SEARCH · Engineering Papers

Results for “Reasoning”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,009 records · Page 56

All sources NASA NTRS DOE OSTI

DOE OSTI2016

FCRD Advanced Reactor (Transmutation) Fuels Handbook

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, important to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, the handbook attempts to provide information about how well the property is known and how much variation exists between measurements. Although it includes some results from models, its primary focus is experimental data. The Handbook is organized in two sections: one with information about the U-Pu-Zr ternary and one with information about other elements and binary and vi ternary alloys in the U-Np-Pu-Am-La-Ce-Pr-Nd-Zr system. Within each section, information about elements is presented first, followed by information about binary alloys, then information about ternary alloys. The order in which the elements in each alloy are mentioned follows the order in the first sentence of this paragraph. Much of the information on the U-Pu-Zr system repeats information from the FCRD Transmutation Fuels Handbook 2015. Most of the other data has been published elsewhere (although scattered throughout numerous references, some quite obscure); however, some data from Idaho National Laboratory is presented here for the first time. As the FCRD programmatic mission evolves, future editions of this handbook will begin to include other advanced reactor fuel designs and compositions. Hence, the title of the handbook will transition to the Advanced Reactor Fuels Handbook.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Crystallization and Thermoelectric Transport in Semiconductor Micro- and Nanostructures Under Extreme Conditions

This project focused on thermoelectric transport in semiconductor micro and nanostructures where moderate and typical operating voltages and currents lead to extreme thermal gradients and current densities. Models that describe behavior of semiconducting materials typically assume an equilibrium condition or slight deviations from it. In these cases the generation-recombination processes are assumed to have reached a local equilibrium for a given temperature. Hence, free carrier concentrations and their mobilities, band-gap, thermal conductivity, thermoelectric properties, mobility of atoms and mechanical properties of the material, can be described as a function of temperature. In the case of PN junctions under electrical bias, carrier concentrations can change up to ~ 1020 cm-3 and a drift-diffusion approximation is typically used to obtain the carrier concentrations while assuming that the material properties do not change. In non-equilibrium conditions, the assumption that the material properties remain the same may not be valid. While the increased conduction-band electron concentration may not have a drastic effect on the material, large hole concentration is expected to soften the material as ‘a hole’ comes into existence as a broken bond in the lattice. As the hole density approaches 1022 cm-3, the number of bonds holding the lattice together is significantly reduced, making it easier to break additional bonds, reduce band-gap and inhibit phonon transport. As these holes move away from where they were generated, local properties are expected to deviate significantly from the equilibrium case. Hence, temperature alone is not sufficient to describe the behavior of the material. The behavior of the solid material close to a molten region (liquid-solid interfaces) is also expected to deviate from the equilibrium case as a function of hole injection rate, which can be drastically increased or decreased in the presence of an electric field. In the past years we have investigated the possible thermoelectric explanation of asymmetric melting of self-heated Si micro-structures using equilibrium materials’ properties that exist in the literature. We have analyzed the contribution of the electrons and the holes and identified the generation-transport-recombination of minority carriers (GTR) as the reason for an extreme change in the thermal profile in presence of strong generation and electric field. A more complete analysis required construction of models that capture the individual generation and recombination processes to understand the thermal profile as well as the possibility of electronic softening and non-equilibrium melting of the structure below melting temperature. The possibility of melting a material at a lower temperature breaks the correlation between the atomic mobility and the kinetic energy in the system for a given temperature and may allow alternative growth processes. This may also be the mechanism behind ‘amorphization-without-melting in layered structures heated with laser pulses’ that has been reported earlier. The conventional models for semiconductors are constructed for low temperature operation and their projections to higher temperatures do not yield reasonable carrier concentrations. Using these models, the free hole concentrations are calculated to be on the order of 1019 cm-3 at melting, which also do not correlate well with the latent heat of fusion. The melt is expected to correspond to broken bond concentrations on the order of the atomic density (~5x1022 cm-3 for Silicon). Hence, using conventional models the thermoelectric contribution expected from the GTR process is estimated to be much smaller than it likely is. Our work focused on improving the computational models and electrical characterization of materials and devices to better understand thermoelectric trabsport under extremen thermal gradients and current densities. Specifically, during this project, we have - Expanded our computational models to include self-consistent solution of Poisson charge equation (together with current and heat equations currently solved) for improved accuracy of role of bipolar conduction, - Developed a crystallization model incorporating experimentally determined nucleation rates and growth velocities to enable simulation of grain growth, growth-from-melt, filament formation and retention, - Performed high-temperature characterization of relevant materials (including metal contacts, interfacial and insulation layers); electrical and thermal conductivities, Seebeck coefficient, carrier mobility and concentration, - Performed High-speed device-level characterization of metastable amorphous and crystalline phases, crystallization and amorphization dynamics, melting and crystalline growth-from-melt, - Observed and characterized formation of microplasmas in electrically stressed ZnO nanoforests.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Hydrolysis and Radiation Chemistry of the DGAs

This document was prepared to meet FCR&D level 3 milestone M3FT-16IN030104052, “continue the study of the short chain compounds,” and, “ study the radiolysis of the non-symmetrical DGA (D3DODGA)….,” under the Radiation Chemistry FCR&D work package. Toward these goals, the short chain DGA, tetraethyldiglycolamide (TEDGA) was investigated for its hydrolytic stability in HNO3 solution, with comparisons to the less oxidizing mineral acid HCl. Initial gamma-irradiations were also performed on DGA solutions, to inform a more detailed investigation of several short chain compounds anticipated for FY17. The hydrolytic and radiolysis behavior of TEDGA is of interest for two reasons. First, previous long chain DGA radiolysis was conducted in dodecane solution since the long chain compounds are soluble in that diluent, and new radiation chemistry is expected in the aqueous environment due to reactions with •OH radical. The second reason is that this water-soluble DGA has been proposed for use as a stripping or holdback agent in both European and American fuel cycle scenarios. Therefore, this work was performed in collaboration Forschungszentrum Jülich (FZJ), and the European SACSESS program. Additionally, results are presented here regarding the radiation of chemistry the non-symmetrical DGA didodecyldioctyldiglycolamide (D3DODGA), for comparison to previous work with symmetrical DGAs such as TODGA and TEHDGA. This was also conducted in collaboration with researchers in the SACSESS program.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2012

The Design of the Parallel Box Cavity with New Radial RF Coupler Design

The design of a E/H parallel cavity has been described. It has been designed with the RF simulation program HFSS. HFSS has been very successfully been used to design and fabricate the Orthogonal E/H box cavity. The main reason for this design choice is to move the RF coupling aperture to surface of minimum electric field gradient. It is suspected that this change will remove the damaged observed in the coupling area experimentally observed in the LBL Pillbox cavity. Another reason for this change is that it corresponds to the location of the coupling aperture in a future real experimental cooling channel. Also, with this design the coupling factor to the cavity can be made adjustable externally, a very useful feature.

43 PARTICLE ACCELERATORS↗

DOE OSTI2018

Feasibility Study of Large-Scale Energy Storage in the Earth

Energy storage systems on a large scale are needed when there is a mismatch between electricity generation and demand rates. The mismatch may be due to a variety of reasons: 1. Generation rates of solar are cyclic and are often out of phase with the demand cycles. 2. Generation rates of wind are unpredictable. Furthermore, high-efficiency, low-carbon power plants (e.g., nuclear) typically have high capital cost and low operating cost. Such plants are designed for continuous base-load operation and require a high capacity factor for economic viability (Forsberg, 2017). For these reasons, it is advantageous to be able to store energy when not needed and use it when needed. Our proposed process, called the Air Earth Battery, has been developed to accomplish the task of energy storage and dispatch, using a combination of compressors, expanders, heat exchangers, and subsurface porous, permeable rock formations. This technology utilizes two issued U.S. Patents (Buscheck, 2017a; 2017b), one filed U.S. Patent Application (Buscheck, 2017c), and one filed U.S. Provisional Patent Application (Buscheck and Upadhye, 2018).

25 ENERGY STORAGE↗

DOE OSTI2017

Final Report for the Contractor Readiness Assessment of the Transient Reactor Test Facility Resumption of Transient Testing Activity at the Idaho National Laboratory

The purpose of this review was to verify readiness for the TREAT restart of transient testing activity at the INL and inform the startup authorization authority decision-making for startup authorization by providing a deliberate and managed determination of the state of readiness preparation and determination of whether operations can be performed safely. Verification of readiness included assuring that equipment operability, procedure viability, and personnel knowledge, skills, and performance are adequate for operations. Because the TREAT (including restart of transient testing activity) is a hazard category-2 nuclear facility and because reactor operation and transient testing activities were suspended for programmatic and funding reasons in excess of a year, DOE O 425.1D required that both a contractor and DOE RA be satisfactorily completed prior to startup of the TREAT and restart of transient testing activity. To this end, DOE directed that a CRA and federal RA be conducted. The CRA IP defined the depth of the CRA through the objectives and CRADs, delineated in Appendix A of the IP; established roles and responsibilities for CRA team activities; and established processes, practices, and protocols for preparation, conduct, documentation, and reporting of the CRA. The approaches and methodologies identified in the IP were used to verify that the BEA management team has established SSCs, developed personnel capabilities and competencies, and implemented safety features, programs, and procedures in a manner that provides reasonable assurance for adequate protection of workers, the public, and the environment in preparation for and execution of the TREAT restart of transient testing activity. This CRA was intended to confirm that (1) the facility is in a state of readiness to safely conduct operations in accordance with the safety basis; and (2) the management control programs are in place to ensure safe operations can be sustained. This equated to mature implementation of the integrated safety management (ISM) in conjunction with implementation of the individual facility authorization basis. This CRA was structured to verify the readiness to conduct restart of transient testing activity safely, and that facility programs were in place to support the restart of transient testing activity.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Development of a LOCA Experimental Benchmark for BISON - Milestone L2:FMC.P15.08

Since 2013 CASL has been planning and working towards a PWR LOCA Challenge Problem. Since BISON is the CASL fuel rod simulation platform, it plays a central role in that activity. The goal of the FY-17 CASL milestone reported here was to develop an Experimental Benchmark for LOCA analysis using BISON. Activities encompassed both code development and validation. This report provides a current snapshot of BISON’s capability for LOCA behavior including: 1) a summary of code extensions to facilitate accident analysis, 2) a series of separate effects tests, and 3) initial code validation to complex integral rod LOCA behavior. Conclusions from each of these main activities are summarized below. The key material and behavior models required to address transient high-temperature phenomena occurring during LOCAs in a standard PWR have now been implemented in BISON. These apply specifically to UO2 fuel, Zircaloy cladding and water coolant. During FY-17 important new capability to address axial UO2 fuel relocation (for 1.5D geometry) and account for oxidation energy deposition in cladding, were included. Planned future development efforts include extending the axial fuel relocation model from 1.5D to 2D/3D and improving BISON’s fission gas release model to include transient gas release associated with the high burnup structure (HBS) in high burnup fuel. Additionally, investigation of potential anisotropic creep behavior in Zircaloy cladding is planned, assuming sufficient experimental data are available to support this effort. A substantial number of separate effects validation cases (42 tests from 3 experimental series) have been completed to compare BISON predictions to measured ballooning and burst behavior for Zircaloy cladding. Such experiments include a wide variety of pressures, temperatures and loading rates. In general, BISON predictions of burst temperature, pressure and burst time are very reasonable. For one experimental series, however, involving both very high temperatures and strain rates, BISON systematically over-predicts the cladding hoop strain. Investigation of this discrepancy is an important and planned activity. With another experiment an effort was made to investigate 3D cladding response due to an azimuthal temperature variation. Results indicated 3D effects are potentially important in fuel rod analysis during LOCAs and will be further investigated in the future. BISON validation to a series of integral fuel rod experiments has also been completed. These experiments involve all fuel and cladding phenomena relevant to LOCA conditions, and can include complexities associated with irradiated fuel relative to fresh fuel. Such experiments also generally include complex thermal-hydraulic boundary conditions. Four experiments (6 rods) have been considered to date including simulated fuel (ZrO2) and both fresh and high-burnup UO2. Test rods ranged from rodlets to full length commercial PWR fuel rods. As with the separate effects experiments, BISON predictions of burst temperature, pressure and burst time are generally very reasonable. Comparisons to cladding peak strain and rod outer diameter axial profiles are less satisfactory, and identify material models and possibly modeling approximations (e.g., 2D-RZ vs 3D geometries) requiring additional investigation. Validation of BISON for integral rod LOCA behavior is by no means complete, with additional cases planned.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Implementation of ANL’s Mechanics Based Evolutionary Fatigue Modeling Through ABAQUS-WARP3D Based High-Performance Computing Framework

This report presents an update on the environmental fatigue research that is being conducted at Argonne National Laboratory in support of the Department of Energy’s Light Water Reactor Sustainability (LWRS) program. Argonne is developing a fully mechanistic fatigue evaluation approach without using empirical fatigue (S~N) curves. This approach is based on the fundamental concept of the time evolution of progressive fatigue damage rather than on the conventional S~N curve approaches using end-of-life data. In FY 2017, we performed extensive validation of this approach with respect to fatigue test data for 316 stainless steel [1]. This validation was performed for different loading cases, including constant, variable, and random amplitude. In the present FY 2018 semi-annual report, we present the further advances of Argonne’s environmental fatigue research work in the context for more practical applications. In this report, we discuss a methodology for fully mechanistic (i.e., not using S~N curves) fatigue life evaluation of reactor components subjected to realistic loading cycles, namely, design-basis loading cycles. The loading cycles include plant heat-up, full-power, and cool-down operations. As a test case, we considered a typical pressurized water reactor surge line, which is made of 316 SS. To perform the fatigue simulation for thousands of fatigue cycles in a computationally cost effective way, we modified our previous desktop-based finite element (FE) modeling approach to work in a high-performance computing (HPC) framework. For the HPC implementation, we developed a hybrid framework based on commercial FE software (ABAQUS), open-source FE software (WARP3D), and Argonne-developed evolutionary cyclic-plasticity modeling methods. We validated this HPC-based cycle-by-cycle damage model for the entire fatigue life of a Pressurized Water Reactor (PWR) surge line (SL) pipe with respect to assumed loading cycles. The simulated fatigue life was found to be 5855 cycles, which is 85% accurate as compared to the corresponding small-specimen-based experimental fatigue life (6914 cycles). Also, the simulated stress history captures the cyclic hardening and softening behavior of the material for entire fatigue cycles. The FE simulation of the PWR SL pipe was conducted in a reasonable time of 12.5 days. These results show the promise that a fully mechanistic (not using S~N curves) fatigue life evaluation of a safety-critical nuclear reactor component (or even other safety critical components like those in aircraft, aero-engines, etc.) is possible. We anticipate that this type of methodology will drastically reduce the uncertainly associated with conventional fatigue life estimates based on empirical S~N curves. We also proposed an FE model that is based on a hybrid full-component and single-element approach and that can readily be used by industry if HPC resources are not available. In this approach, a single-cycle FE simulation has to be performed first for the required loading cycle. Then, the resulting strain/stress profile at the hotspot (highest strain/stress location) is used as input strain/stress loading to the single-element FE model. The single-element FE model is run under multiple fatigue cycles as long as the failure criteria are achieved. The cycle number at failure gives the fatigue life of the component in question. This cost-effective framework does not require expensive HPC hardware and can be performed in a reasonable time by using a desktop computer. We demonstrated this approach with respect to PWR SL fatigue evaluation under two realistic design-basis loading cycles (simple and detailed). The associated results were validated with respect to test data obtained through fatigue testing of uniaxial specimens under the in-air condition at 300 °C. The approach is in the development stage and requires further improvement and validation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Local Radiation Shielding Design Methodology

The NSLS-II accelerators are housed within a radiation enclosure of normal concrete (and in some places high density concrete) that has been designed to reduce the radiation generated from an assumed operational beam loss or bulk shielding current (BCS) from each accelerator. The beam loss model used to estimate the thickness of this bulk shielding, assumed a beam energy at the expected operating energy and a current loss based on reasonable assumptions for the fraction of the expected operating current. The thickness of the shield was determined to be the amount required to reduce the radiation outside the shield to a maximum dose rate of <0.5mrem/hr. for occupied areas for these normal operational losses. These assumed beam losses and the bulk shielding thicknesses required are given in reference [1]. Abnormal operating conditions generating higher than normal beam losses must also be considered, and additional engineering controls will be required depending on the magnitude of the radiation levels generated during the abnormal condition. Additional local shielding subject to configuration control may be utilized to reduce radiation levels associated with these abnormal loss conditions. Interlocked area radiation monitors are another means to prevent continued operation with higher than an acceptable level. The NSLS-II Shielding Policy requires that an engineered system be utilized to mitigate an abnormal condition producing radiation levels greater than 100 mR/h, and that two systems are required if the projected radiation levels during a fault condition are > 2000 mR/h. It is intended that shielding limits the dose rate to < 0.5 mR/h in occupied areas for normal operational losses, and the combination of administrative and engineered controls for abnormal losses will maintain radiation exposure to trained personnel working at the NSLS-II facility to as low as reasonably achievable. It is anticipated that all individual exposures will be well within the planned administrative control level of 100mrem/year. This technical note presents the method used to identify these potential high beam loss locations, evaluate the potential radiation risk these losses create outside the enclosure shield and the design process that specifies the required local radiation shielding needed to reduce this risk below the maximum dose rate of the bulk shield design, listed above. If shielding can’t be adequately provided or if other fault initiated losses could occur, then the necessary engineering controls will be indicated that will limit the energy, current or duration of the beam transport that constitute this high beam loss risk.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Sampling for Confidence: Guidance for Evaluating the Quality of Canister Sampling Plans

There is a need for an inspection frequency that demonstrates reasonable assurance that the confinement function of dual-purpose-canisters (DPCs) is maintained as these DPCs are returned to storage service following transportation. The DPCs contain spent fuel from shutdown reactor sites. The currently-proposed 100% DPC inspection frequency is overly conservative and unnecessary to demonstrate reasonable assurance that the confinement function of all DPCs is maintained for return to storage service. Testing all DPCs does not appropriately reflect the very low risk of leakage (i.e., low probability based on design and low consequences based on analysis). Currently, there are 1066 DPCs forecast to be loaded at reactor sites which either are shutdown or plan to cease operations by 2025. These DPCs may be transported by rail in sets of 3 to 5 DPCs per shipment. Assuming each rail consist originates at a single reactor site, an alternative proposal to test one canister/consist/site would result in 356 to 214 canisters, respectively, being tested given the population of 1,066 canisters This Savannah River National Laboratory (SRNL) report outlines statistically-based, random sampling plans that provide guidance on the number of samples of these canisters that are needed to establish bounds on the quality of the outcome of such sampling. The quality of each proposed sampling plan is expressed in the form of a statement that the number of undetected, unacceptable canisters in the population is less than some number, call it D, at a high level of confidence. In addition, evaluations using a similar approach for random sampling at the proposed levels of 214 and 356 canisters are provided.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Sampling for Confidence: Guidance for Evaluating the Quality of Canister Sampling Plans

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Rapid Crossing of a Resonance During Tune Modulation by Synchrotron Oscillations

It had become a standard practice to constrain the particle’s tune footprint while designing the ring lattice so that the particle tunes fit between harmful resonances that limit ring dynamic aperture (DA). This approach, known as “tune confinement”, puts tight limits on the magnitude of the tune shifts with amplitude and with momentum. The latter requires labor-intensive optimization of the off-energy dynamic aperture and control over the corresponding tune footprint for the large energy deviations and is a prerequisite to achieving reasonable lifetime if it is dominated by the Touschek scattering. As nonlinearities of the modern ring lattices are much enhanced as compared with the previous generation of synchrotron light sources, it is becoming more and more difficult to keep the off-momentum tune footprint confined. While one could reach reasonable on-energy DA suitable for injection into the machine, the off-energy DA optimization is much more complex and needs to satisfy more constraints so to reach large energy acceptance and secure decent lifetime at the high ring current. In certain cases when the lattice solution with the confined off-energy tune footprint cannot be found, one may ask whether crossing of a resonance stopband from off-momentum particle will lead to loss of this particle. Intuitively, if the particle crosses the stopband quickly, one may expect that the betatron oscillation amplitude will not increase substantially thereby keeping the particle within the machine acceptance. At the same time the tight tolerances of the modern lattice elements as designed and produced afford much narrower resonance stopbands being compared with the machines built decades ago. Under these circumstances the question above may be asked as what are the conditions for operating a ring with a large chromatic tune footprint, which crosses resonances, while not observing particle losses. In this paper we try to answer the questions above by studying the model of a particle with large initial momentum deviation (as, for example, driven by the Touschek scattering event), which crosses stopband of a half-integer resonance while executing synchrotron oscillations in the highly chromatic ring lattice.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Gamma Ray Spectrometry Assay Results of 242 Pu Samples at the Chemistry and Metallurgy Research Facility

Thirty individual items at the Chemistry and Metallurgy Research (CMR) facility with accountable amounts of 242 Pu required assay by gamma ray spectrometry in order to establish their masses with reasonable precision. There were two phases to this measurement campaign: 1) a preliminary screening phase in which each item was measured for a short 600 second count time to establish the relative masses between similar or like items, and 2) a second phase in which groups of items were counted for a considerably longer time in order to establish the group mass of 242 Pu in a reasonable time frame. The data from the two phases were assessed in a manner that allowed the project team to assign a 242 Pu mass to each individual item in the inventory, with the exception of three column tubes that could not be measured individually because of radiological contamination issues. The project team performed technical analyses with the SNAP™ analysis routine, PC-FRAM plutonium isotopic routine, and the NEN-2 PeakEasy routine. When total measurement uncertainties are factored in, the individual mass quantities calculated for all samples are statistically equivalent to the expected mass of 0.88 grams of total plutonium per sample.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Local Radiation Shielding Design Methodology

The NSLS-II accelerators are housed within a radiation enclosure of normal concrete (and in some places high density concrete) that has been designed to reduce the radiation generated from an assumed operational beam loss or bulk shielding current (BCS) from each accelerator. The beam loss model used to estimate the thickness of this bulk shielding, assumed a beam energy at the expected operating energy and a current loss based on reasonable assumptions for the fraction of the expected operating current. The thickness of the shield was determined to be the amount required to reduce the radiation outside the shield to a maximum dose rate of <0.5mrem/hr. for occupied areas for these normal operational losses. These assumed beam losses and the bulk shielding thicknesses required are given in reference. Abnormal operating conditions generating higher than normal beam losses must also be considered, and additional engineering controls will be required depending on the magnitude of the radiation levels generated during the abnormal condition. Additional local shielding subject to configuration control may be utilized to reduce radiation levels associated with these abnormal loss conditions. Interlocked area radiation monitors are another means to prevent continued operation with higher than an acceptable level. The NSLS-II Shielding Policy requires that an engineered system be utilized to mitigate an abnormal condition producing radiation levels greater than 100 mR/h, and that two systems are required if the projected radiation levels during a fault condition are > 2000 mR/h. It is intended that shielding limits the dose rate to < 0.5 mR/h in occupied areas for normal operational losses, and the combination of administrative and engineered controls for abnormal losses will maintain radiation exposure to trained personnel working at the NSLS-II facility to as low as reasonably achievable. It is anticipated that all individual exposures will be well within the planned administrative control level of 100mrem/year.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Increasing Compositions In NBSR Neutronics Model – Part 2

Calculations in support of the NIST research reactor (NBSR) use a Monte Carlo neutronics model in conjunction with a burnup capability. The representation of compositions within the fuel plates for the current reactor has assumed that with burnup there is a single uniform composition within all plates in each half fuel element. That model has been validated using operational data and leads to acceptable conservative results for safety analyses. However, for the conversion of the reactor to the use of low-enriched uranium (LEU) fuel, there has been an effort to change the modeling for several reasons. The use of large composition zones is inconsistent with the rigor in the rest of the neutronics model. There is also the possibility that the conservatism in maximum fission density may lead to excessive requirements for LEU fuel qualification. A new composition model has been proposed that would break each half fuel element into 15 composition zones improving the rigor while still allowing reasonable running time. This approach was tested, and results show that although it is an improvement over the previous model that had one composition zone (and an interim model with six zones) in each half fuel element, it falls short of being the ideal composition model. This report documents the objectives for the new model and the analysis of the resulting calculations.

42 ENGINEERING↗

DOE OSTI2019

Achieving Verifiable and High Integrity Instrumentation and Control Systems through Complexity Awareness and Constrained Design. Final Report

While most process control industries have been able to embrace digital Instrumentation and Control (I&C) technology to improve performance, reliability, maintainability, and efficiency of production, the nuclear industry has been relatively slow to adopt digital I&C for safety critical plant functions. There are a variety of reasons for this, but in general the nuclear industry takes a conservative stance that new technologies must mature before adoption. A deeper underlying reason for not fully realizing digital I&C and automation in nuclear power is associated with the complexity that can manifest in software-based systems that is different in nature from the complexity of their analog predecessors. Accordingly, this research pursued the development of innovative architectures and methods to “constrain complexity” to enhance verifiability of digital I&C devices. These constraints can positively impact the reduction of Software Common Cause Failures (SCCF) and other systematic errors. The architectural concept developed in this research effort is called SymPLe. SymPLe is specifically aimed at safety critical functions within Nuclear Power Plants and other critical infrastructure where constraining device complexity to support verifiability and safety case arguments is important. The SymPLe architecture is designed to promote “accessible verification” between OEM/utility engineers and the regulatory agencies who review and license nuclear power plant I&C systems. The major contributions of this work to the nuclear industry are: Demonstration of SymPLe as a viable approach to “complexity aware” I&C systems and Embedded Digital Devices.Comprehensive use of model-based design and engineering to produce verifiable I&C systems for the nuclear industry with clear findings on the usability of model-based engineering. Employ comprehensive use of model-based engineering, testing, and verification in an IEC 61508 workflow in support of design assurance. Gathered evidence of “difficult to find” design flaws found using model-based design and testing methods. Detailed the nature of these SCCFs including where they were introduced, the triggers, and how they were found. Gathered evidence on the synergistic use of practical formal methods and testing and how these methods can be used to significantly increase assurance of reduction and avoidance of design flaws.

Testing↗

DOE OSTI2019

IAEA Coordinated Research Project on HTGR Physics, Thermal Hydraulics, and Depletion Uncertainty Analysis: PHISICS/RELAP5-3D Results for the Phase III Coupled Core Exercises

This report details the Parallel and Highly Innovative Simulation for INL Code System (PHISICS)/Reactor Excursions and Leak Analysis Program (RELAP5)-3D results obtained for the neutronics stand-alone core exercises defined for Phase II of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on high-temperature gas cooled reactor (HTGR) uncertainty analysis in modeling (UAM). The Phase II models and results are linked to the earlier Standardized Computer Analyses for Licensing Evaluation (SCALE)/Sampler/New ESC-based Weighting Transport (NEWT) data generated for the lattice physics (lattice) stage Phase I of the CRP. The focus of this report is the propagation of cross-section uncertainties from the lattice to the core calculation (core) phase of the prismatic 350 megawatt (MW) General Atomics (GA) Modular HTGR (MHTGR). The Phase II neutronics exercises consist of two sub-cases (II-2a and II-2b) specified for the fresh and mixed core loading of the MHTGR-350 prismatic high-temperature reactor design. Exercises II-2a and II-2b are performed without temperature feedback using isothermal temperature profiles. For the nominal lattice physics (i.e., best-estimate) model, a series of fresh and depleted single block and super cell SCALE/KENO and SCALE/NEWT models were developed at Idaho National Laboratory (INL) to assess the impact of spectral effects on the few-group cross sections. A combination of these lattice cells were used in a super cell calculation to construct 8-group cross-section libraries for use in the INL-developed PHISICS/RELAP5-3D code suite. The block-level homogenized libraries were assigned to the 220 fuel zones utilized in the MHTGR-350 PHISICS core model, according to the fresh and mixed fresh/depleted core loading pattern. In order to evaluate the effect of uncertainty propagation, several sets consisting of 1,000 perturbed NEWT lattice and PHISICS/RELAP5-3D core calculations were performed using the SCALE 6.2 ENDF/B-VII.1 cross-section co-variance data and the stochastic sampling module Sampler in SCALE 6.2.0. The Figures of Merit (FOM) investigated include the mean and standard deviation uncertainty values obtained for the eigenvalues, control rod worth values, block-averaged power profiles, power axial offsets, and isothermal temperature feedback coefficients. A sensitivity study on the main nuclide reaction contributors to the FOM uncertainties has not yet been performed since the SCALE/PHISICS/RAVEN coupling with ENDF/B-VII co-variance data is still being developed. It was found that uncertainties in the cross sections lead to uncertainties in the local power values of up to 1.5%. The variances between the fresh and mixed/depleted core models are not as large as the axial variations between the top and bottom within each core model. The core eigenvalue uncertainty is approximately 0.5%, which is significant in comparison to typical eigenvalue difference targets (less than ~0.1% for LWRs). The power peaking and axial offset uncertainties are smaller (0.10% – 0.15%) and probably below the level of significance in typical core design metrics. The use of an 8-group energy structure slightly under-estimates some of the uncertainties, but in absolute terms, the uncertainty results are reasonably close to the 26-group results. The 26- and 8-group mean and nominal results differ by a larger degree, and if it can be afforded in terms of calculation resources, the 26-group structure should be used for HTGR calculations. The 8-group structure can, however, still be useful for reasonably accurate uncertainty assessments of transient cases where computational time is important. It was furthermore found that the use of a super cell during the cross-section generation process for better representation of the spectral environment at the core periphery did not lead to significant changes in the core eigenvalues, axial offset, and power peaking uncertainties. The impact of the super cell cross sections on the nominal and mean eigenvalues is, however, important enough (0.6% – 1.1% lower) to justify the use of these super cells for HTGR core simulation to represent the softer spectral environment in the core peripheral regions for nominal best-estimate calculations. The use of super cells for HTGR uncertainty studies is, therefore, not strictly necessary, but the improvements in nominal results could justify their use.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Micro Reactor Simulation Using the PROTEUS Suite in FY19

Two micro reactor cores were simulated using the MOC solver of PROTEUS: MegaPower for a heat pipe micro reactor and Holos-Quad for a high-temperature gas-cooled micro reactor. Verification of PROTEUS standalone was performed by simulating 2D and 3D core benchmark problems developed based on MegaPower and Holos-Quad. Simulation results indicated that PROTEUS successfully modelled 2D and 3D cores of the MegaPower and Holos-Quad micro reactors and PROTEUS eigenvalues agreed well with the SERPENT Monte Carlo solutions. For a heat pipe cooled micro reactor, multiphysics simulations were performed using PROTEUS / FLUENT / ANLHTP. The coupling system of the three codes were developed using Python-based drivers, which exchange the data (power, temperature, and heat rate) required for the coupling and control the boundary conditions of FLUENT and ANLHTP. The coupled system was qualitatively verified using a 3D unit assembly problem composed of six fuel rods and seven heat pipes, which was developed based on the MegaPower compositions. The steady-state problem was solved using FLUENT / ANLHTP with given asymmetric fuel powers, demonstrating the reasonable convergence of temperatures of 7 heat pipes. Transient simulations were also performed by making one out of 7 heat pipes fail. For the transient, the coupled system of PROTEUS / FLUENT / ANLHTP demonstrated reasonable changes of total power and heat pipe temperatures with time, accounting for temperature feedback effects. In this simulation, no effect of structural change was accounted for, which will be performed using ANSYS-mechanical in future along with further verification with larger or whole-core 3D problems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗