Very high burnup vapor transport fuel pin concept for long life nuclear reactors
Very high burnup vapor transport fuel pin concept for long life nuclear reactors
Finnegan, P. M.↗
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Thermal mechanical assessment of a SiC-SiC-composite clad fuel pin concept in a light water reactor environment
Accident Tolerant Fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady state reactor operational performance. A potential ATF concept is SiC-SiC composite claddings. Fuel performance simulations were conducted on a SiC-SiC based cladding concept utilizing a multilayered approach for improved performance. This cladding concept referred to as the Duplex concept is a duplex structure composed of a monolithic SiC layer placed on the outside of a SiC-SiC composite. A liquid metal is added to fuel-cladding gap for improved heat dissipation from the fuel. The monolithic SiC layer is used to improve the coolant corrosion characteristics and protect the SiC-SiC composite layer from exposure to the coolant. The fuel performance code BISON was used to conduct fuel performance simulations on the cladding concepts. Comparisons are made with a current prototypic fuel rod design (UO 2 fuel enclosed in Zircaloy-4 cladding). Representative steady-state cases were considered for normal power and two cycle power histories. Additionally, a PCI ramp case was simulated to analyze potential anticipated operational occurrences. Transient response during a Loss of Coolant Accident and a Reactivity Initiated Accident were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard power cases evaluated. The ramping evaluations indicate potential fracturing of the SiC-SiC composite of the composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In conclusion, the rod fails early at low enthalpy for RIA but survives a LOCA with minimal material loss due to high temperature steam corrosion.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Fabrication of capsule assemblies, phase 3
Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.
Keeton, A. R.↗
TEV-3349 "ATF-2 2A Loop Pin Cladding Requirements"
Industry partners, including Westinghouse, Framatome, General Electric, General Atomics, Toshiba, etc. require requirements for fuel pins they insert in the reactor for testing. This document contains those requirements. These requirements represent an agreement between ATR and NS&T, and include the requirements for pins fabricated at the INL. These include inspections such as radiography, helium leak check, etc.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept
An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.
Bowles, K. J.↗
Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C
The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.
Bowles, K. J.↗
Specification of EBR-II Fabricated Fuel Diameter Data
All IFR EBR-II fuel pins were fabricated at one of the related fuel fabrication facilities at Argonne National Laboratory-West, at the EBR-II site [1]. U-Pu-Zr experimental pins were fabricated at the Experimental Fuels Laboratory (EFL), which was established in April of 1984 after the positive acceptance of the IFR concept, while the U-Zr experimental pins were fabricated at the Fuel Manufacturing Facility (FMF), along with the driver fuel pins [2]. The injection-casting fabrication process is described in detail in ref [1]. All the fuel pins that were loaded into EBR-II were examined after fabrication to ensure that they met the design dimensional requirements and tolerances prior to assembly. Fuel slug diameter was measured with either a laser profilometer (at FMF) or a hand micrometer (at EFL), at a minimum of 1-inch intervals over the length of the fuel slug, as shown in Figure 1 [2]. In all cases the average fuel diameter is reported, in some cases with additional summary statistics (e.g., minimum, maximum), and in rare cases the complete record of axially-varying fuel diameters is available. It should be noted that in cases where axial varying data is available, in general the orientation of the slug was not recorded or carried over to loading of the pin, so a diameter measurement at the bottom of the fabricated slug does not necessarily correspond to the bottom of the loaded slug, as it may have been loaded in either orientation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
ATF-2 Ramp Conceptual Design Report
The following report documents the conceptual design for the ATF-2 Ramp experiment. The experiment aims to fill the in-pile irradiation testing gap to conduct integral ramp testing, which was created by the closure of R2, Osiris, and Halden test reactors. The concept involves the simultaneous ramping of three fuel pins using a power axial locator mechanism in the Loop-2A testing facility in the center flux trap of the advanced test reactor (ATR). The three fuel pins are ramped in individual coolant channels containing a prototypic PWR environment. Rod failure will be detected using a fuel rod elongation sensor attached to the fuel pin upper end cap. An axial stack of concentric hafnium and zirconium shrouds will be used to shape the flux around the test pins to create different power levels in each pin and to ensure the peak power location of each pin remains in the center of the test rod. Monte Carlo simulations are used to demonstrate the viability of this design concept.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Fabrication of fuel pin assemblies, phase 3
Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.
Keeton, A. R.↗
Concept of Operations for Microreactor Transportation
Microreactors are very small nuclear reactors with a power output of 20 MWe or less and are designed to be factory-built, modular in nature, and highly portable. The Idaho National Laboratory (INL) has been identified as a potential location to demonstrate microreactors and two microreactor concepts currently under consideration for demonstration at the INL include a tristructural isotropic (TRISO) fueled high-temperature gas reactor and a sodium cooled reactor using a SNAP-10A style fuel pin design. The objective of this report is to identify a concept of operations (CONOPS) for shipping these microreactors.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Emergency cooling analysis for the loss of coolant malfunction
This report examines the dynamic response of a conceptual space power fast-spectrum lithium cooled reactor to the loss of coolant malfunction and several emergency cooling concepts. The results show that, following the loss of primary coolant, the peak temperatures of the center most 73 fuel elements can range from 2556 K to the region of the fuel melting point of 3122 K within 3600 seconds after the start of the accident. Two types of emergency aftercooling concepts were examined: (1) full core open loop cooling and (2) partial core closed loop cooling. The full core open loop concept is a one pass method of supplying lithium to the 247 fuel pins. This method can maintain fuel temperature below the 1611 K transient damage limit but requires a sizable 22,680-kilogram auxiliary lithium supply. The second concept utilizes a redundant internal closed loop to supply lithium to only the central area of each hexagonal fuel array. By using this method and supplying lithium to only the triflute region, fuel temperatures can be held well below the transient damage limit.
Peoples, J. A.↗
Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor
This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.
Godfroy, Thomas↗
Development of thermoplastic composite aircraft structures
Efforts focused on the use of thermoplastic composite materials in the development of structural details associated with an advanced fighter fuselage section with applicability to transport design. In support of these designs, mechanics developments were conducted in two areas. First, a dissipative strain energy approach to material characterization and failure prediction, developed at the Naval Research Laboratory, was evaluated as a design/analysis tool. Second, a finite element formulation for thick composites was developed and incorporated into a lug analysis method which incorporates pin bending effects. Manufacturing concepts were developed for an upper fuel cell cover. A detailed trade study produced two promising concepts: fiber placement and single-step diaphragm forming. Based on the innovative design/manufacturing concepts for the fuselage section primary structure, elements were designed, fabricated, and structurally tested. These elements focused on key issues such as thick composite lugs and low cost forming of fastenerless, stiffener/moldine concepts. Manufacturing techniques included autoclave consolidation, single diaphragm consolidation (SDCC) and roll-forming.
Renieri, Michael P.↗
Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing
This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.
Wright, S. A.↗
Advanced High-Temperature Sodium-Cooled Thermal Reactors Using Less Than 10% Enriched UO 2 Fuel
Here, this paper presents a 1200-MW(thermal) advanced sodium-cooled thermal reactor concept that uses online refueling of 3.5% to 9.95% enriched UO2 fuel pin bundles; uses either graphite or beryllium oxide (BeO) as a neutron moderator; reaches outlet temperatures of 650°C enabling a thermal efficiency of at least 45%; has a high specific power of 133 W/g U; has average power densities of 16.4 and 43.2 W/cm 3 with graphite and BeO, respectively; reaches an average discharge burnup of 100 MWd/kg U; and generates 52% less spent fuel volume, 28% less fission products, and 47% to 64% less transuranics than a typical large pressurized water reactor for the same amount of electricity produced.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Fast-spectrum space-power-reactor concepts using boron control devices
Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.
Mayo, W.↗
Effect of nitrogen on tensile properties and structures of T-111 (tantalum, 8 percent tungsten, 2 percent hafnium) tubing
The effect of controlled nitrogen additions was evaluated on the mechanical properties of T-111 (Ta-8W-2Hf) fuel pin cladding material proposed for use in a lithium-cooled nuclear reactor concept. Additions of 80 to 1125 ppm nitrogen resulted in increased strengthening of T-111 tubular section test specimens at temperatures of 25 to 1200 C. Homogeneous distributions of up to 500 ppm nitrogen did not seriously decrease tensile ductility. Both single and two-phase microstructures, with hafnium nitride as the second phase, were evaluated in this study.
Buzzard, R. J.↗
High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results
Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.
Bragg-Sitton, Shannon M.↗