Development of Radiation Endurance Ultrasonic Transducer for Nuclear Reactors
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46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
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Summary of Previous Mechanical Test Data on ODS Alloys 14YWT and OFRAC up to 1000ºC
The Nanostructured Ferritic Alloys (NFA) 14YWT and OFRAC were developed for future fission and fusion nuclear energy reactors requiring high-temperature mechanical properties that are tolerant to extreme neutron irradiation environments. The NFA contain a high concentration of Ti-, Y- and O-enriched nanoclusters (NC) and ultra-fine grains to achieve high temperature strength and creep properties and high sink strength for trapping irradiation induced point defects to minimize hardening and swelling and transmutated He atoms to form intragranular nano-size bubbles that prevent formation of coarse bubbles on grain boundaries that cause embrittlement. The mechanical properties of 14YWT and OFRAC have been acquired from tensile and creep tests conducted in the past at Oak Ridge National Laboratory. The development of 14YWT started in 2000, resulting in the production of numerous heats. Tensile properties were obtained from eleven heats of 14YWT from room temperature to 800ºC. Tensile data for the SM10 heat of 14YWT was extended to 1,000ºC. Initial development of OFRAC occurred in 2016. Tensile tests conducted from room temperature to 800ºC revealed similar properties of OFRAC with those obtained from the newer generation of 14YWT heats. The creep properties of 14YWT-SM10 evaluated using constant stress tensile and load time-to-failure tests at 800ºC showed low minimum creep rates with stresses of 200 and 100 MPa. The single time-to-failure test of 14YWT-SM10 at 800ºC and 100 MPa was terminated after 20,357 hours with no specimen failure and a low creep strain of ~0.24 %. The creep properties of OFRAC determined from strain-rate jump tests at were similar those of 14YWT-SM10. The stress exponent for 14YWT and OFRAC at 800ºC are similar and are consistent with threshold stress behavior. Since both 14YWT and OFRAC are candidates for fuel cladding in future fast reactors, several fabrication studies were recently conducted and have successfully demonstrated that thin wall tubes can be fabricated from 14YWT and OFRAC by cold pilger rolling and high precision tube rolling. The high temperature mechanical properties and feasibility of fabricating thin wall tubes make 14YWT and OFRAC candidates for application as heat pipes in advanced micro-reactors. The purpose of this report is to summarize the previously obtained tensile and creep data at temperatures up to 1000ºC for NFA 14YWT and OFRAC that have been acquired over the past 20 years at ORNL.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Development of a core coolant flow distribution calculation code for HTGRs - 18484
In order to ensure the thermal integrity of fuel in high temperature gas-cooled reactors (HTGRs), it is necessary that the maximum fuel temperature in the normal operation is to be lower than the thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power and neutron fluence distributions, and core coolant flow distribution. The core coolant flow distribution calculation code used in the design stage of High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and not user-friendly. Therefore, a new core coolant flow distribution calculation code with a user-friendly system such as simple and easy operations and execution procedures has been developed. In this paper, the outline of the new code is described and the simulation result of an out-of-pile test with one fuel column are shown as the first step of the code validation. The simulation results provide good agreement with the test one. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
AGR-1, AGR-2, AGR-3/4, and AGR-5/6/7 DimensionalChange Analysis
A series of fuel irradiation experiments have been planned in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to support the licensing and operation of the Advanced Reactor Technologies high temperature gas-cooled reactor. The advanced gas reactor (AGR) experiments are comprised of multiple independent capsules containing multiple cylindrical fuel compacts, placed inside of a graphite cylinder in ATR. The purpose of the AGR experiments is to provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for accident testing, and support the development of fuel performance and fission product transport models. The advanced graphite creep (AGC) experiments provide irradiation creep data for design and licensing. To date, six irradiation campaigns have been completed: AGR-1 (December, 2006 – November, 2009); AGR-2 (June, 2010 – October, 2013); AGR-3/4 (December, 2011 – April, 2014); AGC-1 (September, 2009 – January, 2011); AGC-2 (April, 2011 – May, 2012); and AGC-3 (November, 2012 – April, 2014).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Test facility for fission chamber assemblies - 239
The Institution -Project Center ITER- with participation of the National Research Nuclear University MEPhI are developing the neutron diagnostics -Divertor Neutron Flux Monitor- (DNFM) for the International Thermonuclear Experimental Reactor (ITER). This diagnostics shall ensure measurement of the total neutron yield and ITER fusion power. Assemblies of fission chambers (FC) with different fissile materials and sensitivities are planned to be used for the neutron flux measurement. Study of fission chambers characteristics and acceptance testing of the diagnostic system will be carried out at the testing facility of the neutronic laboratory of the 'TRINITI' Institute. The facility includes: compact powerful neutron generator with the power, control and monitoring system; remote handling system for FCs positioning and movement; novel three channels data acquisition and processing system. Each acquisition channel contains some measuring circuits with signal processing. This set of circuits provides neutron flux measurement in the range of at least 10 decades for each single FC. Data acquisition and processing system features are: signal digitizing with sample rate up to 125 MHz; real time data processing with computation and data update cycle from 0.1 ms. This paper examines the structure, composition and specification of the test facility. Application software, hardware solutions and the results of fission chambers prototypes tests are discussed. (authors)
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Assessment of passively actuated in-situ cyclic surveillance test specimens for advanced non-light water reactors
The use of surveillance specimens was identified in a roadmap (Sims, 2010, ASME HTGR Code Development Roadmap, STP-NU-045, American Society of Mechanical Engineers, New York, NY) developed for the U.S. Nuclear Regulatory Commission for non-light water reactors that are intended to have very long design lives on the order of 500,000 hours. Creep rupture, long term strain accumulation, creep-fatigue damage, and their interaction with environmental effects require significant extrapolation from shorter term tests. Surveillance specimens provide a means for confirmation of these extrapolations in-situ. The effects of cyclic service on creep-fatigue damage would be particularly challenging based on conventional mechanical testing. Proposed herein is a passive cyclic testing methodology that lends itself well to in-situ surveillance applications. The concept is based on utilizing the difference in the thermal expansion coefficient of candidate materials to generate cyclic loading based on operational thermal transients. Prototypical specimen designs are proposed and their in-situ response to representative plant operation are evaluated.
Messner, M. C.↗
COMPUTATION FLUID DYNAMICS ANALYSIS FOR GENERIC SMALL MODULAR REACTOR CONTAINMENT SEPARATE EFFECTS TEST
It is desirable for fourth-generation Small Modular Reactors to be passively cooled in standard and accident operations. Passive Containment Cooling Systems can reject heat from the containment structure, without using pumps or blowers. The targeted design containment structure is a large, domed, stainless steel, cylindrical vessel. In a postulated Design Basis Accident, steam will flash inside containment. Steam condensation occurs on the inner containment wall and transfers heat through the steel containment into a large body of water known as the annular reservoir (AR) surrounding the vessel serving as the ultimate heat sink. Natural circulation drives the flow in the AR and heat will be released to the environment by evaporation of water. Unique containment geometry requires a separate effects test (SET) facility for the verification and validation of the computer code and evaluation model development and assessment for reactor licensing efforts. In this study, STAR-CCM+, a computational fluid dynamics (CFD) code was used to inform the decision-making process on the design of the SET. The CFD simulation modeled, a two-phase turbulent flow with fluid film development and heat transfer for different containment geometries. The Reactor Excursion and Leak Analysis Program will also be used in a code-to-code verification against the CFD results.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
AGR-5/6/7 Irradiation Test Predictions using PARFUME
PARFUME, (PARticle FUel ModEl) a fuel performance modeling code used for high temperature gas-cooled reactors (HTGRs), was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using predicted physics and thermal hydraulics data. The AGR-5/6/7 test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter north east flux trap (NEFT) position of Advanced Test Reactor (ATR). Each capsule contains compacts filled with uranium oxycarbide (UCO) unaltered fuel particles. This report documents the calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated source term from the driver fuel. The calculations include modeling of the AGR-5/6/7 irradiation that is scheduled to occur from October 2017 to April 2021 over a total of 13 ATR cycles, including nine normal cycles and four Power Axial Locator Mechanism (PALM) cycle for a total between 500 – 550 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 2, and 4. Fuel particle failures were predicted in Capsule 3 due to internal particle pressure. These failures were predicted in the highest temperature compacts. Capsule 5 fuel particle failures were due to inner pyrolytic carbon (IPyC) cracking causing localized stresses concentrations in the SiC layer. This capsule predicted the highest particle failures due to the lower irradiation temperature. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. The two capsules at the two ends of the test train, Capsules 1 and 5 experienced the smallest buffer-IPyC gap formation due to the lower irradiation fluences and temperatures. Capsule 3 experienced the largest buffer-IPyC gap formation of just under 24 µm. The release fraction of fission products Ag, Cs, and Sr silver (Ag), cesium (Cs), and strontium (Sr) vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 3, reaching up to 84.8% for the TRISO fuel particles. The release fraction of the other two fission products, Cs and Sr are much smaller and, in most cases, less than 1%. The notable exception is again in Capsule 3, where the release fraction for Cs and Sr reach up to 9.7% and 19.1%, respectively.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Wide-range fission chambers signal simulator - 373
Fission chambers (FC) signal simulators are a useful means to test neutron flux instrumentation channels of the reactors control and protection systems (CPS). Well-known signal simulators do not provide a secondary equipment testing in all modes of measurement. A wide range simulator that simulates a full range of FC signal changes (pulse, fluctuation, current) was designed. The simulator takes into account the statistical nature of neutrons detection in the FC, including the shape of the pulse and the spectrum of the pulse amplitude distribution. The simulator uses the principle of mathematical simulation of the fission chambers signal with realtime calculations using the FPGA. The wide-range simulator reproduces a FC signal in the range of ten orders without switching ranges or playback modes while ensuring very good linearity. It allows to generate both static and dynamic test signals such as: stair-step, linearly and exponential varying signals, exponential varying signals with a linearly changing period or a signal change simulating the response of the nuclear reactor power to a given reactivity response. Continuity of reproduced signals and the ability to playback dynamic effects makes it possible the simulator using to check the algorithms of cross-linking of measuring paths, reactivity and period measuring tools and to monitor the operation of interlocking and alarm CPS functions. This article discusses the simulator's software and hardware, its features and specifications, user interface and example of the simulator using to check the neutron flux monitoring channels. (authors)
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Interrelation of technologies for RW preparation and sites for final isolation of the wastes from pyrochemical processing of SNF
For the justification of engineering solutions and practical testing of the radiochemical component of the perspective nuclear power complex with on-site variant of nuclear fuel cycle (NFC), it is planned to establish a multi-functional research-development complex (MFCRC) for radiochemical processing of spent nuclear fuels (SNF) from fast reactors. MFCRC is being established at the NIIAR site, it comprises technological process lines, where innovation pyro-electrochemical and hydrometallurgical technologies are realized, with an option for closing the inter-chain material flows for testing the combined radiochemically converted materials. The technological flowchart for processing at the MFCRC is subdivided into 3 segments: -) complex of the lead operations for dismantling the fuel elements (FE) and fuel assemblies (FA), -) pyrochemical extraction flowchart for processing SNF, and -) hydrometallurgical flowchart for processing SNF. The engineered solutions for the management and disposition of the radioactive wastes from MFCRC are reviewed.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Mass Transport of Condensed Species in Aerodynamic Fallout Glass from a Near-Surface Nuclear Test
Abstract not provided
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Hybrid super homogenization and discontinuity factor method for continuous finite element diffusion
This paper presents a novel homogenization equivalence technique aiming to simultaneously leveragethe simplicity of the Super Homogenization (SPH) method to reproduce reference reaction rates andthe ability to preserve reference leakage rates at desired surfaces through Discontinuity Factors (DF).The need for this new class of methods arises from the inability of the current state-of-the-art SPH technology to properly reproduce the reactor multiplication factor for problems with significant leakage. This work shows that this defect lies in the use of normalization factors in the SPH algorithm: while they are necessary with purely reflecting problems to ensure uniqueness of the solution, they introduce homogenization inconsistencies if at least one vacuum boundary condition is present. Two solutions to this problem are presented in this work: (i) to simply remove the normalization factor or (ii) to introduce additional degrees of freedom in the form of DFs in such a way that the normalization factors can still constrain the problem. While the former clearly offers unrivaled simplicity and can lead to very satisfactory results for leakage-dominated cores or if spatial restriction of the SPH regions is applied – as demonstrated for the High Temperature Test Reactor, the latter is more robust and does not require to scale the fission terms with the reference eigenvalue.
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Oxidation behavior of Zry-4 in steam-air mixtures at high temperature
This study dealt with the oxidation behavior of Zry-4 nuclear fuel cladding under a severe nuclear reactor accident scenario occurring in a light-water reactor. Influence of the reaction extent on cladding mechanical properties was also treated. Zry-4 fuel cladding segments were exposed to steam atmosphere containing different amount of air in composition ranging from 0 to 100 vol. %. Exposure times covered both pre- and post-breakaway regimes. Metallographic observations along with microhardness testing were carried out to study the effect of air fraction on the material microstructure. Hydrogen pick-up was measured and residual mechanical properties were assessed employing Ring Compression Tests. The results show that the addition of air in steam can lead to more than 300 % increase of weight gain in case of ∼50 % or higher air fraction after the exposure of 60 min., namely, 'late post-breakaway regime'. Approximately three times thicker oxide scales were observed in such cases, indicating that most of the weight gain was due to the oxide layer increase. Nitride phase was preferentially observed close to the interface between oxide and metal layers. The α-Zr(O) sublayer thickness decreases with the increasing fraction of air in steam. Microhardness, measured within the prior β-phase region, slightly depends on the air fraction. Hydrogen uptake, on the other hand, depends strongly on the air fraction in steam. In the case of shorter exposure times, the effect of air addition in steam was suppressed. Eventually, specimen plastic strains drastically reduced in the late post-breakaway regime. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Integrated Simulation Capabilities for Analysis of Experiments in the Versatile Test Reactor
The DOE-initiated Versatile Test Reactor (VTR) program aims to establish a domestic fast-neutron irradiation testing capability that meets a variety of domestic and international nuclear energy needs. Currently, integrated tools capable of mechanistic modeling of VTR and a test vehicle do not exist. To address this modeling and analysis deficiency, Argonne's SAS4A/SYSSYS-1 safety analysis code has been coupled with SAM (System Analysis Module) to provide a novel modeling capability that supports VTR and other fast test facilities. This report documents the design, implementation, and testing of this integrated capability. A coupling boundary has been identified at the test vehicle and primary coolant interface, where SAS4A/SASSYS-1 treats primary coolant thermal hydraulics outside the test vehicle, while SAM treats all thermal hydraulic behavior within the test vehicle, including the vehicle walls. Essential to this integrated tool is its newly developed capability to properly model the conjugate heat transfer process which ensures equality of temperatures and heat fluxes at the vehicle wall interface while ensuring energy conservation. A range of testing has been completed to support demonstration of this interface. The testing scope includes steady-state verification using a series of increasingly complex analytical solutions and transient demonstration for a range of design basis accident scenarios using prototypic VTR and test vehicle (i.e. cartridge loop) configurations. Results of this testing confirm verification of the steady-state solutions and provide reasonable and expected transient behavior. The coupling interface has been developed to be robust yet flexible. The implementation within SAS4A/SASSYS-1 supports coupling to any external software or module. The use of decomposed domains and the existing in-core thermal-hydraulic calculational framework within SAS4A/SASSYS-1 enable treatment of arbitrary test vehicle geometries with nonuniform meshing between SAS4A/SASSYS-1 and the coupled code. Modeling of multiple test locations is also supported. Furthermore, computational efficiency is enhanced by utilizing named pipes (or FIFOs), an interface for interprocess communication (IPC).
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ECAR-5633 AGR-5/6/7 Daily As-Run Thermal Analyses
The purpose of this engineering and calculations report is to document the results of the thermal analyses performed to calculate the Advanced Gas-Cooled Reactor (AGR)-5/6/7 as-run daily temperatures of the fuel compacts. Temperature data provided by this report will be used to evaluate fuel performance. The AGR-5/6 portion of the experiment was for fuel qualification at prototypic temperatures, while the AGR-7 portion of the experiment was a margin test at very high temperatures.
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FY19 Transient Prescriptions in TREAT Full-Slotted Core
The transient reactor test facility (TREAT) is a graphite-based transient test reactor at the Idaho National Laboratory. TREAT started operation in 1959 and performed thousands of transient irradiations for the primary purpose of nuclear fuel safety research. TREAT operations were suspended in 1994 and then, several years later, resumed reactor operations in 2017 to support development of advanced nuclear fuels. During TREAT’s first era of operation the transient rod drive system and eventual Automatic Reactor Control System (ARCS) were upgraded and improved to create a transient reactor capable of unparalleled power shaping capabilities. TREAT’s first year of modern experiment work was performed in 2018 and saw the demonstration of new kinds of transients focused on modern experiment needs; especially focused on light water reactor (LWR) category transients including reactivity-initiated accidents (RIA) and loss of coolant accidents (LOCA). These transients were performed using a non-fueled neutronic mockup of a historical sodium loop which was not a perfect representation of future core configurations, but was adequate for demonstrating capabilities, successfully troubleshooting transient programming parameters, and calibrating nuclear models of the core. These physics tests were termed “transient prescription tests”. Later in 2018 the core was reconfigured to a modern fueled experiment arrangement and TREAT began performing irradiations on LWR rodlets where crucial fuel specimen nuclear heating data were obtained, and several supporting systems were commissioned. This successful effort gave way to additional fueled experiments which were performed throughout 2019.
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Development of Digital Twin Predictive Model for PWR Components: Updates on Multi Times Series Temperature Prediction Using Recurrent Neural Network, DMW Fatigue Tests, System Level Thermal-Mechanical-Stress Analysis
The long-term operation (LTO) of nuclear power plant (NPP) beyond their original design life of 40 years, can lead to more material damage associated with cyclic fatigue under thermal-mechanical loading cycles and associated long-term exposure of reactor material to the deleterious reactor-coolant environments. However, under this LTO condition the reactor components can still safely operate but may require more frequent Nondestructive Evaluation (NDE) of reactor components. Frequent NDE requirement may lead to frequent shutdown of the NPP. This in turn can lead to power outage and additional NDE-inspection-cost related economic loss. The economic loss can be minimized by reducing uncertainty in life estimation of safety-critical pressure boundary components and by implementing more digital approach such as by using upcoming digital-twin (DT) technology for predicting the structural states (e.g., time and location dependent inside/outside thickness temperature, stress, strain, plastic deformation, etc.) and associated fatigue life of a component in real time. Towards this goal Argonne National Laboratory (ANL) with the sponsorship of DOE Light Water Reactor Sustainability (LWRS) program is working on the development of a DT framework that can be used for real time environmental fatigue prediction of reactor components. The DT framework is based on limited experiment-data, Artificial-intelligence (AI) – Machine-Learning (ML) - Deep-Learning (DL) based techniques and Multiphysics-computational-mechanics such as finite element (FE) based modeling tools. Towards this overall goal, following are some of the major contributions made during the FY21: 1) Multiple 82/182 dissimilar metal weld (DMW) specimens (both solid-weld and joint-weld representing the actual reactor multi-metal nozzles) were fatigue tested. The resulting fatigue lives were compared to the NUREG-6909 based best-fit and design fatigue curves. Additionally, the results of 52/152 DMW fatigue specimens (which were recently tested at Republic of Korea under the sponsorship of International Nuclear Energy Research Initiative - INERI program) were compared to the NUREG-6909 based best-fit and design fatigue curves. From the comparison of 82/182 and 52/152 DMW test data with NUREG-6909 best-fit curve, most of the reported test data fall way away from the NUREG-6909 suggested best-fit or mean curve. The NUREG-6909 suggested best-fit curve is the best-fit curve of austenitic stainless steel and due to lack of enough data on Nickel-based welds, this is currently being used for predicting the life of Nickel-alloy-based welded components. However, the above observation may require higher scaling factor (e.g., ASME suggested factor of 20 on cycles rather than the current NUREG-6909 suggested factor of 12 on cycles) for scaling the austenitic-stainless-steel best-fit-curve for estimating the design or safe-life of a welded component. Accordingly, for example, if a DMW component experience a strain amplitude of 0.6% the PWR-water life of the component would be 52 cycles instead of 85 cycles. However, more DMW tests are required to further ascertain the above-mentioned observations. 2) A system level CAD and finite element model were developed which consists of reactor pressure vessel (RPV), part of steam generator (SG), part of pressurizer (PRZ), hot leg (HL), and surge line (SL). This is with detailed nozzle geometry and thermal-mechanical material properties of different metals to simulate realistic thermal-mechanical stress under connected system global thermal-mechanical boundary conditions. 3) Different system level heat transfer analyses were performed with estimation of relevant heat transfer coefficients. The resulting data were used in subsequent system level thermal-mechanical stress analysis and for generating spatial-temporal training and validation data for a system level digital-twin based temperature predictor. Transient heat transfer analyses were performed considering thermal boundary condition under design-basis (DB) loading and EDF (Électricité de France) data-based grid-load-following (EDF-GLF) loading cycles. 4) System level thermal-mechanical stress analysis was performed for identifying damage-prone hotspots and for future extension of the model for cyclic state prediction. From the system-level model simulation under DB loading cycle it is found that HL and the SL nozzle that connect to the HL can experience significant stress and strain and could be one of the weakest links in the overall reactor coolant system (RCS). 5) An AI/ML based DT model was developed for multi-time-series temperature prediction at any inside/outside thickness locations of PWR pressure boundary components. This is by using Recurrent-neural-network (RNN) and keras machine learning libraries. The RNN model was validated against two laboratory test-based data sets with one obtained through ANL’s in-air fatigue test system and other through PWR-water test loop. The experimentally validated DT model further validated against FE model results to predict thermal scarification related spatialtemporal temperatures at random locations of a component. The well validated DT model was then used for demonstrating spatial-temporal temperature prediction under 100+ years of reactor operation subjected to combined DB, EDF-GLF and randomized grid-load-following (RANDOMGLF) loading Cycles. The expert-elicitation DT model framework was developed assuming field/input/process measurements can be available from a few existing plant sensors and can readily be used by the NPP operators. The above temperature prediction model will feed to the next-step stress analysis model based on which the life of a component can be predicted in realtime, which is one of our future works.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Proposal of Methodology and Approach for the Coupling of CTF and TRACE within the Framework of VERA-CS
This document discusses strategies for coupling the Consortium for Advanced Simulation of Light water reactors (CASL) thermal-hydraulics code, CTF, with the Nuclear Regulatory Commission (NRC) system code, TRACE. A high-level plan is presented for how the coupling will be implemented, how TRACE will work inside of the VERA-CS environment, and how the coupling will be tested and expanded.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗