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Trellue, Holly R.

Publications and source records attributed to Trellue, Holly R..

16 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Complete Survey of Fuel Candidates for Microreactor Purposes

This report summarizes various fuel types that may be applicable to several microreactor concepts, which are defined as (1) very high temperature reactors (VHTR), (2) sodium fast reactors (SFR), (3) system for nuclear auxiliary power (SNAP) reactors, (4) gas fast reactors (GFR), and (5) molten salt reactors (MSR). The fuel systems that were assessed include: uranium mononitride (UN), uranium monocarbide (UC), uranium dioxide (UO 2 ), uranium oxycarbide (UCO) Tristructural Isotropic (TRISO), UN TRISO, mixed oxide (MOX), metallic fuels, and metal hydrides. While UCO TRISO has undergone significant testing through the advanced gas reactor (AGR) program, the very high cost necessitates consideration of other fuel types. UN and UC were identified as fuels that should receive further investigation due to their thermophysical and mechanical properties. Minimal irradiation performance data shows that these fuels are good candidate fuels for microreactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

The National Criticality Experiments Research Center and its role in support of advanced reactor design

The National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site (NNSS) in the Device Assembly Facility (DAF) and operated by Los Alamos National Laboratory (LANL) is the only general purpose critical experiments facility in the United States. Experiments from subcritical to critical and above prompt critical are carried out at NCERC on a regular basis. In recent years, NCERC has become more involved in experiments related to nuclear energy, including the Kilopower/KRUSTY demonstration and the recent Hypatia experiment. Multiple nuclear energy related projects are currently ongoing at NCERC. This paper discusses NCERC’s role in advanced reactor design and how that role may change in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification

The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup, under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes. The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.

Jacobsson-Svard, Staffan↗
DOE OSTI2017

Determination of origin and intended use of plutonium metal using nuclear forensic techniques

Nuclear forensics techniques, including micro-XRF, gamma spectrometry, trace elemental analysis and isotopic/chronometric characterization were used to interrogate two, potentially related plutonium metal foils. These samples were submitted for analysis with only limited production information, and a comprehensive suite of forensic analyses were performed. Resulting analytical data was paired with available reactor model and historical information to provide insight into the materials’ properties, origins, and likely intended uses. Both were super-grade plutonium, containing less than 3% 240 Pu, and age-dating suggested that most recent chemical purification occurred in 1948 and 1955 for the respective metals. Additional consideration of reactor modelling feedback and trace elemental observables indicate plausible U.S. reactor origin associated with the Hanford site production efforts. In conclusion, based on this investigation, the most likely intended use for these plutonium foils was 239 Pu fission foil targets for physics experiments, such as cross-section measurements, etc.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Maintaining a critical spectra within Monteburns for a gas-cooled reactor array by way of control rod manipulation

Our burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4×4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach - where the amount of fissile material in a set configuration is slowly altered until criticality is attained - in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. Finally, while the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Spent Fuel NDA Research Path for the Sweden Encapsulation-Repository

This set of slides provides a description of research performed to date on spent fuel NDA: Next Generation Safeguards Initiative Spent Fuel Project, and NDA analysis and research planned for CLINK. The general purpose is strengthening the technical toolkit of safeguard inspectors. Data mining is being applied to determine the optimal mathematical structure to match the complexity of spent fuel NDA signals and to enable a range of quantities to be estimated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Analysis of Differential Die Away Instrument Simulated Performance Using Boiling Water Reactor Spent Fuel Assemblies

As part of evaluating the performance of various Non-Destructive Assay (NDA) instruments measuring the plutonium content and other information about spent nuclear fuel, the Next Generation of Safeguards Initiative (NGSI) Spent Fuel (SF) Project generated spent fuel libraries containing the composition of spent fuel assemblies originating from various reactors, initial enrichments (IE), burnups (BU), and cooling times (CT). Spent Fuel Library (SFL) number 5 (SFL5), representing Boiling Water Reactor (BWR) assemblies, was created as part of a “Gentleman’s Agreement” with Sweden to examine if the Differential Die Away (DDA) instrument performs similarly for BWR assemblies as for the Pressurized Water Reactor (PWR) ones. The difference between PWRs and BWRs is that BWRs have a large axial variation in moderator density because it is in the form of water at the bottom of the reactor but turns into steam as it gets warmer at the top of the reactor; thus, more plutonium builds up at the bottom of each assembly than at the top. In a PWR, the moderator density varies slightly from the bottom to the top, but since it is pressurized, the changes are not as significant with respect to the burnup and plutonium distributions. BWR assemblies are also typically smaller than PWR ones, which could affect detector performance. Since Sweden has both PWR and BWR assemblies, it is anticipated that we will be deploying the DDA instrument to measure both spent fuel assembly (SFA) types.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Spent Fuel Assembly Modeling for Unattended Gamma Emission Tomography (UGET) Project [Slides]

NGSI-SF project provided good basis for spent fuel assemblies. UGET assemblies have both DU pins and water replacing fuel in the center of the bundle. Both MCNP geometry files and CINDER90 isotopic compositions files were created and used by LLNL and/or others. Four different spent fuel assembly types were represented: PWR, BWR, VVER-440, and the VVER-1000. Not enough information was known about irradiation of assemblies in Forsmark or Olkiluoto (JNT 1510) benchmarks for detailed calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Development of Database of SFL2a Signals for Neutron Generator Techniques Proposed for Integration

The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and International Security (NIS), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) quantify the plutonium (Pu) mass of a spent fuel assembly, (2) detect partial defects in assemblies, and (3) verify the initial enrichment (IE), burnup (BU) and cooling time (CT) of an assembly. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Ref. 1 and 2]. A subset of 5 NDA techniques involving passive counting and/or a neutron generator were researched for integration potential with the focus on the first of the three goals, quantifying Pu mass.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

The Use of Non Destructive Assay Measurements to Estimate Burnup and Plutonium Inventory in Pressurized Water Reactor Spent Fuel Assemblies

One of the overarching goals of the Next Generation of Safeguards Initiative (NGSI) Spent Fuel (SF) Non-Destructive Assay (NDA) Project is to determine the Pu mass in spent fuel assemblies with NDA instruments.1 Other goals include verifying initial enrichment, burn-up and cooling time of declaration, detecting diversion or replacement of pins, and/or determining heat emitted from an assembly (not a safeguards goal but important nonetheless). In pursuit of these goals, various NDA instruments of interest need to be built, tested, and then calibrated. Ideally simple calibration sources can be used such as isotopic sources or fresh assemblies. However, given the fact that many of the measured signals are influenced by the complex, energy-dependent interaction of neutrons with a range of isotopes, the connection between the measured signals and the Pu mass is difficult to make.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Description of Axial Detail for ROK Fuel

For the purpose of NDA simulations of the ROK fuel assemblies, we have developed an axial burnup distribution to represent the pins themselves based on gamma scans of rods in the G23 assembly. For the purpose of modeling the G23 assembly (both at ORNL and LANL), the pin-by-pin burnup map as simulated by ROK is being assumed to represent the radial burnup distribution. However, both DA and NDA results indicate that this simulated estimate is not 100% correct. In particular, the burnup obtained from the axial gamma scan of 7 pins does not represent exactly the same 'average' pin burnup as the ROK simulation. Correction for this discrepancy is a goal of the well-characterized assembly task but will take time. For now, I have come up with a correlation for 26 axial points of the burnup as obtained by gamma scans of 7 different rods (C13, G01, G02, J11, K10, L02, and M04, neglecting K02 at this time) to the average burnup given by the simulation for each of the rods individually. The resulting fraction in each axial zone is then averaged for the 7 different rods so that it can represent every fuel pin in the assembly. The burnup in each of the 26 axial zones of rods in all ROK assemblies will then be directly adjusted using this fraction, which is given in Table 1. Note that the gamma scan data given by ROK for assembly G23 included a length of ~3686 mm, so the first 12 mm and the last 14 mm were ignored to give an actual rod length of ~366 cm. To represent assembly F02 in which no pin-by-pin burnup distribution is given by ROK, we must model it using infinitely-reflected geometry but can look at the effects of measuring in different axial zones by using intermediate burnup files (i.e. smaller burnups than 28 GWd/MTU) and determining which axial zone(s) each burnup represents. Details for assembly F02 are then given in Tables 2 and 3, which is given in Table 1 and has 44 total axial zones to represent the top meter in explicit detail in addition to the other 26 zones. Note that the MCNP files for F02 were created using the Monte Carlo burnup linkage code Monteburns, which saves MCNP input files with detailed compositions as a function of burnup. The 'intermediate burnup files' produced for F02 include a cooling time of 27 years. The axial location of 5 spacers was also included in the ROK F02 assembly in which each spacer contained a length of 3.81 cm. Note that due to the nature of Monteburns, which was run in a special fashion for this problem, the step number increments after the 27 year decay, so the second column of Table 2 refers to the step number that should be used in the Monteburns files.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗