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At least 379 records · Page 21

Overview of recent studies related to lead-alloy-cooled fast reactors

The recent progress of the studies related to lead alloy-cooled reactors (LFR) and the accelerator driven system (ADS) is summarized. The compatibility of materials with lead alloys has been clarified under steady and transient temperature conditions. Higher Cr content, Si and Al addition and Al-Fe alloy-coating improved the corrosion resistance of steels. The Al-Fe alloy-coated steel was not corroded even under high temperature transient conditions. The ceramics of SiC and Si{sub 3}N{sub 4} are expected to be used as cladding material for high temperature LFR. For the analytical consideration of mass transport in lead alloy circuit, the diffusion coefficient of Ni was measured using the capillary methods. A new bubble visualization method in LBE using gamma-ray radiography was developed. The thermal interaction of lead-bismuth eutectic (LBE) and lead droplets with sub-cooled water, and the behaviors of droplet fragmentation were investigated, and the visualization of volatile liquids in high temperature liquids was achieved.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Quantitative void fraction measurement with an eddy current flowmeter for generation IV Sodium cooled Fast Reactor

This study was carried out to understand the response of an eddy current type flowmeter in two phase liquid-metal flow. We use the technique of ellipse fit and correlate the fluctuations in the angle of inclination of this ellipse with the void fraction. The effects of physical parameters such as coil excitation frequency and flow velocity have been studied. The results show the possibility of using an eddy current flowmeter as a gas detector for large void fractions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Feasibility of Lead Fast Reactor Heat Exchanger Tube Online Monitoring

Within virtually all reactor systems with heat exchangers, the thinnest barriers between primary and secondary fluids are the heat exchanger tubes. Consistent with light water reactor experience, the advanced reactor heat exchangers are the most susceptible regions for corrosion and leakage (NRC, 2019). Moreover, traditional inspections and repairs using robotics will be expensive or impractical. On-line structural health corrosion monitoring (OLSHM) is therefore desirable to detect tube degradation prior to leaks that may allow mixing of heat exchanger fluids. Rhetorically many molten salt reactor (MSR) designers plan for on-line monitoring of critical reactor components including heat exchangers but there are no proven OLSHM sensors or systems for MSR heat exchangers. Guided wave ultrasound from the tube-end or ID using conventional low-temperature piezoelectric sensors has been shown to be sensitive to pits and cracks less than 50% through-wall in long pipes and tubes (Narayanan M.M., 2019) . Mode and frequency selection are important for practical sensor design decisions and ultimate performance. The L-0 wave-mode is easily excited by L-mode transducers but they have surface normal wave motion that will be sensitive to and more attenuative because of fluid interaction with the tube surface (Rose J.L., 1994). Shear Horizontal (SH) wave modes have material motion parallel to the tube surface therefore are virtually unaffected by fluid interaction. For SH-0 to be nondispersive in steel, the frequency-thickness product (MHz-mm) must be less than approximately 1.5. SH mode transducers must be coupled by a material that can support shear stress; a thin layer of honey works well for temporary sensors and an adhesive or a brazed metallurgical bond for permanent sensors. If sensors are incorporated into the fabrication process, the same kinds of SH-0 waves can be generated from the stagnant flow area of the tube OD without compromising the heat-exchange tube function. Piezoelectric materials are also available that can withstand harsh temperatures of molten salt heat exchangers (Stevenson T., 2015) (Machura et al., 2008). Though they may not have as high-performance characteristics, their performance only needs to be good enough to detect flaws of interest. This project attempts to demonstrate feasibility to develop a sensor that can be sensitive to flaws of interest and perform long-term monitoring of a molten salt reactor heat exchanger. Benefits to the advanced reactor system include: • Minimizing the expense of a reactor shut down for periodic inspection—only shutdown for cause. • Reduce risk of unexpected tube rupture by detecting flaws before they grow to through-wall leaks. • Cost reduction by minimizing need to design for manual/robotic inspectability. Transducer design, characterization of the beam profile, wedges vs. direct application of piezoelectric sensors, and ultrasonic pulse-echo responses from a representative heat exchange tube with fabricated flaw types are discussed.

On-Line Tube Monitor, Guided Wave Ultrasound, SH-0↗

Impact of nuclear data covariance libraries on uncertainty quantification of sodium cooled fast reactor simulation

Uncertainty quantification in core modelling to assess reactor performance is a regular practice to identify design safety limits. Best-estimate plus uncertainty (BEPU) methods are commonly used to propagate uncertainties to core responses from input parameters, characterized by variance-covariance libraries. This study evaluates the impact of different covariance libraries on the uncertainty associated with core safety parameters and their propagation through reactor transient modelling for a metallic core (ABR-1000). The COMAC-1.0, COMMARA-2.0, and COMMARA-2.1 covariance libraries are used to propagate nuclear data uncertainties to neutron feedback coefficients using the Generalized Perturbation Theory, then through transient safety analyses to establish the confidence interval for safety performance of the reactor using stochastic sampling. The considered transients include Unprotected Transient Over Power and Unprotected Loss Of Flow. Most influential nuclide and reaction pairs in different variance-covariance libraries are identified and their impact on the uncertainty of safety parameters are evaluated. Significant differences were observed between uncertainties from COMAC compared to COMMARA libraries. Uncertainties from COMMARA libraries were consistently larger than COMAC for all cases. For structure, coolant density feedback coefficients and control rod worth, total nuclear data uncertainties were approximately 50% larger for COMMARA than COMAC. Differences were also observed in identification of top contributors of uncertainty in terms of reaction-cross sections. For instance, largest contribution to total uncertainty of radial expansion coefficient from COMMARA-2.0 and 2.1 originates from Na{sup 23} P1 Elastic cross section (0.81%) but from Pu{sup 239} Fission (0.53%) with COM C-1.0. Such differences are evaluated in this research for all quantities of interest. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Modeling and Analysis of the BUK/BES-5 Fast Reactor Using MCNP [Slides]

The Soviet Union launched a series of satellites designated RORSAT that were designed to essentially collect information on NATO and US Navy. 33 satellites were launched between 1967 and 1988, the majority (31) of which were powered by the BUK reactor design or minor variants of BUK, The motivation for this effort is to better understand this small space reactor design. In addition, there is much to be learned from the issues encountered while modeling this reactor concept and its agreement with the real reactor deployed in the space.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Tensile Testing on FFTF Irradiated Fast Reactor Cladding

This report covers the first tensile tests on samples oriented perpendicular to the extrusion direction on the FFTF irradiated ACO3 duct, and comparison to similar dose samples oriented in the extrusion direction. Control tests were performed to test the same orientation dependency in control lots of HT9 material. This report also presents data from a recently obtained a cold FFTF HT9 duct and compares its tensile properties to a previously used ACO3 archive control lot. Finally, comparison studies of different thicknesses of SSJ type tensile samples were performed.

36 MATERIALS SCIENCE↗

Assessment of Applicability of Standards Endorsed by Regulatory Guides to Sodium Fast Reactors

The benefits of the use of voluntary consensus standards and industry standards is evident in the design and licensing of the current generation of nuclear power plants. The use of voluntary consensus standards and industry standards would be helpful in the design and licensing of advanced reactors to improve the effectiveness and efficiency of the licensing and regulation of non-LWR technologies. Most of the regulations, guidance, and standards applicable to nuclear power plants were developed for water-cooled plants, so they may not adequately address the coolants, materials, temperatures, operations, testing, maintenance, etc., proposed for advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Status of the NEAMS and ARC neutronic fast reactor tools integration to the NEAMS Workbench

The Workbench initiative was launched in FY-2017 within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to facilitate the transition from conventional tools to high-fidelity tools. The Workbench provides a common user interface for model creation, real-time validation, execution, output processing, and visualization for integrated codes. The integration of the Argonne Reactor Computation (ARC) suite of codes into the NEAMS Workbench was initiated in FY-2017.

97 MATHEMATICS AND COMPUTING↗

Gamma Irradiation of NaCl-UCl 3 Salt for the Molten Chloride Fast Reactor

This report documents the results of the gamma ray irradiation of NaCl-UCl 3 eutectic salt in the Advanced Test Reactor (ATR) spent fuel pool gamma tube and subsequent analysis of capsules utilizing the Gas Assay Sample and Recharge (GASR) system. NaCl-UCl 3 salt capsules at four different temperatures, 75C, 150C, 300C and 600C, were irradiated shortly after the core internal changeout (CIC) of the ATR, beginning April 28 th of 2021, and ending on August 18 th of 2021. Re-positioning of the fuel elements in the spent fuel pool to move more fresh fuel elements around the gamma tube occurred on May 11 th of 2021, resulting in an approximate 20% increase in the dose rate and consequently the total absorbed dose. In total, the salt-filled capsules underwent 2638 hours of gamma irradiation from the adjacent freshly discharged ATR spent fuel. Capsule internal pressure measurements were taken via the Gas Assay Sample and Recharge (GASR) system in the INL Hot Fuels Examination Facility (HFEF) following completion of the irradiation. Based on the GASR analysis results, the primary conclusion from this experiment to-date is that radiolytic generation of chlorine gas is insignificant for the dose rates and total absorbed dose that these samples experienced. To elucidate more detailed effects of radiolysis on NaCl-UCl 3 eutectic salts and reach a more definitive conclusion, further advanced analyses such as electron microscopy, simultaneous thermal analysis, and electron paramagnetic resonance are recommended to analyze the capsule wall and salt samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗