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The Lagrangian in eq. (2.1) of JHEP 11 (2020) 005 has an erroneous factor of 1/2. The analysis code used for the results reported in the paper does not have this factor. The results remain unchanged.
The Advanced High-Temperature Reactor (AHTR) is a conceptual design of a Fluoride salt-cooled High-temperature Reactor (FHR) utilizing 7 LiF-BeF 2 (66-34 mol%) as its primary coolant. To identify key phenomena that impose potential challenges on thermal hydraulics modeling and simulation of such a reactor system, a thermal hydraulics phenomena identification and ranking table (TH-PIRT) study was performed for the AHTR in a Department of Energy Nuclear Energy University Program (NEUP) integrated research project led by Georgia Institute of Technology. A panel of experts from regulators, industries, national laboratories, and academia was assembled for the study. In this paper, the TH-PIRTs identified by the panel for two events, namely, station blackout and simultaneous withdrawal of all control rods, are summarized and discussed in detail. In addition, the key phenomena that warrant further study and research for AHTR analysis are identified to support the validation of thermal hydraulics system-level analysis codes and computational fluid dynamics simulation tools, as well as future FHR reactor licensing.
The Reactor Core Isolation Cooling (RCIC) system consists of a turbine, pump and wetwell and is used in many US Boiling Water Reactors (BWRs) as an important heat removal system. The wetwell is a major heat sink within containment. Experimental investigations have shown that thermal stratification may occur in the wetwell during RCIC operation. Current systems-level analysis codes lack stratified wetwell models. Here, this paper describes the development of a stratified wetwell model for the RELAP-7 code. This paper uses a simple two-zone model for the wetwell water that can capture thermal stratification. This model assumes that a buoyant plume transports heat and mass from the steam injection site to the upper water zone. The buoyant plume assumption requires that steam is injected at a low enough rate so it adds negligible momentum. Model validation against experimental data shows similar trends of temperature field development in the suppression pool.
A new large natural circulation FLiBe loop has operated for the first time in 40 years. Reaching temperatures above 700 C with stable flow rates around 5 cm/s, the loop exhibits a range of flow phenomena pertinent to high temperature natural circulation safety systems for nuclear reactors. This paper discusses the system and several operating conditions spanning a wide range of behavior to aid the development of other systems, simulation benchmarks, or system-codes. Analysis of heat transport phenomena indicates that standard Nusselt number correlations apply to a first approximation. More accurate salt-freezing models may be necessary for air-cooled salt-systems due to the insulating affect of the low thermal-conductivity frozen salt layer on the cold pipe walls.
Thermal hydraulic instability can challenge the control and the safety of nuclear reactors. This experimental research captures and investigates the instability in a low-pressure (1–10 atmospheres) natural circulation loop with an adiabatic chimney installed above the heated section. Direct transient local measurement of two-phase flow is performed at multiple axial locations along the flow channel with thermocouples and conductivity probes traversed by customized high-precision mechanisms. Based on the periodicity of oscillations, a post-processing method is proposed to align the data captured non-simultaneously to reconstruct the instability behavior in one representative period. The averaging operation over multiple oscillation cycles is performed to reduce the random errors. Cyclic phenomena are clearly demonstrated by the post-processed data, and rapid void fraction change induced by the flashing is observed to be significant in the driving mechanism for low pressure oscillations. Parametric effects of inlet subcooling, heat flux, inlet flow restriction and system pressure are studied within the experiment range, and qualitative results are explained and compared with existing literature. The ratio between the oscillation period and the travelling time of the fluid particle through the test section is compared with the theoretical value for the density wave oscillation. The new dataset contains directly measured transient multi-dimensional two-phase data along the flow channel for natural circulation under density wave oscillations. The generated dataset is valuable for the validation of system analysis codes and CFD codes in the simulation of natural circulation phenomena and dynamic two-phase flow.
Demonstration of hermetic SiC fiber–reinforced SiC matrix composite cladding under normal operating environments has been identified as one of the most critical feasibility issues for accident-tolerant fuel cladding in light-water reactors. This study provides critical experimental data needed for understanding the effects of irradiation on hermeticity. SiC composite and monolithic tubes were neutron-irradiated to 2 displacements per atom with and without a nominal radial heat flux of 0.6 MW/m2 to produce a simulated in-pile stress state for the normal operation of a light-water reactor. The through-thickness temperature gradient under irradiation results in a gradient in swelling, which causes a significant stress buildup. Such irradiation-induced stress was modeled using a commercial finite element analysis code. The radial heat flux–irradiation synergism was experimentally investigated by constructing a special irradiation capsule and evaluating the helium hermeticity of the specimens. The simulated stress state exhibited a near equi-biaxial tensile axial and hoop stress of ~150 MPa at the inner surface of the SiC composite tube. This stress level is potentially beyond the matrix cracking stress. Degradation of hermeticity of the SiC composite tubes was observed after irradiation, indicating irradiation-induced cracking, whereas the irradiated monolithic SiC tubes remained hermetic. In conclusion, the results indicate that loss of hermeticity caused by radiation-induced microcracking is a potential issue for SiC composite cladding, depending on the magnitude of the temperature gradients. Coating the outer surface of the cladding was identified as a mitigation strategy that might overcome the cracking issue.
The declared inventory of 235 U in fresh, low enriched uranium, nuclear fuel assemblies is routinely verified by nuclear safeguards inspectorates using the UNCL (Uranium Neutron Collar – Light Water Reactor Fuel) nondestructive assay instrument. The trend in modern fuels is towards higher initial enrichment, which in turn requires a larger number of burnable poison pins with higher Gadolinia concentration to hold down the initial reactivity. UNCL assay error for modern fuel assemblies typically exceeds 10%. The traditional algorithm used to correct the response for burnable poison content needs revision to achieve accuracy comparable to the $\sigma_\text{R}$ of 4.5% expected for poison-free assemblies. Here we review the problem and use a large set of 287 Monte Carlo simulations based on an experimentally benchmarked model corresponding to the 16×16 PWR array used in Brazilian Angra type II and III fuel. Simulated relative responses are used to evaluate the functional form of the poison correction and determine parameters optimal for 16×16 PWR assemblies. This update is conducted using between 4–24 Gadolinia pins ranging 2–11 wt% Gd 2 O 3 with assemblies having mean enrichments ranging 2.5%–5%. Experimental validation of the updated coefficients using seven measurements shows bias is reduced by an order of magnitude and $\sigma_\text{R}$R reduced from 10% to less than 2%. Two updated sets of coefficients are provided, the most accurate of which is directly useable in the existing analysis code (INCC) in use by inspectorate at fuel fabrication facilities.
Multi-layered Silicon Carbide (SiC) cladding has been proposed to replace the current Zircaloy (Zr) cladding in existing nuclear power plants. Due to its ceramic property, SiC may have a different failure mechanism from that of metallic Zr clad. This characteristic should be considered in designing the plant’s safety system performance, such that the fuel’s integrity is reasonably assured during accident scenarios. This paper aims at optimizing the Emergency Core Cooling System (ECCS) performance criteria in mitigating LBLOCA accidents at a Pressurized Water Reactor (PWR) using the uranium dioxide fuel and duplex or triplex SiC cladding structures. Clad failures by brittle fractures and thermal degradation were modeled in a fuel analysis code. This fuel model was coupled with the RELAP5 plant model by adapting the fuel’s mesh configuration and formulating the effective heat transfer coefficients. A model to quantify the likelihood of clad functional failure as a combination of individual layer fractures at various axial positions was proposed. Updated SiC material properties were gathered from recent reference studies. Clad reliability estimates using the proposed model were compared with the estimates obtained using other models. The results were used to relax the criterion of required actuation timing of ECCS, such that the risk of the new plant design is less or equal to that of the initial plant configuration. Triplex clad was more reliable than duplex clad and Zr, reducing the conditional core damage probability by three orders of magnitude in the reference LBLOCA scenario. Compressive stresses due to irradiation swelling played the determining role in the vulnerabilities of each clad design. The duplex clad was vulnerable to pressure-loading stresses while triplex clad was prone to fail due to thermal shocks experienced during the LOCA reflood phase. Additionally, it was concluded that the most vulnerable axial fuel rod location in the reference LBLOCA scenario was located at the highest axial power peaking location. A sensitivity study on the ECCS timing density was conducted. The results showed that ECCS actuation timing could be extended from the initial technical specification, without increasing the fuel damage likelihood. The time extension was distributed in the order of several hundreds of seconds, which was comparable with time extensions offered by FeCrAl and Cr-coated-Zr clad designs in LBLOCA scenarios. This time extension may potentially reduce the failure probabilities of EDGs and pumps, hence improving the plant safety further.
In order to accurately predict the behavior of the Reactor Core Isolation Cooling System (RCIC) in the context of the FLEX strategies under the extended station black-out accident (SBO), we present a mechanistic model of the RCIC in this work. The control volume formulation of angular momentum conservation is employed to model the Terry turbine in the RCIC system. The two-phase nozzle flow model developed in the previous work is utilized to simulate the convergent-divergent nozzles in the Terry turbine. A nozzle flow table is generated to cover the operation range of the RCIC during beyond design basis accidents (BDBAs). The mechanistic RCIC turbine-pump model and the nozzle flow table are incorporated in a TRACE BWR model. Simulations of short-term and long-term SBO transients are performed to demonstrate dynamic RCIC responses. A system analysis code with the proposed mechanistic model of the RCIC system will have a better prediction of system safety response during SBO accidents. With the addition of such more realistic models, the system code will be capable of providing expanded understanding of safety margin, which can help help to form the technical basis for the successful implementation of the FLEX and SAMG measures under Extended Loss of AC Power (ELAP) conditions for BWR or PWR reactor designs.
For this work, Westinghouse partnered with Argonne National Laboratory to design, model and optimize UO 2 - and MOX-fueled core designs for a medium size (950 MWt) Lead Fast Reactor that was pursued by Westinghouse. Using Argonne’s suite of reactor analysis codes together with Westinghouse fuel cost economic models, thousands of candidate cores were considered to achieve the economics-optimized cores presented in this paper. This optimization process considered detailed reactor physics, fuel performance, transient performance, and fuel economics models. The reactor performance of the resulting optimized UO 2 - and MOX-fueled core designs are described and compared in this paper. Both cores show fuel performance and transient behavior that is considered acceptable for the optimization presented herein, while further testing campaigns on material performance in high-temperature liquid lead will be required to confirm acceptability at the operating conditions chosen. A multi-batch strategy was selected for the UO 2 core for best fuel utilization with minimum fuel inventory costs. A single-batch fuel management was instead selected for the MOX core to maximize cycle length and minimize the impact of the longer refueling outage resulting from the higher decay heat of the discharged MOX fuel relative to the discharged UO 2 fuel, requiring a longer cooling time before dry-lift of discharged fuel could take place.
Here, the FASt TEst Reactor (FASTER) plant is a sodium-cooled fast spectrum test reactor design that was developed as part of an advanced demonstration and test reactor study, and provides high levels of fast and thermal neutron flux for scientific research and development. The 120 MWe FASTER plant has been designed with extended testing capabilities in mind, while keeping the design as simple as possible in order to make it efficient and cost attractive. The main function of the reactor is to provide neutrons for irradiation testing utilizing existing proven technologies to ensure a high technology readiness level. A preliminary safety analysis was performed of the FASTER design using the fast reactor safety analysis code SAS4A/SASSYS-1 to assess the reactor's safety performance during beyond design basis accidents. Three types of transient scenarios were analyzed for the FASTER design: unprotected transient overpower, unprotected loss of heat sink, and unprotected station blackout. These transients were simulated at beginning of equilibrium cycle and end of equilibrium cycle conditions. For each transient, failures of both the primary and secondary reactor scram systems were assumed. Because of the favorable reactivity feedbacks, relatively low peak fuel, cladding, and coolant temperatures were obtained during the transients and the reactor vessel temperature remained well below the limit for ensuring structural stability.
In order to unravel the reaction pathways relevant to anisole co-oxidation within a fuel blend, a detailed study of isooctane/anisole blends was performed with the ULille RCM. Ignition delays as well as mole fraction profiles were measured during a two-stage ignition delay using sampling and GC techniques. These results are used to validate a kinetic model developed from ab initio calculations for the most relevant rate constants which included H-atom abstraction reactions from anisole, and reactions on the potential energy surfaces of methoxyphenyl + O 2 and anisyl + O 2 . Pressure dependent rate constants were computed for the methoxyphenyl + O 2 and anisyl + O 2 reactive systems using master equation code analysis. The new kinetic model shows good agreement with the experimental data. Dual brute-force sensitivity analysis was performed, on both first- and second-stages of ignition, allowing the identification of the most important reactions in the prediction of both ignition delays. It was observed that while pure anisole does not show NTC behavior, a 60/40 isooctane/anisole blend displays such behavior, as well as two-stage ignition. This suggests anisole addition may not be as beneficial to knock resistance as expected from its high octane number. The kinetic modeling results demonstrate the importance of H-abstraction reactions both from the methoxy group and from the aryl ring in ortho-position and the addition of the resultant radicals to O2, mostly leading to the formation of polar or non-aromatic products.
We present a detailed exposition of how first-principles methods can be used to guide alkali superionic conductor (ASIC) study and design. Using the argyrodite Li 6 PS 5 Cl as a case study, we demonstrate how modern information technology (IT) infrastructure and software tools can facilitate the assessment of alkali superionic conductors in terms of various critical properties of interest such as phase and electrochemical stability and ionic conductivity. The emphasis is on well-documented, reproducible analysis code that can be readily generalized to other material systems and design problems. For our chosen Li 6 PS 5 Cl case study material, we show that Li excess is crucial to enhancing its conductivity by increasing the occupancy of interstitial sites that promote long-range Li + diffusion between cage-like frameworks. The predicted room-temperature conductivities and activation barriers are in reasonably good agreement with experimental values.
This work exploits gas adsorption and small-angle X-ray scattering (SAXS) to determine the morphology of complex nanoporous materials. We resolve multiple classes of porosity including previously undetected large-scale texture that significantly compromises the canonical interpretation of gas adsorption. Specifically, a UVM-7 class mesoporous silica was synthesized that has morphological features on three length scales: macropores due to packing of 150 nm globules, 1.9 nm radius spherical mesopores inside the globules, and >7 nm pockets on and between the globules. The total and external surface areas, as well as the mesopore volume, were determined using gas adsorption (α s -plot) and SAXS. A new approach was applied to the SAXS data using multilevel fitting to determine the surface areas on multiple length scales. The SAXS analysis code is applicable to any two-phase system and is freely available to the public. The total surface area determined by SAXS was 12% greater than that obtained by gas adsorption. The macropore interfacial area, however, is only 30% of the external surface area determined by the α s -plot. The overestimation of the external surface area by the αs-plot method is attributed to capillary condensation in nanoscale surface irregularities. Here, the discrepancy is resolved assuming that the macropore–globule interfaces harbor fractally distributed nooks and crannies, which lead to gas adsorption at pressures above the mesopore filling pressure.
Ferroelectric materials have technological applications in information storage and electronic devices. The ferroelectric polar phase can be controlled with external fields, chemical substitution and size-effects in bulk and ultrathin film form, providing a platform for future technologies and for exploratory research. In this work, we integrate spin-polarized density functional theory (DFT) calculations, crystal structure databases, symmetry tools, workflow software, and a custom analysis toolkit to build a library of known, previously-proposed, and newly-proposed ferroelectric materials. With our automated workflow, we screen over 67,000 candidate materials from the Materials Project database to generate a dataset of 255 ferroelectric candidates, and propose 126 new ferroelectric materials. We benchmark our results against experimental data and previous first-principles results. The data provided includes atomic structures, output files, and DFT values of band gaps, energies, and the spontaneous polarization for each ferroelectric candidate. Finally, we contribute our workflow and analysis code to the open-source python packages atomate and pymatgen so others can conduct analogous symmetry driven searches for ferroelectrics and related phenomena.
The quantitative analysis of dislocation-type defects in irradiated materials is critical to materials characterization in the nuclear energy industry. The conventional approach of an instrument scientist manually identifying any dislocation defects is both time-consuming and subjective, thereby potentially introducing inconsistencies in the quantification. This work approaches dislocation-type defect identification and segmentation using a standard open-source computer vision model, YOLO11, that leverages transfer learning to create a highly effective dislocation defect quantification tool while using only a minimal number of annotated micrographs for training. This model demonstrates the ability to segment both dislocation lines and loops concurrently in micrographs with high pixel noise levels and on two alloys not represented in the training set. Inference of dislocation defects using transmission electron microscopy on three different irradiated alloys relevant to the nuclear energy industry are examined in this work with widely varying pixel noise levels and with completely unrelated composition and dislocation formations for practical post irradiation examination analysis. Code and models are available at https://github.com/idaholab/PANDA.
The RNTuple I/O subsystem is ROOT’s future event data file format and access API. It is driven by the expected data volume increase at upcoming HEP experiments, e.g. at the HL-LHC, and recent opportunities in the storage hardware and software landscape such as NVMe drives and distributed object stores. RNTuple is a redesign of the TTree binary format and API and has shown to deliver substantially faster data throughput and better data compression both compared to TTree and to industry standard formats. In order to let HENP computing workflows benefit from RNTuple’s superior performance, however, the I/O stack needs to connect efficiently to the rest of the ecosystem, from grid storage to (distributed) analysis frameworks to (multithreaded) experiment frameworks for reconstruction and ntuple derivation. With the RNTuple binary format soon arriving at its first production release, we present RNTuple’s feature set, integration efforts, and its performance impact on the time-to-solution. We show the latest performance figures of RDataFrame analysis code of realistic complexity, comparing RNTuple and TTree as data sources. We discuss RNTuple’s approach to functionality critical to the HENP I/O (such as multithreaded writes, fast data merging, schema evolution) and we provide an outlook on the road to its use in production.
Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.