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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

SAS4A/SASSYS-1 Verification Testing for Sodium Fast Reactor Applications (Acceptance Testing Report)

SAS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal- cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over fifty years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. To support demonstration of software pedigree and confirm key functional requirements, this report has been generated to provide detailed verification of the software. Although SAS was developed to support the analysis of any liquid-metal-cooled nuclear reactor, the acceptance testing described in this document focuses on the verification of SAS capabilities as they relate to pool-type Sodium Fast Reactors (SFRs). This report includes documentation of the test problem definition, analytical solution(s), computational solution(s), acceptance criteria, comparisons of analytical/computational solutions, and determination of acceptance of the computational solution(s). Deviations from acceptance criteria are noted.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Leachate Treatment for Radioactive Waste Containment Mound: Pilot Test and Design - 19098

Canadian Nuclear Laboratories plans to construct an engineered containment mound (ECM) at its Chalk River Laboratories site in Ontario for safe management of radioactive and mixed waste at the facility. The ECM is designed for containment of 1,000,000 cubic meters of waste, and will be operated over a period of 50 years. Contact storm water, leachate, and other wastewater will be generated during operation of the ECM, and must be treated prior to discharge to the environment. The wastewater is projected to contain low concentrations of radionuclides and metals, as well as other cations and anions, and low concentrations of organic chemicals. Through comparison with surface water discharge criteria, radionuclides cesium-137 and strontium-90, and metals including aluminum, barium, copper, iron, lead, and manganese were projected to be present in the wastewater at concentrations that may exceed surface water discharge criteria, and were therefore targeted for treatment. Several technologies, including chemical precipitation, membrane filtration, ion exchange, and reverse osmosis were reviewed and tested on a laboratory scale using simulated wastewater prepared to replicate the concentrations of both target and non-target constituents. Based upon the results of the laboratory-scale tests, a six-week pilot test was conducted to simulate the performance of the technologies under continuous-flow conditions and determine full-scale design criteria. This paper describes the methodology and results of the six-week pilot test, and how the test results were used to inform the full-scale wastewater treatment system design. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2011

Fission Power System Technology for NASA Exploration Missions

Under the NASA Exploration Technology Development Program, and in partnership with the Department of Energy (DOE), NASA is conducting a project to mature Fission Power System (FPS) technology. A primary project goal is to develop viable system options to support future NASA mission needs for nuclear power. The main FPS project objectives are as follows: 1) Develop FPS concepts that meet expected NASA mission power requirements at reasonable cost with added benefits over other options. 2) Establish a hardware-based technical foundation for FPS design concepts and reduce overall development risk. 3) Reduce the cost uncertainties for FPS and establish greater credibility for flight system cost estimates. 4) Generate the key products to allow NASA decisionmakers to consider FPS as a preferred option for flight development. In order to achieve these goals, the FPS project has two main thrusts: concept definition and risk reduction. Under concept definition, NASA and DOE are performing trade studies, defining requirements, developing analytical tools, and formulating system concepts. A typical FPS consists of the reactor, shield, power conversion, heat rejection, and power management and distribution (PMAD). Studies are performed to identify the desired design parameters for each subsystem that allow the system to meet the requirements with reasonable cost and development risk. Risk reduction provides the means to evaluate technologies in a laboratory test environment. Non-nuclear hardware prototypes are built and tested to verify performance expectations, gain operating experience, and resolve design uncertainties.

Mason, Lee↗
DOE OSTI2019

Length Dependence of Severe Accident Test Station Integral Testing

The safety of accident-tolerant fuel (ATF) candidates must be tested under simulated loss-of-coolant accident (LOCA) conditions to evaluate their improved safety margins. This evaluation eventually must include testing of irradiated materials. This report describes the fiscal year (FY) 2019 activities to (1) improve the information yield of integral design–basis LOCA tests by using digital image correlation (DIC) calculations on the engraved specimens’ surfaces and (2) to investigate the effects of reducing the length of nuclear-grade FeCrAl specimens tested in the Severe Accident Test Station (SATS) to improve the costeffectiveness of irradiation campaigns for future work. Analysis showed the DIC technique to be effective at calculating strains, and that specimens as short as 4 inches could be used for LOCA experiments while maintaining similar balloon and burst behavior as 12-inch specimens. With these results, an irradiation vehicle design effort was initiated to encapsulate multiple 4-inch thin-walled tube cladding specimens and test them under irradiation in the flux trap of the High Flux Isotope Reactor. By utilizing thermal modeling and optimizing the design parameters, calculated specimen temperatures were predicted to vary less than 2% during irradiation. Future work is planned to finalize the design and safety basis approvals and initiate a HFIR irradiation in support of in-cell integral LOCA tests at the Irradiated Fuels Examination Laboratory (IFEL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Integral LOCA fragmentation test on high-burnup fuel

Increasing the fuel burnup limit in light water reactors is sought to enhance fuel cycle economics and requires establishing a technical basis. Experimental observations of severe fuel fragmentation under loss-of-coolant-accident conditions at Halden and Studsvik had raised the need for additional considerations during the development of this technical basis. These test data suggested that the burnup threshold for high burnup fuel fragmentation may be influenced by pre-transient power. Additional loss-of-coolant-accident test data is therefore valuable to complement these tests and enhance the current state of understating. Oak Ridge National Lab has developed the Severe Accident Test Station capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident and beyond design basis accident scenarios. Severe Accident Test Station provides various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident testing, defueled oxidation quench testing, and high-temperature beyond design basis accident testing. Severe Accident Test Station has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory at Oak Ridge National Lab. Furthermore, descriptions of the in-cell re-fabrication capabilities and assembly of the loss-of-coolant-accident test train are provided. Installation of the Severe Accident Test Station system and in-cell re-fabrication restores United States capability to examine postulated and extended loss-of-coolant-accident conditions on spent fuel and cladding and provides a platform for evaluating advanced fuel and accident-tolerant fuel cladding concepts. Lastly, three in-cell integral loss-of-coolant-accident test were performed in the Severe Accident Test Station and compared to the Nuclear Regulatory Commission sponsored loss-of-coolant-accident test as well as the original loss-of-coolant-accident test performed at Argonne National Lab. Finally, the results of these tests as well as all publicly available integral loss-of-coolant-accident test were used to validate a threshold for high burnup fuel fragmentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Initial Findings in Qualifying the Instrument Thimble 11 Position in the Advanced Test Reactor Critical

The Advanced Test Reactor (ATR) is a versatile nuclear research reactor located at the Idaho National Laboratory (INL) in Eastern Idaho, United States of America. It is one of the most powerful and flexible research reactors in the world and is primarily used for materials testing, isotope production and basic nuclear science research. The ATR is a light-water-cooled, beryllium-moderated reactor with a nominal thermal power capacity of 250 MW utilizing specially designed fuel, arranged in a serpentine pattern to create 9 flux traps. The special design of ATR allows for the neutron flux within each of the flux traps providing flexibility for experiments, programs and allowing experiments with different dose requirements to be irradiated simultaneously. The Advance Test Reactor Critical (ATRC) is a full-scale replica of the ATR core but is housed in a pool instead of a pressure vessel. ATRC is usually operated at less than 600 watts [1]. There are 12 dry instrument thimbles located around the outside of the core tank where a variety of measurement equipment is housed to facilitate reactor operations. Currently Instrument Thimble 11 (IT-11) is not utilized in ATR or ATRC. A range of measurements is ongoing to qualify IT-11 as a test platform for nuclear instrumentation research, development and further experimentation.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2019

Compositions and methods for monitoring actinides

Compositions and methods for monitoring the quantity of actinides present in a test sample are disclosed. Compositions and methods for monitoring the motion of special nuclear materials through space are also described. Compositions and methods for monitoring the quantity of a fissile special nuclear material present in a test sample are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 30-year cool down are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 180 day cool down are also disclosed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Exploring nuclear structure with multiparticle azimuthal correlations at the LHC

Details of the nuclear structure of Xe, such as the quadrupole deformation and the nuclear diffuseness, are studied by extensive measurements of anisotropic-flow-related observables in Xe–Xe collisions at a centre-of-mass energy per nucleon pair $\sqrt{s_{NN}}$ = 5.44 TeV with the ALICE detector at the LHC. The results are compared with those from Pb–Pb collisions at $\sqrt{s_{NN}}$ = 5.02 TeV for a baseline, given that the 208 Pb nucleus exhibits a very weak deformation. Furthermore, comprehensive comparisons are performed with a state-of-the-art hybrid model using IP-Glasma+MUSIC+UrQMD. It is found that among various IP-Glasma+MUSIC+UrQMD calculations with different values of nuclear parameters, the one using a nuclear diffuseness parameter of α 0 = 0.492 and a nuclear quadrupole deformation parameter of β 2 = 0.207 provides a better description of the presented flow measurements. These studies represent the first systematic exploration of nuclear structure at TeV energies, utilizing a comprehensive set of anisotropic flow observables. The measurements serve as a critical experimental benchmark for rigorously testing the interplay between nuclear structure inputs and heavy-ion theoretical models.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Comparison of Extended and Accelerated VOG Tests

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas streams of a processing plant, including ¹²⁹I. The dissolver off-gas and more dilute vessel off-gas streams (VOG) must both be targeted when mitigating ¹²⁹I environmental release because each contains some amount of iodine. Iodine-129 will likely be found as elemental iodine (I₂) and methyl iodide (CH₃I) in these off-gas streams. Reduced silver-exchanged mordenite (AgZ or Ag⁰Z) has been investigated as a potential sorbent for iodine abatement and is studied here under prototypic VOG conditions. Because of the relatively low iodine abundance and high flow rates of the VOG, total iodine concentrations are expected to be in the parts per billion range. Thus, VOG experiments need to run for extended durations at low concentrations to reach sorbent saturation. Because the sorbent will be exposed to oxidizing gas streams for extended periods of time, aging effects and sorbent degradation need to be considered when designing a sorbent-based abatement system. Three sets of experiments were performed with the aim of determining how the adsorption of iodine by AgZ is affected by differing test durations and gas compositions. The first tested the capacity of sorbent aged for 8 months under a humid air stream. The second tested differences in sorbent capacity and mass transfer zone (MTZ) length during high-concentration (1200 ppb v ) CH₃I loading in a humid nitrogen gas (N₂) stream and a humid air gas stream over 28 days. The third tested sorbent capacity and MTZ length during low concentration (< 200 ppb v ) CH₃I and I₂ loading in a humid air stream over 9 months. The results of these tests suggest that AgZ capacity drops by ~50% after 1 month of aging, by ~60% after 2 months of aging, and then does not continue to decrease significantly at up to 8 months of aging. One-month aging tests that compared N₂ and air as the gas stream diluent resulted in similar maximum loading capacities and overall loading curves, indicating that the effects of aging cannot be mitigated by avoiding air as the balance gas. The 9-month extended VOG tests did not result in clear sorbent saturation at the inlet, but it can be inferred using an assumption of a maximum capacity of 45 mg I/g AgZ. Applying this assumption, then the MTZ of CH₃I is 12.8 cm, and the MTZ of I₂ is 12.4 cm. These results are similar to previous tests. Comparisons to tests completed at Idaho National Laboratory suggest that the CH₃I MTZ may be dependent on concentration. Future work should re-evaluate the design of the VOG iodine capture system with this updated data and should seek to understand fundamental characteristics of CH₃I adsorption by AgZ, such as the effect of concentration, gas velocity, and gas composition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

An Architectural Survey of the Test Cell C Historic District, Area 25, Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) has authorized an updated recording and evaluation of Test Cell C. This facility is part of the former Nuclear Rocket Development Station (NRDS) in Area 25 of Nevada National Security Site (NNSS) in Nye County, Nevada (formerly the Nevada Test Site [NTS]). Future alterations to the facility are anticipated to meet current and future National Weapons Science, Global and Homeland Security Programs, and Environmental Management mission requirements. Therefore, this survey was done to comply with requirements of Section 106 of the National Historic Preservation Act and its implementing regulations, 36 C.F.R. Part 800. An architectural survey of 35.75 acres was completed by Desert Research Institute (DRI). This area also represents the potential Test Cell C Historic District and is identified by NNSA/NFO as the Area of Potential Effect (APE) within which the undertaking may directly cause alterations to the character of historic properties. One resource, Building 25-3210 (SHPO Resource No. B2444), within this area was previously surveyed by DRI. It was found eligible for inclusion in the National Register of Historic Places (NRHP or National Register) under Criteria A and C, and Consideration G. Mitigation of this building was completed prior to its demolition. The APE for reasonably foreseeable effects that may indirectly cause alterations later in time, be farther removed in distance, or be cumulative is the unrecorded NRDS historic district. The present survey updates the previous survey to present standards and expands it to include the entire Test Cell C complex. This includes updated recording and evaluation of the remains of 25-3210 and its 32 previously unrecorded and unevaluated Accessory Resources. The entire complex was recorded as the Test Cell C Historic District (SHPO Resource No. D346), which is eligible for the National Register at the National Level under Criteria A, C, and D. All nine architectural resources in the district are recommended as contributing elements of the Test Cell C Historic District (Figure 2). Of these, four (25-3210 Resource No. B2444, 25-3220 SHPO Resource No. B18110, 25-3226 SHPO Resource No. B18111, and 25-3230-32 SHPO Resource No. B18114) are recommended as individually eligible to the National Register under Criteria A and C, and one (KIWI-TNT [Transient Nuclear Test] SHPO Resource No. S2287) is recommended as individually eligible under Criteria A, C, and D. The remaining four resources (25-3214 SHPO Resource No. B18109, 25-3228 SHPO Resource No. B18112, 25-3229 SHPO Resource No. B18113, and 25-3233 SHPO Resource No. 18115) are not recommended as individually eligible. In addition, the Test Cell C Historic District is regarded as a contributor to the potential national significance of the NRDS Historic District. In 2010, the NTS became the NNSS. Because of the nature of the discussion in the Historic Context below, “NTS” is used throughout that section, otherwise “NNSS” is used to refer to the site. The survey area is located in unplatted sections of T13S R51E Mount Diablo Baseline and Meridian (projected) in Area 25 of the NNSS. It is within the NRDS in Jackass Flats, which has long been recognized as a potential historic district. An oblique aerial photo of the Test Cell C complex is shown on the cover of this report.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Design and out-of-pile testing of a novel irradiation experiment vehicle to support qualification of nuclear thermal propulsion components

Fission power systems have demonstrated higher thermal energy density per unit mass than chemical systems. High mass specific energy density offers advantages in space thermal propulsion applications. In the 1960s, nuclear thermal propulsion (NTP) development included the ROVER/NERVA program, in which full-scale prototype reactors were tested on the ground using hydrogen as the reactor coolant and propellant. NTP systems are being revisited, with expectations of improved performance resulting from improved fuel materials, allowing propellant temperatures near 2,800 K and over 20 thermal propulsion cycles. An apparatus has been developed to test NTP fuel materials in a research reactor at temperature and in a radiation environment typical of the NTP application. Here, this paper presents the design constraints, preliminary scoping thermal analysis, and basic thermal qualification testing for the out-of-pile experiment set apparatus, or OUTSET.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2025

ORNL Testing of Multiple Graphite Benchmarks [Slides]

Integral criticality safety and reactor physics benchmark experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Reactor Physics Experiments (IRPhE Handbook) are essential for nuclear data validation and testing

ENDF↗
DOE OSTI2018

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operating in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report

The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach has not been to create testing conditions that are fully prototypic of a fast reactor, but to create a thermal profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of limited fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The available evidence supports the assertion that use of cadmium-filtering creates irradiation conditions in the ATR that are both relevant to, and suitable for, fast reactor fuel performance testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

FY19 Report for Instrumentation Development for the Transient Testing Program

This report details the technical activities and accomplishment carried out under funding from the Department of Energy (DOE) Nuclear Technology Research and Development (NTRD) program for in-pile instrumentation supporting the transient testing program in FY19. These activities were performed in support of cross-cutting transient testing experiment objectives. The purpose of this report is to provide a summary of key technical work, of particular interest to nuclear irradiation test experimenters and in-pile instrumentation engineers. During FY19, development activities are focused on deployment of devices to perform online measurement of neutron flux, temperature, and mechanical behaviors in nuclear fuels experiments. Specifically, these R&D activities include in-pile investigations at the Transient Reactor Test (TREAT) facility throughout the year. Other in-pile instrumentation R&D activities are being carried out under other DOE programs, which may be recognized but not the focus of this document. A brief summary of activities and accomplishments is first provided for each major activity. More detailed summaries are presented in appendices.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2010

Kennedy Space Center's Partnership with Graftel Incorporated

NASA Kennedy Space Center (KSC) has recently partnered with Graftel Incorporated under an exclusive license agreement for the manufacture and sale of the Smart Current Signature Sensor. The Smart Current Signature Sensor and software were designed and developed to be utilized on any application using solenoid valves. The system monitors the electrical and mechanical health of solenoids by comparing the electrical current profile of each solenoid actuation to a typical current profile and reporting deviation from its learned behavior. The objective of this partnership with Graftel is for them to develop the technology into a hand-held testing device for their customer base in the Nuclear Power Industry. The device will be used to perform diagnostic testing on electromechanical valves used in Nuclear Power plants. Initially, Graftel plans to have working units within the first year of license in order to show customers and allow them to put purchase requests into their next year's budget. The subject technology under discussion was commercialized by the Kennedy Space Center Technology Programs and Partnerships Office, which patented the technology and licensed it to Graftel, Inc., a company providing support, instrumentation, and calibration services to the nuclear community and private sector for over 10 years. For the nuclear power industry, Graftel designs, manufacturers, and calibrates a full line of testing instrumentation. Grafters smart sensors have been in use in the United States since 1993 and have proved to decrease set-up time and test durations. The project was funded by Non-Destructive Engineering, and it is felt that this technology will have more emphasis on future vehicles. Graftel plans to market the Current Signature Sensor to the Electric Utility industry. Graftel currently supplies product and services to the Nuclear Power Industry in the United States as well as internationally. Product and services sold are used in non-destructive testing for valves, penetrations and other applications. Graftel also supplies testing services to an industrial customer base. The customer base includes 90 percent of the U.S. Nuclear plants and plants in Brazil, Europe, and Asia. Graftel works internationally with two representative groups and employees and has ten people at the principle location and a group of contract engineers around the country.

Dunn, Carol Anne↗
DOE OSTI2018

Nondestructive Testing Research and Development Efforts at Argonne National Laboratory—An Overview

The Nuclear Engineering (NE) Division of Argonne National Laboratory (ANL) has been involved for many years with research and development efforts associated with the nondestructive testing/evaluation (NDT/NDE) of critical components and structures in power generation (nuclear and fossil plants), aerospace, defense, homeland security, medical, and scientific applications. A wide range of NDT methods including electromagnetic, thermal, radiographic, acoustic and ultrasonic methods have been used for a wide range of material evaluation applications. Argonne’s expertise in analytical and numerical modeling of thermal, acoustic and electromagnetic interaction with complex media as well as advanced signal processing and data analysis have been utilized extensively for NDT applications. Three active areas of research, all in connection with NDT of materials and systems utilized in nuclear energy applications, are discussed in detail in this paper.

Eddy current testing↗