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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Submersible Multistage Centrifugal Pump for Versatile Test Reactor Cartridge Test Loop

Submersible multistage centrifugal pumps are ideal for pumping in narrow confined spaces and achieving necessary head pressures and flow rates. Once a diameter is determined then manipulation of the number of stages and motor speed are all that are required to meet desired flow conditions. The Versatile Test Reactor (VTR) closed loop cartridge systems will need forced convection cooling independent of the main reactor. A multistage centrifugal pump can meet the necessary flow rates and pumping pressures while minimizing space taken. The pump considered for this work was based off a deep well submersible pump, a variation of a multistage centrifugal pump. We experimented with two pump sizes, 5 cm (2 inch) and 7.5 cm (3 inch) diameters. These diameters were chosen to fit into the inner diameter of standard 5 and 7.5 (2 and 3 inch) Schedule 40 pipe, respectively. This made the design for the test loop both simpler and less expensive as the need for an engineered pump housing was eliminated. Initial test cartridge planning indicated space for only a 5 cm (2 inch) diameter pump, though early testing of this size showed the need for an abnormally high-speed and high-power motor. Fine tuning of the cartridge design allowed a pump size increase to 7.5 cm (3 inches), which was the pump size most extensively tested in this work. The test loop is composed of various sizes of PVC and aluminum piping components in a loop configuration. The pump is driven by a Pittman 250 W (1/3 horsepower) electric motor with maximum speed of 3,450 RPM. Testing consisted of running the pump at a constant motor speed while varying a control valve to restrict flow through the loop, with differential pressure and flow rate recorded. This was done for one and two stage configurations for the 5 cm (2 inch) diameter impeller design and one, two, and three stage configurations for the 7.5 cm (3 inch diameter) impeller design, respectively. Due to pumping power requirements, two and three stage 7.5 cm (3 inch) diameter impeller testing at higher flowrates lowered the motor speed substantially. In regions where motor speed could not be maintained constant, the data were discarded. The test loop was also reconfigured to allow for the pump to be tested for pressure drop in a stalled or inoperable (0 RPM) flow condition. Demonstration of adequate natural convection cooling of the test cartridge fuel type is necessary under accident conditions, and this will depend upon the flow resistance through the impeller assembly when the pump is not operating. Thus, accurate knowledge of the effective impeller assembly loss coefficient is important for safety evaluations. The test loop was modified to provide water inlet and outlets on either side of the pump impeller stack, and a metered flow of lab water was provided in order to measure the pressure drop across the cartridges as a function of flowrate. Data from the pump head curve testing developed as part of this work and supported by analysis using pump head affinity laws indicates that a three stage 7.5 cm (3 inch) pump impeller design will meet target requirements for coolant flow within the VTR cartridge sodium cartridge at full power conditions [1] with margin; this corresponds to a flowrate of 45 l/min (12 gpm) at a pressure drop of 6.1 m (20 feet) of water head. The results of the pressure loss measurements across the impeller assembly when the pump is stationary (i.e., at 0 RPM) indicate that the pressure loss coefficient is 0.921 for a two impeller stack configuration; this value is calculated based on the flow velocity through the minimum available flow area within a single stage of the impeller which corresponds to 1.4 cm2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program (Rev. 2)

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 and SHRT-6 were protected loss of flow tests where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R, SHRT-39, and SHRT-43R were unprotected loss of flow tests where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. PICT-5/6 demonstrated how core power could be increased and decreased via changes in the intermediate sodium and feedwater flow rates. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for all tests simulated there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1989

Transient Pressure Test Article Test Program

The Transient Pressure Test Article (TPTA) test program is being conducted at a new test facility located in the East Test Area at the National Aeronautics and Space Administration's (NASA's) Marshall Space Flight Center (MSFC) in Huntsville, Alabama. This facility, along with the special test equipment (STE) required for facility support, was constructed specifically to test and verify the sealing capability of the Redesigned Solid Rocket Motor (RSRM) field, igniter, and nozzle joints. The test article consists of full scale RSRM hardware loaded with inert propellant and assembled in a short stack configuration. The TPTA is pressurized by igniting a propellant cartridge capable of inducing a pressure rise rate which stimulates the ignition transient that occurs during launch. Dynamic loads are applied during the pressure cycle to simulate external tank attach (ETA) strut loads present on the ETA ring. Sealing ability of the redesigned joints is evaluated under joint movement conditions produced by these combined loads since joint sealing ability depends on seal resilience velocity being greater than gap opening velocity. Also, maximum flight dynamic loads are applied to the test article which is either pressurized to 600 psia using gaseous nitrogen (GN2) or applied to the test article as the pressure decays inside the test article on the down cycle after the ignition transient cycle. This new test facility is examined with respect to its capabilities. In addition, both the topic of test effectiveness versus space vehicle flight performance and new aerospace test techniques, as well as a comparison between the old SRM design and the RSRM are presented.

Vibbart, Charles M.↗
NASA NTRS1991

A portable Ka-band front-end test package for beam-waveguide antenna performance evaluation. Part 2: Tests on the antenna

In part one of this article, a description was given of a Ka-band test package developed to enable testing of the Deep Space Station (DDS) 13 34-m beam-waveguide (BWG) antenna at 32 GHz. Test results were given for the Ka-band test package in an on-the-ground test configuration. This article is a companion article concerned with Ka-band test results for the test package in an on-the-antenna test configuration. Included are Ka-band zenith noise-temperature values, tipping-curve data, and subreflector test results obtained at the Cassegrain focal point, as well as at the final BWG focal point (located in a subterranean pedestal room). Test results show that, through the use of the Ka-band test package, the BWG antenna performance was successfully evaluated at Ka-band. The Ka-band test package operated well in all of the different antenna test configurations.

Otoshi, T. Y.↗
NASA NTRS2012

Correlation Results for a Mass Loaded Vehicle Panel Test Article Finite Element Models and Modal Survey Tests

High-fidelity Finite Element Models (FEMs) were developed to support a recent test program at Marshall Space Flight Center (MSFC). The FEMs correspond to test articles used for a series of acoustic tests. Modal survey tests were used to validate the FEMs for five acoustic tests (a bare panel and four different mass-loaded panel configurations). An additional modal survey test was performed on the empty test fixture (orthogrid panel mounting fixture, between the reverb and anechoic chambers). Modal survey tests were used to test-validate the dynamic characteristics of FEMs used for acoustic test excitation. Modal survey testing and subsequent model correlation has validated the natural frequencies and mode shapes of the FEMs. The modal survey test results provide a basis for the analysis models used for acoustic loading response test and analysis comparisons

Maasha, Rumaasha↗
NASA NTRS2017

Temporary Thermocouple Attachment for Thermal/Vacuum Testing at Non-Extreme Temperatures - Test Results

Post-test examination and data analysis that followed a two week long vacuum test showed that numerous self-stick thermocouples became detached from the test article. The thermocouples were reattached with thermally conductive epoxy and the test was repeated to obtain the required data. Because the thermocouple detachment resulted in significant expense and rework, it was decided to investigate the temporary attachment methods used around NASA and to perform a test to assess their efficacy. The present work describes the testing that was performed in early and mid-2017. The test article and the temporary thermocouple attachment methods tested are described. During the first test, fully half of the thermocouples detached - although the detachment showed subtly in the data for some. The second test was performed to confirm the data from the first test and to investigate the effect of test article and thermocouple grounding. The results of the testing over temperatures ranging from -150 to 200degF are detailed and preliminary recommendations are made for temporary thermocouple attachment methods.

Wright, Sarah E.↗
DOE OSTI2012

Interpretation of Colloid-Homologue Tracer Test 10-03, Including Comparisons to Test 10-01

This presentation covers the interpretations of colloid-homologue tracer test 10-03 conducted at the Grimsel Test Site, Switzerland, in 2010. It also provides a comparison of the interpreted test results with those of tracer test 10-01, which was conducted in the same fracture flow system and using the same tracers than test 10-03, but at a higher extraction flow rate. A method of correcting for apparent uranine degradation in test 10-03 is presented. Conclusions are: (1) Uranine degradation occurred in test 10-03, but not in 10-01; (2) Uranine correction based on apparent degradation rate in injection loop in test 11-02 seems reasonable when applied to data from test 10-03; (3) Colloid breakthrough curves quite similar in the two tests with similar recoveries relative to uranine (after correction); and (4) Much slower apparent desorption of homologues in test 10-03 than in 10-01 (any effect of residual homologues from test 10-01 in test 10-03?).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. Thirteen additional tests would be performed in the CABRI reactor over the next decade on fuel rods with higher burnups [1]. Additionally, a larger testing program at the NSRR reactor in Japan with high burnup fuel would uncover a similar trend of pre-DNB ruptures in high burnup test rods at lower-than-expected peak enthalpies [2]. Numerous out of pile testing programs involving a variety of innovative mechanical testing techniques have been employed in an attempt to better quantify the failure thresholds of corroded zirconium alloy cladding in these rapid heating and loading conditions [3][4][5]. While previously interim guidance had been issued, in June of 2020 the NRC officially published updated regulatory guidance to account of these pre-DNB failures in Regulatory Guide 1.236 [6]. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Optimal Nested Test Plan for Combinatorial Quantitative Group Testing

We consider the quantitative group testing problem where the objective is to identify defective items in a given population based on results of tests performed on subsets of the population. Under the quantitative group testing model, the result of each test reveals the number of defective items in the tested group. The minimum number of tests achievable by nested test plans was established by Aigner and Schughart in 1985 within a minimax framework. Furthermore, the optimal nested test plan offering this performance, however, was not obtained. In this paper, we establish the optimal nested test plan in a closed form. This optimal nested test plan is also order optimal among all test plans as the population size approaches infinity. Using heavy-hitter detection as a case study, we show via simulation examples orders of magnitude improvement of the group testing approach over two prevailing sampling-based approaches in detection accuracy and counter consumption. Other applications include anomaly detection and wideband spectrum sensing in cognitive radio systems.

97 MATHEMATICS AND COMPUTING↗
NASA NTRS2018

NASA Plum Brook Station In-Space Propulsion Facility Test Stand Characterization Hot Fire Test

A test facility modification to enable small scale altitude propulsion testing at the NASA Glenn Research Center's In-Space Propulsion (ISP) Facility was verified with a hot fire test campaign. As the facility's primary steam supply system undergoes refurbishment, the alternate facility configuration, known as the "vacuum accumulator" mode, would enable rocket engine testing up to 10,000 lbf thrust. The NASA Johnson Space Center developed the vehicle for the verification test campaign: the Integrated Cryogenic Propulsion Test Article (ICPTA). Constructed primarily from assets of the former Morpheus Project, the ICPTA provided an integrated liquid oxygen (LOX) / liquid methane (LCH4) propulsion system including a 2,800 lbf thrust main engine. The ISP Facility's vacuum accumulator configuration leveraged the large test volume of the facility and a diffuser insert to maintain altitude conditions. During hot fire, the ICPTA main engine "started" the diffuser insert constructed for the test campaign. As a result, the test chamber upstream of the diffuser insert remained at altitude conditions throughout the hot fire. Upon engine shut down, a backflow deflector mitigated blow back into the test chamber by restricting the mass flow and redirecting it away from the test article. The test campaign successfully characterized the performance of the vacuum accumulator configuration. In addition, it provided an opportunity to collect data for an integrated LOX / LCH4 propulsion system in an altitude and thermal vacuum environment.

Rocket Engine Testing↗
DOE OSTI2014

Methodology for Mechanical Property Testing of Fuel Cladding Using a Expanded Plug Wedge Test

An expanded plug method was developed earlier for determining the tensile properties of irradiated fuel cladding. This method tests fuel rod cladding ductility by utilizing an expandable plug to radially stretch a small ring of irradiated cladding material. The circumferential or hoop strain is determined from the measured diametrical expansion of the ring. A developed procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves, from which material properties of the cladding can be extracted. However, several deficiencies existed in this expanded-plug test that can impact the accuracy of test results, such as that the large axial compressive stress resulted from the expansion plug test can potentially induce the shear failure mode of the tested specimen. Moreover, highly nonuniform stress and strain distribution in the deformed clad gage section and significant compressive stresses, induced by bending deformation due to clad bulging effect, will further result in highly nonconservative estimates of the mechanical properties for both strength and ductility of the tested clad. To overcome the aforementioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. By optimizing the specific geometry designs, selecting the appropriate material for the expansion plug, and adding new components into the testing system, a modified expansion plug testing protocol has been developed. A general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor, -factor, was used to convert the ring load Fring into hoop stress , and is written as _ = F_ring/tl , where t is the clad thickness and l is the clad length. The generated stress-strain curve agrees well with the associated tensile test data in both elastic and plastic deformation regions.

expansion plug test↗
DOE OSTI2012

Deep Vadose Zone Treatability Test for the Hanford Central Plateau: Soil Desiccation Pilot Test Results

Over decades of operation, the U.S. Department of Energy (DOE) and its predecessors have released nearly 2 trillion L (450 billion gal) of liquid into the vadose zone at the Hanford Site. Much of this liquid waste discharge into the vadose zone occurred in the Central Plateau, a 200 km2 (75 mi2 ) area that includes approximately 800 waste sites. Some of the inorganic and radionuclide contaminants in the deep vadose zone at the Hanford Site are at depths below the limit of direct exposure pathways, but may need to be remediated to protect groundwater (DOE 2008a; Dresel et al. 2011). The Tri-Party Agencies (DOE, U.S. Environmental Protection Agency, and Washington State Department of Ecology) established Milestone M-015-50, which directed DOE to submit a treatability test plan for remediation of Tc-99 and uranium in the deep vadose zone. These contaminants are mobile in the subsurface environment and have been detected at high concentrations deep in the vadose zone, and at some locations have reached groundwater. Testing technologies for remediating Tc-99 and uranium will also provide information relevant for remediating other contaminants in the vadose zone. The desiccation test described herein was conducted as an element of the test plan published in March 2008 to meet Milestone M-015-50 (DOE 2008a). This field test report was prepared and submitted in response to Milestone M-015-110D, which set a date for desiccation field test activities to be reported to the Tri-Party Agencies. The desiccation field test was conducted at the Hanford Site 200-BC-1 Operable Unit. This waste site contains 26 cribs and trenches that received about 110 million L (29 million gal) of liquid waste primarily in the mid-1950s. The waste contained about 410 curies of Tc-99 (Corbin et al. 2005). There is no evidence the contamination has reached groundwater, located about 100 m (330 ft) below ground surface (bgs) in this area. Initial characterization efforts indicated the Tc-99 inventory is located mostly at a depth in the vadose zone of between about 30 and 70 m (98 and 230 ft) bgs. However, transport model predictions have indicated the potential for this contamination to adversely impact groundwater in the future (Ward et al. 2004). The test was conducted in two steps to provide information about desiccation that is intended for use in subsequent feasibility studies for waste sites with inorganic and radionuclide contaminants in the deep vadose zone. The first step was field-scale test site characterization conducted as described in a characterization work plan (DOE 2008b). Results of the characterization effort have been previously reported in DOE (2010a) and Um et al. (2009). A field test plan (DOE 2010b) was prepared and used to guide the desiccation field testing effort. Laboratory and numerical modeling efforts (Truex et al. 2011; Ward et al. 2008; Oostrom et al. 2009, 2011, 2012, In Press) preceded and accompanied the field test and are incorporated herein as their results pertain to assessment of desiccation for future feasibility studies.

desiccation↗
DOE OSTI2017

Test Methodologies for Hydrogen Sensor Performance Assessment: Chamber vs. Flow Through Test Apparatus: Preprint

Certification of hydrogen sensors to standards often prescribes using large-volume test chambers [1, 2]. However, feedback from stakeholders such as sensor manufacturers and end-users indicate that chamber test methods are often viewed as too slow and expensive for routine assessment. Flow through test methods potentially are an efficient, cost-effective alternative for sensor performance assessment. A large number of sensors can be simultaneously tested, in series or in parallel, with an appropriate flow through test fixture. The recent development of sensors with response times of less than 1s mandates improvements in equipment and methodology to properly capture the performance of this new generation of fast sensors; flow methods are a viable approach for accurate response and recovery time determinations, but there are potential drawbacks. According to ISO 26142 [1], flow through test methods may not properly simulate ambient applications. In chamber test methods, gas transport to the sensor can be dominated by diffusion which is viewed by some users as mimicking deployment in rooms and other confined spaces. Alternatively, in flow through methods, forced flow transports the gas to the sensing element. The advective flow dynamics may induce changes in the sensor behaviour relative to the quasi-quiescent condition that may prevail in chamber test methods. One goal of the current activity in the JRC and NREL sensor laboratories [3, 4] is to develop a validated flow through apparatus and methods for hydrogen sensor performance testing. In addition to minimizing the impact on sensor behaviour induced by differences in flow dynamics, challenges associated with flow through methods include the ability to control environmental parameters (humidity, pressure and temperature) during the test and changes in the test gas composition induced by chemical reactions with upstream sensors. Guidelines on flow through test apparatus design and protocols for the evaluation of hydrogen sensor performance are being developed. Various commercial sensor platforms (e.g., thermal conductivity, catalytic and metal semiconductor) were used to demonstrate the advantages and issues with the flow through methodology.

14 SOLAR ENERGY↗
DOE OSTI2025

Test Results of the LQXFA/B02 and LQXFA/B03 Cryo-Assemblies for the High Luminosity LHC Upgrade

Here, the US High-Luminosity LHC Accelerator Upgrade Project (AUP) is responsible for delivering cryo-assemblies for the Q1/Q3 quadrupole optical components of the High Luminosity LHC upgrade at CERN. Total of 10 cryo-assemblies containing two Nb3Sn quadrupole magnets per cold mass will be delivered within this program. After the successful test of the first pre-series cryo-assembly in 2023, two more cryo-assemblies were tested at Fermilab's horizontal test facility. Production overview and the test results of the LQXFA/B02 and LQXFA/B03 cryo-assemblies are summarized in this paper. After the first test, to increase the capability of the horizontal test facility, various improvements have been made. These improvements are also described in this paper.

Chlachidze, Guram [Fermi National Accelerator Labo↗
DOE OSTI2013

Improving the Sandia Test Protocols with Advanced Inverter Functionality Testing of INV3, VV11, FW21, and L/HVRT

Sandia National Laboratories has created a test protocol for IEC TR 61850-90-7 advanced distributed energy resource (DER) functions, titled "Test Protocols for Advanced Inverter Interoperability Functions," often referred to as the Sandia Test Protocols. This document is currently in draft form, but has been shared with stakeholders around the world with the ultimate goal of collaborating to create a consensus set of test protocols which can be then incorporated into an International Electrotechnical Commission (IEC) and/or Underwriters Laboratories (UL) certification standard. The protocols are designed to ensure functional interoperability of DER (primarily photovoltaic (PV) inverters and energy storage systems) as specified by the IEC technical report through communication and electrical tests. In this report, Sandia exercises the electrical characterization portion of the test protocols for four functions: constant power factor (INV3), volt-var (VV11), frequency-watt (FW21), and Low and High Voltage Ride Through (L/HVRT). The goal of the tests reported here was not to characterize the performance of the equipment under test (EUT), but rather to (a) exercise the draft Sandia Test Protocols in order to identify any revisions needed in test procedures, conditions, or equipment and (b) gain experience with state-of-the-art DER equipment to determine if the tests put unrealistic or overly aggressive requirements on EUT operation. In performing the work according to the current versions of the protocols, Sandia was able to identify weaknesses in the current versions and suggest improvements to the test protocols.

14 SOLAR ENERGY↗
DOE OSTI2021

Controlled Pyroshock Test Transients Can Be Used To Better Match Operational Shock Transients and Reduce Artificial Pyroshock Test Failures [Slides]

Shock analysis is typically done via SRS (shock response spectrum). Oftentimes shock analysis is not performed because most loads specifications only supply SRS requirements with no representative shock transients that can be used for analysis. SRS analysis is too conservative. There is no structural dynamics involved as all modes are assumed to have peak responses simultaneously and in-phase with each other. It is possible that all other loads and combination of loads show the design to be good, whereas SRS analysis (1000 g’s are involved) results may require redesign. Multiple shock transients can satisfy a SRS, yet most loads documents don’t provide a representative shock transient that could be used for shock analysis. Shock testing is less common than harmonic and random vibration testing. Engineers typically go into shock tests with very little knowledge of what to expect. A pass/fail type of approach with very little engineering done to prepare for the test or to control the type of shock transient is used. Most engineers will not model the shock test fixture due to lack of knowledge about the details of the test setup. Shock analysis and testing is often an open loop type situation with no test setup modeling, no test correlation, and no control over the test shock transient. In this work, if we can create applied test force transient that comes close to matching calculated force transient then a more reasonable acceleration transient will be created at the test reference accelerometer that better matches operational shocks.

42 ENGINEERING↗
DOE OSTI2022

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 was a protected loss of flow test where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R was an unprotected loss of flow test where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for SHRT-17, SHRT-45R, BOP-301, and BOP-302R there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗