Welding and aging of bimetallic refractory metal joints
Bimetallic refractory metal joints electron beam welding and aging for applications to in-pile thermionic converters
Phillips, W. M.↗
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An out-of-core thermionic-converter system for nuclear space power
Design of the nuclear thermionic space power system, 40 50 70 Kw(e) power range, are given. The design configuration (1) meets the constraints of readily available launch vehicles; (2) allows for off-design operation including startup, shutdown, and possible emergency conditions; (3) provides tolerance of failure by extensive use of modular, redundant elements; (4) incorporates and uses heat pipes in a fashion that reduces the need for extensive in-pile testing of system components; and (5) uses thermionic converters, nuclear fuel elements, and heat transfer devices in a geometrical form adapted from existing incore thermionic system designs. Designs and in some cases performance data for elements and groups of the elements of the system are included. Benefits of the highly modular system approach to reliability, safety, economy of development, and flexibility are discussed.
Breitwieser, R.↗
Results of a preliminary experimental investigation of a vapor transport fuel pin
In-Pile experimental tests were conducted on two fuel pins mounted side-by-side in a holder assembly. The fuel pins consisted of stacked fully enriched UO2 pellets enclosed in a type 316 stainless steel clad 1/2 inch outside diameter and 3 inches in length. Each fuel pin contained four pellets for a total of 22 grams of fuel weight. The experiment was operated to maintain the maximum clad temperature at 395 K corresponding to a fuel temperature at the central void surface of 2350 K. Total primary coolant water flow past the pin holder was calculated to be 110 gallons per minute. Forty-four gallons per minute at a velocity of 36.5 feet per second flowed around each fuel pin. Calculations indicated the maximum heat flux was at the point of the highest temperature and was equal to 0.24 kW/sq cm or 7.62 Btu/hr sq ft. A series of high resolution neutron radiographs of the nuclear fueled vapor transport capsule was taken. The thermal neutrons emitted from the core of the 60 megawatt Plum Brook reactor facility were used.
Lanzo, C. D.↗
TLD measurements of gamma heating in heavy elements.
Measurements and calculations of gamma heating in polyethylene and lead containers were done and compared. The objective was to provide a workable method of getting good values for gamma heating in in-pile experiments containing materials of high atomic numbers. It was inferred that a combination of thermoluminescent dosimeter measurements, using Bragg-Gray theory, with photon transport calculations using the ANISN computer program, would meet this objective.
Reilly, H. J.↗
Material development for thermionic fuel-cladding systems.
Fuel-cladding systems capable of maintaining stable dimension and electron emission characteristics at high temperatures under irradiation are essential to the successful operation of nuclear thermionic power sources. Studies regarding two possible systems, involving tungsten clad uranium carbide and uranium oxide, are considered. Out-of-pile developmental efforts and the in-pile evaluation results for the two systems are described. Prototypical emitters have been tested for 8000 hours for the carbide system and 4000 hours for the oxide system.
Yang, L.↗
An out-of-core thermionic-converter system for nuclear space power.
Reexamination of designs of nuclear thermionic space power systems with the converter outside the reactor in the perspective of recent advances in heat-transfer methods, materials, converter performance, and radiation design. The 40- to 70-kW(e) power range is treated. The configuration is found to meet the constraints of readily available launch vehicles. It allows for off-design operation including startup, shutdown, and possible emergency conditions; provides tolerance of failure by extensive use of modular, redundant elements; incorporates and uses heat pipes in a fashion that reduces the need for extensive in-pile testing of system components; and uses thermionic converters, nuclear fuel elements, and heat-transfer devices in a geometrical form adapted from existing incore thermionic system designs.
Breitwieser, R.↗
Developmental status of thermionic materials.
Description of the reference materials selected for the major components of the unit cell of a thermionic pile element (TFE), the out-of-pile and in-pile test results, and current efforts for improving the life and performance of thermionic fuel elements. The component materials are required to withstand the fuel burnup and fast neutron fluence dictated by the thermionic reactor system. Tungsten was selected as the cladding material because of its compatibility with both the carbide and the oxide fuel materials. Niobium was selected as the collector material because its thermal expansion coefficient matches closely with that of the thin aluminum oxide layer used to electrically insulate the collector from the TFE sheath. An unfueled converter has performed stably over 41,000 hr. Accelerated irradiation tests have attained burnups equivalent to that for 40,000 hr of the thermionic reactor under consideration.
Yang, L.↗
Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements
The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.
Phillips, W. M.↗
Developmental status of thermionic materials.
Reference materials selected for the major components of the unit cell of a thermonic fuel element are described. Out-of-pile and in-pile evaluation results are given. Current efforts for improving the life and performance of thermionic fuel element are outlined.
Yang, L.↗
Thermionic fuel element technology status
The results of research, conducted between the mid-1960s and 1973, on the multiconverter thermionic fuel elements (TFEs) that comprise the reactor core of an SP-100 thermionic reactor system are presented. Fueled-emitter technology, insulator technology and cell and TFE assembly technology of the prototypical TFEs which were tested in-pile and out-of-pile during these years are described. The proto-TFEs have demonstrated reproducible performance within 5 percent and no premature failures within the 1.5 yr of operation (with projected 3-yr lifetimes). The two primary life-limiting factors had been identified as thermionic emitter dimensional increase due to interactions with the fuel and electrical insulator structural damage from fast neutrons. Multiple options for extending TFE lifetimes to 7 yr or longer are available and will be investigated in the 1984-1985 SP-100 program for resolution of critical technology issues. Design diagrams and test graphs are included.
Holland, J. W.↗
Testing in Support of Fission Surface Power System Qualification
The strategy for qualifying a FSP system could have a significant programmatic impact. The US has not qualified a space fission power system since launch of the SNAP-10A in 1965. This paper explores cost-effective options for obtaining data that would be needed for flight qualification of a fission system. Qualification data could be obtained from both nuclear and non-nuclear testing. The ability to perform highly realistic nonnuclear testing has advanced significantly throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modern FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.) and extensive data to be taken from the core region. For transient testing, pin power during a transient is calculated based on the reactivity feedback that would occur given measured values of test article temperature and/or dimensional changes. The reactivity feedback coefficients needed for the test are either calculated or measured using cold/warm zero-power criticals. In this way non-nuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. FSP fuels and materials are typically chosen to ensure very high confidence in operation at design burnups, fluences, and temperatures. However, facilities exist (e.g. ATR, HFIR) for affordably performing in-pile fuel and materials irradiations, if such testing is desired. Ex-core materials and components (such as alternator materials, control drum drives, etc.) could be irradiated in university or DOE reactors to ensure adequate radiation resistance. Facilities also exist for performing warm and cold zero-power criticals.
Houts, Mike↗
Nuclear Safety [Vol. 16, No. 1, January-February 1975]
Nuclear Safety covers significant developments in the field of nuclear safety. The scope is limited to topics relevant to the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, safety considerations in regard to the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 AEC Working Paper on Population Density Around Nuclear Power-Plant Sites Adapted by J. Ft. Buchanan, 8 Water-Reactor Safety-Research Information Meeting by Wm. B. Cottrell, D. O. Hobson, and G. D. Whitman; ACCIDENT ANALYSIS: 29 Burnout in Boiling Heat Transfer. Part I: Pool-Boiling Systems by A. E. Bergles; PLANT SAFETY FEATURES: 43 Radionuclide Behavior During Normal Operation of Liquid-Metal-Cooled Fast Breeder Reactors. Part 1: Production by C. A. Erdman and A. B. Reynolds; CONSEQUENCES OF EFFLUENT RELEASE: 60 The Control, Monitoring, and Reporting of Radioactivity in Effluents. Part 1: AEC-Owned Facilities by M. B. Biles and F. E. Coffman, 71 The Control, Monitoring, and Reporting of Radioactivity in Effluents. Part 2: AEC-Licensed Facilities by L. B. Higginbotham and J. T. Collins; OPERATING EXPERIENCES: 76 Incident at the Lucens Reactor by J. M. Miller, 79 IAEA Symposium on Experience from Operating and Fueling of Nuclear Power Reactors by William A. Pryor, 87 Recent Occurrences at Nuclear Reactors and Their Causes Compiled by William R. Casto, 98 Summary of Operating U. S. Power Reactors as of Nov. 1. 1974, 101 Selected Safety-Related Occurrences Reported in September and October 1974 Compiled by William R. Casto; CURRENT EVENTS: 104 General Administrative Activities Compiled by Wm. B. Cottrell, 110 Action on Power-Reactor Projects Undergoing Regulatory Review or Consideration Compiled by Wm. B. Cottrell, 119 Action on Nonreactor Projects Undergoing Regulatory Review or Consideration Compiled by Wm. B. Cottrell, 121 Proposed Rule Changes as of Oct. 31, 1974; MISCELLANY: 97 ASTM—Euratom Symposium on Radiation Dosimetry: Developments and Standardization (Announcement), 118 Workshops on In-Pile Filter Testing (Announcement), 119 IAEA Symposium on Radiological Impacts from Releases from Nuclear Facilities into Aquatic Environments (Announcement).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Sensor Qualification for the Development of Advanced Reactors
A temperature sensor qualification device is undergoing development and testing in order to qualify temperature sensors for advanced reactor metrology applications. This device will provide National Institute of Standards and Technology (NIST) traceable performance testing of temperature sensors while under the high radiation environments experienced while in-core. This is completed by utilizing retractable sensors which transfer the measurements from ex-core to in-core sensors. A detailed discussion of the uncertainty propagation associated with this comparative process is provided. Experimental testing of the temperature distribution occurring in the device has begun and is reported on. It is pertinent to minimize this temperature distribution to prevent variation in sensor measurements. This variation will be indistinguishable from sensor drift for in-pile applications and would result in unnecessarily conservative uncertainty bounds.
46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
AGR-5/6/7 Experiment Preliminary Findings
AGR-5/6/7 in-pile data have been published: Fuel irradiation conditions, Fission gas release-rate-to-birth-rate (R/B) ratios, PIE and safety testing are still in progress and only limited data have been published, and preliminary results.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Irradiation of an enriched uranium (NaCl-UCl 3 ) fuel salt capsule, summary of nondestructive post-irradiation examinations, and solidification modeling
Molten salt reactors (MSRs) are gaining attention due to their potential for safe, carbon-free nuclear energy with reduced waste. However, licensing these reactors is hindered by limited experimental data on fueled salts, both pre- and post-irradiation. Here, the novel Molten-salt Research Temperature-controlled Irradiation (MRTI) vehicle was designed to address knowledge gaps in irradiating enriched‑uranium-bearing salts. The MRTI experiment irradiated 13 cm 3 of UCl₃-NaCl (93 % U-235) salt in the Neutron Radiography (NRAD) Reactor, achieving a burnup of 0.196 GWd/MTU over 390 h. Despite a heater failure, the thermocouple data suggested fission heat kept the salt molten. The MRTI assembly was remotely disassembled for nondestructive post-irradiation examination (PIE), which included precision gamma-ray scanning (PGS) and neutron radiography. Radiograph images showed the location of the salt and the solidification pattern. PGS results provided an early indication that activated materials of construction did not increase in relative intensity in the region of the capsule where the salt was in contact with the material of construction. Additionally, PGS data showed the presence of several gamma emitting fission products, such as Nb-95, Zr-95, Ru-103, Ce-141, and La-140, where Ru-103 had the highest counts at the bottom of the capsule. Computational fluid dynamics modeling supported observations of salt solidification patterns and proved to be a valuable tool to inform PIE activities. The MRTI experiment has thus far provided critical data and lessons learned for fuel salt PIE activities, essential for advancing the technical readiness of MSRs.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
M3AS-25IN1002073: Analysis of data from irradiation testing of printed strain gauges in prototypic nuclear environments
Advancement in additively manufactured strain gauges help address critical technology gaps to accurately monitor real-time materials behavior in reactor experiments. This is critical as it provides data to inform predictive models and simulations that enhance the development of reactors and fuel cycle systems. In this report, additively manufactured strain gauges are exposed to a neutron irradiation environment at the Ohio State University Research Reactor. This report goes over a 2-week campaign for neutron irradiating printed resistive strain gauges and capacitive strain gauges. These results complement the prior separate effects (i.e., mechanical strain, temperature, etc.) testing that were performed on these additive manufactured sensors and presented in prior milestone reports. These results also help progress our understanding of their usage in harsh environment applications. The outcome of developing advanced sensing and instrumentation capabilities plays an important role in increasing the safety, reliability, and energy efficiency of both next-generation and existing nuclear reactors.
36 - MATERIALS SCIENCE↗
Performance Testing of Enhanced Linear Variable Integrated Sensor (ELVIS III) for LVDTs
The Enhanced Linear Variable Intrinsic Sensor (ELVIS) was a 2024 breakthrough discovery by Idaho National Laboratory (INL) in terms of further innovating the usage of all linear variable differential transformers (LVDTs) and enhancing their performance in harsh environments. In irradiation tests conducted within material test reactors (MTRs), LVDTs with an internal temperature sensing capability can address several critical challenges. This report focuses on performance testing of the ELVIS III device with new hybrid LVDTs provided by the Institute for Energy Technology (IFE), which is the world’s sole supplier of nuclear-grade LVDTs. The ELVIS III device was evaluated in terms of temperature showcasing promising results comparable to a typical type-K thermocouple (TC).
46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗