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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

On fast reactor kinetics studies

The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Conceptual design of inverted core lead bismuth eutectic fast reactor for marine applications

The development of an inverted core fast reactor aims to generate 60 MWth for about 30 Effective Full Power Years without refueling. The reactor design is a transportable reactor using UO{sub 2} fuel and lead-bismuth-eutectic cooled designed for marine applications and is intended to improve the reactor performances compared to the normal core design: better condition for passive cooling system capability by lower core pressure drop, taking advantage of potential power uprate from the lower maximum fuel temperature. Systematic design processes are presented in this work: fuel pin geometry selection, fuel assembly (FA) design, and core design. A relationship between pressure drops, coolant velocity, maximum fuel temperature, coolant channel diameter, and fuel volume fraction was introduced in a single graph used as a tool to select fuel pin geometry. Fuel fabrication capability also took place in consideration of FA design which led to 7 holes per FA, and two-dimensional temperature distribution studies were also carried out. Core design processes including radial zoning, axial zoning, and core optimization were conducted using Monte Carlo code MCS, which is UNIST CORE laboratory in-house code. The current core design uses 3 fuel enrichment levels and 3 FA types to control the local power distribution and power shift during its lifetime. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

VASIMR{sup R} Interfaced High Temperature Gas-Cooled Fast Reactor for Space Applications

This project was undertaken to assess the feasibility and advantages of utilizing a gas-cooled fast reactor to supply power to an electric propulsion system for deep space exploration to Europa, one of Jupiter's moons. Specifically, the primary objective of this project was to design a 9 MWe gas cooled fast reactor with an accompanying power conversion system to supply electricity to multiple electrically-powered thrusters for a four year mission duration. The thruster design chosen for this analysis is the Variable Specific Impulse Magnetoplasma Rocket (VASIMR). The VASIMR propulsion system, currently being developed by Ad Astra Rocket Company (AARC), is an electric propulsion rocket that uses electric power to accelerate neutral plasma propellant. Advanced electric space propulsion systems such as the VASIMR offer unique capabilities that allow for higher exhaust velocities that are up to ten times that of traditional chemical rockets. Furthermore, the thrust and specific impulse of the VASIMR system can be varied in order to achieve optimal travel times to given destinations for a desired payload. As a result, electric propulsion systems characteristically have lower ratios of system mass to power output than chemical rockets. This ratio is commonly known as the alpha parameter. Alpha parameters of approximately 10 kg/kWe or less make electric propulsion more attractive for deep space exploration missions in terms of resource efficiency and cost. However, the primary challenge with electric propulsion is to develop a high power density source of electricity in order to minimize the total system mass. The alpha parameter is the main benchmark used in this project to test whether or not the proposed coupled gas cooled reactor and power conversion system design is more competitive than other proposed systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. In addition, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

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 ↗
DOE OSTI2012

Nuclear data uncertainty analysis for the generation IV gas-cooled fast reactor

For the European 2400 MW Gas-cooled Fast Reactor (GoFastR), this paper summarizes a priori uncertainties, i.e. without any integral experiment assessment, of the main neutronic parameters which were obtained on the basis of the deterministic code system ERANOS (Edition 2.2-N). JEFF-3.1 cross-sections were used in conjunction with the newest ENDF/B-VII.0 based covariance library (COMMARA-2.0) resulting from a recent cooperation of the Brookhaven and Los Alamos National Laboratories within the Advanced Fuel Cycle Initiative. The basis for the analysis is the original GoFastR concept with carbide fuel pins and silicon-carbide ceramic cladding, which was developed and proposed in the first quarter of 2009 by the 'French alternative energies and Atomic Energy Commission', CEA. The main conclusions from the current study are that nuclear data uncertainties of neutronic parameters may still be too large for this Generation IV reactor, especially concerning the multiplication factor, despite the fact that the new covariance library is quite complete; These uncertainties, in relative terms, do not show the a priori expected increase with bum-up as a result of the minor actinide and fission product build-up. Indeed, they are found almost independent of the fuel depletion, since the uncertainty associated with {sup 238}U inelastic scattering results largely dominating. This finding clearly supports the activities of Subgroup 33 of the Working Party on International Nuclear Data Evaluation Cooperation (WPEC), i.e. Methods and issues for the combined use of integral experiments and covariance data, attempting to reduce the present unbiased uncertainties on nuclear data through adjustments based on available experimental data. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

A 100 MWe advanced sodium-cooled fast reactor core concept

An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

MOOSE-Based Fast Reactor Core Bowing Capabilities: Coupled Structural Mechanics – Thermal Fluids Demonstration and Related Verification Efforts

Under the U.S. Department of Energy Office of Nuclear Energy’s Advanced Modeling and Simulation (NEAMS) Program, an integrated multiphysics approach is being developed to model the core bowing phenomena important to liquid metal-cooled fast reactors. Core bowing is an important passive safety mechanism in liquid metal-cooled fast reactors and involves multiphysics effects including radiation transport, fluid flow, heat transfer, and mechanical response to temperature and flux gradients. Verification and assessment efforts continued on the Multiphysics Object Oriented Simulation Environment (MOOSE) capabilities relevant for modeling thermo-mechanical core bowing behavior. IAEA Verification Problem 4, which was started in FY23, was further examined with MOOSE capabilities to rectify discrepancies observed in previous years when compared to IAEA benchmark participant data. Meshing and postprocessing capabilities in MOOSE were also advanced by other teams and utilized this year. A thermal fluids-structural mechanical coupling demonstration has performed on 7-assemblyand 19-assembly fast reactor assembly configurations using MOOSE. Subchannel capabilities are used to calculate coolant temperature, and heat conduction capabilities calculate duct wall temperature as well as heat transfer through the inter-assembly gap. Structural mechanical capabilities then deform the mesh, accounting for contact between assemblies, according to the temperature gradients calculated by the thermal solvers. Power distributions are imposed rather than calculated to demonstrate different deformations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Ultra-Long-Life Fast Reactor Cores having Axial Blanket-Driver-Blanket Burnup Strategy with Thorium and PWR Spent Fuels

The sustainable energy over a long-term period can be supplied by ultra-long-life fast reactor cores that are designed by fully utilizing breeding by fast neutrons to be operated over several tens of years without refueling. But the fast reactor cores of this type usually have high positive coolant void reactivity worth resulted from the high TRU (Transuranics) contents in the high burnup, which can degrade the inherent safety through high positive coolant expansion reactivity coefficient. Recently, the authors have designed new small ultra-long-life cores that can be operated over 50{approx}60 EFPYs (Effective Full Power Years) and satisfy the self-controllability conditions under the unprotected accidents. The cores use axial blanket-driver-blanket configuration to achieve ultra-long operational cycle length where the fuel burnup propagates from the outer core to the inner core and from the axially central driver to the outer blanket regions. The objective of this work is to suggest the different design options with thorium addition and TRU and uranium from PWR spent fuels for effectively utilizing the large PWR spent fuel stocks and to analyze the performances of the cores. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Improved Fast Reactor Capability of Griffin in FY23

Griffin is a MOOSE based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE-NE NEAMS program. In FY23, we enhanced capabilities required for fast reactor analysis. This effort included primarily updating the cross-section generation workflow using MC2-3 for various reactor configurations, such as homogeneous, duct-heterogeneous, ring-heterogeneous, and fully-heterogeneous geometries. In addition, we initiated the implementation of a multi-cycle depletion and shuffling capability. To support fast reactor simulation capabilities, we significantly improved the performance of the DFEM-SN-based R-Z transport solver to efficiently solve ultrafine group (over 1000 groups) transport problems. Additionally, the performance of HFEM-PN was improved by introducing red-black iteration, the cmfd acceleration technique, and various optimizations. We also completed the pin power reconstruction capability to support multiphysics simulations while identifying and addressing issues associated with SPH equivalence parameter approach. These enhanced capabilities for fast reactor core simulations, specially HFEM-PN and pin power reconstruction features, were applied to benchmark problems involving ABTR and ABR-1000. These applications showcased excellent agreement with Monte Carlo and other code solutions in terms of eigenvalue, control rod worth, and assembly and pin powers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Overview of the SAS4A Metallic Fuel Models and Extended Analysis of a Postulated Severe Accident in the Prototype Gen-IV Sodium-Cooled Fast Reactor

The SAS4A safety analysis code, originally developed for the analysis of postulated severe accidents in oxide fuel sodium-cooled fast reactors (SFRs), has been significantly extended to allow the mechanistic analysis of severe accidents in metallic fuel SFRs. The SAS4A metallic fuel models simulate the metallic fuel thermomechanical and chemical behavior and track the evolution and relocation of multiple fuel and cladding components during the pretransient irradiation and during the postulated accident, allowing an accurate description of the changes in the local fuel composition. The local fuel composition determines the fuel thermophysical properties, such as freezing and melting temperatures, which in turn affect the fuel relocation behavior and ultimately the core reactivity and power history during the postulated accidents. Models describing the fuel-cladding interaction and eutectic formation, the effects of the in-pin sodium on the in-pin fuel relocation, and the postfailure reentry of the molten fuel and fission gas from the pin plenum have also been added. This paper provides an overview of the SAS4A key metallic fuel models emphasizing the postfailure metallic fuel relocation models included in the LEVITATE-M module of SAS4A. The capabilities of the SAS4A metallic fuel models are illustrated through an extended SAS4A analysis of a postulated unprotected loss-of-flow and transient-overpower accident in the metallic fuel prototype Gen-IV sodium fast reactor. The results show that the maximum relative power reached during the postulated accident is 1.19 P 0 . The favorable characteristics of the metallic fuel cause a significant decrease in net reactivity and relative power due to prefailure in-pin fuel relocation. Furthermore, negative net reactivity values persist after cladding failure, and the postfailure fuel relocation events occur at low and decreasing power levels.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Assessment of Regulatory Technology Gaps for Advanced Small Modular Sodium Fast Reactors

As part of an ongoing Regulatory Technology Development Plan (RTDP) for advanced small modular reactors, a review of the Sandia National Laboratories (SNL) report Sodium Fast Reactor Safety and Licensing Research Plan was conducted to identify technology-related barriers that would impact the licensing of a small sodium fast reactor (SFR), which could be resolved with technology development performed in conjunction with key additional regulatory interactions. While the general conclusions of the SNL report are still valid for small SFRs, a subsequent evaluation of the report recognized several areas that would benefit from regulatory clarification of requirements, as described in the following table.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can be accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized. Results will be presented showing the potential for thorium to reach a high TRU transmutation rate over a wide variety of fuel types (oxide, metal, nitride and carbide) and transmutation schemes (recycle or partition of in-bred U-233). In addition, a sustainable scheme has been devised to burn the TRU accumulated in the core inventory once the legacy TRU supply has been exhausted, thereby achieving long-term virtually TRU-free. A comprehensive 'back-to-front' approach to the fuel cycle has recently been proposed by Westinghouse which emphasizes achieving 'acceptable', low-radiotoxicity, high-level waste, with the intent not only to satisfy all technical constraints but also to improve public acceptance of nuclear energy. Following this approach, the thorium fuel cycle, due to its low radiotoxicity and high potential for TRU transmutation has been selected as a promising solution. Additional studies not shown here have shown significant reduction of decay heat. The TRU burning potential of the Th-based fuel cycle has been illustrated with a variety of fuel types, using the Toshiba ARR to perform the analysis, including scenarios with continued LWR operation of either uranium fueled or thorium fueled LWRs. These scenarios will afford overall reduction in actinide radiotoxicity, however when the TRU supply is exhausted, a continued U- 235 LWR operation must be assumed to provide TRU makeup feed. This scenario will never reach the characteristically low TRU content of a closed thorium fuel cycle with its associated potential benefits on waste radiotoxicity, as exemplified by the transition scenario studied. At present, the cases studied indicate ThC as a potential fuel for maximizing TRU burning, while ThN with nitrogen enriched to 95% N-15 shows the highest breeding potential. As a result, a transition scenario with ThN was developed to show that a sustainable, closed Th-cycle can be achieved starting from burning the legacy TRU stock and completing the transmutation of the residual TRU remaining in the core inventory after the legacy TRU external supply has been exhausted. The radiotoxicity of the actinide waste during the various phases has been characterized, showing the beneficial effect of the decreasing content of TRU in the recycled fuel as the transition to a closed Th-based fuel cycle is undertaken. Due to the back-to-front nature of the proposed methodology, detailed designs are not the first step taken when assessing a fuel cycle scenario potential. As a result, design refinement is still required and should be expected in future studies. Moreover, significant safety assessment, including determination of associated reactivity coefficients, fuel and reprocessing feasibility studies and economic assessments will still be needed for a more comprehensive and meaningful comparison against other potential solutions. With the above considerations in mind, the potential advantages of thorium fuelled reactors on HLW management optimization lead us to believe that thorium fuelled reactor systems can play a significant role in the future and deserve further consideration. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of a Thermochemical Database for Sodium Fast Reactor Mechanistic Source Term Calculations

An accurate assessment of the radionuclide source term for sodium fast reactors requires a predictive understanding of the chemical speciation and phase transformations of key dose contributing radionuclides, including those released to the sodium pool and cover gas region. Developing models for chemical speciation and phase equilibria requires the compilation of thermodynamic constants and experimentally determined functions that account for non-ideal solution behavior. For the present study, a thermochemical database, including parameters accounting for non-ideal chemical mixing, was compiled and used for a set of mechanistic source term trial calculations for a metal-fueled, pool-type sodium fast reactor [1].

Jerden, James↗
DOE OSTI2011

Upgrade of the Resonance Ionization Mass Spectrometer for Precise Identification of Failed Fuel in a Fast Reactor

Isotopic analysis of krypton (Kr) and xenon (Xe) by resonance ionization mass spectrometry (RIMS) is an effective tool for identification of failed fuel in fast reactors to achieve their safety operation and high plant availability. Reliability of the failed fuel detection and location (FFDL) system depends on the precise determination of {sup 78}Kr/{sup 80}Kr, {sup 82}Kr/{sup 80}Kr and {sup 126}Xe/{sup 129}Xe isotopic ratios, which is mainly hampered by statistical errors for detection of the corresponding isotopes except {sup 82}Kr generated in large amounts during operation of fast reactors. In this paper, we report on improvements of the laser optical system of our spectrometer to increase the resonance ionization efficiency of Kr and Xe atoms, focusing on (i) utilization of the uniform YAG laser beam to improve the wavelength conversion efficiency of sum frequency generation and (ii) reflection of the ultraviolet light by a concave mirror to increase the photon density. The results indicate that our upgraded resonance ionization mass spectrometer has enough performance for isotopic analysis of Kr and Xe required in the Monju FFDL system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

High Temperature Ultrasonic Transducers for In-Service Inspection of Liquid Metal Fast Reactors

In-service inspection of liquid metal (sodium) fast reactors requires the use of ultrasonic transducers capable of operating at high temperatures (>200°C), high gamma radiation fields, and the chemically reactive liquid sodium environment. In the early- to mid-1970s, the U.S. Atomic Energy Commission supported development of high-temperature, submersible single-element transducers, used for scanning and under-sodium imaging in the Fast Flux Test Facility and the Clinch River Breeder Reactor. Current work is building on this technology to develop the next generation of high-temperature linear ultrasonic transducer arrays for under-sodium viewing and in-service inspections.

Fast Reactors↗
DOE OSTI2013

An improved resonance self-shielding method for heterogeneous fast reactor assembly and core calculations

An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗