Development and Characterization of SiC-Clad Control Rods & SiC Guide Tubes for Accident-Tolerant PWR Fuel Assemblies
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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Control rod damping system
A damping area or “dash pot” on the upper ends of control rods absorb energy from dropped control rod assemblies without narrowing the diameter of guide tubes. As a result, coolant can freely flow through the guide tubes reducing boiling water issues. The dampening area reduces a separation distance between an outside surface of the control rod and an inside surface of the guide tubes decelerating the control rods when entering a top end of the guide tubes. In another example, the dampening area may be located on a drive shaft. The dampening area may have a larger diameter than an opening in a drive shaft support member that decelerates the drive shaft when dropped by a drive mechanism.
Snuggerud, Ross Douglas↗
Control rod damping system
A damping area or “dash pot” on the upper ends of control rods absorb energy from dropped control rod assemblies without narrowing the diameter of guide tubes. As a result, coolant can freely flow through the guide tubes reducing boiling water issues. The dampening area reduces a separation distance between an outside surface of the control rod and an inside surface of the guide tubes decelerating the control rods when entering a top end of the guide tubes. In another example, the dampening area may be located on a drive shaft. The dampening area may have a larger diameter than an opening in a drive shaft support member that decelerates the drive shaft when dropped by a drive mechanism.
Snuggerud, Ross Douglas↗
Towards a Neutronic and Thermal-Hydraulic Safety Analysis of PWR Fuel Bow with SIMULATE5 and COBRA-TF
In the 1990's, collective fuel assembly bow was observed in the Ringhals PWRs in Sweden. Back then fuel vendors reinforced the fuel assemblies' dashpot, decreased hold-down spring forces, increased guide tube wall thickness, and introduced guide tube materials with higher creep resistance in order to improve dimensional stability. Nevertheless, twenty years later in 2010, the same kind of collective bow with S and C shape patterns started to appear in most of Germany's PWRs. At this time the SIMULATE5 (S5) core simulator with its delta-gap bow model (DGA) which was originally developed for safety analysis at Ringhals, could be used as an off-the-shelf solution to assess the impact of bow on Fq and FH. However, these two quantities are of secondary importance in German PWRs in normal operation because the most restricting licensing criteria set limits on DNBRmin values. For this purpose, a method was developed to extract S5 pin powers from DGA calculations and automatically generate COBRA-TF input files (CTF) to analyze DNBR values in the cores. In the following two sections, it is shown how the S5 model for DGA calculations has been verified with independent MCNP6 calculations in order to check its predictions on pin power changes for delta gaps up to 1 cm in size and how the existing licensed DNBR calculation methodology has been updated to bowed fuel assembly situations with the help of CTF. Finally, the applicability of the Doroschuk CHF correlation, which is the licensing standard in most German PWRs, is discussed. For the fuel bow patterns considered, it is assumed that only the gap between fuel assemblies change and that the pitch between fuel rods inside the assemblies remains largely unchanged. This approximation can be justified by the visual inspections conducted at the plants during the outages. Every calculation of the pin-power impact of delta-gaps in the core under operating conditions is only as good as the knowledge of the bow pattern. Measurements of fuel assembly bow at cold conditions have been made for some German PWRs for all the fuel assemblies in the core during plant outages. Together with methods developed by the fuel assembly vendors, a best-estimate bow pattern during operation could be reconstructed. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
APR1400 In-Core Instrumentation Penetration Failure by Zirconia Melt
Light water reactors have many penetrations at the reactor vessel lower head which are called ICI (in-core instrumentation tubes), ICM-GT (in-core monitoring guide tubes), CRGT (control rod guide tubes) and drain tubes. For Korean PWRs (pressurized water reactors), there are dozens of bottom-up type ICI penetration tubes (45 for OPR1000 and 61 for APR1400). For the typical BWRs (boiling water reactors), hundreds of penetrations are positioned at the reactor vessel lower head due to the structural complexity at the reactor upper head. These penetrations are regarded as the most vulnerable structures during a severe accident as they could be failed by high temperature of melt (or debris) and its decay heat. That is, tube rupture and/or tube ejection could occur because there exist an empty hole inside the penetration tube which can be a melt flow path, and they are attached to the inside of the reactor vessel by a partial weld with a gap between the tubes and reactor vessel. Therefore, the research on the penetration failure is of importance, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. This paper focuses on the experimental investigation on the tube ejection failure for the APR1400 ICI penetration. The tube ejection experiments were performed as it was exposed to high temperature of zirconia melt and high pressure. The tube ejection failure experiments were performed for the APR1400 in-core instrumentation penetration nozzle. Zirconia melt was used as a simulant of molten corium and interacted with the penetration specimen. The specimen was pressurized up to 2.6 bar during the interaction with melt to induce the tube ejection. The penetration weld was heated up to its melting temperature, and the penetration tube and weld above the reactor vessel surface were eroded severely by the melt. However, even though it was estimated that the weld failure occurred and the reactor vessel with stainless steel cladding was ablated severely by the melt, the tube ejection did not take place, and consequently the integrity for the APR1400 ICI penetration was confirmed in the present experimental conditions. The goal of this research is to improve the penetrations failure models in MAAP5 codes based on the experimental data. To achieve this goal, it is necessary to increase the heat flux and pressure acting on the penetration tube to find the conditions for the tube ejection failure. Moreover, the penetration failure experiments will be performed by delivering melt into the penetration specimen to simulate the melt relocation process in a real severe accident situation. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Detecting pin diversion from pressurized water reactors spent fuel assemblies
Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Design of a Continuous Pellet Fueling System for Wendelstein 7-X
A continuous pellet fueling system (CPFS) is currently being designed at the Oak Ridge National Laboratory (ORNL) for the long pulse operation of the Wendelstein 7-X (W7-X) stellarator. The purpose of the CPFS is to provide deep continuous fueling for feedback-controlled high-density operation and mitigation of predicted hollow density profiles. As described here, the system will provide the capability to inject cylindrical pellets of solid hydrogen or deuterium into the plasma core, with flexibility to vary the pellet size, velocity, and injection frequency. Pellets are nominally of 3 mm in diameter and have a length between 1 and 4 mm. The heart of the CPFS is a vertically oriented, twin-screw extruder, cooled by three Gifford-McMahon cryocoolers in parallel, designed to form a continuous filament of hydrogen or deuterium. The filament width, which determines the pellet length, can be adjusted by means of a variable nozzle driven by a linear actuator at the base of the extruder. Coupled to the nozzle is a solenoid-operated gas gun and cutter assembly. A pneumatic propellant valve pulses a burst of ~60 bar helium to accelerate the cut pellet into W7-X. Three gaps in the guide tubes provide pumping locations to remove the helium propellant before it reaches the plasma. The maximum velocity of the pellet is limited by its ability to survive navigating the guide tube trajectory intact. A microwave cavity located within the guide tube provides the capability to measure the pellet size and velocity. The mechanical and thermal designs of the W7-X extruder, adjustable nozzle, and gun and cutter assembly design are described. A guide tube design and experimental pellet survivability test results are presented.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Response of a Pressurized Water Reactor Dashpot Region to Commercial Drying Cycles
The purpose of this report is to document updates to the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents testing updates for the Dashpot Drying Apparatus (DDA), an apparatus constructed at a reduced scale with multiple Pressurized Water Reactor (PWR) fuel rod surrogates and a single guide tube dashpot. This apparatus is fashioned from a truncated 5×5 section of a prototypic 17×17 PWR fuel skeleton and includes the lowest segment of a single guide tube, often referred to as the dashpot region. The guide tube in this assembly is open and allows for insertion of a poison rod (neutron absorber) surrogate.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
U.S. implementation of Q12{sup TM} zirconium alloy for improved stability of structural components
AREVA NP undertook the Q12{sup TM} research and development program to optimize the contribution to fuel assembly dimensional stability of structural components fabricated from ultra-low tin Zr-1Nb-Sn-Fe quaternary alloys. Because Q12{sup TM} is an evolutionary development of the current M5 alloy metallurgy, it benefits from the extensive industrial experience already gained with the latter. The Q12{sup TM} alloy is processed in the same way as M5, with slight increases in tin (Sn) and iron (Fe) contents. For structural components, these chemistry modifications provide higher irradiation creep strength and improved corrosion performance over Zircaloy-4. Fuel assemblies with quaternary alloy guide tubes are thus expected to increase robustness with respect to fuel assembly bow, improve growth characteristics, and maintain high burnup capability. The basic material properties have been determined through various in-core and out-of-core test programs. Components made from Q12{sup TM} have undergone irradiation in lead assemblies and in batch fuel for 13 Pressurized Water Reactor (PWR) units worldwide. Fuel with Q12{sup TM} guide tubes and grids has achieved a fuel assembly average burnup of 60 GWd/tU. Post irradiation examinations of these components have demonstrated that Q12{sup TM} performs in a stable, predictable manner. AREVA NP has submitted a topical report (TR) for Q12{sup TM} as a structural material to the United States Nuclear Regulatory Commission (US-NRC). This TR outlines the material behavior and performance of Q12{sup TM} for both guide tubes and grids. Once approved, implementation of Q12{sup TM} material throughout US PWR fuel designs will enhance dimensional stability of structural components. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Small bore magnetic measurement system, method for measuring small bore magnets
The invention provides a system for measuring magnets, the system comprising a guide tube removably received by a beam chamber tunnel, wherein said guide tube has a first end terminating at a first end of the tunnel and a second end terminating at a second end of the tunnel; a flexible substrate removably and slidably received by the tube, wherein the flexible substrate has a first end and a second end; a first rotary stage removably attached to the first end of said flexible substrate and a second rotary stage removably attached to the second end of said flexible substrate wherein the first rotary stage and the second rotary stage effects movement of the substrate through the tube; and a Hall probe supported by the flexible substrate. Also provided is a method for measuring magnetic fields, the method comprising extending a measuring probe into a magnetic core while maintaining the probe at ambient temperature and pressure.
Kasa, Matthew T.↗
FY 2016 Status Report: Documentation of All CIRFT Data including Hydride Reorientation Tests (Draft M2)
The first portion of this report provides a detailed description of fiscal year (FY) 2015 test result corrections and analysis updates based on FY 2016 updates to the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) program methodology, which is used to evaluate the vibration integrity of spent nuclear fuel (SNF) under normal conditions of transport (NCT). The CIRFT consists of a U-frame test setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages connecting to a universal testing machine. The curvature SNF rod bending is obtained through a three-point deflection measurement method. Three linear variable differential transformers (LVDTs) are clamped to the side connecting plates of the U-frame and used to capture deformation of the rod. The second portion of this report provides the latest CIRFT data, including data for the hydride reorientation test. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNF samples into a single zone. A detailed examination revealed that, at the same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), LMK, and mixed uranium-plutonium oxide (MOX). Just looking at the strain, LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The third portion of this report provides finite element analysis (FEA) dynamic deformation simulation of SNF assemblies . In a horizontal layout under NCT, the fuel assembly’s skeleton, which is formed by guide tubes and spacer grids, is the primary load bearing apparatus carrying and transferring vibration loads within an SNF assembly. These vibration loads include interaction forces between the SNF assembly and the canister basket walls. Therefore, the integrity of the guide tubes and spacer grids critically affects the vibration intensity of the fuel assembly during transport and must be considered when developing the multipurpose purpose canister (MPC) design for safe SNF transport.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Magnetic Mapping of MAGIS-100 Tubes to Guide Experimental Redesign
The MAGIS-100 atom interferometer requires an exceptionally uniform, low-magnetic-field environment along its 100-meter baseline. While characterization of the raw structural tubes established a clean baseline, the original surrounding experimental design proved overly complex, ultimately introducing complications that degraded data precision. To resolve these systemic issues, a comprehensive redesign of the surrounding apparatus is underway. This poster details our magnetometry data collection efforts on the baseline tubes and highlights how these measurements are being used to guide a streamlined external design. By leveraging this background data, the new configuration aims to eliminate the complexities of the initial setup and ensure the environmental stability required for high-precision quantum sensing.
Savoia Cooley, Patricia [North Central Coll.]↗
Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS
Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~ 1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥ 1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line'' paths, to inject high-speed pellets (in the 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. As a result, a possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Spent nuclear fuel system dynamic stability under normal conditions of transportation
In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Dynamic modeling of heat pipe integrated thermal battery latent heat storage system experiment validation
A heat pipe integrated thermal battery system has been constructed to investigate a high-temperature latent heat thermal energy storage technology that takes advantage of near isothermal operation of latent heat storage and heat pipes to potentially enable high-energy isothermal heat storage. A dynamic model constructed in Modelica has been validated, showing errors between 2.5 °C–39.7 °C across 10-h to 47-h simulations against experiment results, showing good prediction capability of experiment output, especially against phase change time. Model calibrations showing vessel heat-up capability of 3 kW and heat pipes combining to provide 600 W each during experiment operation validate experiment circumstances including reduced material loading and reduced power capability. The experiment configuration uses an Al-Mg-Zn eutectic metal as the storage material, heated via heat tape wrapped around the vessel and guide tubes to bring the system to operation range (>400 °C) and to simulate charging heat exchange, respectively, with heat rejection occurring through the surfaces of the material and facilitated via guide tubes with less insulation wrapping. The model is available in the open-source repository HYBRID on Github.
25 - ENERGY STORAGE↗
Phenomena-based Uncertainty Quantification in Predictive Coupled- Physics Reactor Simulations
This project has sought to develop methodologies, tailored to phenomena that govern nuclearreactor behavior, to produce predictions (including uncertainties) for quantities of interest (QOIs) in the simulation of steady-state and transient reactor behavior. Examples of such predictions include, for each QOI, an expected value as well as a distribution around this value and an assessment of how much of the distribution stems from each major source of uncertainty. The project has sought to test its methodologies by comparing against measured experimental outcomes. The main experimental platform has been a 1-MW TRIGA reactor. This is a flexible platform for a wide range of experiments, including steady state with and without temperature feedback, slow transients with and without feedback, and rapid transients with strong feedback. The original plan was for the primary experimental data to come from in-core neutron detectors. We made considerable progress toward this goal but did not get as far along as we had planned. We have designed, developed, installed, and tested vertical guide tubes, each able to accept a detector or stack of detectors that can be moved axially inside the tube, and we have tested several new detector designs. One of these shows considerable promise.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Excitation Function Calculations of Neutron-Induced Reactions of Some Zirconium Target Isotopes
Nuclear structures with high performance materials will be critical for the success of future fusion energy. Zirconium alloys are extensively used for nuclear reactor core materials as fuel cladding and guide tubes due to excellent radiation and corrosion resistant properties. In this paper, the excitation functions of {sup 90}Zr(n,2n){sup 89}Zr, {sup 90}Zr(n,α){sup 87}Sr, {sup 91}Zr(n,p){sup 91}Y and {sup 92}Zr(n,p){sup 92}Y nuclear reactions were calculated for energies up to 25 MeV. Besides, ALICE/ASH and TALYS 1.6 codes were used for the cross section calculations. The obtained calculations were compared with the experimental data in the literature.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Assessment of Core Status of Fukushima Daiichi Nuclear Power Plant
There were some events and phenomena that occurred which have been deemed specific to the Fukushima Daiichi NPP. The primary issues were as follows: - Direct steam leakage from the reactor pressure vessels (RPVs) to the dry-wells through the gaskets of the safety relief valves and/or due to creep damage of the main steam line, and damage of the in-core monitor guide tubes due to buckling. - Direct corium leakage from the RPVs to the pedestal floors through the melt portions of the in-core monitor housings. - Alternative water injection by the fire trucks with large amount of leakages through the branch lines. - Partial load operation of the Reactor Core Isolation Cooling system and the High Pressure Coolant Injection system. - Gas leakage from the primary containment vessels to the reactor buildings due to over pressure than the design maximum value. - Thermal stratification in the suppression pool resulting in incomplete steam condensation in the pool. The accident progressions were analyzed with the newly developed physical models for the above issues. And also considering the qualitatively estimated core status based on the plant parameters measured after the accident, the current core status was assessed.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗