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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Readout system of TPC/MPD NICA project

The time-projection chamber (TPC) is the main tracking detector in the MPD/NICA. The information on charge-particle tracks in the TPC is registered by the MWPG with cathode pad readout. The frontend electronics (FEE) are developed with use of modern technologies such as application specific integrated circuits (ASIC), field-programmable gate arrays (FPGA), and data transfer to a concentrator via a fast optical interface. The main parameters of the FEE are as follows: total number of channels, ∼95 000; data stream from the whole TPC, 5 GB/s; low power consumption, less than 100 mW/ch; signal to noise ratio (S/N), 30; equivalent noise charge (ENC), <1000e{sup –} (C{sub in} = 10–20 pF); and zero suppression (pad signal rejection ∼90%). The article presents the status of the readout chamber construction and the data acquisition system. The results of testing FEE prototypes are presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Transient Reactor Test Facility (TREAT) Low Enriched Uranium (LEU) Conversion Progress

The National Nuclear Security Administration Office of Material Management and Minimization (NNSA M3) is working towards converting the Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory (INL) from its original high-enriched uranium (HEU) core to a low-enriched uranium (LEU) core. TREAT reactor operations were restarted in November of 2017 and data obtained from the initial testing will be used to validate the existing ANL reactor modeling. Fuel block development has progressed with the completion of 2 and 4 inch crack free surrogate fuel blocks. Alternative fabrication methods for the fuel blocks continue to be explored and are showing promising results. Additionally, cladding oxidation studies have been performed and analyzed to initiate the down selection of cladding material to be used for the TREAT LEU fuel elements. This paper describes the current status of the above mentioned activities, and discusses the anticipated path forward for the conversion program as it progresses toward a final fuel design and fuel fabrication process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Chemistry and Reactivity of Oxide Films on Aluminum Clad Test Reactor Fuels

This paper describes recent work performed to assess the degradation of aluminum spent nuclear fuel (ASNF) as part of an ongoing Department of Energy project. As part of a multi-task effort, this paper specifically describes efforts initiated in 2018 to better understand chemical phenomena and degradation in ASNF). In this paper experimental and modeling activities are described. The modeling effort presents initial results in two areas potentially influencing stability. Using an established electrolyte model based on experimental data, the solubility of aluminum as a function of temperature and pH was examined. This work describes interesting results for the minimum pH providing relevant data that was not available from the literature. It was found that solubility increases with temperature and that the minimum pH shifts to low pH with increase in temperature. The second area involves understanding the metallurgy of aluminum alloys during fuel drying. In this study, the temperature range from 100°C to 300°C was analyzed. Thermodynamic models predict that AA1100 will not experience significant phase changes while the more highly alloyed AA6061 and AA5052 may experience significant phase changes. These thermodynamic measurements establish a potential for phase changes during drying, however they do not predict the kinetics for such processes. Phase 1 of the experimental program has initiated and results are forthcoming. The test program will examine aluminum oxidation under anticipated dry storage conditions (temperature and humidity). Examination of localized corrosion due to tight crevices and galvanic contact (with stainless steel) will be included. Finally, water deliquescence will be assessed using thermogravimetric analysis with dynamic humidity profiles.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Testing of the Authenticatable Container Tracking System (ACTS)

The Authenticatable Container Tracking System (ACTS) is being designed to provide a secure container tag that can be used to optimize chain-of-custody monitoring for packaged nuclear materials as they are being stored, processed, and transported. ACTS is an active device that uses an almost-universal core platform that can be appropriately configured to provide the necessary data acquisition, data logging, and communications functions needed for 21st century material accountancy, monitoring, and tracking applications. The core design supports active monitoring of containment, movement, and location and provides secure data transmission via selectable communication methods that can be used throughout the world. This core architecture enables appropriately designed modules to be easily interfaced to the basic system, thus providing an integration path for current and new technologies. ACTS contains a built-in set of sensors but also supports an expansion bus for up to six additional communications, tracking, or sensing modules that may enhance monitoring and tracking of particular containers for specific applications. ACTS stores events in memory, communicates activity to Argonne National Laboratory’s ARG-US TransPort tracking and information server, and provides options for data encryption and authentication. We are currently beginning a testing phase that will field five devices outdoors in a controlled environment to allow evaluation of the functionality of the systems and the hardness of the devices to outdoor conditions. Partners for this system are Oak Ridge National Laboratory, Argonne National Laboratory, Savannah River National Laboratory, Pacific Northwest National Laboratory, and commercialization partner The Aquila Group.

Britton Jr, Chuck↗
DOE OSTI2016

EPRI Software for Estimating Decommissioning Waste Volumes - 16046

EPRI has adapted computer software that can estimate the volume and activity content of waste resulting from the dismantlement of nuclear facilities. This software, designated as Automatic Estimation of the Radiological Inventory or AERI, was originally developed by Iberdrola for use by ENRESA in Spain. The software estimates the radiological inventory for all the components and civil structures of the plant or other facility that are included in the physical inventory. The radiological inventory shows the internal and external surface contamination, as well as the volumetric contamination. The processes of the software are able to estimate both the specific activity levels (in Bq/g or pCi/g) for each radionuclide and the surface activity (Bq/cm{sup 2} or dpm/100 cm{sup 2}) for components and civil structures located in the radiological areas. AERI allows the user to extrapolate, through a statistical approach, the contamination levels obtained by sampling data in specific components and civil structures to the entire facility. The main benefits and values obtained with AERI are the following: - The radiological inventory of the plant can be obtained from a limited set of radiological sampling data. Using the mathematical models allows for the optimization of the number of sampling and analysis that needs to be done. The uncertainties of the calculations are reduced with more data. The user must balance the level of uncertainty of the calculations and the number of sample data they will use. - The automatic management of radiological data in short computing times allows for the execution of several scenarios in a reasonable time. Therefore, different radioactive waste and decommissioning hypotheses can be tested. For example, inventory and data can be split into two different set. One of them can be used for obtaining radiological inventory for early dismantling (3 years, as example) and the second one for deferred dismantling (25 years.) - The software supports the planning and implementation of decommissioning strategies for nuclear plants. For instance, several decontamination factors can be evaluated, assuming different level for decontamination of materials, and results for materials from several streams and use of different container types may be included in each evaluation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Evaluation of Spent Nuclear Fuel Disposition in Salt (FY18)

This report summarizes the 2018 fiscal year (FY18) field, laboratory, and modeling work funded by the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) campaign as part of the Sandia National Laboratories Salt Research and Development (R&D) and Salt International work packages. This report satisfies level-two milestone M2SF-18SNO10303031and comprises three related but stand-alone sections. The first section summarizes the programmatic progress made to date in the DOE-NE salt program and its goals going forward. The second section presents brine composition modeling and laboratory activities related to salt evaporation experiments, which will be used to interpret data collected during the heater test. The third section presents theoretical and numerical modeling work done to investigate the effects brine composition have on dihedral angle and the permeability of salt.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Integrated sampling and data flow for Hanford vitrification - 15337

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is a one-of-a-kind nuclear facility being designed, built, and commissioned for the U.S. Department of Energy (DOE) by Bechtel National, Inc. and its subcontractor URS Corporation. The WTP facilities are under construction to complete the design and construction, as well as planning to support the startup, commissioning, and operation for vitrifying the radioactive waste stored in aging underground tanks at the Hanford Site. The work progress includes development of waste feed acceptance and qualification program, including methodologies for processing feed batches. Qualification data from staged feed samples will support the verification that waste acceptance criteria (WAC) and processability requirements are met. This verification is required to ensure safe operation of WTP processes for vitrifying the waste, while establishing specific campaign and batch sheets to comply with glass form acceptance requirements. Sampling and data integration are critical for process control and monitoring of the feed through unit operations. Data collection begins with process knowledge of staged feed for determining the required WAC analyses and testing for unit operations. The waste feed qualification data form the baseline information for WTP operation. The qualification data will be tracked and compared with sample analyses at different steps in vitrification operation. All aspects of data life cycle will be managed by the WTP Plant Engineering organization. The data management approach is established in this manner to allow maintaining campaign traceability through glass production records, while safely processing the corresponding feed campaigns and complying with permitting requirements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

A near infra-red video system as a protective diagnostic for electron cyclotron resonance heating operation in the Wendelstein 7-X stellarator

The Wendelstein 7-X stellarator is a large nuclear fusion device based at Max-Planck-Institut für Plasmaphysik in Greifswald in Germany. The main plasma heating system for steady state operation in W7-X is electron cyclotron resonance heating (ECRH). During operation, part of plama facing components will be directly heated by the non-absorbed power of 1 MW rf beams of ECRH. In order to avoid damages of such components made of graphite tiles during the first operational phase, a near infra-red video system has been developed as a protective diagnostic for safe and secure ECRH operation. Both the mechanical design housing the camera and the optical system are very flexible and respect the requirements of steady state operation. The full system including data acquisition and control system has been successfully tested in the vacuum vessel, including on-line visualization and data storage of the four cameras equipping the ECRH equatorial launchers of W7-X.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS2015

Affordable Development and Demonstration of a Small NTR Engine and Stage: A Preliminary NASA, DOE, and Industry Assessment

In FY11, NASA formulated a plan for Nuclear Thermal Propulsion (NTP) development that included Foundational Technology Development followed by system-level Technology Demonstrations. The ongoing NTP project, funded by NASA's Advanced Exploration Systems (AES) program, is focused on Foundational Technology Development and includes 5 key task activities: (1) Fuel element fabrication and non-nuclear validation testing of heritage fuel options; (2) Engine conceptual design; (3) Mission analysis and engine requirements definition; (4) Identification of affordable options for ground testing; and (5) Formulation of an affordable and sustainable NTP development program. Performance parameters for Point of Departure designs for a small criticality-limited and full size 25 klb (sub f) (kip, or 1000 pound-force)-class engine were developed during FYs 13-14 using heritage fuel element designs for both Rover NERVA (Nuclear Engine for Rocket Vehicle Applications) Graphite Composite (GC) and Ceramic Metal (Cermet) fuel forms. To focus the fuel development effort and maximize use of its resources, the AES program decided, in FY14, that a leader-follower down-selection between GC and cermet fuel was required. An Independent Review Panel (IRP) was convened by NASA and tasked with reviewing the available fuel data and making a recommendation to NASA. In February 2015, the IRP recommended and the AES program endorsed GC as the leader fuel In FY14, a preliminary development schedule DDTE (Design Development Test and Evaluation) plan was produced by GRC (Glenn Research Center), DOE (Department of Energy) and industry for the AES program. Assumptions, considerations and key task activities are presented here. At the direction of NASA HQ (32515), NASA and DOE are to work together to formulate a detailed development plan and schedule allowing the affordable development of a small (7.5-16.5 klbf) GC engine for possible flight technology demonstration (FTD) mission within a 10-year timeframe.

Lunar Flyby Mission↗
DOE OSTI2013

A Historical Evaluation of the U16a Tunnel, Nevada National Security Site, Nye County, Nevada Volume 2

This report presents a historical evaluation of the U16a Tunnel on the Nevada National Security Site in southern Nevada. The work was conducted by the Desert Research Institute at the request of the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office and the U.S. Department of Defense, Defense Threat Reduction Agency. The U16a Tunnel was used for underground nuclear weapons effects tests in Shoshone Mountain in Area 16 of the Nevada National Security Site. Six nuclear tests were conducted in the U16a Tunnel from 1962 to 1971. These tests are Marshmallow, Gum Drop, Double Play, Ming Vase, Diamond Dust, and Diamond Mine. The U.S. Department of Defense Threat Reduction Agency, with participation from Lawrence Livermore National Laboratory and Las Alamos National Laboratory, sponsored the tests. Fifteen high explosives tests were also conducted at the tunnel. Two were calibration tests during nuclear testing and the remaining were U.S. Department of Defense, Defense Threat Reduction Agency tunnel defeat tests. The U16a Tunnel complex is on the top and slopes of Shoshone Mountain, encompassing an area of approximately 16.7 hectares (41.1 acres). Major modifications to the landscape are a result of three principal activities, road construction and maintenance, mining activities related to development of the tunnel complex, and site preparation for activities related to testing. Forty-seven cultural features were recorded at the portal and on the slopes of Shoshone Mountain. At the portal area, features relate to the mining, construction, testing, and general every day operational support activities within the tunnel. These include concrete foundations for buildings, equipment pads, and rail lines. Features on the slopes above the tunnel relate to tunnel ventilation, borehole drilling, and data recording. Feature types include soil-covered bunkers, concrete foundations, instrument cable holes, drill holes, and ventilation shafts. The U16a Tunnel complex is eligible to the National Register of Historic Places under criteria a and c, consideration g of 36 CFR Part 60.4 as a historic landscape. Scientific research conducted at the tunnel has made significant contributions to the broad patterns of our history, particularly in regard to the Cold War era that was characterized by competing social, economic, and political ideologies between the former Soviet Union and the United States. The tunnel also possesses distinctive construction and engineering methods for conducting underground nuclear tests. The Desert Research Institute recommends that the U16a Tunnel area be left in place in its current condition and that the U16a Tunnel historic landscape be included in the Nevada National Security Site monitoring program and monitored on a regular basis.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

Delayed Critical and Subcritical Experiments with Polyethylene Moderated Unreflected Thin 15 in. Diameter HEU Metal Plates

The thin ~15 in. diameter highly enriched uranium (HEU) metal plates were assembled to delayed criticality at the Oak Ridge Critical Experiments Facility (ORCEF) in 1969 with various thicknesses of polyethylene (varying from 1/16 to 2$\frac{3}{8}$ inches) between uranium metal plates. The average 235 U enrichment was 93.27 wt. %. These unreflected critical configurations contained 4$\frac{2}{3}$ to 20$\frac{5}{6}$ thin 15 in. diameter HEU metal plates (on loan from Los Alamos National Laboratory [LANL] and shipped to Oak Ridge National Laboratory [ORNL] on June 3, 1969). Depending on the thickness of polyethylene, the enriched uranium masses varying from 28,053 to 135,148 grams. Fractional plate sections consisted of the appropriate number of 60° pie sections. In addition to the measurement at delayed criticality, subcritical measurements were also performed by the inverse kinetic rod drop method. Prompt neutron decay constant measurements were also performed by the Rossi alpha and randomly pulsed neutron method using a time-tagged spontaneous fission californium neutron source; these are briefly reported here. At the time of these measurements in 1969, the thin HEU metal plates were in near-pristine condition with extremely little oxidation, allowing better descriptions of the uranium plates than the use of these plates in a heavily oxidized and deteriorated condition in recent reflected benchmark experiments at the LANL facility at the Nevada Test Site with these same thin highly enriched uranium metal plates. This report documents the experimental information for the measurements performed so that later researchers can perform the required uncertainty and calculational analyses and documentation to use these data for an International Nuclear Criticality Safety Benchmark Evaluation Program (ICSBEP) or a Nuclear Energy Agency (NEA) benchmark. Data from the experiments described should be acceptable for use as criticality safety benchmark experiments for the ICSBEP and the NEA nuclear criticality safety benchmark program once the uncertainty analysis on the measured neutron multiplication factors is completed. Additional data—such as the dimensional inspection reports, uranium isotopic information, and other relevant particulars—should be retrieved from the Y-12 Plant or LANL and incorporated in the final ICSBEP benchmark. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, the uncertainties in $k_{eff}$ could be as low as ± 0.0002 for some configurations. Other experiments with smaller-diameter than 15 in. diameter HEU metal plates have been benchmarked in HEU-METFAST-001. The prompt neutron time decay measurements could be the basis for an International Reactor Physics Benchmark Program. Preparation of the present report is part of an effort at ORNL to document more than 15 undocumented critical and subcritical experiments enumerated in ORNL/TM-2019/18 and performed by ORNL at ORCEF and other US Department of Energy critical experiments facilities using more than 500 operational days of critical facility time. This work for this report publication was supported by the Nuclear Criticality, Radiation Transport and Safety NCSP Program at ORNL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1989

Conference on Real-Time Computer Applications in Nuclear, Particle and Plasma Physics, 6th, Williamsburg, VA, May 15-19, 1989, Proceedings

Papers on real-time computer applications in nuclear, particle, and plasma physics are presented, covering topics such as expert systems tactics in testing FASTBUS segment interconnect modules, trigger control in a high energy physcis experiment, the FASTBUS read-out system for the Aleph time projection chamber, a multiprocessor data acquisition systems, DAQ software architecture for Aleph, a VME multiprocessor system for plasma control at the JT-60 upgrade, and a multiasking, multisinked, multiprocessor data acquisition front end. Other topics include real-time data reduction using a microVAX processor, a transputer based coprocessor for VEDAS, simulation of a macropipelined multi-CPU event processor for use in FASTBUS, a distributed VME control system for the LISA superconducting Linac, a distributed system for laboratory process automation, and a distributed system for laboratory process automation. Additional topics include a structure macro assembler for the event handler, a data acquisition and control system for Thomson scattering on ATF, remote procedure execution software for distributed systems, and a PC-based graphic display real-time particle beam uniformity.

Pordes, Ruth↗
DOE OSTI2021

Integrated FY21 Elevated Temperature Mechanical Testing Results for Alloy 709 Code Case

This report provides the status of creep, fatigue, and creep-fatigue testing that transpired in Fiscal Year 2021 at Argonne National Laboratory, Idaho National Laboratory and Oak Ridge National Laboratory. This testing is being conducted to develop the data package to qualify Alloy 709 in Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. This would permit the use of Alloy 709 for elevated-temperature nuclear construction. Preliminary results continue to demonstrate the improved creep and fatigue resistance of Alloy 709 compared to 316H stainless steel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Comparison of Error Rate Depending on Operator Expertise and Simulator Complexity

This paper analyzes operator's error rate from experiments, depending on the expertise and simulator complexity. This study uses the Rancor Microworld, a simplified simulator developed by INL, and the Compact Nuclear Simulator (CNS), a less simplified simulator developed by the Korea Atomic Energy Research Institute (KAERI). The error rates were measured from simulation data of a total of 72 participants, and the collected error rate data were analyzed using analysis of variance (ANOVA) test.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

BISON TRISO Modeling Advancements and Validation to AGR-1 Data

BISON is a finite element-based nuclear fuel performance code. Among its unique characteristics are its ability to model 1D, 2D, and 3D geometries and its applicability to a wide variety of nuclear fuels. For the last eight years, BISON has included a beginning capability to model tri-structural isotropic (TRISO) fuel. Recently, interest in TRISO fuel has grown, and a significant effort has been made to improve BISON’s capabilities in this area. Capability development has occurred for each material present in TRISO fuel particles: the buffer, inner pyrolytic carbon, silicon carbide, and outer pyrolytic carbon layers, as well as the fuel kernel. New elastic, creep, swelling, thermal expansion, thermal conductivity, and fission gas release (FGR) models are available. New models for the graphite matrix are also now available. Another important addition is the ability to perform statistical failure analysis of large samples of fuel particles. This new capability, which continues to grow, enables evaluation of failure due to pressure or crack formation by analyzing many thousands of particles. This enables realistic calculations of fission product release from the many particles in a TRISO-fueled reactor. These capabilities were checked via regression and verification tests. A large number of code benchmarking problems were also run, showing that BISON’s results closely match those of other software tools. Finally, a significant validation effort was completed in which fission product release, measured as part of the AGR-1 capsule experiments, was compared to BISON outputs. BISON outputs compared very well to the experimental data and to PARFUME results. Interest in BISON’s TRISO capabilities is growing, with the U.S. Nuclear Regulatory Commission (NRC) and Westinghouse Electric Company receiving training during the past year. Multiple other entities have expressed interest in or are actively using BISON. Kairos Power, LLC, has a strong partnership with Idaho National Laboratory (INL) regarding the use of BISON for TRISO analysis. While its capabilities still continue to grow, BISON has already become a powerful tool for TRISO analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Chemistry and Reactivity of Oxide Films on Aluminum Clad Test Reactor Fuels - 19294

This paper describes recent work performed to assess the degradation of aluminum spent nuclear fuel (ASNF) as part of an ongoing Department of Energy project. As part of a multi-task effort, this paper specifically describes efforts initiated in 2018 to better understand chemical phenomena and degradation in ASNF). In this paper experimental and modeling activities are described. The modeling effort presents initial results in two areas potentially influencing stability. Using an established electrolyte model based on experimental data, the solubility of aluminum as a function of temperature and pH was examined. This work describes interesting results for the minimum pH providing relevant data that was not available from the literature. It was found that solubility increases with temperature and that the minimum pH shifts to low pH with increase in temperature. The second area involves understanding the metallurgy of aluminum alloys during fuel drying. In this study, the temperature range from 100 deg. C to 300 deg. C was analyzed. Thermodynamic models predict that AA1100 will not experience significant phase changes while the more highly alloyed AA6061 and AA5052 may experience significant phase changes. These thermodynamic measurements establish a potential for phase changes during drying, however they do not predict the kinetics for such processes. Phase 1 of the experimental program has initiated and results are forthcoming. The test program will examine aluminum oxidation under anticipated dry storage conditions (temperature and humidity). Examination of localized corrosion due to tight crevices and galvanic contact (with stainless steel) will be included. Finally, water deliquescence will be assessed using thermogravimetric analysis with dynamic humidity profiles. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

First results of the KM3NeT multi-PMT DOM

The future KM3NeT neutrino telescope will consist of several thousand digital optical modules (DOMs), each of which will be equipped with 31 3-inch photo-multiplier tubes (PMTs). This design has various advantages over the use of one large PMT per optical module, e.g. concerning effective photocathode area per module, improved background suppression and directional reconstruction. Currently, the KM3NeT collaboration is testing a prototype DOM deployed on the instrumentation line of the ANTARES neutrino telescope. The DOM has been operational since mid-April 2013. First data are presented and compared to simulation results. The results are very encouraging and indicate that muon identification and a coarse direction estimation are possible event with a single DOM.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗