An Investigation On The Brazilian Disc Test As An Indirect Tensile Strength Test
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Engineering - Mechanical and civil engineering↗
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Benchmark Mechanical Test Results of Westinghouse Zirconium Cladding and Irradiated Testing Readiness
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Garrison, Ben [ORNL] (ORCID:0000000316880682)↗
Evaluation of Zero-Net-Rate Pumping Tests
Accurately estimating the distribution of aquifer properties is key to understanding contaminant movement in the subsurface. The distribution of aquifer properties is typically addressed using slug or constant-rate well tests, and the pros and cons of these tests are well known. Slug tests are appealing because they avoid removing contaminated water, but their results are affected by well skin and the small volume of displaced water limits the volume of aquifer that can be evaluated. Constant-rate well tests have the disadvantage of requiring disposal of potentially contaminated water, but they can generate properties that are more representative than slug tests, and they can be used to estimate well efficiency and storativity, which are difficult to characterize using slug tests. Periodic pumping tests are appealing because they have many of the advantages and few of the disadvantages of slug and constant-rate well tests. Periodic pumping tests involve cycling the pumping rate with a regular period and measuring the resulting response in monitoring wells. Some of these tests involve imposing a periodic rate on a constant mean rate. However, other tests involve moving water out and back into the well at rates that are balanced so the net rate after each period is zero. Some zero-net-rate (ZNR) well tests use rates that follow a sinusoidal pattern, whereas others use a square wave pattern that switches between a constant rate of pumping and a constant rate of injection to achieve a zero-net rate. ZNR well tests appear to be a useful compromise between slug and constant rate tests, but methods for conducting and analyzing the results from these tests have received limited evaluation. Periodic pumping tests are typically evaluated through recorded pressure responses to a disturbance source. Recent work has demonstrated that the strain field in the vadose zone shows a response to disturbances in the underlying aquifer. Various instrument designs are available that can record vertical and horizontal strain to very precise resolutions (10-9 ε). Measuring the strain in the vadose zone can present lower costs than measuring pressure in a monitoring well, so using strain could improve the resolution of ZNR tests, as well as make these tests potentially cheaper. A ZNR periodic pumping system was constructed and used to perform tests in the Clemson, South Carolina area. The system was designed to pump and inject with a periodic square-wave, so pumping occurs at a constant rate for half the period and is followed by injection at the same rate for half the period resulting in no net gain or loss of water from the aquifer. The system is designed to generate flow rates from 1 to 3.5 gallons per minute (gpm) with a capacity of 875-gallon per half period. Pressure and strain monitoring points are located in the vicinity of the pumping well. A series of 10 ZNR periodic pumping tests were conducted at the field site, with periods ranging from 2 to 540 minutes. Pressure data were collected and recorded in four monitoring wells in various locations around the pumping well. In addition, vertical and horizontal strain and tilt data were recorded in various locations around the pumping well. Traditional aquifer tests, including slug tests and constant-rate pumping tests were conducted at the field site to provide a baseline of aquifer parameter estimates that are compared the results of ZNR tests. Pressure data measured at monitoring wells during ZNR tests were analyzed utilizing an analytical solution (Streltsova, 1988) that assumes a confined aquifer and estimates hydraulic diffusivity and transmissivity using the time lag and amplitude of the pressure signal. The storativity can be separated out from the hydraulic diffusivity using the transmissivity estimate. The time-lag and amplitude of the pressure signals were estimated using a Fourier transform. The time-lag of the primary period of the pressure increases as a linear function of distance from the pumping well, and it increases as the square root of the period of the ZNR test over a range of periods spanning two orders of magnitudes. The first few harmonics follow similar trends, but the higher frequency harmonics diverge from this trend. Strain and tilt data from various instruments in the vadose and saturated zones are periodic during ZNR tests. The time lags of the strain components increase roughly linearly with distance from the well. Many of the time lags of the strain data from the vadose zone are similar to the time lags observed in the pressure data from a similar distance and pumping period. Theoretical experiments were performed to understand how pressure and strain responds to ZNR periodic pumping. The time-lags of the simulated pressure responses match the field data trends. The results were used to validate the Streltsova (1988) solution for periodic pumping to estimate hydraulic diffusivity. Hydraulic diffusivity was estimated to be 1.8 x 10-2 m2s-1 < Dh < 9.0 x 10-2 m2s-1 using pressure data from all the wells during ZNR tests. Transmissivity was estimated to be 0.8 x 10-4 m2s-1 < T < 4.3 x 10-4 m2s-1. Assuming Storativity = T/Dh gives 1.1 x 10-3 < S < 19 x10-3. Transmissivity is used to estimate the hydraulic conductivity, K, by dividing by the assumed aquifer thickness. Two constant-rate pumping tests were conducted and analyzed using two conceptual models: The Hantush (1961) solution was used to analyze data assuming confined conditions and the Neuman (1974) solution was used to analyze data assuming unconfined conditions. The Hantush (1961) solution estimated transmissivity to be 1.6 x 10-4 m2s-1 < T < 1.8 x 10-4 m2s-1 when using data from all the wells. Storativity was estimated to be 4.4 x 10-3 < S < 8.2 x 10-3. Assuming Dh = T/S gives 0.22 x 10-2 m2s-1 < Dh < 3.6 x 10-2 m2s-1. The Neuman (1974) solution estimated transmissivity to be 0.80 x 10-4 m2s-1 < T < 0.84 x 10-4 m2s-1. Total storativity (S + Sy) was estimated to be 22 x 10-3 < S < 110 x 10-3. Storativity excluding the Sy term was estimated to be 2.2 x 10-3 < S < 2.5 x 10-3. Assuming Dh = T/S gives 0.073 x 10-2 m2s-1 < S < 0.36 x 10-2 m2s-1. Two slug tests were performed on the pumping well, PW-2. The Bouwer and Rice (1976) solution for slug tests was used to analyze the data. This solution works for both confined and unconfined conditions, so both conceptual models were covered through the analysis. Hydraulic conductivity was estimated to be 1.95 x 10-6 ms-1 < K < 2.49 x 10-6 ms-1. Statistical comparison show generally no significant difference in the estimates of Dh, K, or S made using ZNR and conventional pumping tests.. However, the total storage estimated by the Neuman (1974) solution for unconfined settings is larger than that estimate using the Streltsove analysis applied to the ZNR tests. This is likely because the Neuman solution considers delayed yield from storage that the water table, whereas the Streltsova (1988) solution assumes confined conditions. The analysis also indicates that there is a statistically significant difference between parameters estimated with conventional slug tests and those measured with either constant-rate pumping tests or ZNR tests. Hydraulic diffusivity is estimated from the time-lag and the distance from the well. The time-lags of the strain data from the vadose zone are similar to the lags of the pressure data from a similar distance. This indicates that hydraulic diffusivities estimated from the strain data measured in the vadose zone would be similar to the estimates from the pressure data measured in the aquifer. There appear to be sign reversals in some of the strain data that were corrected to estimate the time lag. The dissolved oxygen (CO) concentration was measured during several ZNR tests to evaluate the feasibility of using the procedure to increase contaminant degradation kinetics that are related to DO. The results indicate that DO can be increased during the cycling of ZNR tests in some cases.
Smith-Jones, Austin↗
An automatic input/output test experience of FPGA-based logic controller with I/O stimulator - 282
This paper describes an automatic testing approach instead of a traditional manual test for an FPGA-based Logic Controller (FLC). It mainly describes an automatic testing of FLC using an I/O stimulator with the initiated manual action. The FLC was developed by Doosan Heavy Industry and KAERI in S. Korea. We also developed a test-bed like I/O stimulator for a burning test based on the operational scenario. Signal sources from the I/O stimulator contain an analog input, an analog output, a digital input and a digital output, a programmable power supply, an RS232C serial interface, and a dual port Ethernet interface. LabVIEW Programming was used for the automatic testing. The FLC consists of a general processor module (FPM-01), a complex processor module (FPM-02), analog input modules (FAI-01), analog output modules (FAO-01), digital input modules (FDI-01), digital output modules (FDO-01), a data link module (FDL-01), and a network module (FNW-01). Among these, in this paper, automatic digital I/O testing and automatic analog I/O testing are described. In addition, for the accuracy of the automatic testing, two graphic recorders are used. One is the output results of general processor based testing, and the other is the output results of complex processor based testing. These are different company products. Redundancy recording was found in the test results. The test results were synchronized internally with the existing I/O stimulator and externally with two graphic recorders. Calibrations of all of the equipment were made before the system testing. Before the testing started, a test plan and the test procedure were set up in advance, and we performed test scenario-based automatic functional testing for the FLC as the qualification process. It can be several executed depending on the scenario and conditional parameters of the test. For validation of the accuracy and reliability in our automated test, manual testing was also performed. Manual tests were conducted in two ways. The first one was a manual test using Fluke 754 and Fluke 715. The other one was using embedded manual features of the I/O stimulator. By performing the automatic input and output testing using the I/O stimulator, the time and expense of FLC testing can be reduced significantly compared with a conventional manual test. We rechecked the automatic test results using Fluke754 manually. (authors)
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Performance Test on Polymer Waste Form - 12137
Polymer solidification was attempted to produce stable waste form for the boric acid concentrates and the dewatered spent resins. The polymer mixture was directly injected into the mold or drum which was packed with the boric acid concentrates and the dewatered spent resins, respectively. The waste form was produced by entirely curing the polymer mixture. A series of performance tests was conducted including compressive strength test, water immersion test, leach test, thermal stability test, irradiation stability test and biodegradation stability test for the polymer waste forms. From the results of the performance tests for the polymer waste forms, it is believed that the polymer waste form is very stable and can satisfy the acceptance criteria for permanent disposal. At present, performance tests with full scale polymer waste forms are being carried out in order to obtain qualification certificate by the regulatory institute in Korea. Polymer waste forms were prepared with the surrogate of boric acid concentrates and the surrogate of spent ion exchange resins respectively. Waste forms were also made in lab scale and in full scale. Lab. scale waste forms were directly subjected to a series of the performance tests. In the case of full scale waste form, the test specimens for the performance test were taken from a part of waste form by coring. A series of performance tests was conducted including compressive strength test, thermal stability test, irradiation stability test and biodegradation stability test, water immersion test, leach test, and free standing water for the polymer waste forms. In addition, a fire resistance test was performed on the waste forms by the requirement of the regulatory institute in Korea. Every polymer waste forms containing the boric acid concentrates and the spent ion exchange resins had exhibited excellent structural integrity of more than 27.58 MPa (4,000 psi) of compressive strength. On thermal stability testing, biodegradation testing and water immersion testing, no degradation was observed in the waste forms. Also, by measuring the compressive strength after these tests, it was confirmed that the structural integrity was still retained. A leach test was performed by using non radioactive cobalt, cesium and strontium. The leaching of cobalt, cesium and strontium from the polymer waste forms was very low. Also, the polymer waste forms were found to possess adequate fire resistance. From the results of the performance tests, it is believed that the polymer waste form is very stable and can satisfy the acceptance criteria for permanent disposal. At present, Performance tests with full scale polymer waste forms are on-going in order to obtain qualification certificate by the regulatory institute in Korea. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Gripper Test Assembly Status Report (FY2020)
Refueling systems for fast reactors are designed to handle fresh and used core assemblies (fuel, reflector, and shield core assemblies) within the reactor vessel in an opaque coolant environment without visual reference. These refueling machines are designed to work in a sodium (or other fast reactor coolant) and argon vapor space environment and are engineered with the rotatable plug system to allow for the movement of fresh and spent fuel into and out of the reactor core. The refueling machines are a critical component in any reactor and thus need to undergo extensive testing in a prototypic environment to ensure that they will meet all of the system functions and requirements. Argonne has developed an innovative compact refueling system design for the Advanced Fast Reactor-100 that is based upon some mechanisms used in previous reactor designs, such as the U.K.’s Prototype Fast Reactor (PFR) and some mechanisms that have not been used in sodium. This compact refueling machine supports the reduction in size of the AFR-100’s reactor vessel, and if fully developed, would support and inform the development of the in-vessel refueling machines for such commercial reactors as the GEH PRISM reactor plant, the ARC Clean Energy’s ARC-100 reactor, and the Natrium reactor, among others. This refueling system is a vital component of a fast reactor that supports reducing the cost of the reactor and increasing its reliability. During the development of the compact fuel handling machine conceptual design for the AFR-100, a lack of testing data for many mechanical components in sodium under typical in-reactor loads and conditions was discovered. The reduction in lifetime of the various mechanical components in the liquid sodium environment needs to be quantified versus the calculated component lifetimes under normal conditions in the testing while they are subjected to typical loading profiles experienced in the past. The Gripper Test Assembly discussed here includes a full-size gripper device with appropriate mechanical features that will be tested in sodium to provide this testing data. The Gripper Test Assembly is used to test various mechanical fuel handling components submerged in high temperature liquid sodium. These mechanisms are gears, bearings, gripper jaws and head, universal joints and shafts, ball screws, among others. These components will be tested under the typical sodium environmental conditions experienced during refueling operations with appropriate loading conditions that simulate the removal and insertion of core assemblies into a fast reactor grid plate structure. This Gripper Test Assembly is the second in a series of refueling system mechanisms developed for testing in sodium. The first test assembly is the Gear Test Assembly (GTA) which was used to test the performance of gears and bearings operating in sodium. Because of the successful testing conducted with the GTA, it was decided to continue with the development of the Gripper Test Assembly which uses the same gears and bearings tested in GTA. Using the data collected during operation of this gripper test assembly, lifetime reduction factors of the various mechanical components can be calculated for the material combinations selected. These lifetime reduction factors can be used in the design of future mechanical systems which operate in these environments to accurately predict component end of life. In addition, an understanding of the ability of these components and mechanisms to operate under-sodium with the chosen materials will be accomplished. Again, this Gripper Test Assembly is the follow-on test article to the Gear Test Assembly which was testing the ability of gears and bearings operating in a sodium environment. Once fabricated and qualified, it will be tested in the Mechanism Engineering Test Loop (METL) facility located in Building 308 at Argonne National Laboratory.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
August 5, 1963-President Kennedy's Nuclear Test Ban Treaty signed in Moscow, Russia
On August 5, 1963, after more than eight years of negotiations, the United States, the United Kingdom, and the Soviet Union signed the Limited Nuclear Test Ban Treaty. The destruction of Hiroshima and Nagasaki by atomic bombs marked the end of World War II and the beginning of the nuclear age. As tensions between East and West settled into a Cold War, scientists in the United States, Great Britain, and the Soviet Union conducted tests and developed more powerful nuclear weapons. In 1959, radioactive deposits were found in wheat and milk in the northern United States. As scientists and the public gradually became aware of the dangers of radioactive fallout, they began to raise their voices against nuclear testing. Leaders and diplomats of several countries sought to address the issue. In May 1955, the United Nations Disarmament Commission brought together the United States, the United Kingdom, Canada, France, and the Soviet Union to begin negotiations on ending nuclear weapons testing. Conflict soon arose over inspections to verify underground testing. The Soviet Union feared that on-site inspections could lead to spying that might expose the Soviets' vastly exaggerated claims of the number of deliverable nuclear weapons. As negotiators struggled over differences, the Soviet Union and the United States suspended nuclear tests—a moratorium that lasted from November 1958 to September 1961. John F. Kennedy had supported ban on nuclear weapons testing since 1956. He believed a ban would prevent other countries from obtaining nuclear weapons, and took a strong stand on the issue in the 1960 presidential campaign. Once elected, President Kennedy pledged not to resume testing in the air and promised to pursue all diplomatic efforts for a test ban treaty before resuming underground testing. He envisioned the test ban as a first step to nuclear disarmament. President Kennedy met with Soviet Premier Khrushchev in Vienna in June 1961, just five weeks after the humiliating defeat of the US-sponsored invasion of Cuba at the Bay of Pigs. Khrushchev took a hard line at the summit. He announced his intention to cut off Western access to Berlin and threatened war if the United States or its allies tried to stop him. Many US diplomats felt that Kennedy had not stood up to the Soviet premier at the summit and left Khrushchev with the impression that he was a weak leader. President Kennedy's political and military advisers feared that the Soviet Union had continued secret underground testing and made gains in nuclear technology. They pressured Kennedy to resume testing. And, according to a Gallup poll in July 1961, the public approved of testing by a margin of two-to-one. In August 1961, the Soviet Union announced its intention to resume atmospheric testing, and over the next three months it conducted 31 nuclear tests. It exploded the largest nuclear bomb in history—58 megatons—4,000 times more powerful than the bomb dropped on Hiroshima. In his commencement address at American University on June 10, 1963, Kennedy announced a new round of high-level arms negotiations with the Russians. He boldly called for an end to the Cold War. "If we cannot end our differences," he said, "at least we can help make the world a safe place for diversity." The Soviet government broadcast a translation of the entire speech, and allowed it to be reprinted in the controlled Soviet press. The Limited Nuclear Test Ban treaty was signed in Moscow on August 5, 1963, by US Secretary Dean Rusk, Soviet Foreign Minister Andrei Gromyko, and British Foreign Secretary Lord Home—one day short of the 18th anniversary of the dropping of an atomic bomb on Hiroshima. Over the next two months, President Kennedy convinced a fearful public and a divided Senate to support the treaty. The Senate approved the treaty on September 23, 1963, by an 80-19 margin. Kennedy signed the ratified treaty on October 7, 1963. The treaty: prohibited nuclear weapons tests or other nuclear explosions under water, in the atmosphere, or in outer space allowed underground nuclear tests as long as no radioactive debris falls outside the boundaries of the nation conducting the test pledged signatories to work towards complete disarmament, an end to the armaments race, and an end to the contamination of the environment by radioactive substances. Thirty-three years later, the United Nations General Assembly adopted the Comprehensive Nuclear Test Ban Treaty. Signed by 71 nations, including those possessing nuclear weapons, the treaty prohibited all nuclear test explosions including those conducted underground. Though it was signed by President Bill Clinton, the Senate rejected the treaty by a vote of 51 to 48.
98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
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↗
Definition of Capabilities Needed for a Single Event Effects Test Facility
The Federal Aviation Administration (FAA) is contemplating new regulations mandating testing of the vulnerability of flight-critical avionics to single event effects (SEE). A limited number of high-energy neutron test facilities currently serve the SEE industrial and institutional research community. The FAA recognizes that existing facilities have insufficient test capacity to meet new demand from such mandates; it desires more flexible irradiation capabilities to test complete, large systems and would like capabilities to address greater concerns for thermal neutrons. For this reason, the FAA funded this study by Spallation Neutron Source (SNS) staff with the ultimate aim of developing options for SEE test facilities using high-energy neutrons at the SNS complex. After an investigation of current SEE test practices and assessment of future testing requirements, three concepts were identified covering a range of test functionality, neutron flux levels, and fidelity to the atmospheric neutron spectrum. The costs and times required to complete each facility were also estimated. SEE testing is generally performed by accelerating the event rate to a point where the effects are still dominated by single events and double event causes of failures are negligible. In practice, acceleration factors of as high as 10 6 are applicable for component testing, whereas for systems testing acceleration factors of 10 4 seem to be the upper limit. It is strongly desirable that the irradiation facility be tunable over a large range of high-energy neutron fluxes of 10 2 - 10 4 n/cm²/s for systems testing and from 10 4 - 10 7 n/cm²/s for components testing. The most capable, most flexible, and highest-test-capacity option is a new stand-alone target station named the High-Energy neutron Test Station (HETS). It is also the most expensive option, with a cost to complete of approximately $100 million. Dual test enclosures would allow for simultaneous testing activity effectively doubling overall test capacity per HETS operating hour. Using about 1 kilowatt (kW) of proton power extracted from the accelerator before injection in the accumulator ring, its operation would be unnoticeable by neutron scattering users at the SNS target station. The H beam laser stripping technique would allow for control of beam power on the HETS target independent from power delivered to the SNS. Large systems with frontal areas of up to 1 x 2 m² could be accommodated with integral high-energy flux values (above 10 megaelectron-volt, or MeV) to at most 10 4 n/cm²/s; components could also be tested with flux levels to at most 10 7 n/cm²/s on beam sizes of up to 0.2 x 0.2 m². Selectable moderating material and neutron filters would allow tailoring of the neutron spectrum to user demands; charged particle deflectors could be switched to allow or deflect protons, pions, and muons. It is estimated that HETS would take 5 years to complete after award of contract, including engineering design and construction. Commissioning would take at least another 6 months. Interference with SNS principal operations was not considered in the construction time estimate; connection of the proton transport line and tunnel from the accelerator high energy beam transport (HEBT) and construction around existing site utilities would require careful planning and coordination with beam operations at the SNS. A high-energy (HE) neutron test facility using an available beam line on the SNS target station is a technically and financially attractive option. Inspired by the new ChipIR instrument on the ISIS TS 2 spallation source in the UK, a similar facility could be placed on an unused beam line in the SNS instrument hall [e.g., on beam line 8 (both A and B channels would be needed) or on beam line 10]. The performance would approach that of an HETS (~80%), but it would be operationally more limited, with only a single user at a time. Space is more limited, so the maximum system size would be about half of that in an HETS. Flexibility to tailor the spectrum would be somewhat more limited. While this concept was not as fully developed and characterized, preliminary work indicates very high HE flux levels should be possible, with ample thermal neutrons as well. Flux control would be more difficult than at HETS because proton power on target be whatever the SNS was operating at for neutron scattering. Neutron attenuation devices would have to be employed with as-yet undetermined control resolution. However, no new buildings would be needed, and the necessary utilities are already present in the SNS Experiment Hall. The estimated cost for a beam line option is around $15 million; the time to complete would be 3 years after award of contract, plus at least 6 months for commissioning. Interference of construction activities with SNS operations should be negligible. This option would require negotiation with the Department of Energy Basic Energy Sciences (BES) office -- the primary stakeholder of SNS -- for an application outside the usual scope of neutron scattering sciences. Furthermore, these presently open beam lines are highly desirable locations for proposed neutron scattering instruments and obtaining one of them for an SEE test facility will come only with persuasive and timely arguments to SNS leadership and the DOE BES. The third option is a tunnel extension/target cave facility providing the most basic system-level irradiation capability with minimal flexibility. Again not as well developed a concept as HETS, it would use a laser-stripping technique like an HETS, redirecting protons to a tunnel similar to the initial HETS proton transport tunnel. Indeed, this concept is intended to be upgradable to a full HETS facility. Only a small fraction of a watt of proton power would be used in this basic configuration, though. An uncooled target and primitive shielding arrangement would provide beam on modestly sized systems that must be placed in close proximity to the target. The neutron fluence would be less uniform over the system than with the HETS or the beam line option. A data acquisition room and support area would be located on the ground level; access to the target cave would be via elevator and/or stairway. As a result of the required excavation, new tunnel construction, shielding, data acquisition building, utilities, and other items, the estimated cost is $30 million. The time to complete is expected be more than 3 years; here again construction interference with SNS operations has not been accounted for, but it could have a significant impact.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254
Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Advanced Vehicle Testing and Evaluation Final Technical Report Encompassing Project Activities From October 1, 2011 to April 30, 2018
The Advanced Vehicle Testing and Evaluation (AVTE) project incorporated the conduct of advanced technology vehicle and infrastructure testing in the U.S. Department of Energy (DOE) Advanced Vehicle Testing Activity (AVTA). It was the only activity tasked by DOE to conduct independent field evaluations of vehicle technologies that use advanced technology systems and subsystems in vehicles that reduce petroleum consumption. The results are available through reports published on the AVTA website at http://avt.inl.gov/. The AVTE project provided insight into advanced vehicle technology's long-term performance, durability, maintenance, and lifecycle costs that cannot be found elsewhere. The site is a resource for DOE stakeholders and the public to gain knowledge on advanced technology vehicle progress and innovation. Vehicles tested included light and heavy-duty vehicles with conventional, hybrid, plug-in hybrid, and all-electric propulsion systems. Vehicles were tested on closed tracks and chassis dynamometers, as well as operated on public roads, in fleet operations, and over prescribed routes. All testing was controlled by procedures developed specifically to support such testing. Testing and evaluations were conducted in the following phases: Development and refinement of test procedures, which established testing procedures; Baseline performance testing, which established a performance baseline; Accelerated reliability testing, which determined vehicle reliability; Fleet testing, used to evaluate vehicle economics in fleet operation, and End of test performance evaluation. Test results from all aspects of testing were posted by Idaho National Laboratory (INL) to their website. This final report documents work conducted for the entirety of contract DE-EE0005501 by Intertek Testing Services NA, Inc. (Intertek). The contract was performed from October 1, 2011 through April 30, 2018, and includes 18 amendments to the original award with the most recent Amendment 18 dated March 28, 2018. The final budget for the AVTE project totaled 20,220,873 dollars. As a cooperative agreement, Intertek was obligated to provide cost share of 4,044,175 dollars, leaving 16,176,698 dollars as the Federal share of the agreement. Actual total expenditure for the AVTE project was 20,286,850 dollars. The full 16,176,698 dollars Federal share was expended and 4,110,152 dollars of cost share was provided by Intertek. This resulted in an over-share of 65,977 dollars. There were 106 light-duty on-road vehicles vehicles tested in both fleet and accelerated reliability testing. One additional vehicle was rented to obtain performance testing only. Three heavy-duty truck and trailer combinations were tested as part of the semi-autonomous driving with heavy-duty vehicles in a platoon. Overall project scope and direction was determined between Intertek, DOE, and technical teams from Idaho National Laboratory (INL), Argonne National Laboratory (ANL), and National Renewable Energy Laboratory (NREL). The project was administered by the DOE’s National Energy Technology Laboratory (NETL).
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
RELAP5-3D Restart and Backup Verification Testing
Existing testing methodology for RELAP5-3D employs a set of test cases collected over two decades to test a variety of code features and run on a Linux or Windows platform. However, this set has numerous deficiencies in terms of code coverage, detail of comparison, running time, and testing fidelity of RELAP5-3D restart and backup capabilities. The test suite covers less than three quarters of the lines of code in the relap directory and just over half those in the environmental library. Even in terms of code features, many are not covered. Moreover, the test set runs many problems long past the point necessary to test the relevant features. It requires standard problems to run to completion. This is unnecessary for features can be tested in a short-running problem. For example, many trips and controls can be tested in the first few time steps, as can a number of fluid flow options. The testing system is also inaccurate. For the past decade, the diffem script has been the primary tool for checking that printouts from two different RELAP5-3D executables agree. This tool compares two output files to verify that all characters are the same except for those relating to date, time and a few other excluded items. The variable values printed on the output file are accurate to no more than eight decimal places. Therefore, calculations with errors in decimal places beyond those printed remain undetected. Finally, fidelity of restart is not tested except in the PVM sub-suite and backup is not specifically tested at all. When a restart is made from any midway point of the base-case transient, the restart must produce the same values. When a backup condition occurs, the code repeats advancements with the same time step. A perfect backup can be tested by forcing RELAP5 to perform a backup by falsely setting a backup condition flag at a user-specified-time. Comparison of the calculations of that run and those produced by the same input w/o the spurious condition should be identical. Backup testing is more difficult the other kinds of testing described above because it requires additional coding to implement. The testing system constructed and described in this document resolves all of these issues. A matrix of test features and short-running cases that exercise them is presented. A small information file that contains sufficient data to verify calculations to the last decimal place and bit is produced. This testing system is used to test base cases (called null testing) as well as restart and backup cases. The programming that implements these new capabilities is presented.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Is the testing effect ready to be put to work? Evidence from the laboratory to the classroom.
The testing effect refers to the benefits to retention that result from structuring learning activities in the form of a test. As educators consider implementing test-enhanced learning paradigms in real classroom environments, we think it is critical to consider how an array of factors affecting test-enhanced learning in laboratory studies bear on test-enhanced learning in real-world classroom environments. As such, this review discusses the degree to which test feedback, test format (of formative tests), number of tests, level of the test questions, timing of tests (relative to initial learning), and retention duration have import for testing effects in ecologically valid contexts (e.g., classroom studies). Attention is also devoted to characteristics of much laboratory testing-effect research that may limit translation to classroom environments, such as the complexity of the material being learned, the value of the testing effect relative to other generative learning activities in classrooms, an educational orientation that favors criterial tests focused on transfer of learning, and online instructional modalities. We consider how student-centric variables present in the classroom (e.g., cognitive abilities, motivation) may have bearing on the effects of testing-effect techniques implemented in the classroom. We conclude that the testing effect is a robust phenomenon that benefits a wide variety of learners in a broad array of learning domains. Still, studies are needed to compare the benefit of testing to other learning strategies, to further characterize how individual differences relate to testing benefits, and to examine whether testing benefits learners at advanced levels.
60 APPLIED LIFE SCIENCES↗
Component Testing Supporting SHINE Medical Technologies
SHINE is going to be implementing a tritium gas processing system to support their Mo-99 work. They have requested that the Hydrogen Process Group of SRNL does some required testing on the system components before startup. The lab plans on conducting both seismic and valve testing. Test Plan: - Testing 10 1C valves 750,000 cycles or until failure; - Testing intervals will be every couple thousand; - 1C valves will have metal stem tips instead of Polymer due to being in a tritium prone area; - Hope that the valves will still have enough force to create a sufficient seal with metal stem tips; - Goal is for a more cost and space efficient alternative to the 5C valves. Status: - Received half the parts, waiting on the other half due to long procurement times; - Waiting on the LabVIEW program which will perform the testing. Path Forward: - R and DE to change out the stem tips (July 2018); - Assemble Manifold (July 2018); - Perform Testing (August 2018); - Report (September 2018). Since SRNL does not own a shaker table, the test will be conducted at a facility like the NTS Facility in Huntsville, Alabama. Test Plans: - Leak Testing; - Use LabVIEW to monitor for leaks before, during and after seismic testing with an Electronic Pressure Transducer; - If leak is detected, sniff for location using Helium Leak Detector; - Connector Testing; - Use LabVIEW to test for electrical continuity before, during and after seismic testing; - If a connectors continuity is interrupted, further testing will be performed; - Each Pin in faulty connector will be testing to determine specific location of disruption. Path Forward: - Pick a Qualified Vender for testing (July 2018); - Procure the connector plugs (July 2018); - Send components and test plan to facility for testing (August 2018); - Perform Testing (August-September 2018); - Report Results (September 2018)
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
High-Mileage Courier Fleet Vehicle Laboratory Battery Pack Testing
For one of each of the AVTA's plug-in hybrid electric vehicles, battery electric vehicles, and some hybrid electric vehicles tested in high-mileage courier fleets, the high-voltage traction battery packs were removed from the vehicle and tested at the beginning and end of fleet testing. For some vehicles, batteries also were tested at periodic intervals during fleet testing. Standard reference performance tests were conducted to characterize battery degradation over time. This dataset contains results from two or more rounds of battery tests for 21 distinct year/make/model vehicles (see reference ["INL Advanced Vehicle Testing Activity: On-road Logger and Laboratory Battery Pack Testing Vehicle List"](https://avt.inl.gov/sites/default/files/pdf/reports/DatasetVehicleList.pdf) for full list of vehicles). Each round of battery testing included the "Static Capacity Test" and the "Hybrid Pulse Power Characterization (HPPC) Test", conducted according to test procedures published in the United States Advanced Battery Consortium ["Battery Test Manual For Power-Assist Hybrid Electric Vehicles"](https://www.uscar.org/commands/files_download.php?files_id=57), ["Battery Test Manual For Plug-In Hybrid Electric Vehicles"](https://www.uscar.org/commands/files_download.php?files_id=168), and ["Electric Vehicle Battery Test Procedures Manual"](https://www.uscar.org/commands/files_download.php?files_id=5) prior to the time of testing. These tests were performed by Intertek Testing Services, North America. (This dataset will be shared by API; a small sample of the vehicle battery and test data has been extracted and is available for download while the API is being developed.)
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
High-Mileage Courier Fleet Vehicle Laboratory Battery Pack Testing
For one of each of the AVTA's plug-in hybrid electric vehicles, battery electric vehicles, and some hybrid electric vehicles tested in high-mileage courier fleets, the high-voltage traction battery packs were removed from the vehicle and tested at the beginning and end of fleet testing. For some vehicles, batteries also were tested at periodic intervals during fleet testing. Standard reference performance tests were conducted to characterize battery degradation over time. This dataset contains results from two or more rounds of battery tests for 21 distinct year/make/model vehicles (see reference ["INL Advanced Vehicle Testing Activity: On-road Logger and Laboratory Battery Pack Testing Vehicle List"](https://avt.inl.gov/sites/default/files/pdf/reports/DatasetVehicleList.pdf) for full list of vehicles). Each round of battery testing included the "Static Capacity Test" and the "Hybrid Pulse Power Characterization (HPPC) Test", conducted according to test procedures published in the United States Advanced Battery Consortium ["Battery Test Manual For Power-Assist Hybrid Electric Vehicles"](https://www.uscar.org/commands/files_download.php?files_id=57), ["Battery Test Manual For Plug-In Hybrid Electric Vehicles"](https://www.uscar.org/commands/files_download.php?files_id=168), and ["Electric Vehicle Battery Test Procedures Manual"](https://www.uscar.org/commands/files_download.php?files_id=5) prior to the time of testing. These tests were performed by Intertek Testing Services, North America. This dataset is shared by API; a small sample of the vehicle battery and test data has been extracted and is also available for download.
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
BOP-301 and BOP-302R: Test Definitions and Analyses
Following the conclusion of the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on analysis of two EBR-II tests, Argonne has continued its EBR-II analysis with additional tests from the Shutdown Heat Removal Test (SHRT) series. The tests selected for this follow-up analysis are BOP-301 and BOP-302R, two unprotected loss of heat of sink tests in which the intermediate sodium pump tripped without scramming the control rods or tripping the primary pumps. The IAEA CRP focused on the SHRT-17 and SHRT-45R tests, the most severe protected and unprotected loss of flow tests performed during the SHRT series. With the primary pumps tripping, flow rates decreased to levels where flowmeter uncertainty was very high. Accurately capturing the progressions of the two tests despite this uncertainty was one of the key modeling and simulation challenges during the CRP. BOP-301 and BOP-302R were selected because the primary sodium pumps did not trip for those tests. Primary flow rate measurements are known with greater certainty. Power and temperature discrepancies can be explored without concern that flow predictions match test measurements but not the actual flow rates from the tests. These two tests were performed during the same test window as SHRT-45R and had the same core load configuration. Therefore, core models created for analysis of SHRT-45R were easily adapted for analysis of BOP-301 and BOP-302R. Another reason for analyzing BOP-301 and BOP-302R is to analyze tests in which the core inlet temperature changed significantly, which did not occur during the loss of flow tests. The transient progression of the unprotected SHRT-45R test was driven by a decreasing core flow rate while the core inlet temperature remained relatively constant. For the BOP tests, the transient progression was driven by an increasing core inlet temperature while the core flow rate was stable. This report defines the BOP-301 and BOP-302R tests and provides a selection of measured test data. Initial predictions from SAS4A/SASSYS-1 simulations of these tests are then presented. Following these simulations, additional analysis was performed with the Dakota uncertainty quantification and optimization toolkit to determine which parameters have the greater impact on the transient results and the optimal values of the most important parameters.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
In-Sodium Testing of a Prototype Thermoacoustic Power Sensor for Sodium-Cooled Fast Reactors
The ultimate goal of this project is to develop and demonstrate a thermoacoustic power sensor (TAPS) for Sodium-Cooled Fast Reactors (SFRs), with potential application also envisioned to other nuclear technologies such as Lead-Cooled Fast Reactors (LFRs), Molten Salt Reactors (MSRs), in addition to Light Water Reactors (LWRs). The project was led by Westinghouse Electric Company, LLC (Westinghouse) and carried out in collaboration with Argonne National Laboratory (ANL) and the University of Pittsburgh. A TAPS is a passive (self-powered), non-invasive (wireless) sensor envisioned for measuring key parameters, such as local temperature and neutron flux, in a nuclear reactor core. The sensor generates pressure waves (i.e., sound waves) with a frequency and amplitude dependent upon nuclear operating conditions (coolant temperature or power changes). The acoustic waves are able to travel through the core and associated structures, and reach to the sensor network placed outside and/or inside of the reactor vessel. These sensors require a very small amount of power which, during loss of power events, can be provided, for example, by harvesting gamma radiation energy, thus resulting in a monitoring system that can function both during normal operation and during loss of power events. Westinghouse and the University of Pittsburgh designed and fabricated TAPS prototypes for Argonne National Laboratory (ANL) to carry out in-sodium testing to evaluate the effects of sodium on the TAPS and the performance of the TAPS technique in sodium. Argonne received a TAPS prototype from Westinghouse, and the prototype was modified such that it can be installed into a test vessel and function in sodium at elevated temperature without potentially leaking. A water mockup test apparatus was constructed to validate proper working of the prototype. An instrumentation and control (I&C) system, running on the National Instruments (NI) LabView platform, was developed to: 1) operate both the water mockup test and the in-sodium test facility; and 2) process and analyze the received acoustic signals from an array of accelerometers and the Argonne sodium-submersible high-temperature acoustic sensor. The prototype was successfully tested in a water bath at different temperatures. Water mockup tests demonstrated that the TAPS prototype is working properly and its resonance frequency changes linearly with the coolant (water) temperature. A TAPS test apparatus was constructed and integrated with the upgraded Under-Sodium Viewing (USV) sodium test facility. The integrated USV-TAPS sodium test facility has been operational. The TAPS prototype and a high-temperature sodium-submersible acoustic sensor (SSAS) developed by Argonne were both installed inside the TAPS test vessel. Being operated within argon cover gas under ambient conditions, the TAPS prototype demonstrated that it was functioning properly with a resonance frequency at 1407.2 Hz, which was successfully detected by the accelerometers mounted on the external wall of the vessel and the high-temperature SSAS installed inside the vessel. After successfully transferring sodium into the vessel, in-sodium tests of the prototype were conducted. Tests of the TAPS prototype demonstrated that the resonance frequency of the TAPS changes linearly with respect to the temperature difference between the interior of the TAPS and bulk sodium. The early tests showed that the TAPS prototype could not establish a continuous and consistent resonance in sodium. The resonance diminished before the prototype reached its operating temperature. A signal postprocessor was added to the DAQ module latterly to isolate interferences, enhance signal conditioning, improve peak detection, and generate resonance frequency versus temperature plots. After testing in molten sodium and immersion at higher temperature for several weeks, the TAPS prototype was able to establish a continuous resonance. Performance evaluation of the TAPS prototype was then conducted in sodium. The tests included the investigation of 1) the effects of the temperature difference between the TAPS and bulk sodium, 2) the effects of sodium flowrate; and 3) the performance of the different sensor-receiver systems positioned inside or outside the vessel. Results of a test demonstrated that, with limited sodium circulation, a continuous and consistent resonance of the TAPS prototype was established occasionally. The tests also demonstrated that, because of the nature of detection principles and mounting methods, the high temperature SSAS is more affected by acoustic noise, while accelerometers are more affected by vibrations, in the test environment. It is unknown why the TAPS prototype only occasionally established a continuous and consistent resonance when the TAPS temperature reached its operating temperature in molten sodium, and why it ultimately failed to resonate at all. A failure modes assessment was conducted and a few potential causes of failure were identified. Different post in-sodium tests were conducted to obtain additional information potentially relevant to the cause of the failure. Nondestructive evaluation techniques are suggested to examine the internal integrity as well as the gas mixture of the prototype. If they prove inconclusive, the prototype should be cut open to conduct a thorough inspection of its internal integrity and determine the state of the gas mixture.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗