Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Transient Testing”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

ATF Transient Prescription Tests in TREAT

EXECUTIVE SUMMARY The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s and provided thousands of transient irradiations before being placed in standby mode in 1994. In 2017, it resumed operation in order to reclaim its crucial role in nuclear-heated safety research. TREAT is a transient reactor where inherently-safe core physics, a nimble transient rod drive system, and a philosophy of continual facility improvement work together to enable flexible power maneuvers for fuel safety research. While the facility’s flexible design and multi-mission nature saw historic experiments for numerous transient types, TREAT was not specifically adapted to optimize or demonstrate its full performance capability for Light Water Reactor (LWR) postulated accidents including Hot Zero Power Reactivity Initiated Accidents (HZP-RIA) and Loss of Coolant Accidents (LOCA). The physics of HZP-RIA drive very brief pulses with fairly large energy excursions. TREAT is currently capable of approaching HZP-RIA pulse widths through transients initiated on a large step insertion of reactivity and terminated shortly thereafter by rapid re-insertion of the transient rods (“clipping”). LOCAs are postulated to occur quite differently, starting from normal power operations followed by a long period of lower power decay heat. TREAT is able to simulate LOCA events with an initial flattop transient for tens of seconds, followed by a sizeable power reduction, to a longer segment where specimen fission heating represents internal decay heat generation. Two reactor physics testing campaigns were undertaken to address the HZP-RIA type transients, neither of which had been performed in any great measure historically to address data gaps in the timing, repeatability of clipping and effect of dynamic transient rod motions on core-to-specimen fission power coupling. The first campaign, referred to as the Accident Tolerant Fuel (ATF) transient prescription tests, employed dosimeters in the core to characterize the effect of rod motions on specimen power coupling for clipped, unclipped, and low-level steady state operations each with approximately the same moderate core energy release. This campaign was accompanied by efforts to measure dosimetry and compare results to analytic predictions. The other HZP-RIA campaign, referred to as Narrow Pulse Width (NPW) transient prescriptions, employed higher step insertions with increased core energy release to demonstrate TREAT’s capability in achieving the narrowest-possible pulse width. Additionally, a series of LOCA transient prescriptions were also performed to demonstrate how well TREAT could perform long-duration shaped transients. Both NPW and LOCA transient prescriptions did not include dosimeters but used existing plant power instrumentation to produce transient power data. All three transient prescription campaigns were successfully carried out in TREAT. The ATF-3 transient prescriptions showed that TREAT can perform clipped transients with remarkable repeatability. Measurement and modeling both showed that the power coupling in small, fissile dosimeters does not change by more than 8% between clipped, unclipped, and steady state runs while uncertainty sources in both the modeling and measurement methods were identified and found to be on the order of 20% in total. Overall, the ATF-3 transient prescription tests were successful in determining repeatability and consistency of dosimeter power coupling, developing processes for modeling and measuring this crucial parameter, and demonstrating the need for continued efforts in transient prescription work at TREAT in support of future experiment design and post test data interpretation. The NPW and LOCA transient prescriptions were also successfully performed to exhibit TREAT’s capability for these important LWR accident categories. The NPW transients showed pulse widths of ~90 ms, which were the narrowest ever achieved in TREAT to date and relevant for the upper end range of HZP-RIA simulation. This effort provided crucial data for calibrating kinetic models to be used in enhanced pulse-narrowing strategies in upcoming irradiations. Similarly, The LOCA transient prescriptions showed that TREAT can simulate the needed nuclear heating and event time scales for LOCA fuel safety research and provided essential data for future experiment design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

Critical↗
DOE OSTI2024

Development and demonstration of a BISON–Griffin modeling framework for the design of targeted TRISO transient experiments in the Transient Reactor Test Facility

Uranium oxycarbide (UCO)-bearing tri-structural isotropic (TRISO) particle fuels are expected to be used in numerous U.S. commercial reactor applications within the next decade. Here, in this work, we reviewed historical particle fuel transient experiments to identify gaps in TRISO fuel performance transient testing. A BISON–Griffin modeling framework was then developed to conduct preliminary TRISO transient analyses and begin to address these gaps. The framework was demonstrated using limiting-case transient conditions from a prototypic high-temperature gas-cooled reactor (HTGR). It was then applied to develop a matrix of experiments that could be performed in the Transient Reactor Test Facility (TREAT) to (1) evaluate UCO-fueled particle performance at moderate and high heat rates, (2) assess whether historical testing involving UO 2 -fueled particles is applicable to modern UCO-fueled particles, (3) deconvolute the impacts of temperature and heat rate on particle transient response, and (4) collect the data needed for fuel performance model validation and/or further development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1991

The new Section 23 of DO160C/ED14C lightning testing of externally mounted electrical equipment

The new Section 23 is introduced which has only very recently been fully approved by the RTCA for incorporation into the first revision of DO160C/ED14C. Full threat lightning direct effects testing of equipment is entirely new to DO160, the only existing lightning testing is transient testing for LRU's (Line Replaceable Units) by pin or cable bundle injection methods, for equipment entirely contained within the airframe and assumed to be unaffected by direct effects. This testing required transients of very low amplitude compared with lightning itself, whereas the tests now to be described involve full threat lightning testing, that is using the previously established severe parameters of lightning appropriate to the Zone, such as 200 kA for Zone 1A as in AC20-136. Direct effects (i.e., damage) testing involves normally the lightning current arc attaching to the object under test (or very near to it) so submitting it to full potential for the electric, mechanical, thermal and shock damage which is caused by high current arcing. Since equipment for any part of the airframe require qualification, tests to demonstrate safety of equipment in fuel vapor regions of the airframe are also included.

Burrows, B. J. C.↗
NASA NTRS1988

Transient Pressure Test Article (TPTA) 1.1 and 1.1A, volume 1

This final test report presents the results obtained during the static hot firing and cold-gas high Q tests of the first Transient Pressure Test Article (TPTA) 1.1. The TPTA consisted of field test joints A and B, which were the original RSRM J-insulation configuration, with a metal capture feature. It also consisted of a flight configuration nozzle-to-case test joint (Joint D) with shorter vent slots. Fluorocarbon O-rings were used in all the test joints. The purpose of the TPTA tests is to evaluate and characterize the RSMR field and nozzle-to-case joints under the influence of ignition and strut loads during liftoff anf high Q. All objectives of the cold-gas high Q (TPTA 1.1A) test were met and all measurements were close to predicted values. During the static hot-firing test (TPTA 1.1), the motor was inadvertently plugged by the quench injector plug, making it a more severe test, although no strut loads were applied. The motor was depressurized after approximately 11 min using an auxiliary system, and no anomalies were noted. In the static hot-firing test, pressure was incident on the insulation and the test joint gaps were within the predicted range. During the static hot-firing test, no strut loads were applied because the loading system malfunctioned. For this test, all measurements were within range of similar tests performed without strut loads.

Rebells, Clarence A.↗
DOE OSTI2017

Modeling the Transient Reactor Test Loop in TRACE for PWR Critical Heat Flux Benchmarking

Understanding transient critical heat flux (CHF) resulting from a Reactivity Insertion Accident (RIA) in nuclear systems is an extremely undeveloped field. Though many experiments have been done for CHF, remarkably few have been done for rods at PWR operating temperature and pressure (300 deg. C, 2200 psia) with a short time constant ramp in power. This leads to many concerns regarding understanding how RIA scenarios develop and evolve in commercial nuclear reactors. With a limited understanding of the phenomenology governing transient CHF, there is little hope in being able to accurately model these scenarios using today's thermal hydraulic codes. The Accident Tolerant Fuels (ATF) Program established by the Department of Energy (DOE) is seeking out new fuel and cladding systems that can be used to reduce the probability and consequences of severe accidents. To better understand this phenomenon, the DOE is restarting the Transient Reactor Test (TREAT) Facility at the Idaho National Laboratory (INL). TREAT will be able to provide representative neutronic boundary conditions for testing coupled thermal hydraulics on the Transient Water Environment Recirculating Loop (TWERL) to study transient CHF with the best representation possible. To support this task, Oregon State University (OSU) is designing and building a representative flow loop that will be able to mimic the conditions of TWERL (sans the neutronic boundary conditions) to create a benchmark study for existing lumped parameter thermal-hydraulic codes such as RELAP5 and TRACE. This flow loop, the Transient Reactor Test Facility (TRTL), will support the ATF program by providing much needed NQA-1 data for transient CHF at operating temperature and pressure.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2019

A Simple Transient Thermal Test Assembly for Insulation Materials

A simple transient thermal test was developed for evaluating the thermal performance of high-temperature flexible insulation materials in atmospheric pressure air. The test setup was inspired by the test assembly used for evaluating the performance of fire shelters and fire protective clothing. The heating source used was a burner supplied with propane gas that can generate relatively uniform heating over a wide area. The ratio of the in-plane width to thickness of the insulation test sample in the test assembly was selected to be greater than ten so that the heat transfer in the center of the test assembly was nearly one-dimensional. A rigid, thin Inconel plate was used as the septum plate directly exposed to the flame to provide a relatively uniform, known-temperature boundary condition for the test assembly, and to prevent convective heat transfer through the test insulation from the burner. The test sample, flexible alumina paper insulation, was placed between the Inconel plate and a thin titanium witness plate. The overall assembly was further insulated using a combination of rigid and flexible ceramic insulations to minimize heat losses from the periphery of the test assembly. Thermocouples installed on the septum and witness plates and inside the insulation test sample provided temperature measurements at various locations. For the evaluation tests, the Inconel septum plate reached temperatures between 780°C and 850°C for tests with flame exposure durations of 120 sec to 300 sec. The measured temperatures for various tests with similar flame exposure times were repeatable. A one-dimensional numerical heat transfer model of the test assembly was developed. The close agreement between measured and predicted temperatures on the titanium witness plate and inside the insulation test sample indicated that this test may provide a one-dimensional thermal testing capability for evaluation of thermal performance of similar insulations.

Daryabeigi, Kamran↗
DOE OSTI2017

Spark Plasma Sintering (SPS) Update for the Fuel Conversion Effort at the Transient Reactor Test Facility

Over the lifetime of the Transient Reactor Test Facility (TREAT) reactor from 1959 to 1994, a series of historical tests disclosed certain material specific requisites for irradiation testing variables resulted in negligible structural damage after exposure to the reactor and multiple experiments. Comparable fuel design, material testing, and qualifications needs to occur for the high-enriched uranium (HEU) to low enriched uranium (LEU) fuel conversion (<20% U235) of TREAT. This includes, retaining a uniform precipitate dispersion of fueled UO2 micron-size particles throughout a graphite-moderating matrix, where high-density graphite is in direct contact with the fuel and acts as a moderator. On that note, the design of future LEU reactor TREAT fuel core will consider the relative sizing and spacing of fuel as well as the retention of graphite in accordance with the expected thermal, neutron, and energy portfolio for future conversion. In light of the above, LEU conversion fuel blocks currently being manufactured must undergo critical materials testing, irradiation, and examination to determine effects on the manufactured replacement fuel blocks at the millimeter to sub-micron size scale. Herein, we will report and update the community on value-added studies focusing on spark plasma sintering as an alternative technique to fabricate high density graphite dispersion fuel.

36 MATERIALS SCIENCE↗
NASA NTRS2015

Transient Thermal Testing and Analysis of a Thermally Insulating Structural Sandwich Panel

A core configuration was devised for a thermally insulating structural sandwich panel. Two titanium prototype panels were constructed to illustrate the proposed sandwich panel geometry. The core of one of the titanium panels was filled with Saffil(trademark) alumina fibrous insulation and the panel was tested in a series of transient thermal tests. Finite element analysis was used to predict the thermal response of the panel using one- and two-dimensional models. Excellent agreement was obtained between predicted and measured temperature histories.

Blosser, Max L.↗
DOE OSTI2016

Safety Analysis Challenges With Restart of the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) Facility

Safety Analysis Challenges With Restart of the Idaho National Laboratory Transient Reactor Test (TREAT) Facility Dr. James Parry David Broussard Anthony LaPorta Douglas Gerstner Battelle Energy Alliance Idaho National Laboratory P.O. Box 1625/MS6194 Idaho Falls, ID 83415 The Transient Reactor Test Facility, commonly referred to as TREAT, is an air-cooled, thermal spectrum test reactor facility designed to: • Induce intense fission heating in the nuclear fuel being tested. • Test nuclear reactor fuels under severe reactor-accident conditions. • Study fuel meltdowns, metal-water reactions, interactions between overheated fuel and coolant, and the transient behavior of fuels for high-temperature systems. • Provide nondestructive test data through neutron radiography of fuel samples. • Allow for the detailed monitoring of the experiments during the test. TREAT was specifically designed to test prototypic-sized reactor fuel pins and bundles under transient power conditions. During its 35 year operating history, it conducted thousands of transient experiments. Today, these capabilities are required for predicting safety margins for next generation fuels and advanced reactor design safety. Even though TREAT has not operated since 1994, return to operable conditions and resumption of testing will occur in FY 2018. This paper discusses the unique challenges with performing and documenting the safety analyses that demonstrate that the TREAT restart may be performed safely. This includes a discussion of 1) the TREAT capability, design, and operations, 2) identification of the format and content of the required safety basis document changes, and 3) demonstration that the consequences of potential accidents during TREAT operations meet the consequence guidelines.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Neutronics Analysis of TREAT Multi-SERTTA Calibration Test Vehicle (Multi-SERTTA CAL)

Transient testing is a necessary step in the performance evaluation of Accident Tolerant Fuels (ATF) developed as part of the Fuel Cycle Research and Development Advanced Fuels Campaign initiated and supported by the United States Department of Energy (DOE-NE). These advanced fuels are expected to be used in future reactors or replace the current fuel design used in existing power reactors for enhanced performance and safety. Upon completion of a process proscribed by the National Environmental Policy Act, the Transient Reactor Test (TREAT) facility, located at Idaho National Laboratory (INL), was chosen by DOE-NE to provide the desired specialized testing. TREAT has successfully operated from start-up in 1959 to 1994 when it was placed on operational stand-by. Careful preparations are being made for operational readiness and TREAT is projected to resume testing no later than 2018. Past TREAT testing primarily focused on sodium-cooled reactor fuels (SRF). Test vehicle designs varied to accommodate testing needs ranging from separate effects tests to integral tests approaching simulation of actual reactor environment conditions including forced convection cooling. The current focus on light water reactor (LWR) ATF requires design of new experiment vehicles to accommodate the water/steam environment and associated complexities regarding desired energy deposition into the specimen and proper containment. The first set of experiments in TREAT restart is expected to be conducted with the Multi-SERTTA (Static Environment Rodlet Transient Test Apparatuses). Difficult to quantify and calculate integral and time dependent effects inherent in transient testing limit predictive modeling capability; therefore, calibration testing in TREAT prior to the actual experiment is necessary to empirically determine the effectiveness of core-to-specimen power coupling in order to achieve the desired energy deposition into the test specimen during a nuclear power transient. This has been done in the past, in part, by evaluating the difference in energy deposition into dosimeter wires subjected to various TREAT time dependent power conditions. These calibration tests were typically conducted in a test vehicle neutronically similar to the actual test vehicle, but more simplistic in design in order to quickly retrieve and analyze dosimeter wires and minimize containment structure costs due to several trials. Prior calibration testing for SRF experiments could be conducted without coolant as sodium has no notable neutronic effect on the core-to-specimen power coupling factor (PCF). Such is not the case with LWR fuel where water serves as both moderator and coolant. This presents a design challenge as the calibration test vehicle must retain heated, high-pressure water to preserve neutronic equivalence representative of near prototypic reactor conditions while allowing for quick retrieval of dosimeters. The solution to this challenge comes in the form of a thin walled dry well positioned through the center of the basic Multi-SERTTA design. Test vehicle reactivity worth and PCF are two important factors with respect to ascertainment of neutronic equivalence. Similarity in test vehicle reactivity worth is desired to avoid the need to change operational core rod positions in TREAT which could significantly alter neutron flux distribution and consequently power distribution throughout the core and into the test specimen. The geometry and composition of various materials within the experiment containment structure can also significantly alter the quality of neutron signal delivery from the core to the test specimen. Introduction of an Inconel tube, albeit very thin, is expected to impact PCF. Simple modifications to the CAL design may compensate for the anticipated change. Evaluation of these factors and proposed modifications through neutronics analysis serves to inform experiment design and is a valuable tool towards achieving the optimal CAL design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1997

A Technique for Transient Thermal Testing of Thick Structures

A new open-loop heat flux control technique has been developed to conduct transient thermal testing of thick, thermally-conductive aerospace structures. This technique uses calibration of the radiant heater system power level as a function of heat flux, predicted aerodynamic heat flux, and the properties of an instrumented test article. An iterative process was used to generate open-loop heater power profiles prior to each transient thermal test. Differences between the measured and predicted surface temperatures were used to refine the heater power level command profiles through the iteration process. This iteration process has reduced the effects of environmental and test system design factors, which are normally compensated for by closed-loop temperature control, to acceptable levels. The final revised heater power profiles resulted in measured temperature time histories which deviated less than 25 F from the predicted surface temperatures.

Horn, Thomas J.↗
DOE OSTI2016

Uncertainty and Sensitivity Analysis of Reactivity Insertion Accidents using BISON Fuel Performance Code

The development of Accident Tolerant Fuels (ATF) has expanded into a considerable research program in the wake of the events at the Japanese Fukushima Daiichi nuclear power plant in 2011. The ATF goal is to insert a test assembly into a commercial reactor by 2022. To support this, testing of the ATF design under off-normal conditions is needed to assess the 'enhanced accident tolerance' capability of the fuel rod design over standard UO{sub 2}-Zircaloy concepts. Transient testing capabilities are necessary for off-normal testing of events such as loss of coolant accidents (LOCA) and reactivity insertion accidents (RIA). A transient testing program has been developed to support the testing of ATF designs. The Transient Reactor Test (TREAT) facility has been chosen as the site to perform the transient testing of the ATF designs. TREAT is a versatile test facility capable of subjecting fuel specimens to various off-normal conditions. Historic TREAT experiments were performed in a variety of experimental test vehicles including static environment capsules, once-through flowing systems, and also recirculating flowing systems. The planned TREAT experiments will be operated under PWR conditions. Therefore, a static capsule vehicle is under development to handle required temperatures and pressures. The static capsule, termed the Static Environment Rodlet Transient Test Apparatus (SERTTA), will be used for irradiation tests in TREAT to subject ATF fuel rodlets to a range of power pulse conditions and environments. To aid in the design and analysis of future tests in TREAT, the fuel performance code BISON is being used. Fuel performance codes capable of accident modeling have been developed for some time and they have shown an ability to reproduce experimental results with some degree of adequacy, but they still require extensive verification and validation. Future experiments in TREAT will add to a limited database of past transient tests performed in transient reactors to support LWR systems. These tests can be used for future verification and validation of fuel performance codes. Understanding the impact of the uncertainty of influential parameters is critical for not only analyzing the results of the experiment but also for modeling and validation of codes. For example, uncertainties associated with the initial manufactured state of the sample, instrument measurements, material property models, and planned experiment conditions can provide important insight into model performance, experimental data, and play an important role in experimental design. The objective of this work is to perform an uncertainty and sensitivity analysis on a simplified RIA case of a standard UO{sub 2}-Zircaloy sample using expected nominal conditions in the SERTTA capsule in TREAT. The BISON fuel performance code is used in conjunction with DAKOTA (Design Analysis Kit for Optimization and Tera-scale Applications) to perform the uncertainty and sensitivity analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗