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DOE OSTI2021

Proton Nuclear Magnetic Resonance (H NMR) of Glycolate in Real Waste: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

In preparation for implementing the NG acid flowsheet for the SRS LWS, analytical methods for determining glycolate at low concentration are desired to support system management and safety. Previous work documented that the IC method performed well for samples with low to moderate ionic strength. That work also included a scoping effort to determine if an alternate analytical strategy using H NMR would complement and extend the capabilities of the IC method. The H NMR scoping results indicated that the method had the potential to expand glycolate analysis in LWS samples to higher ionic strength tanks/solutions. The scope of this work is to develop innovative proton NMR techniques, including demonstrating ion exchange decontamination protocols needed for application of the technique to real waste samples. Part of radioactive waste processing at SRS uses formic acid to reduce oxidized (Hg 2+ ) to more volatile elemental Hg for steam striping, collecting, and disposal. Under acidic conditions found in the Chemical Processing Cell (CPC) at the Defense Waste Processing Facility (DWPF), formic acid has a much higher hydrogen generation rate than an alternative reductant, glycolic acid. Thus, an NG acid flowsheet has been developed utilizing glycolic acid with the benefit of easing the need for headspace monitoring requirements for hydrogen and ammonia at DWPF. Low concentrations of glycolate are conservatively assumed to be in the recycle stream, which will collect in the Recycle Collection Tank (RCT). The DWPF recycle stream collected in the RCT has a distinct pathway to the LWS waste tanks that feed the 2H and 3H Evaporator. This route involves transfer of the DWPF recycle to Tank 22 in the Concentration, Storage, and Transfer Facilities (CSTF) followed by transfer to the LWS tank farm/evaporators.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

In the wake of the Fukushima Daiichi nuclear power plant accident in 2011, the accident tolerant fuels (ATF) program was initiated to enhance the safety of light-water reactor fuels, placing significant emphasis on cladding. A crucial step for the broad implementation of ATF in commercial reactors involves irradiation testing of the fuel designs. ATF-2D, the latest experiment in the ATF series, is slated to undergo irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). ATF-2D is a joint effort of the INL with industry partners General Electric Global Research; Framatome; General Atomics; the Japan Atomic Energy Agency; Hitachi-GE Nuclear Energy, Ltd; Global Nuclear Fuel-Japan Co., Ltd; Nippon Nuclear Fuel Development Co., Ltd; and Mitsubishi Heavy Industries, Ltd. The ATF-2D test train design consists of four tiers, each housing six rodlets. The device will be inserted in Loop 2A within the central flux trap of the ATR, and is anticipated to undergo irradiation throughout three 60-day cycles. Typical pressurized water reactor conditions will be emulated during the irradiation. The objective of the work presented here is to dimension neutron-absorbing hafnium (Hf) components surrounding the fuel rodlets, such that the axial power profile is flattened, while simultaneously ensuring that the total fission power output of the entire test train remains below 200 kW.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Achieving reasonable confidence in the validation of control room designs and modifications: A summary of the 2015 NEA experts' workshop - 124

Defining and achieving reasonable confidence in the validation of nuclear power plant control room designs is a complex challenge for many reasons; not the least of which is that validation is a test of strength rather than a test of truth. In 2015, the Nuclear Energy Agency, Committee for Safety of Nuclear Installations' Working Group on Human and Organizational Factors (WGHOF) hosted an experts' workshop to identify and critically examine potential means to address the challenge of achieving reasonable confidence in control room design validations. The experts' principal task was to answer four challenge questions. (1) What are the critical considerations in defining validation objectives and how do these impact achieving reasonable confidence? (2) What methods, approaches, resources, or rationales might be used for deriving performance requirements, selecting measures, and establishing acceptance criteria so as to support reasonable confidence? (3) What methods might be used to develop scenarios that maximize the amount and relevance of information in support of the validation conclusions and achieving reasonable confidence? (4) How should the validation results be aggregated and analyzed to determine the final validation conclusions? This paper provides an overview of the experts' responses to these challenges and briefly touches upon their implications for future directions. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

AMORE Mo-99 Spike Test Results

With funding from the National Nuclear Security Administrations Material Management and Minimization Office, Argonne National Laboratory (Argonne) is providing technical assistance to help accelerate the U.S. production of Mo-99 using a non-highly enriched uranium (non-HEU) source. A potential Mo-99 production pathway is by accelerator-initiated fissioning in a subcritical uranyl sulfate solution containing low enriched uranium (LEU). As part of the Argonne development effort, we are undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Production of Mo-99 and other fission products in the subcritical target solution is initiated by putting an electron beam on a depleted uranium (DU) target; the fast neutrons produced in the DU target are thermalized and lead to fissioning of U-235. At the end of irradiation, Mo is recovered from the target solution and separated from uranium and most of the fission products by using a titania column. The Mo is stripped from the column with an alkaline solution. After acidification of the Mo product solution from the recovery column, the Mo is concentrated (and further purified) in a second titania column. The strip solution from the concentration column is then purified with the LEU Modified Cintichem process. A full description of the process can be found elsewhere [1–3]. The initial commissioning steps for the AMORE project include performing a Mo-99 spike test with pH 1 sulfuric acid in the target vessel without a beam on the target to demonstrate the initial Mo separation-and-recovery process, followed by the concentration column process. All glovebox operations were tested with cold solutions prior to performing the Mo-99 spike tests. Two Mo-99 spike tests with pH 1 sulfuric acid have been performed to date. Figure 1 shows the flow diagram for the remotely operated Mo-recovery system for the AMORE project. There are two separate pumps and flow paths for the acid and base operations. The system contains three sample ladders with eight sample loops per ladder for target mixing; column loading, including acid and water washes; and column stripping, including the final water wash.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Scorpius and The Integrated Test Stand [Slides]

Scorpius will provide (DARHT scale) flash-radiography of dynamic nuclear material. DARHT radiographs non-nuclear (surrogate) materials at a cadence of 8 to 10 hydrodynamic shots per year. Scorpius will radiograph nuclear material at an initial cadence of about 2 to 3 subcritical shots per year in a more challenging environment. Both machines are Linear Induction Accelerators (LIAs) and generate X-rays with high-current, relativistic electron beams. Their use is termed Flash-Radiography.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2020

Test for Reconditioning RA Waste with Simulated Bitumen and Concrete in a 1,2 MW Plasma Test Facility - 20092

The operation and maintenance of nuclear power plants, the non nuclear fuel cycle, etc generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000 deg. C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified, oxidized and purified in an off-gas cleaning system. First the paper describes the new full-scale Plasma Melting Facility (PMF) at the Kozloduy Nuclear Power Plant in Bulgaria which was taken in nuclear operation In May 2018. The plant has a capacity of 250 tons per year and the maximum contact dose rates of the incoming waste is 2 mSv/h. Different mixtures of radioactive waste packed in 200 l drums were successfully treated resulting in a glassy slag free from liquids and organic material with an important volume reduction factor (VRF). The Project was co-financed through a grant by Kozloduy International Decommissioning fund (KIDSF) administrated by the EBRD through Bulgarian national funding. Plasma is a suitable technology for treatment of problematic waste or even reconditioning waste so Belgoprocess was contracted to do plasma tests with simulated conditioned waste types. One can do tests on a laboratory scale on smaller samples and torch capacities of e.g. 50 kW but Belgoprocess wanted to do more realistic and reliable tests. So Belgoprocess contracted Phoenix Solutions Co who has a full-scope test facility equipped with a 1200 kW plasma torch for full-scope treatment of simulated conditioned waste. For a first confidential contract simulated 200 l (55 gallon) bitumen drums were treated. The drums contained different pucks of compacted waste such as rags, used filters, granulates, etc. The pucks were stacked in the 200 l drums and subsequently embedded with bitumen. A total of 6 drums were treated in the plasma facility. For a second contract simulated homogeneous 200 l (55 gallon) concrete drums with on the one hand concentrates and on the other hand spent resins were selected. A total of 6 drums with concrete and spent resins were treated and melted in the plasma testing facility. The paper describe the test facility, volume reduction factor (VRF) of different waste streams and most important parameters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS2006

Stainless Steel NaK Circuit Integration and Fill Submission

The Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system, which was originally designed to hold a eutectic mixture of sodium potassium (NaK), was redesigned to hold lithium; but due to a shift in focus, it is once again being prepared for use with NaK. Changes made to the actively pumped, high temperature loop include the replacement of the expansion reservoir, addition of remotely operated valves, and modification of the support table. Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This document summarizes the integration and fill of the pumped liquid metal NaK flow circuit.

Garber, Anne E.↗

DOE OSTI2023

The 2022 Hunga eruption exercises the nuclear explosion monitoring community

The 2022 Hunga eruption was a global event recorded by all three Comprehensive Test Ban Treaty Organization (CTBTO) International Monitoring System (IMS) geophysical waveform sensor technologies: seismic, infrasound, and hydroacoustic. Large volcanic eruptions are rare and additionally present an opportunity to test seismoacoustic detection, location, and identification methods for nuclear explosion monitoring.

58 GEOSCIENCES↗

DOE OSTI2021

Stochastic Approaches for Calculating and Aggregating Detection Probabilities for Nuclear Material Diversion

The authors built and tested a stochastic simulation to estimate achieved detection probabilities (DPs) on a stratum basis, over a tailorable range of diverted amounts from 0 to 2 significant quantities (SQ), using typical International Atomic Energy Agency (IAEA) inspection data: i.e., SQ in stratum, number of items, number of gross/partial/bias defect measurements conducted, and realistic relative standard deviation (RSD) values for typical IAEA verification measurements. For bulk strata, the model calculates achieved DP at 0.01 SQ diversion increments; for item strata, the model calculates DP using the smallest realistic diversion increment (e.g., a plate, pin, or coupon). After successfully benchmarking against IAEA deterministic models, the simulation was used to test the sensitivity of DP to certain standard assumptions and selected input parameters. First, the equal defect assumption was tested; the results suggest significant complexity in the effectiveness of partial defect measurements. Next, the authors explored the sensitivity of DP to the assumed RSD of attribute tests. Then, the authors compared non-normal models for instrument performance (e.g., logistic, step, or arbitrary functions) to the typical results from a normal distribution (characterized by RSD). This last comparison was supplemented with experimentally derived performance data for an HM-5 gamma spectrometer. The HM-5 was used to make enrichment measurements on both LEU and HEU MTR fuel elements as plates were removed, and the results fit with logistic and step curves and applied in the simulation. These stochastic DP results were compared to DP estimates from a deterministic model assuming a normal curve and typical RSD, yielding insights that could improve effectiveness in the field. These early results illustrate the potential of stochastic models to better understand achieved DP and to improve safeguards effectiveness.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2017

Neutron and Ion Irradiation Studies on Advanced Steels Using the Nuclear Science User Facilities

For nuclear reactor structural materials, irradiation is capable of inducing a series of microstructural modifications, such as dislocation loops/network, selective elemental segregation and precipitation, voids and helium bubbles, leading to the degradation in structural materials' performance through irradiation hardening and embrittlement, swelling, irradiation creep, etc. Since radiation poses challenges for structural applications in nuclear reactors in multiple aspects and the underlying mechanisms are often interconnected, radiation effect studies need well-designed test matrix and proper characterization techniques. The Nuclear Science User Facilities (NSUF) offer extensive capabilities for studying radiation effects in nuclear fuels and structural materials. This research covers both neutron and ion irradiation studies carried out using a variety of instruments through the NSUF. The neutron irradiation ferritic ODS steel MA957 was accomplished utilizing the neutron irradiation capability of the Advanced Test Reactor (ATR) and the post-irradiation examination (PIE) capabilities of the Center for Advanced Energy Studies (CAES). The neutron-irradiated specimens were characterized using transmission electron microscopy (TEM) and atom probe tomography (APT). The in situ TEM ion irradiation study on austenitic stainless steel (SS) 316H was completed using the IVEM-Tandem facility at Argonne National Laboratory.

36 MATERIALS SCIENCE↗

DOE OSTI2019

The Nab experiment: A precision measurement of unpolarized neutron beta decay

Neutron beta decay is one of the most fundamental processes in nuclear physics and provides sensitive means to uncover the details of the weak interaction. Neutron beta decay can evaluate the ratio of axial-vector to vector coupling constants in the standard model, λ = gA/gV, through multiple decay correlations. The Nab experiment will carry out measurements of the electron-neutrino correlation parameter a with a precision of δa/a = 10 -3 and the Fierz interference term b to δb = 3 × 10 -3 in unpolarized free neutron beta decay. These results, along with a more precise measurement of the neutron lifetime, aim to deliver an independent determination of the ratio λ with a precision of δλ/λ = 0.03% that will allow an evaluation of Vud and sensitively test CKM unitarity, independent of nuclear models. Nab utilizes a novel, long asymmetric spectrometer that guides the decay electron and proton to two large area silicon detectors in order to precisely determine the electron energy and an estimation of the proton momentum from the proton time of flight. The Nab spectrometer is being commissioned at the Fundamental Neutron Physics Beamline at the Spallation Neutron Source at Oak Ridge National Lab. We present an overview of the Nab experiment and recent updates on the spectrometer, analysis, and systematic effects.

Fry, Jason↗

DOE OSTI2023

Mechanical testing data from neutron irradiations of PM-HIP and conventionally manufactured nuclear structural alloys

Not Available

Wharry, Janelle P. (ORCID:0000000177914394)↗

DOE OSTI2019

Verification of MCNP6.2 with ENDF/B-VIII.0 Nuclear Data for Nuclear Criticality Safety Applications

MCNP6.2 was released in 2018 and is the latest production release of MCNP ® available from RSICC. MCNP6.2 includes all of the standard features for nuclear criticality safety (NCS) calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods using Whisper for NCS validation. MCNP6.2 verification using ENDF/B-VII.1 nuclear data was previously reported in. This report focuses on the initial testing and verification of MCNP6.2 using the ENDF/B-VIII.0 nuclear data that were released in 2018. The sections that follow provide guidance for NCS users on modifying existing MCNP input files to use the new ENDF/B-VIII.0 ACE nuclear data, discussion of results for several ICSBEP-based benchmark suites using MCNP6.2 with ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

NASA NTRS1999

Teflon FEP Analyzed After Retrieval From the Hubble Space Telescope

During the Hubble Space Telescope (HST) Second Servicing Mission, 6.8 years after the telescope was deployed in low Earth orbit, degradation of unsupported Teflon FEP (DuPont; fluorinated ethylene propylene), used as the outer layer of the multilayer insulation (MLI) blankets, was evident as large cracks on the telescope light shield. A sample of the degraded outer layer (see the photograph) was retrieved during the second servicing mission and returned to Earth for ground testing and evaluation. Also retrieved was a Teflon FEP radiator surface from a cryogen vent cover that was exposed to the space environment on the aft bulkhead of the HST. NASA Goddard Space Flight Center directed the efforts of the Hubble Space Telescope MLI Failure Review Board, whose goals included determining the FEP degradation mechanisms. As part of the investigations into the degradation mechanisms, specimens retrieved from the first and second HST servicing missions, 3.6 and 6.8 years after launch, respectively, were characterized through exhaustive mechanical, optical, and chemical testing. Testing led by Goddard included scanning electron microscopy, optical microscopy, tensile testing, solar absorptance measurements, time-of-flight secondary ion mass spectroscopy (TOF-SIMS), Fourier transform infrared microscopy (m-FTIR), attenuated total reflectance infrared microscopy (ATR/FTIR), and x-ray diffraction (XRD). The NASA Lewis Research Center contributed significantly to the analysis of the retrieved HST materials by leading efforts and providing results of bend testing, surface microhardness measurements, x-ray photoelectron spectroscopy, solid-state nuclear magnetic resonance spectroscopy, and density measurements. Other testing was conducted by Nano Instruments, Inc., and the University of Akron.

Dever, Joyce A.↗

DOE OSTI2019

Feasibility Testing of Delayed Scintillator Screens for Indirect Digital Neutron Radiography of Irradiated Nuclear Fuel

Neutron radiography provides information about the internal condition of irradiated nuclear fuels. Traditionally, neutron radiography of highly radioactive irradiated nuclear fuels has been done with film using indirect radiography methods.[1] While indirect film neutron radiography is insensitive to gamma radiation from radioactive samples, it is a time-consuming and labor-intensive process. Digital neutron radiography requires a fraction of the time to produce a neutron radiograph compared to film-based processes. However, electronic components of digital radiography systems can easily be damaged by gamma radiation, which typically precludes use of digital systems for examination of highly radioactive objects, though recent work has demonstrated that some digital neutron radiography systems can indeed directly examine such samples.[2,3] This paper describes experiments performed using a hybrid transfer-digital method with an activated scintillator screen that can record an image that is subsequently read by a digital camera. The screens used in this study use a dysprosium neutron converter that records a latent image in the form of activation while in a neutron beam, then is physically transported away from the beam and sample to a digital camera elsewhere to record the image while the activated dysprosium decays and the adjacent scintillator produces light from interactions with the decay radiation. This approach could accelerate radiographic examination of irradiated nuclear fuels compared to the current transfer method with film. Current indirect imaging infrastructure was modified and augmented with a digital system. A digital neutron radiograph of a highly-radioactive irradiated nuclear fuel pins was produced in ~20 minutes, which is a significant improvement compared to the current process that takes almost a full day to produce a radiograph.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2020

ZPPR Test Bed (ZTB) Pre-Conceptual Design Report

Authorized by the Nuclear Energy Innovation Capabilities Act (NEICA), the National Reactor Innovation Center (NRIC) provides private sector technology developers access to strategic infrastructures and assets for commercial nuclear energy research, development, demonstration, and deployment activities. The mission is to support a timely and cost-effective path to the licensing and commercialization of new nuclear energy systems. To meet these needs, NRIC is developing two reactor demonstration test beds at Idaho National Laboratory (INL), the ZPPR Test bed (ZTB) and the EBR-II Test bed (ETB). ZTB will support the demonstration of systems that operate at less than 500 kWt. The baseline objective is for the ZPPR Cell to act as a confinement structure capable of siting reactors that utilize Safeguards Category 1 material for operations. The major areas addressed in the pre-conceptual design include: • Installation of an access door • Electrical Power • Heat Removal • Ventilation in the Cell • Reactor Installation Along with the design for ZTB, a concept of operations (COP) has also been developed. The COP is intended to facilitate a common understanding of ideas, challenges, and issues. As systems continue to evolve in complexity System Engineers and Project Directors will utilize and update the COP to develop and sustain a common vision of the system for stakeholders.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Advanced Materials Program Summary

Materials solutions - enabling design, construction, and operation of licensable Advanced Nuclear Material, design, fabrication, installation, examination, testing, overpressure protection, inspection, stamping, and certification

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗