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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Development and Integration of a Stochastic Clad Damage Propagation Model into PRONGHORN-SC Subchannel Analysis Code

The failure of fuel pins in nuclear reactors is intrinsically stochastic. Typically, a combination of variation in manufacturing that affects the material characteristics and the fuel assembly dimensions, variation in operating conditions, such as local power, coolant flow rate, and irradiation induced changes in material properties lead to a large uncertainty in failure margin of the fuel pins. Failure, therefore, may occur in exceptional pins with adverse combinations of these variations. Upon a metal fuel pin (U-Pu-Zr/HT9) failure, depressurization of the fuel pin takes place by release of fission gas, liquid sodium bond, and potentially solid fuel particles or molten/eutectic fuel droplets through the hole in cladding. The effect of a fission gas jet on neighbor fuel pins and possible propagation of a clad damage during normal operation was studied experimentally in 1970s and it was found that the post-failure fission gas jet insulates the jet impingement area of the target fuel pin surface and could increase the target pin’s surface temperature by as much as 100 – 200 K during the failed pin depressurization. It was concluded that the effect should not lead to fuel pin failure propagation during normal operation. In accident scenarios of sodium and lead fast reactors such as Unprotected Loss-Of-Flow (ULOF) or Unprotected Transient Over Power (UTOP), the fuel pins can be subjected to higher clad temperatures and fuel pin pressures or fuel clad mechanical/chemical interaction where thermal creep margin becomes significantly lower compared to the normal operation conditions. Therefore, possible stochastic failure and the post-failure fission gas/fuel jet impingement could be critical in order to predict fuel pin failure propagation. Pin depressurization due to fission gas release may degrade the heat transfer by formation of a gas blanket on a neighboring pin surface, which is a local phenomenon, and by causing coolant flow deceleration and starvation, which could affect a surrounding region as well. Furthermore, the potential presence of solid fuel particles or molten fuel at the time of clad failure could boost post-failure jet induced degradation even further. The present study models the U-Pu-Zr/HT9 metal fuel pin failure and stochastic clad damage propagation by biased sampling based on a Cumulative Damage Fraction (CDF) type clad failure criterion and the normal distribution of fuel failure probability density as a function of logarithm of Cumulative Damage Fraction. In addition, the effect of post-failure fission gas jet on heat transfer degradation is modeled for the target pins. This model is called stochastic Clad Damage Propagation (CDAP). The CDAP model is now fully integrated into developmental version of PRONGHORN-SC subchannel analysis code, allowing for modeling local failures and its propagation potential. Section 2 describes the components of the CDAP models. Section 3 describes the model implementation to PRONGHORN-SC and input specifications. Section 4 describes the CDAP model validation coupled to PRONGHORN-SC.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2005

Crane-Load Contact Sensor

An electronic instrument has been developed as a prototype of a portable crane-load contact sensor. Such a sensor could be helpful in an application in which the load rests on a base in a horizontal position determined by vertical alignment pins (see Figure 1). If the crane is not positioned to lift the load precisely vertically, then the load can be expected to swing once it has been lifted clear of the pins. If the load is especially heavy, large, and/or fragile, it could hurt workers and/or damage itself and nearby objects. By indicating whether the load remains in contact with the pins when it has been lifted a fraction of the length of the pins, the crane-load contact sensor helps the crane operator determine whether it is safe to lift the load clear of the pins: If there is contact, then the load is resting against the sides of the pins and, hence, it may not be safe to lift; if contact is occasionally broken, then the load is probably not resting against the pins, so it should be safe to lift. It is assumed that the load and base, or at least the pins and the surfaces of the alignment holes in the load, are electrically conductive, so the instrument can use electrical contact to indicate mechanical contact. However, DC resistance cannot be used as an indicator of contact for the following reasons: The load and the base are both electrically grounded through cables (the load is grounded through the lifting cable of the crane) to prevent discharge of static electricity. In other words, the DC resistance between the load and the pins is always low, as though they were always in direct contact. Therefore, instead of DC resistance, the instrument utilizes the AC electrical impedance between the pins and the load. The signal frequency used in the measurement is high enough (.1 MHz) that the impedance contributed by the cables and the electrical ground network of the building in which the crane and the base are situated is significantly greater than the contact impedance between the pins and the load. The instrument includes a signal generator and voltage-measuring circuitry, and is connected to the load and the base as shown in Figure 2. The output of the signal generator (typically having amplitude of the order of a volt) is applied to the load via a 50-resistor, and the voltage between the load and the pins is measured. When the load and the pins are not in contact, the impedance between them is relatively high, causing the measured voltage to exceed a threshold value. When the load and the pins are in contact, the impedance between them falls to a much lower value, causing the voltage to fall below the threshold value. The voltage-measuring circuitry turns on a red light-emitting diode (LED) to indicate the lower-voltage/ contact condition. Whenever the contact has been broken and the non-contact/higher-voltage condition has lasted for more than 2 ms, the voltage-measuring circuitry indicates this condition by blinking a green LED.

Youngquist, Robert↗
NASA NTRS2011

Electrical Connector Mechanical Seating Sensor

A sensor provides a measurement of the degree of seating of an electrical connector. This sensor provides a number of discrete distances that a plug is inserted into a socket or receptacle. The number of measurements is equal to the number of pins available in the connector for sensing. On at least two occasions, the Shuttle Program has suffered serious time delays and incurred excessive costs simply because a plug was not seated well within a receptacle. Two methods were designed to address this problem: (1) the resistive pin technique and (2) the discrete length pins technique. In the resistive pin approach, a standard pin in a male connector is replaced with a pin that has a uniform resistivity along its length. This provides a variable resistance on that pin that is dependent on how far the pin is inserted into a socket. This is essentially a linear potentiometer. The discrete approach uses a pin (or a few pins) in the connector as a displacement indicator by truncating the pin length so it sits shorter in the connector than the other pins. A loss of signal on this pin would indicate a discrete amount of displacement of the connector. This approach would only give discrete values of connector displacement, and at least one pin would be needed for each displacement value that would be of interest.

Arens, Ellen↗
DOE OSTI2016

TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment while safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional TREAT fuel elements to facilitate the experiment will not inhibit the ability to successfully simulate a RIA for the 2-pin or 3-pin bundle. This new water loop design leaves room for accommodating a larger fuel pin bundle than previously analyzed. The 7-pin fuel bundle in a hexagonal array with similar spacing of fuel pins in a SFR fuel assembly was considered the minimum needed for one central fuel pin to encounter the most correct thermal conditions. The 9-rod fuel bundle in a square array similar in spacing to pins in a LWR fuel assembly would be considered the LWR equivalent. MCNP analysis conducted on a preliminary LWR 9-rod bundle design shows that sufficient energy deposition into the central pin can be achieved well within range to investigate fuel and cladding performance in a simulated RIA. This is achieved by surrounding the flow channel with an additional annulus of water. Findings also show that a highly significant increase in TREAT to specimen power coupling factor (PCF) within the central pin can be achieved by surrounding the experiment with one to two rings of TREAT upgrade fuel assemblies. The experiment design holds promise for the performance evaluation of PWR fuel at extremely high burnup under similar reactor environment conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Switchable rocker arm

A rocker arm includes an outer arm defining a lock pin bore which is centered about a lock pin bore axis and an inner arm which selectively pivots relative to the outer arm about a pivot axis which is parallel to the lock pin bore axis. The inner arm includes a protrusion which defines an inner arm stop surface which is planar. A lock pin is disposed within the lock pin bore, the lock pin having a lock pin slot extending there into and also having a lock pin stop surface which is planar. The lock pin is displaced within the lock pin bore between a coupled position in which the lock pin stop surface is aligned with the inner arm stop surface and a decoupled position in which the lock pin slot is aligned with the protrusion.

42 ENGINEERING↗
DOE OSTI2016

Study on Modeling of Nuclear Fuel End Drop during a Hypothetical Accident

It is well known that shell elements have been used to model the cladding of single nuclear fuel pins. These have been used, for example, to analyze the response of a fuel assembly by assuming that all fuel pins in a fuel assembly deform in the same pattern. It is increasingly significant to model the entire fuel assembly in order to capture more realistically the behavior of each individual nuclear fuel during a hypothetical accident in a manner that can account for interactions among single pins. However, modeling each individual fuel pin in an entire fuel assembly using shell elements will result in models with large amount of elements and degrees of freedom that take correspondingly larger amount of computational times to execute. Therefore, modeling the nuclear fuel assembly with beam elements may become an option. This paper presents the results of a set of analysis with the LS-DYNA finite element software that investigated the differences and similarities of modeling nuclear fuel assemblies using either shell elements or beam elements. This study provides additional insights on the appropriateness of using beam element models of the fuel pins in the simulation of a hypothetical accident. In this study, two single pin models, a single pin model using shell elements and a single beam model using beam elements, are constructed to analyze end drop scenarios with the LS-DYNA code. Each model is then analyzed for a 19 ft end drop scenario (linear behavior) and 30 ft end drop scenario (non-linear behavior). Comparisons are made between the results obtained with the single pin shell model and with the single pin beam model to study the differences and similarities resulting from the two modeling approaches. In this study, a B and W 15x15 PWR fuel is analyzed for both 30 ft and 19 ft end hypothetical accident drops using both single pin finite element model with shell elements and beam elements for the fuel cladding. It is found that for the 19 ft end drop, the difference in the Max/Min total axial strains is about 12% between the two models. For the 30 ft end drop, the maximum total axial strains calculated with both models is similar, and a difference of about 12% is observed for the minimum total axial strains calculated by the two models. Advantages of the single pin shell model are that it permits consideration of internal pressures and may capture local buckling. Therefore, use of beam elements to model the fuel cladding needs to take into account the possible significance of local buckling and internal pressurization for the problem being analyzed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Benchmarking of calculation schemes in APOLLO2 and COBAYA3 for WER lattices

This paper presents solutions of the NURISP WER lattice benchmark using APOLLO2, TRIPOLI4 and COBAYA3 pin-by-pin. The main objective is to validate MOC based calculation schemes for pin-by-pin cross-section generation with APOLLO2 against TRIPOLI4 reference results. A specific objective is to test the APOLLO2 generated cross-sections and interface discontinuity factors in COBAYA3 pin-by-pin calculations with unstructured mesh. The VVER-1000 core consists of large hexagonal assemblies with 2 mm inter-assembly water gaps which require the use of unstructured meshes in the pin-by-pin core simulators. The considered 2D benchmark problems include 19-pin clusters, fuel assemblies and 7-assembly clusters. APOLLO2 calculation schemes with the step characteristic method (MOC) and the higher-order Linear Surface MOC have been tested. The comparison of APOLLO2 vs. TRIPOLI4 results shows a very close agreement. The 3D lattice solver in COBAYA3 uses transport corrected multi-group diffusion approximation with interface discontinuity factors of Generalized Equivalence Theory (GET) or Black Box Homogenization (BBH) type. The COBAYA3 pin-by-pin results in 2, 4 and 8 energy groups are close to the reference solutions when using side-dependent interface discontinuity factors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental Data Sets for CFD Calculations of Flow and Heat Transfer in Deformed Fuel Assemblies

Liquid-metal fast-breeder reactors (LMFBRs) have traditionally used fuel assemblies consisting of a hexagonal array of pins enclosed in a hexagonally shaped duct. Each fuel pin is surrounded by a wire that is wrapped in a helical configuration, which is used to separate the pins sufficiently to allow sodium coolant to flow through the assembly. The helical wire wrap also promotes mixing of the coolant. Experience with these assemblies has shown that they begin to deform with time during normal operation leading to significant changes in core thermal-hydraulics. For example, deformations of assemblies in the Experimental Breeder Reactor-2 and the Fast Flux Test Facility led to pin overheating resulting from pin-to-duct and pin-to-pin contact. This problem is well known and was investigated extensively in the 1970's and early 1980's. This was a period, however, before the development of modern Computational Fluid Dynamics (CFD) computer codes. Therefore, the experiments performed at the time did not record their results in sufficient detail to be useful for compiling validation datasets for CFD codes. Most of the experiments focused only on bulk measurements, such as the pressure drop across the bundle. When the effects of deformation of the fuel assembly were investigated, the studies performed to quantify these effects considered only bare pins without the wire wrap. With the resurgence of interest in LMFBRs in recent years, particularly in designs that leverage long-life performance of the fuel, the need has arisen for data that are suitable for the validation and benchmarking of CFD codes to predict the thermal-hydraulic behavior of flow in wire-wrapped fuel bundles. Such data must include high-resolution, transient, full-field measurements of the fluid velocity and temperature and an accurate record of the test conditions, including a precise description of the geometry. This paper describes a project to produce a validation-quality set of data that can be used to benchmark CFD codes for modeling flow and thermal performance in wire-wrapped fuel assemblies of the kind used by LMFBR designs. In addition, this effort investigates the changes to the flow and other phenomena that occur in the deformed geometries that can form later in a fuel assembly's life. This work is a joint effort of four organizations. The experimental work is being performed by AREVA and Texas A and M University. The computational work, both to support the experiments and to perform blind benchmark calculations, is being done by TerraPower, LLC, and Argonne National Laboratory (ANL). The experiments use a geometry that consists of a complete assembly of 61 wire-wrapped pins, which is surrounded by a hexagonal duct. This configuration is considered the minimum size necessary to adequately capture the important results of interest, such as the thermal gradients across the assembly and the velocity field effects in the assembly resulting from the surrounding duct. These types of effects cannot be captured by a geometry consisting of only a small group of fuel pins. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2001

Thermal Effects on the Bearing Behavior of Composite Joints

Thermal effects on the pin-bearing behavior of an IM7/PET15 composite laminate are studied comprehensively. A hypothesis presents factors influencing a change in pin-bearing strength with a change in temperature for a given joint design. The factors include the change in the state of residual cure stress, the material properties, and the fastener fit with a change in temperature. Experiments are conducted to determine necessary lamina and laminate material property data for the IM7/PET15 being utilized in this study. Lamina material properties are determined between the assumed stress free temperature of 460 F down to -200 F. Laminate strength properties are determined for several lay-ups at the operating temperatures of 350 F, 70 F, and -200 F. A three-dimensional finite element analysis model of a composite laminate subject to compressive loading is developed. Both the resin rich layer located between lamina and the thermal residual stresses present in the laminate due to curing are determined to influence the state of stress significantly. Pin-bearing tests of several lay-ups were conducted to develop an understanding on the effect of temperature changes on the pin-bearing behavior of the material. A computational study investigating the factors influencing pin-bearing strength was performed. A finite element model was developed and used to determine the residual thermal cure stresses in the laminate containing a hole. Very high interlaminar stress concentrations were observed two elements away from the hole boundary at all three operating temperatures. The pin-bearing problem was modeled assuming a rigid frictionless pin and restraining only radial displacements at the hole boundary. A uniform negative pressure load was then applied to the straight end of the model. A solution, where thermal residual stresses were combined with the state of stress due to pin-bearing loads was evaluated. The presence of thermal residual stresses intensified the interlaminar stresses predicted at the hole boundary in the pin-bearing problem. This dissertation shows that changes in material properties drives pin-bearing strength degradation with increasing temperature.

Walker, Sandra Polesky↗
NASA NTRS2009

Innovative Tools Advance Revolutionary Weld Technique

The iconic, orange external tank of the space shuttle launch system not only contains the fuel used by the shuttle s main engines during liftoff but also comprises the shuttle s backbone, supporting the space shuttle orbiter and solid rocket boosters. Given the tank s structural importance and the extreme forces (7.8 million pounds of thrust load) and temperatures it encounters during launch, the welds used to construct the tank must be highly reliable. Variable polarity plasma arc welding, developed for manufacturing the external tank and later employed for building the International Space Station, was until 1994 the best process for joining the aluminum alloys used during construction. That year, Marshall Space Flight Center engineers began experimenting with a relatively new welding technique called friction stir welding (FSW), developed in 1991 by The Welding Institute, of Cambridge, England. FSW differs from traditional fusion welding in that it is a solid-state welding technique, using frictional heat and motion to join structural components without actually melting any of the material. The weld is created by a shouldered pin tool that is plunged into the seam of the materials to be joined. The tool traverses the line while rotating at high speeds, generating friction that heats and softens but does not melt the metal. (The heat produced approaches about 80 percent of the metal s melting temperature.) The pin tool s rotation crushes and stirs the plasticized metal, extruding it along the seam as the tool moves forward. The material cools and consolidates, resulting in a weld with superior mechanical properties as compared to those weld properties of fusion welds. The innovative FSW technology promises a number of attractive benefits. Because the welded materials are not melted, many of the undesirables associated with fusion welding porosity, cracking, shrinkage, and distortion of the weld are minimized or avoided. The process is more energy efficient, safe (no toxic smoke or shielding gas, liquid metal splatter, arcing, dangerous voltage, or radiation), and environmentally sound (no consumables, fumes, or noise) than fusion welding. Under computer control, an automated FSW machine can create welds with high reproducibility, improving efficiency and overall quality of manufactured materials. The process also allows for welding dissimilar metals as well as those metals considered to be "unweldable" such as the 7xxx series aluminum alloys. Its effectiveness and versatility makes FSW useful for aerospace, rail, automotive, marine, and military applications. A downside to FSW, however, is the keyhole opening left in the weld when the FSW pin tool exits the weld joint. This is a significant problem when using the FSW process to join circumferential structures such as pipes and storage containers. Furthermore, weld joints that taper in material thickness also present problems when using the conventional FSW pin tool, because the threaded pin rotating within the weld joint material is a fixed length. There must be capability for the rotating pin to both increase and decrease in length in real time while welding the tapered material. (Both circumferential and tapered thickness weldments are found in the space shuttle external tank.) Marshall engineers addressed both the keyhole and tapered material thickness problems by developing the auto-adjustable pin tool. This unique piece of equipment automatically withdraws the pin into the tool s shoulder for keyhole closeout. In addition, the auto-adjustable pin tool retracts, or shortens, the rotating pin while welding a weld joint that tapers from one thickness to a thinner thickness. This year, the impact of the Marshall innovation was recognized with an "Excellence in Technology Transfer Award" from the Federal Laboratory Consortium.

Source record↗
DOE OSTI2025

Assessing High Burnup U-19Pu-10Zr Fuel Performance against Historical and Modeled Behavior

Advancing the deployment of sodium-cooled fast reactors (SFRs) requires thorough testing of metallic fuel pins under accident conditions to establish safe operational limits of high burnup fuel. To conduct transient testing, a comprehensive understanding of steady-state fuel behavior obtained through both experimental characterization and accurate predictive capabilities is needed. This study comparatively assesses the steady-state irradiation performance of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, irradiated under prototypic fast reactor conditions in preparation for planned safety testing at the Transient Reactor Test Facility. Since DP-40 was designated for use in the test and DP-36 serves as its sibling pin, non-destructive, engineering-scale post-irradiation examinations (PIE) were conducted on both pins while destructive examinations were performed exclusively on DP-36. The results were then assessed against historical performance data from similar fuel pins irradiated in the Experimental Breeder Reactor-II. Additionally, the steady-state irradiation of each pin was modeled using the BISON fuel performance code to assess the accuracy of current modeling capabilities in predicting the baseline irradiation behavior. Non-destructive examinations included neutron radiography to measure fuel column elongation, gamma scanning to verify pin integrity and fission product migration, and profilometry to assess dimensional changes. Benchmarking against existing PIE data revealed consistent patterns in axial fuel column growth and cladding diametral strain, though both pins exhibited longer low-density “fluff” structures, which can have implications for core reactivity and source term calculations. Destructive examinations on DP-36 included fission gas release analysis and sectioning for optical microscopy, which showed more complex constituent redistribution patterns than the traditionally accepted 3-ring model. The axial evolution of fractional areas and porosities of each of the redistributed zones were quantified and presented. Modeling comparisons showed agreement in fractional fission gas release but consistently overestimated axial and radial swelling and disagreed with measured axial porosity patterns. These conservative overpredictions suggested that the pins would appear closer to failure or operational limits at the start of transient tests, potentially leading to higher strain accumulation during the transient. While conservative estimates provide safety margins, they can negatively impact fuel economics. A review of the swelling models identified areas for improvement in the gaseous swelling, solid swelling, and fuel hot-pressing models when applied to ternary fuel. The results of this study highlight the critical importance of conducting pre-test characterization on both test and sibling pins to accurately capture steady-state fuel behavior, providing a precise baseline for post-test evaluations and essential inputs for transient modeling of the planned experiments. The analysis also revealed significant data gaps that require further investigation to enhance the understanding and prediction of fuel swelling and pore dynamics. Collecting comprehensive data across different irradiation conditions, burnup levels, and fuel compositions are essential for refining existing models and developing mechanistic models for both binary and ternary metallic fuels, ultimately improving the integration of modeling and experimental approaches in accident testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of Critical Experiments to Benchmark Moderator Temperature Reactivity Worth

The simulation of nuclear reactors requires being able to accurately model several different interrelated physical processes including neutronics, heat production, fluid/thermal physics, structural mechanics, fuel behavior, chemistry, and balance-of-plant. In order to design and use new types of nuclear reactors, modern codes have been developed which couple together the behavior of these different physical models to accurately consider the feedback effects between different physical effects. The SHARP toolset, part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project, is an example of an effort to combine models of neutronics, thermal-hydraulics, and structural mechanics in a coupled simulation to model any type of nuclear reactor as accurately as possible. The validation of multi-physics coupling techniques requires new experimental benchmark experiments specifically designed for that purpose. As a preliminary step towards design of these experiments, a series of measurements have been performed to demonstrate the temperature-dependent neutronic behavior of a well-characterized reactor assembly for comparison with the available codes. The facility adopted for these experiments is Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI). The RCF is unique in several aspects and provides several advantages to this type of experiment. Built in 1954 by the American Locomotive Company (ALCO), the RCF has been used to design small compact reactors. RPI acquired the facility in 1963, and the highly-enriched uranium fuel was later swapped for low-enriched fuel pins that were excess from the SPERT facility experiments. Nowadays, The RCF is regarded as a teaching and research facility, permitting the training of licensed Senior Reactor Operators, and offering a critical experiments laboratory course to students. The reactor core standard configuration consists of 332 or 333 UO{sub 2} fuel pins, containing a total of 35.2 grams of U- 235 each (4.81 % enriched). The pins are arranged in a regular lattice with a pitch of 1.6256 cm. The number and configuration of these pins are flexible, and the reactor's overall thermal power is limited to no more than 15 W. As the low thermal output of the reactor does not induce any measurable temperature changes, the effect of temperature in the reactor is observed by applying two 18- kW electric heaters to the water moderator, with mechanical agitators to maintain uniformity in moderator temperature. Before beginning experiments, each fuel pin inside the core was surveyed for mechanical defects (bending or warping) that may contribute to uncertainty in core behavior. Every fuel pin has been individually removed from the core and checked for any irregularities on its surface (caused by shocks or usage). About 12 % of the 333 pins were found to have a small degree of warping and were replaced with excess pins available in the facility. In doing this, we made sure all the pins are identical, thereby increasing the accuracy of future models. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA

In this paper, we present post-irradiation examination results on two type of annular mixed oxide fuel pins irradiated in the Fast Flux Test Facility (FFTF) sodium cooled reactor to an average burnup between 4% and 5% fission of initial heavy atom (FIMA). The pins differed only from the initial Pu content, which was 22 wt% and 26 wt%, respectively. The overall performance of the pins was excellent, in line with previous historical results. The pins with higher Pu content experienced higher irradiation temperatures which influenced the fission gas release, fuel swelling, and Cs distribution compared to the other pins. All the post-irradiation examinations results are discussed against the irradiation parameters. In particular, the pins with higher initial Pu content, i.e., 26 wt%, experienced higher power that resulted in enhanced fission gas release compared to the other two pins with 22 wt% initial Pu content. For the pins with higher fission gas release, onset of Cs redistribution was observed. The two pins that had lower initial Pu content and burnup showed a Cs axial distribution similar to the as-produced one.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Coupled Neutron and Gamma Heating Calculation Based on VARIANT Nodal Transport Method

Accurate determination of temperature distributions in fuel pins and duct walls is essential for fuel-pin performance and assembly bowing analyses. Currently, the multi-assembly, sub-channel analysis code SE2-ANL is used in thermalhydraulic analysis of sodium cooled fast reactors (SFR). SE2- ANL is interfaced with the DIF3D/GAMSOR/DIF3D procedure for coupled neutron and gamma heating calculations, developed at Argonne National Laboratory (ANL). In this procedure, the neutron flux distribution is first calculated using the finite-difference diffusion theory option of the DIF3D code. Using the neutron flux solution, the GAMSOR code computes the gamma source distribution for the subsequent gamma flux calculation with DIF3D. However, in the current heating calculation scheme, the neutron and gamma flux calculations are limited to six triangular meshes per hexagonal assembly. In addition, the pin power distributions are estimated by assuming that the power distribution within an assembly is separable in the radial and axial directions. In each assembly, the axial power profile is obtained by averaging the power densities of six triangular meshes at each axial mesh interval. The radial power profile is evaluated at a single axial mesh (e.g., at the core mid-plane) by approximating the neutron and gamma flux distributions by a linear or quadratic polynomial of two variables x and y, depending on the polynomial order specified by the user. The coefficients of the linear and quadratic polynomials are determined by the least squares fit of six and 24 triangular mesh fluxes, respectively. Furthermore, the total heat generation rate in an assembly duct wall is simply determined by a user-specified fraction of the total gamma heating within that assembly. With the advancement of computer technology, the nodal transport theory code VARIANT is routinely used in core design analyses of SFR instead of DIF3D. It is therefore necessary to develop a new heating calculation procedure based on VARIANT neutron and gamma transport calculations. In this study, a new, coupled neutron and gamma heating calculation scheme has been developed to determine the pin power distributions more accurately by eliminating the aforementioned limitations of the current procedure. For this, the GAMSOR code has been extended to generate the intra-nodal gamma source distribution to be consistent with the trial functions of VARIANT. In the new heating calculation procedure, the neutron flux distribution is determined first by solving an eigenvalue transport problem using VARIANT. The obtained neutron flux is then used in GAMSOR to calculate the gamma source distribution. Using this gamma source distribution, the gamma flux distribution is determined by solving a fixed-source transport problem using VARIANT again. Using the computed neutron and gamma flux distributions, the power distribution in each fuel pin and duct wall is determined using a new utility code CURVE, which has been developed to evaluate the VARIANT flux solution at each fuel pin position and along the duct mid-wall. For each pin segment contained in a VARIANT node, the axial power shape is represented as a quadratic profile. The performance of the new heating calculation scheme based on VARIANT transport solutions has been tested against the existing method of SE2-ANL, using the ABR-1000 metal core design. The gamma source distribution and the assembly, fuel pin and duct wall power distributions were calculated and compared. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Impact of Radial Reflector Fidelity on Neutronics and Vessel Fluence Simulations

The Consortium for Advanced Simulation of Light Water Reactors is developing the Virtual Environment for Reactor Applications (VERA), and the MPACT code, which is the primary deterministic neutron transport solver in VERA, provides sub-pin level flux and power distributions as part of full-scale cycle depletion and analysis. In such calculations, an important aspect is the radial reflector treatment. To improve the fidelity of the radial reflector treatment, MPACT was extended to approximate the modeling of the reactor’s structural components such as the core shroud, barrel, neutron pads, and vessel. This work explores several modeling configurations with varying levels of fidelity and computational burden and assesses the importance of modeling fidelity on the eigenvalue and pin power distribution. Two two-dimensional (2-D) problems were analyzed to assess the impact on eigenvalue and pin power distributions with low-fidelity, coarse square cell reflector representations: (1) a Watts Bar Nuclear Plant Unit 1 (WBN1) quarter-core slice with depletion and (2) an AP1000 quarter-core slice. In this work, the analyses showed that the effect on eigenvalue is fairly small, but the effect on pin power is more pronounced, especially locally in the assemblies closest to the periphery, where the maximum pin power difference is nearly 3.5% in the AP1000 case. Two additional 2-D problems were used to assess the comparison between the low-fidelity coarse square cell treatment and a high-fidelity geometric representation that uses subpin material specification: (1) the same WBN1 quarter-core slice and (2) a representative model of the NuScale small modular reactor (SMR), which features a solid reflector design with moderator holes. These results demonstrate that even a coarse, low-fidelity representation adequately captures the necessary simulation characteristics. Last, these capabilities were applied to the 2-D WBN1 quarter-core depletion to assess the impact on vessel fluence using VeraShift. From adjoint calculations, pins along the periphery were observed to be of highest importance for fluence calculation, so the impact of the reflector representation in MPACT could theoretically substantially affect the predicted result. However, it was observed that the change in pin powers along the periphery minimally impacts the maximum vessel fluence with a difference within the statistical uncertainty but provides terrific insight on the sensitivity of the peripheral pins.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Report on UQ Assessments to support SESAME wire-wrapped bundle experiment

This work assesses the influence of assumptions made when generating a mesh of a wire-wrapped geometry. The contact region between a wire and its adjacent pin is commonly modeled by either embedding the wire to the adjacent pin or trimming the wire so that a gap separates the wire from its adjacent pin. These models are referred to as close-gap and open-gap approaches herein and are applied to two geometries. The first geometry consists of a single pin wire-wrapped subchannel. A polyhedral mesh and a hexahedral mesh are generated. The second and third geometry are a 7-pin and a 19-pinwire-wrapped bundles meshed with polyhedral elements only. Pressure drops are obtained with the STAR-CCM+computational fluid dynamic package. Sensitivity analyses of the mesh density, the mesh type, and the turbulent models are performed. Numerical results show that the best match to the experimental data and to the Cheng-Todreas correlation is obtained with the combination of a hexahedral mesh, the shear stress transport (SST) turbulent model, and the open-gap approach. In the case of the 7-pin geometry, the best results are obtained with the open-gap approach and the SST turbulent model. The 19-pin geometry yields contradictory results to the 7-pin geometry results, and thus will require further investigations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1992

Coupling device with improved thermal interface

The primary object of the present invention is to provide a simple, reliable, and lightweight coupling that will also have an efficient thermal interface. A further object of the invention is to provide a coupling that is capable of blind mating with little or no insertion forces. Another object of the invention is to provide a coupling that acts as a thermal regulator to maintain a constant temperature on one side of the coupling. Another object of the invention is to increase the available surface area of a coupling thus providing a larger area for the conduction of heat across the thermal interface. Another object of the invention is to provide a fluidic coupling that has no fluid passing across the interface, thus reducing the likelihood of leaks and contamination. The foregoing objects are achieved by utilizing, as in the prior art, a hot area (at an elevated temperature as compared to a cold area) with a need to remove excess heat from the hot area to a cold area. In this device, the thermal interface will occur not on a planar horizontal surface, but along a non-planar vertical surface, which will reduce the reaction forces and increase the thermal conductivity of the device. One non-planar surface is a surface on a cold pin extending from the cold area and the other non-planar surface is a surface on a hot pin extending from the hot area. The cold pin is fixed and does not move while the hot pin is a flexible member and its movement towards the cold pin will bring the two non-planar surfaces together forming the thermal interface. The actuating member for the device is a shape-memory actuation wire which is attached through an aperture to the hot pin and through another aperture to an actuation wire retainer. By properly programming the actuation wire, heat from the hot area will cause the actuation wire to bend the hot wire. Heat from the hot area will cause the actuation wire to bend the hot pin towards the cold pin forming the coupling and the desired thermal interface. The shape-memory actuation wire is made of a shape-memory-effect alloy such as Nitinol.

Milam, Malcolm Bruce↗
NASA NTRS2008

Gimballed Shoulders for Friction Stir Welding

In a proposed improvement of tooling for friction stir welding, gimballed shoulders would supplant shoulders that, heretofore, have been fixedly aligned with pins. The proposal is especially relevant to self-reacting friction stir welding. Some definitions of terms, recapitulated from related prior NASA Tech Briefs articles, are prerequisite to a meaningful description of the proposed improvement. In friction stir welding, one uses a tool that includes (1) a rotating shoulder on top (or front) of the workpiece and (2) a pin that rotates with the shoulder and protrudes from the shoulder into the depth of the workpiece. In conventional friction stir welding, the main axial force exerted by the tool on the workpiece is reacted through a ridged backing anvil under (behind) the workpiece. When conventional friction stir welding is augmented with an auto-adjustable pin-tool (APT) capability, the depth of penetration of the pin into the workpiece is varied in real time by a position- or forcecontrol system that extends or retracts the pin as needed to obtain the desired effect. In self-reacting (also known as self-reacted) friction stir welding as practiced heretofore, there are two shoulders: one on top (or front) and one on the bottom (or back) of the workpiece. In this case, a threaded shaft protrudes from the tip of the pin to beyond the back surface of the workpiece. The back shoulder is held axially in place against tension by a nut on the threaded shaft. Both shoulders rotate with the pin and remain aligned coaxially with the pin. The main axial force exerted on the workpiece by the tool and front shoulder is reacted through the back shoulder and the threaded shaft into the friction-stir-welding machine head, so that a backing anvil is no longer needed. A key transmits torque between the bottom shoulder and the threaded shaft, so that the bottom shoulder rotates with the shaft. This concludes the prerequisite definitions of terms.

Carter, Robert↗