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At least 1,729 records · Page 96

Experimental Methods for the Performance Test of Environment-Assisted Cracking (EAC) in Nuclear Industry and Examples through NSUF

Environmentally assisted cracking (EAC) is a critical failure mode in nuclear reactor environments, where materials are exposed to high temperatures, radiation, and corrosive media. Specific EAC studies in nuclear environments include stress corrosion cracking in primary water chemistry, irradiation-assisted stress corrosion cracking (IASCC) under neutron exposure, and hydrogen induced cracking (HIC) in reactor pressure vessels. The experimental methodologies of EAC include slow strain rate testing (SSRT), constant load, and cyclic loading techniques, conducted in autoclaves designed to simulate the reactor’s operational environments. This paper highlights several experimental case studies, especially on IASCC, conducted within the Nuclear Science User Facilities (NSUF).

36 - MATERIALS SCIENCE↗

Effects of Geometric Discontinuities on Creep Behavior of Alloy 617

Due to the excellent long-term high-temperature performance of Alloy 617, the Alloy has been identified as the primary structural material for the intermediate heat exchanger of the Very High Temperature Reactor (VHTR). Both the base and welded alloy has been qualified to be used up to 950 °C up to 100,000 hours by the American Society of Mechanical Engineer (ASME) Boiler and Pressure Vessel Code (BPVC). The qualification properties of the material were exclusively determined by uniaxial creep tests. The components may, however, experience multiaxial stress conditions in service due to the reactor geometry. This study used notched creep tests to investigate the effect of geometric discontinuities on creep behavior of both the base and welded Alloy 617. The V-notch tests reveal that the geometric discontinuities do not reduce the creep rupture live under high and intermediate stress conditions. The U-notch tests show that a larger radius notch behave in a similar way as a straight gauge section of a uniaxial test. The creep rupture live of the welded alloy was slightly longer than the base metal.

36 - MATERIALS SCIENCE↗

ECAR-7547 Rev 0 MARVEL Intermediate Heat Exchanger Liner ASME Analysis

This ECAR analyzes the Secondary Coolant System liner, otherwise known as the Intermediate Heat Exchanger (IHX) Liner, as an ASME Section VIII, Division 1 pressure vessel. This analysis is only necessary in the event of a failure of the primary coolant boundary, in which case the IHX liner acts as a secondary confinement boundary to prevent primary coolant leakage. The liner main body or vessel is analyzed, as well as the liner flange ring.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

STANDARD REFERENCE MATERIAL CHARPY TESTING FROM HIGH FLUENCE SURVEILLANCE CAPSULE

The Palisades Nuclear Generating Station included in its original surveillance program a surveillance capsule, designated A-60. The capsule was removed from its surveillance position in early 1995 and has been resident in the spent fuel pool since that time. It was harvested to perform characterization of surveillance specimens in this capsule in 2023. This capsule was irradiated to a fluence of 1.96x1020 n/cm2 (E> 1MeV) that is equivalent for more than 150 effective full power years for the current US reactor pressure vessel (RPV) fleet. This capsule contained several materials, including Charpy specimens of standard reference material (SRM) from Heavy-Section Steel Technology (HSST) A533-B Plate 01. To perform Charpy testing of this highly irradiated material, the new Charpy specimen transfer system was designed and implemented on Charpy impact testing machine in the hot cell to accommodate remote testing of these specimens. This new transfer system includes integrated environmental chamber to cool or heat Charpy specimens to the desired temperature. Testing of this highly irradiated material revealed very large shift of Charpy transition temperature, 188oC.

Sokolov, Mikhail [ORNL] (ORCID:0000000256373346)↗

Measurement of Fatigue Crack Growth Relationships in Hydrogen Gas for Pressure Swing Adsorber Vessel Steels

We measured the hydrogen-assisted fatigue crack growth rates (da/dN) for SA516 Grade 70 steel as a function of stress-intensity factor range (ΔK) and load-cycle frequency to provide life-prediction data relevant to pressure swing adsorber (PSA) vessels. For ΔK values up to 18.5 MPa m 1/2 , the baseline da/dN versus ΔK relationship measured at 1Hz in 2.8 MPa hydrogen gas represents an upper bound with respect to crack growth rates measured at lower frequency. However, at higher ΔK values, we found that the baseline da/dN data had to be corrected to account for modestly higher crack growth rates at the lower frequencies relevant to PSA vessel operation.

36 MATERIALS SCIENCE↗

Non-invasive method and apparatus for measuring pressure within a pliable vessel

A non-invasive method and apparatus is disclosed for measuring pressure within a pliable vessel such as a blood vessel. The blood vessel is clamped by means of a clamping structure having a first portion housing a pressure sensor and a second portion extending over the remote side of the blood vessel for pressing the blood vessel into engagement with the pressure sensing device. The pressure sensing device includes a flat deflectable diaphragm portion arranged to engage a portion of the blood vessel flattened against the diaphragm by means of the clamp structure. In one embodiment, the clamp structure includes first and second semicylindrical members held together by retaining rings. In a second embodiment the clamp structure is of one piece construction having a solid semicylindrical portion and a hollow semicylindrical portion with a longitudinal slot in the follow semicylindrical portion through which a slip the blood vessel. In a third embodiment, an elastic strap is employed for clamping the blood vessel against the pressure sensing device.

Shimizu, M.↗

Energy efficient continuous flow ash lockhopper

The invention relates to an energy efficient continuous flow ash lockhopper, or other lockhopper for reactor product or byproduct. The invention includes an ash hopper at the outlet of a high temperature, high pressure reactor vessel containing heated high pressure gas, a fluidics control chamber having an input port connected to the ash hopper's output port and an output port connected to the input port of a pressure letdown means, and a control fluid supply for regulating the pressure in the control chamber to be equal to or greater than the internal gas pressure of the reactor vessel, whereby the reactor gas is contained while ash is permitted to continuously flow from the ash hopper's output port, impelled by gravity. The main novelty resides in the use of a control chamber to so control pressure under the lockhopper that gases will not exit from the reactor vessel, and to also regulate the ash flow rate. There is also novelty in the design of the ash lockhopper shown in two figures. The novelty there is the use of annular passages of progressively greater diameter, and rotating the center parts on a shaft, with the center part of each slightly offset from adjacent ones to better assure ash flow through the opening.

Collins, Earl R., Jr.↗

ASME Design Code Rule Changes for Nuclear Graphite

The American Society of Mechanical Engineers Boiler Pressure and Vessel Code (ASME BPVC) Section III, Division 5, Article HHA-3000 outlines graphite core component and graphite core assembly design guidelines. Graphite core components are defined as ?components manufactured from graphite that are installed to form a graphite core assembly within the reactor pressure vessel of a high temperature, graphite moderated fission reactor.? (p. 413) Graphites? inherent defect distributions do not allow for deterministic material reliability. Rather, graphite has variable strength distributions which change by grade. Article HHA-3000 outlines two semi-probabilistic methods, the full and simplified assessments, which set design load limit targets for each of three component structural reliability classes. The Design Task Group was officially recognized as a specialized task group within ASME November of 2023, though we?ve been collaborating since 2022. The purpose of the Design Task Group is to correct, clarify, and make HHA-3000 function as intended. The Design Task Group will sunset once we?ve achieved our objectives. The Design Task Group was specifically told to not write new Code. While there may be more precise and more accurate methods to determine reliability targets, the current methods are conservative, relatively simple to implement, and have thus far been considered satisfactory for setting design reliability targets. Much of the ground-work to write proposal files and background documents for records to make the changes needed to achieve our objective have been completed. The Design Task Group has documented much of their work through papers, presentations, and memorandums. Three memorandums in which INL team members had substantial contributions are found in the Appendices: FEA Modeling for the Baseline Program, Evaluating the Effects on Margin of Updating the Threshold and Shape Parameters in the Full Assessment, and Interpretations of the Full and Simplified Assessments in ASME BPVC. Most of the on-going work to achieve the Design Task Group?s objective will be addressing comments on existing records and moving records through the balloting process. The Design Task Group met bi-weekly mostly through the end of FY2023. Since February 2024, the Design Task Group has mostly been completed with solving and documenting the technical issues associated with the assessments. Unless new tasks are identified, the remaining work of the Design Task Group will be political and editorial.

97 MATHEMATICS AND COMPUTING↗

2021 NE-XF Pathways Internship Student Showcase Presentation

Branch of Fluid Mechanics concern with the mechanics and properties of gases and its application to engineering systems. Applications of Pneumatics in the Aerospace systems: (1) Flight Hardware - Reaction Control, Systems, Rocket Pressure Fed Systems (2) Ground Operations – Valve Actuation, Inert Environment (Purge), Tube Drying/Cleaning NE-XFPneumatics Engineering provides design & operation expertise to support EGS. Commodities used at KSC systems: Gaseous Helium (GHe) Gaseous Oxygen (GO2) Gaseous Nitrogen (GN2) Breathing Air (BAIR) Support other subsystems: LH2, LO2, Ground Main Propulsion System (GMPS), etc.

Environmental and Life Support Systems↗

Toward Smart, Automated Junctional Tourniquets—AI Models to Interpret Vessel Occlusion at Physiological Pressure Points

Hemorrhage is the leading cause of preventable death in both civilian and military medicine. Junctional hemorrhages are especially difficult to manage since traditional tourniquet placement is often not possible. Ultrasound can be used to visualize and guide the caretaker to apply pressure at physiological pressure points to stop hemorrhage. However, this process is technically challenging, requiring the vessel to be properly positioned over rigid boney surfaces and applying sufficient pressure to maintain proper occlusion. As a first step toward automating this life-saving intervention, we demonstrate an artificial intelligence algorithm that classifies a vessel as patent or occluded, which can guide a user to apply the appropriate pressure required to stop flow. Neural network models were trained using images captured from a custom tissue-mimicking phantom and an ex vivo swine model of the inguinal region, as pressure was applied using an ultrasound probe with and without color Doppler overlays. Using these images, we developed an image classification algorithm suitable for the determination of patency or occlusion in an ultrasound image containing color Doppler overlay. Separate AI models for both test platforms were able to accurately detect occlusion status in test-image sets to more than 93% accuracy. In conclusion, this methodology can be utilized for guiding and monitoring proper vessel occlusion, which, when combined with automated actuation and other AI models, can allow for automated junctional tourniquet application.

60 APPLIED LIFE SCIENCES↗

Supercritical CO2 Cleaning System for Planetary Protection and Contamination Control Applications

Current spacecraft-compatible cleaning protocols involve a vapor degreaser, liquid sonication, and alcohol wiping. These methods are not very effective in removing live and dead microbes from spacecraft piece parts of slightly complicated geometry, such as tubing and loosely fitted nuts and bolts. Contamination control practices are traditionally focused on cleaning and monitoring of particulate and oily residual. Vapor degreaser and outgassing bakeout have not been proven to be effective in removing some less volatile, hydrophilic biomolecules of significant relevance to life detection. A precision cleaning technology was developed using supercritical CO2 (SCC). SCC is used as both solvent and carrier for removing organic and particulate contaminants. Supercritical fluid, like SCC, is characterized by physical and thermal properties that are between those of the pure liquid and gas phases. The fluid density is a function of the temperature and pressure. Its solvating power can be adjusted by changing the pressure or temperature, or adding a secondary solvent such as alcohol or water. Unlike a regular organic solvent, SCC has higher diffusivities, lower viscosity, and lower surface tension. It readily penetrates porous and fibrous solids and can reach hard-to-reach surfaces of the parts with complex geometry. Importantly, the CO2 solvent does not leave any residue. The results using this new cleaning device demonstrated that both supercritical CO2 with 5% water as a co-solvent can achieve cleanliness levels of 0.01 mg/cm2 or less for contaminants of a wide range of hydrophobicities. Experiments under the same conditions using compressed Martian air mix, which consists of 95% CO2, produced similar cleaning effectiveness on the hydrophobic compounds. The main components of the SCC cleaning system are a high-pressure cleaning vessel, a boil-off vessel located downstream from the cleaning vessel, a syringe-type high-pressure pump, a heat exchanger, and a back pressure regulator (BPR). After soaking the parts to be cleaned in the clean vessel for a period, the CO2 with contaminants is flushed out of the cleaning vessel using fresh CO2 in a first-in-first-out (FIFO) method. The contaminants are either precipitating out in the boil-off container or being trapped in a filter subsystem. The parts to be cleaned are secured in a basket inside and can be rotated up to 1,400 rpm by a magnetic drive. The fluid flows within the vessel generate tangential forces on the parts surfaces, enhancing the cleaning effectiveness and shortening the soaking time. During the FIFO flushing, the pump subsystem pushes fresh CO2 into the cleaning vessel at a constant flow rate between 0.01 and 200 mL/min, while the BPR regulates the pressure in the cleaning vessel to within 0.1 bar by controlling the needle position in an outlet valve. The fresh CO2 gas flows through the heat exchanger at a given temperature before entering the cleaning vessel. A platinum resistance thermometer (PRT) reads the cleaning vessel interior temperature that can be controlled to within 0.1 K. As a result, cleaning vessel temperature remains constant during the FIFO flushing. There is no change in solvent power during FIFO flushing since both temperature and pressure inside the cleaning vessel remain unchanged, thus minimizing contaminants left behind. During decompression, both temperature and pressure are strictly controlled to prevent bubbles from generating in the cleaning vessel that could stir up the contaminants that sank to the bottom by gravity.

Lin, Ying↗

Development of safety requirements for HTGRs design - 18467

The safety requirements for the design of High Temperature Gas-cooled Reactors (HTGRs), which is analogous to the IAEA specific safety requirements (SSR) for Light Water Reactor (LWR) design (IAEA SSR-2/1), has been developed by the research committee established in the Atomic Energy Society of Japan so as to incorporate the HTGR safety features demonstrated by High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the coupling of the hydrogen production plants with nuclear plant. The safety design approach was determined to establish a high level of safety design standards by utilizing inherent safety features of HTGRs. The safety requirements is then developed using the objectives provisions tree methodology based on the safety design approach. The requirements for a structural integrity of fuel during accident conditions, for a confinement structure instead of a leak-tight pressure resistance containment vessel, for a passive residual heat removal from outside of reactor pressure vessel, etc., which are significantly different from those of LWR, are stated in the developed safety requirements. This paper describes overview of the proposed HTGR specific safety requirements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Passive techniques for long-term reactor cooling

In a pressurized water reactor (PWR), emergency core cooling (ECC) responds to depressurization due to a vessel penetration break at the top of the pressure vessel by draining water from a body of water through an injection line into the pressure vessel. A barrier operates concurrently with the ECC to suppress flow of liquid water from the pressure vessel out the vessel penetration break. The barrier may comprise one or more of: (1) an injection line extension passing through the central riser to drain water into the central riser; (2) openings in a lower portion of a central riser to shunt some upward flow from the central riser into a lower portion of the downcomer annulus; and (3) a surge line providing fluid communication between a pressurizer volume at the top of the pressure vessel and the remainder of the pressure vessel which directs water outboard toward the downcomer annulus.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗