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The current phase of the SP-100 program was initiated in 1986 with the objectives of developing the technology for Space Reactor Power Systems (SRPS) in the 10-1000k. We power range and perfoming ground system tests of both nuclear and non-nuclear major assemblies.
This Technical Manual describes a parametric study of general purpose heat source (GPHS) powered Stirling converter configurations. This study was performed in support of MSFC s efforts to establish the capability to perform non-nuclear system level testing and integration of radioisotope power systems. Six different GPHS stack configurations at a total of three different power levels (80, 250, and 500 W(sub e) were analyzed. The thermal profiles of the integrated GPHS modules (for each configuration) were calculated to determine maximum temperatures for comparison to allowable material limits. Temperature profiles for off-nominal power conditions were also assessed in order to better understand how power demands from the Stirling engine impact the performance of a given configuration.
Under the Exploration Technology Development Program, the National Aeronautics and Space Administration (NASA) and the Department of Energy (DOE) are jointly developing Fission Surface Power (FSP) technology for possible use in human missions to the Moon and Mars. A preliminary reference concept was generated to guide FSP technology development. The concept consists of a liquid-metal-cooled reactor, Stirling power conversion, and water heat rejection, with Brayton power conversion as a backup option. The FSP project has begun risk reduction activities on some key components with the eventual goal of conducting an end-to-end, non-nuclear, integrated system test. Several power conversion and heat rejection hardware prototypes have been built and tested. These include multi-kilowatt Stirling and Brayton power conversion units, titanium-water heat pipes, and composite radiator panels.
Fusion-based nuclear propulsion has the potential to enable fast interplanetary transportation. Due to the great distances between the planets of our solar system and the harmful radiation environment of interplanetary space, high specific impulse (Isp) propulsion in vehicles with high payload mass fractions must be developed to provide practical and safe vehicles for human spaceflight missions. The Z-Pinch dense plasma focus method is a Magneto-Inertial Fusion (MIF) approach that may potentially lead to a small, low cost fusion reactor/engine assembly1. Recent advancements in experimental and theoretical understanding of this concept suggest favorable scaling of fusion power output yield 2. The magnetic field resulting from the large current compresses the plasma to fusion conditions, and this process can be pulsed over short timescales (10(exp -6 sec). This type of plasma formation is widely used in the field of Nuclear Weapons Effects testing in the defense industry, as well as in fusion energy research. A Decade Module 2 (DM2), approx.500 KJ pulsed-power is coming to the RSA Aerophysics Lab managed by UAHuntsville in January, 2012. A Z-Pinch propulsion concept was designed for a vehicle based on a previous fusion vehicle study called "Human Outer Planet Exploration" (HOPE), which used Magnetized Target Fusion (MTF) 3 propulsion. The reference mission is the transport of crew and cargo to Mars and back, with a reusable vehicle.
NASA Glenn Research Center (GRC) is developing fission power system technology for future Lunar surface power applications. The systems are envisioned in the 10 to 100kW(sub e) range and have an anticipated design life of 8 to 15 years with no maintenance. NASA GRC is currently setting up a 55 kW(sub e) non-nuclear system ground test in thermal-vacuum to validate technologies required to transfer reactor heat, convert the heat into electricity, reject waste heat, process the electrical output, and demonstrate overall system performance. Reducing the radiator mass, size, and cost is essential to the success of the program. To meet these goals, Advanced Cooling Technologies, Inc. (ACT) and Vanguard Space Technologies, Inc. (VST) are developing a single facesheet radiator with heat pipes directly bonded to the facesheet. The facesheet material is a graphite fiber reinforced composite (GFRC) and the heat pipes are titanium/water. By directly bonding a single facesheet to the heat pipes, several heavy and expensive components can be eliminated from the traditional radiator design such as, POC(TradeMark) foam saddles, aluminum honeycomb, and a second facesheet. A two-heat pipe radiator prototype, based on the single facesheet direct-bond concept, was fabricated and tested to verify the ability of the direct-bond joint to withstand coefficient of thermal expansion (CTE) induced stresses during thermal cycling. The thermal gradients along the bonds were measured before and after thermal cycle tests to determine if the performance degraded. Overall, the results indicated that the initial uniformity of the adhesive was poor along one of the heat pipes. However, both direct bond joints showed no measureable amount of degradation after being thermally cycled at both moderate and aggressive conditions.
NASA Glenn Research Center (GRC) is developing fission power system technology for future Lunar and Martian surface power applications. The systems are envisioned in the 10 to 100kWe range and have an anticipated design life of 8 to 15 years with no maintenance. NASA GRC is currently setting up a 55 kWe non-nuclear system ground test in thermal-vacuum to validate technologies required to transfer reactor heat, convert the heat into electricity, reject waste heat, process the electrical output, and demonstrate overall system performance. The paper reports on the development of the heat pipe radiator to reject the waste heat from the Stirling convertors. Reducing the radiator mass, size, and cost is essential to the success of the program. To meet these goals, Advanced Cooling Technologies, Inc. (ACT) and Vanguard Space Technologies, Inc. (VST) are developing a single facesheet radiator with heat pipes directly bonded to the facesheet. The facesheet material is a graphite fiber reinforced composite (GFRC) and the heat pipes are titanium/water Variable Conductance Heat Pipes (VCHPs). By directly bonding a single facesheet to the heat pipes, several heavy and expensive components can be eliminated from the traditional radiator design such as, POCO"TM" foam saddles, aluminum honeycomb, and a second facesheet. As mentioned in previous papers by the authors, the final design of the waste heat radiator is described as being modular with independent GFRC panels for each heat pipe. The present paper reports on test results for a single radiator module as well as a radiator cluster consisting of eight integral modules. These tests were carried out in both ambient and vacuum conditions. While the vacuum testing of the single radiator module was performed in the ACT's vacuum chamber, the vacuum testing of the eight heat pipe radiator cluster took place in NASA GRC's vacuum chamber to accommodate the larger size of the cluster. The results for both articles show good agreement with the predictions and are presented in the paper.
We observed the Oort cloud comet C/2014 Q2 (Lovejoy) on 2015 January 31 and February 1 and 2 at a heliocentric distance of 1.3 au and geocentric distance of 0.8 au during its approach to the Sun. Comet Lovejoy was observed with GIANO, the near-infrared high-resolution spectrograph mounted at the Nasmyth-A focus of the TNG (Telescopio Nazionale Galileo) telescope in La Palma, Canary Islands, Spain. We detected strong emissions of radical CN and water, along with many emission features of unidentified origin, across the 1-2.5 micron region. Spectral lines from eight ro-vibrational bands of H2O were detected, six of them for the first time. We quantified the water production rate [Q(H2O), (3.11+/- 0.14) x 10(exp 29)/s] by comparing the calibrated line fluxes with the Goddard full non-resonance cascade fluorescence model for H2O. The production rates of ortho-water [Q(H2O)ORTHO, (2.33+/- 0.11) x 10(exp 29)/s] and para-water [Q(H2O)PARA, (0.87+/-0.21) x 10(exp 29)/s] provide a measure of the ortho-to-para ratio (2.70+/- 0.76)). The confidence limits are not small enough to provide a critical test of the nuclear spin temperature.
Nondestructive evaluation (NDE), alternately known as nondestructive examination (NDE), (commonly used within the nuclear industry), nondestructive testing (NDT) (commonly used in the aerospace and petrochemical industry), and nondestructive inspection (NDI) (commonly used within the military), are as the name implies test and assessment technologies designed to detect and characterize components without damaging or compromising the component’s ability to perform its design function. The most common application is detection of flaws caused either by manufacturing anomalies, service or environmental stresses, or natural material aging. More generally, applications may include estimation of mechanical and material properties, stress/strain, and dynamic response behavior. Post-manufacturing inspections frequently enjoy unfettered access to the component or part that may allow more thorough examinations, while in-service examinations after components have been in use may impose severe access limitations for in-situ testing. If flaws are detected, they must then be evaluated relative to their size, location, growth rate, growth mechanism, and likelihood that they are or will become a liability to the component’s function or performance. NDE is comprised of a large family of specific test disciplines including visual inspection, dimensional metrology, ultrasound, radiography, penetrant test, magnetic particle test, leak test, eddy current tests, potential drop tests, flash and vibro thermography, shearography, acoustic emissions, and many other methods. This section provides discussion on general NDE science and considerations for specific technique selection.
This document presents the test plan for Task 2 of the Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan. This task’s overarching goal is the resolution of gas generation resulting from the radiolytic degradation of the oxide layers on aluminum. The uncertainty associated with the build-up of gaseous radiolytic products inside the ASNF storage canisters is one of the dominant technical challenges associated with long term dry storage of aluminum clad fuel. The objectives of the Task 2 test plan are to: 1. Understand how radiation chemistry inside the ASNF dry storage canister impacts extended dry storage performance. 2. Understand the radiation induced mechanisms governing the production of radiolytic gases and alteration of gas composition within a canister. 3. Correlate the observed changes in material properties of ASNF with the radiolytic degradation of aluminum oxide in fill environments of variable moisture and gas composition. 4. Compile a predictive kinetic modeling tool, to provide quantitative mechanistic insight into the in-cask radiation chemistry of ASNF. 5. Interface with the canister modeling activity of Task 3 to deliver degradation rates for package performance of both sealed and vented scenarios.
The Advanced Test Reactor (ATR) creates a unique high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector eventually requiring a replacement during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data to model and predict criticality scenarios for the new unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult to find data trends that could be useful for the modeling and prediction process. So nuclear requalification testing is conducted using different flux and irradiation tests to accurately verify the models and to recalibrate nuclear monitoring instrumentation in the ATR core allowing for safe re-start of normal operations.
High-consequence nuclear facilities are required to implement a physical protection system (PPS) capable of preventing a design basis threat (DBT) adversary from completing radiological sabotage or theft of nuclear material. Response forces are a core component of the PPS. A rigorous testing and evaluation program is required to ensure that the response forces are appropriately trained and equipped and are capable of meeting the regulatory and mission requirements.
The Source Physics Experiment (SPE) series is a long-term NNSA research and development effort designed to improve U.S. arms control and nuclear nonproliferation verification and monitoring capabilities. The findings from the SPE will advance the United States’ nuclear explosion monitoring capabilities, particularly with respect to detection, discrimination and determination of yields associated with small nuclear explosions that can be lost amid the noisy seismo-acoustic background from other sources. The data generated from the SPE, a series of well-designed and recorded chemical explosions, will contribute to the development and validation of first-principles explosive source generated seismo-acoustic modeling codes. These codes will then facilitate the update of semi-empirical methods, currently based on historic test site data, such that key explosion observables can be reproduced, thus improving confidence in nuclear test monitoring in new areas and/or under novel emplacement conditions. The overall SPE project is comprised of both the development of the new explosion simulation codes and the chemical explosion test series. The chemical explosion test series will generate the empirical data required to both develop and validate the new simulation codes.
The Source Physics Experiment (SPE) series is a long-term NNSA research and development effort designed to improve U.S. arms control and nuclear nonproliferation verification and monitoring capabilities. The findings from the SPE will advance the United States’ nuclear explosion monitoring capabilities, particularly with respect to detection, discrimination and determination of yields associated with small nuclear explosions that can be lost amid the noisy seismo-acoustic background from other sources. The data generated from the SPE, a series of well-designed and recorded chemical explosions, will contribute to the development and validation of first-principles explosive source generated seismo-acoustic modeling codes. These codes will then facilitate the update of semi-empirical methods, currently based on historic test site data, such that key explosion observables can be reproduced, thus improving confidence in nuclear test monitoring in new areas and/or under novel emplacement conditions. The overall SPE project is comprised of both the development of the new explosion simulation codes and the chemical explosion test series. The chemical explosion test series will generate the empirical data required to both develop and validate the new simulation codes.
New and improved materials are being considered for supporting both existing and next-generation nuclear reactors. Reactor materials can significantly degrade with time, thus limiting or altering their properties in harsh reactor environments. To accurately understand such material degradation, real-time data obtained under prototypic irradiation conditions are required. In particular, understanding the creep behavior of materials exposed to irradiation and elevated temperatures is essential for safety concern evaluations. To provide these capabilities, Idaho National Laboratory (INL)’s High Temperature Test Laboratory (HTTL) developed several instrumented test rigs for obtaining real-time data from specimens in well-controlled pressurized-water reactor (PWR) coolant conditions at the Materials Test Reactor. This technical report focuses on INL’s efforts to evaluate and enhance the former creep test rig prototype that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig can detect changes in the length of creep specimens, which is useful for measuring thermal expansion and creep deformation.
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The H-Canyon facility is currently dissolving spent nuclear fuel, including Material Test Reactor (MTR) and High Flux Isotope Reactor (HFIR) fuel. Dissolution of aluminum spent nuclear fuel produces hydrogen and other NOx gases. A theory long held by H-Canyon Engineering and Facility Technical Advisors that the silver nitrate-coated berl saddles, present in the iodine reactor, used for off-gas treatment were acting as a catalytic hydrogen recombiner was captured during an external review of the Accelerated Basin Deinventory program. Results from a separate SRNL project using a Raman Spectrometer to monitor the offgas from the dissolution process indicated a lack of the expected hydrogen in the off-gas stream. To test this hypothesis, a laboratory scale iodine reactor was assembled and filled with silver nitrate-coated berl saddles. Testing with this laboratory scale reactor confirmed the recombination of hydrogen when a simulated dissolver off-gas was passed through the reactor containing silver nitrate-coated berl saddles at 173-188 °C. Control experiments performed with uncoated berl saddles resulted in no change to the hydrogen concentration after passing through the reactor. The residence time of the gas in the reactor was varied to determine if residence time had an impact on the amount of hydrogen recombination occurring. Results from these experiments indicated that at the shortest residence times tested (~5 seconds) recombination of the hydrogen still exceeded 90%; however, the percent recombination did increase at longer residence times, reaching 97.9% or greater for residence times over 10 seconds. Finally, testing performed with a simplified off-gas composition containing only hydrogen and air gave similar results, indicating that the presence of NO and N 2 O gases in the stream do not play a role in the recombination reaction.
The purpose of the meeting was to review the work performed in fiscal year 1992 in the areas of nuclear thermal and nuclear electric propulsion technology development. These proceedings are an accumulation of the presentations provided at the meeting along with annotations provided by authors. The proceedings cover system concepts, technology development, and system modeling for nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP). The test facilities required for the development of the nuclear propulsion systems are also discussed.
In this work, we tested the degradation of two polymeric materials under conditions routinely encountered in nuclear applications. Polymers in the form of plastic bags are commonly used along with metal containers to store and transport nuclear materials. The role of the bag, also known as bag-out bag, is to prevent contamination in case the metal containers are breached. Polymeric materials of interest are sPVC and aromatic polyether urethane (APU-ether), which in this study were exposed to alpha, gamma and thermal conditions. These conditions were specifically chosen to simulate the harsh environments normally encountered in the storage of nuclear materials. The physical, chemical and mechanical properties of these materials were investigated by a series of characterization methods before and after exposures. Our results highlight the superior stability of APU-ether under thermal aging and ionizing radiation and provide new insights into the durability of polymeric materials used in nuclear applications. In conclusion, these findings will serve as the foundation for a rational selection of new bag-out bag materials to be considered in the nuclear industry.