Based Blackboard System for Reactor Design Optimization

The field of reactor design is rich with opportunities for applications of computational optimization algorithms; these applications can range from preliminary core design to reactor shuffling patterns. Many of these schemes rely on sets of previously generated solutions (sometimes referred to as “generations”) to inform future decisions. While it is important to build upon prior knowledge, this process requires a full generation of solutions to be formed before future solutions can be examined. Rather than relying on a generational scheme to perform an optimization, we propose using an agent-based approach in conjunction with a blackboard framework for performing reactor design optimizations. Utilizing an agent-based approach allows agents to perform tasks independently, while retaining the ability to build off of previous solutions. We develop an agent-based blackboard system (ABBS) for determining the Pareto front (PF) in sodium fast reactor design optimization problems and compared this with the Non-Dominated Sorting Genetic Algorithm II (NSGA-II). Our goal is to evaluate the viability of the ABBS in producing a PF that is comparable with the NSGA-II algorithm. The design space consists of the fuel height, fuel smear, and plutonium fraction in the core, and we seek to minimize the reactivity swing and plutonium mass, while maximizing the burnup. The diversity, coverage, and spread of the PFs generated by the two methods are examined, and the ABBS is able to converge to the same PF as the NSGA-II algorithm. These results show that the ABBS is able to find optimal designs that are similar to those found by the NSGA-II algorithm. We conclude our study by applying the ABBS to the design of a sodium-cooled fast reactor to dispose of weapons-grade plutonium. The ABBS finds a core design that can burn upwards of 17.5 kg of weapons-grade plutonium per year and degrade an additional 195 kg of weapons-grade plutonium per year into non-weapons-grade material.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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 OSTI2019

MCNP AND CFD MODELING OF A SODIUM FAST REACTOR SUB-ASSEMBLY CHANNEL TO CAPTURE LOCALIZED TEMPERATURE PEAKING

Typical evaluation tools for sodium-cooled fast reactors (SFR) use simplified neutronic/thermal models to efficiently compute the power and temperature distributions within the core. However, their modeling capability can be limited in extreme cases such as when sharp inter-assembly power gradients occur, or for non-standard SFR assembly geometries. Higher fidelity tools that rely on Monte Carlo methods for neutron transport, and Computational Fluid Dynamics (CFD) for thermal analysis, are better suited in these instances. The higher accuracy comes at the cost of higher computational expense. MCNP and STAR CCM+ were selected for this analysis. While MCNP is capable of generating pin-level power distributions for a whole-core SFR model, it is prohibitively expensive to then conduct CFD analysis for every single assembly. A semi-empirical model was therefore developed to rapidly estimate pin temperature distributions throughout the whole core. It relies on a geometric correction factor for turbulent flow-mixing. The correction factor is obtained from a CFD simulation of a single fuel assembly with a uniform power distribution and an adiabatic boundary condition. Without using CFD, each sub-channel outlet temperature is calculated separately using 1-D heat-balance and Nusselt number correlations and the MCNP calculated pin-powers. The rudimentary 1-D model is then corrected for the flow-mixing effect based on the CFD data. The temperature calculation is performed by a code written in the Python language called CFD Monte-Carlo Developed Reconstruction of Sub-channel at a Quasi-Representative Level (CMDR-SQRL). CMDR-SQRL was shown to adequately predict CFD evaluated temperature distribution for a wide range of SFR configurations, including an extreme case with fuel assemblies nearby containing moderating material. The agreement demonstrates how geometric and power-driven effects can be decoupled within an SFR assembly. CMDR-SQRL is intended to be mainly used for scoping studies to provide core-wide evaluations efficiently.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Preliminary Evaluation of Fuel Management Strategies for the Versatile Test Reactor (VTR)

Different fuel management approaches have been investigated for the Versatile Test Reactor (VTR). The main design variables were the number of batches, the number of (orifice) zones, and the loading pattern within the core. The optimization criteria are maximizing the fast flux at the centermost position, minimizing the peak assembly power, and reducing fuel consumption. Design constraints were mainly driven by material limits such as peak clad temperature, and maximum fuel burnup. The exploratory study determined that active fuel shuffling only brought very limited benefits. A three-orifice region was found to provide enough flexibility to ensure thermal criteria were met. Lastly, a flexible batch loading termed ‘5-6-8’ was identified as a promising management strategy from a resource utilization standpoint.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time-dependent measures of performance including uranium usage, TRU inventory, and radiotoxicity to evaluate the complex impacts of transition from the current uranium-fueled LWR system, and other more realistic impacts that may not be intuited from the time-independent steady-state conditions of the end-state fuel cycle. These analyses were performed using the Verifiable Fuel Cycle Simulation Model VISION. Compared with static calculations, dynamic results paint a different picture of option space and the urgency of starting a FR fleet. For example, in a static analysis, there is a sharp increase in uranium utilization as CR exceeds 1.0 (burner versus breeder). However, in dynamic analyses that examine uranium use over the next 1 to 2 centuries, behavior as CR crosses the 1.0 threshold is smooth, and other parameters such as the time required outside of reactors to recycle fuel become important. Overall, we find that there is no unambiguously superior value of TRU CR; preferences depend on the relative importance of different fuel cycle system objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent of the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2011

Multi-Reactor Transmutation Analysis Utility (MRTAU,alpha1): Verification

Multi-Reactor Transmutation Utility (MRTAU) is a general depletion/decay algorithm under development at INL to support quick assessment of off-normal fuel cycle scenarios of similar nature to well studied reactor and fuel cycle concepts for which isotopic and cross-section data exists. MRTAU has been used in the past for scoping calculations to determine actinide composition evolution over the course of multiple recycles in Light Water Reactor Mixed Oxide and Sodium cooled Fast Reactor. In these applications, various actinide partitioning scenarios of interest were considered. The code has recently been expanded to include fission product generation, depletion and isotopic evolution over multiple recycles. The capability was added to investigate potential partial separations and/or limited recycling technologies such as Melt-Refining, AIROX, DUPIC or other fuel recycle technology where the recycled fuel stream is not completely decontaminated of fission products prior to being re-irradiated in a subsequent reactor pass. This report documents the code's solution methodology and algorithm as well as its solution accuracy compared to the SCALE6.0 software suite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel

An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗