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DOE OSTI2024

RokBase Core Facility Data Collection

The RokBase Core Facility Data Collection v1.0 is an aggregation of information on the locations and collections of geologic core facilities in the United States. It contains two spatial files, their corresponding attribute tables within a catalog, and an interactive web map and dashboard. It is intended to be used for research and comparison purposes, see full disclaimer and credits. Please contact NETL if revisions / corrections can be made to improve this dataset.

core facility location↗

DOE OSTI2025

Radiation Characterization Summary: Godiva IV Critical Assembly Environments at the In-Core, Top Hat, 1m, and 2m Irradiation Locations

This document presents the facility-recommended characterization of the neutron, prompt gamma ray, and delayed gamma ray radiation fields at the Godiva IV critical assembly at the National Criticality Experiments Research Center (NCERC). The environments assessed include the In-Core location, a location on the Top Hat, 1m away from the assembly, and 2m away from the assembly. The neutron, prompt gamma ray, and delayed gamma ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma ray fluence profiles on the Top Hat surrounding the critical assembly. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Nevada Higher Education Benefits from US Department of Energy (DOE) Environmental Management (EM) Nevada Program's Transfer of Geologic Samples - 20442

The Nevada National Security Site (NNSS), formerly the Nevada Test Site, was the location of 100 historic atmospheric and 828 historic underground nuclear tests from 1951 to 1992. Related to this historic nuclear testing, geologic and hydrologic studies of the site were conducted utilizing the skills and the expertise of the U.S. Geological Survey (USGS) and national laboratories (Los Alamos National Laboratory, Lawrence Livermore National Laboratory), who were principal in leading the development of weapons and were responsible for specific underground testing programs. The subsurface samples and technical/scientific data associated with NNSS geologic studies (past and present) are preserved and stored at the USGS Mercury Core Library and Data Center located at the NNSS. Currently, the facility stores over 2,000,000 linear feet of cores and cuttings from more than 2,600 drill holes that can be accessed for study. Most of the samples and historic geologic work was focused directly on subsurface geologic settings that relate directly to historic underground nuclear testing. Since 1992, the United States has observed a unilateral moratorium on full-scale nuclear testing, and the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program is now responsible for hydrogeologic characterization of the potential impacts to the natural groundwater systems that may have resulted from historic underground nuclear tests. As part of this effort, the EM Nevada Program Underground Test Area (UGTA) Activity has drilled and completed over 58 deep (2,000 - 7,000 ft.) characterization wells, totaling in excess of 170,000 linear feet of cuttings and core samples. Cutting samples were collected as triplicate samples for each respective depth interval, to account for potential later nondestructive/destructive analysis and to preserve samples for regulatory purposes. Recently, it was recognized by the EM Nevada Program that opportunities may exist to reduce the cost and floor space required for the storage of geologic samples at the Mercury Core Library without impacting the integrity and representative nature of the samples necessary for project execution. An initiative was sponsored by EM Nevada Program to evaluate several options: 1) disposal of a portion of cuttings and cores in a land fill setting; 2) reduction in sample volume through skeletonizing core and cuttings from wells to a representative but much smaller number of samples; and 3) solicit potential interest in the academic community where other geoscientists could freely access the samples for studies. The Nevada state university system through the Nevada Bureau of Mines and Geology responded positively to the opportunity to receive these NNSS samples from the EM Nevada Program. In June 2019, approximately 17,000 geologic samples, representing greater than 170,000 linear feet of drilling and weighing over 20,000 pounds, were shipped from the NNSS to the Great Basin Science Sample and Records Library located in Reno, Nevada. The benefits resulting from this transfer were realized by the EM Nevada Program in terms of cost and space savings for geologic sample storage at the NNSS. The transfer did not impact the EM Nevada Program mission as access to representative geologic samples for regulatory and scientific purposes was preserved. Moreover, the Nevada Bureau of Mines and Geology acquired a significant resource of subsurface geologic samples and supporting technical data to support academic and scientific studies in a complex volcanic setting in southwest Nevada. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Successes and Lessons Learned in the Design, Construction and Use of a Tool for Sampling the WR-1 Reactor Core - 1900z

Whiteshell Laboratories, located in Manitoba, Canada, provided research facilities for the Canadian nuclear industry since the early 1960's. The facility was centered on an organically cooled, heavy water moderated nuclear reactor in the world, Whiteshell Reactor 1 (WR-1). WR-1 was safely shut down in 1985, de-fueled and drained, and has been in a safe Storage with Surveillance (Deferred Decommissioning) state until present day. In support of an Environmental Assessment for proposed In Situ Decommissioning of WR-1, detailed characterization information regarding the reactor core components (calandria, calandria tubes, pressure tubes) was necessary to verify inventory estimates and neutron activation models produced circa 1992. The Fuel Channels of WR-1 (pressure tubes) were expected to be the most activated components of the reactor core, and to account for about 60% of the reactor core inventory. The interior surface of the fuel channel is also the most readily accessible surface within the reactor core. As such the decision was made to collect a sample of one of each of 3 Fuel Channel material types (Stainless Steel, Zr-2.5Nb, Ozhenite) for comparison to modelled estimates of neutron activation of various core materials. The expected dose rates from the samples were between 5 and 100 mSv (500-10,000 mrem) on contact based on a 200 mg sample. Sampling the Fuel Channels required verification the Fuel Channel was drained of organic coolant, removal of residual organic coolant from the Fuel Channel wall, removal of material from the interior surface of the Fuel Channel at the centerline of flux (∼25' below the operator position), and extraction of that material from the core. Several options for existing tooling and technologies, including hydraulic scrapers, and horizontal drill technologies, were looked at for their applicability to WR-1. While suitable approaches to higher radiation field sample collection, they were overly complex, costly and time consuming for the configuration at WR-1. In an effort to reduce costs, and to simplify the sampling process, a specialized and simple tool was developed to mechanically scrape a small sample of the interior wall of the Fuel Channels, collect the sample into a vial, and remove the sample from the core for analysis. The tool was designed, constructed, tested and implemented in the field entirely by Canadian Nuclear Laboratories Staff. The tool required no powered components, or electronic guidance or measurements; and was designed and constructed for a fraction of the cost of other commercially available technologies for sample collection within reactor cores. The samples were successfully collected without any injuries, contamination events or equipment failures. The tool remains operational in the event additional sampling is required. This paper provides additional details of the design, construction, testing and use of the tool and the successes and lessons learned during the various stages of design and implementation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Remote Technology to Retrieve Waste Below the 324 Building at the Hanford Site - 19622

The Waste Technology Engineering Laboratory (324 Building) is a Category 2 Nuclear Facility at Hanford. The 324 Building Disposition Project is responsible for the removal and remedial actions under CERCLA. Additionally, there are RCRA Dangerous Waste Management Units which the CERCLA actions are presumed to satisfy. The Radiochemical Engineering Complex (REC) is a group of four hot cells (A, B, C and D Cells) with an interconnecting airlock inside of the 324 Building. Underneath the REC B-Cell are contaminated soils that have an estimated 12,793 R/Hr dose rate and remediation is under CERCLA 300-FF-2 Operable Unit. The current action consists of debris removal from the floor of B-Cell, B-Cell floor removal, and excavation of the soils beneath the floor to a depth of up to twelve feet with four Remote Excavator Arms (REA). This paper provides insight into the preparation and performance of high tolerance core drilling into a hot cell complex. Installation of the REA Through Support (REA TS) mechanism through the 4'-6' concrete shield walls is a complex series of tasks with demanding tolerances at each of four locations around the perimeter of B-Cell. The use of training mock ups was included in the completion of this work. The REA TS is two 11 inch diameter high strength steel sleeves that have expandable collars at each end allowing the interstitial space between the outside of the sleeve and inside of the core drill to be grouted to form a substantial structural connection, and a 10 inch pipe with hydraulic hoses to power the REA. There are three cores required for each REA location. The REAs themselves are customized excavator arms, mounted on support columns to be used in the excavation and transfer of the contaminated soil into storage facilities. The tasks for the installation of the REA TSs are: 1) Characterization of B-Cell Cubicles. 2) Embedded Pipe Fill. 3) Interference Removal. a. Internal Walls, b. External Walls. 4) Core Drilling. 5) Go/No-Go gauge, and video inspection. 6) Installation of REA Through Support Assembly. The 324 Building REC Hot Cells have internal, external, and embedded utilities throughout the shield walls in conflict with the REA TS locations. Some of the embedded utilities have connection points inside of the B-Cell Cubicles. There were two separate issues at the conflicted locations. The first issue was the core drill cooling water would flow along the embedded utility potentially spreading legacy chemical and radiological contamination to areas of the facility that was unwanted. The second issue was that the REA TS required a substantial structural connection to the B-Cell wall, the cored diameter is larger than the REA TS and these interstitial areas required a full grouted connection. The embedded utilities posed a hazard that if unfilled, the interstitial grout would flow into the embedded utility and destabilize the required grouted structural connection. The B-Cell Cubicles had not been opened in over 10 years. A mockup was completed to install a glovebag over the cubicle opening and then to verify functionality and port requirements. The embedded utility pipe connections in the cubicles were then opened inside a glovebag to verify the correct connection points, connection method for future embedded pipe fill material delivery system, radiological characterization, and air flow to show that the embedded utility pipes were not blocked and would not pressurize. The embedded pipe fill material and grout system was mocked up and run several times to verify methodology and functionality. The B-Cell Cubicles were then reentered and the embedded utility pipe fill was completed using a hand pump. External interferences at the core drill locations were removed using standard interference removal techniques. Internal interferences at the core drill locations were removed using a remote cutting tool that was controlled through the use of the B-Cell crane, remote cameras and remote lighting. The core drilling was mocked up two separate times, the project's purpose built 'Mock Up Facility' was core drilled to verify tolerances and methodology prior to subcontract award, the core drill subcontractor built a 'dummy wall' to train and practice on achieving the demanding tolerances, coordinated alignments between the cores, short term and long term dose shielding techniques, and contamination control techniques. A core drill template, and additional anchoring techniques were used to maximize accuracy and precision of the REA TS core drills. A 'Go/No-Go' gauge was built and installed at each of the four REA TS locations to verify that the core drill met tolerances. The core drill locations were visually inspected using a boroscope. The REA TS installation was mocked up at the projects 'Mock Up Facility' beginning with basic hazard controls, then with intermediate hazard controls, and finally with full occupational safety, industrial hygiene, and radiological controls. The REA TSs were installed and grouted into place. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

FY2023 NSTTF Annual Report

The National Solar Thermal Test Facility (NSTTF) is a DOE Core Capability and Technology Deployment Center located in Albuquerque, NM. It is operated by Sandia National Laboratories (Sandia) for the U.S. Department of Energy (DOE). The NSTTF is the only multi-mission, multi-use, multi-story test facility of its type in the United States. The NSTTF was founded in 1978 and began testing with high heat flux that same year. Over the past 45 years, the NSTTF has been at the forefront of the research, design, fabrication, and testing of many of the critical Concentrating Solar Power (CSP) technologies. These technologies have allowed costs to be dramatically reduced from over $\$ $0.40 /kWh to $\$$0.12 /kWh since the conception of this renewable energy technology. The NSTTF has worked to make the Solar Energy Generating Systems (SEGS) parabolic trough plants successful, while also working with the Solar One and Solar Two facilities for successful implementation. Over the four decades since its founding, the mission of the NSTTF has grown to include new receiver technologies, like our generation 3 falling particle system (G3P3 Tower), optical metrology techniques like SOFAST, molten salt testing, thermal energy storage, solar thermal chemistry, and more. We continue to expand our capabilities in pursuit of the DOE SETO mission and the DOE SunShot 2030 goals: unsubsidized LCOE of $\$$0.05/kWh for CSP that includes 12 or more hours of thermal energy storage. To support both the DOE SETO mission and support the CSP sector as a whole, we are working to develop our operations and maintenance framework to provide a world class testing facility in support of our technological achievements. To accomplish both of these missions, the NSTTF draws on the decades of experience and expertise of our staff along with the world-class facilities at Sandia National Laboratories to further the science of concentrated solar thermal technologies in diverse applications. We remain a trusted partner for high-quality and impactful research in both fundamental and applied arenas. We are able to provide our partners with both one-of-a-kind testing platforms as well as world-class analytics.

14 SOLAR ENERGY↗

DOE OSTI2020

Technology Assessment for Radioiodine at the Hanford Site - 20240

An evaluation was conducted to assess the practicability of technologies of sufficient promise and maturity to warrant treatability tests for remediating radioiodine (I-129) in groundwater at the 200-UP-1 operable unit at the U.S. Department of Energy Hanford Site. Technologies were evaluated based on recent updated knowledge of subsurface iodine transport behavior at Hanford, and the effectiveness, implementability, and cost of potential treatment technologies for I-129. Because more than two-thirds of the iodine at Hanford is estimated to be present in its oxidized form as iodate, treatment technologies were evaluated with respect to addressing iodate. In situ treatments were identified as having limited implementability because the plume core is located directly beneath the Environmental Restoration Disposal Facility disposal cells, a landfill used for the disposal of low-level radioactive, hazardous, and mixed wastes generated from Hanford cleanup activities. Further limitations in treatment technologies for radioiodine were due to the isotopic distribution of subsurface iodine. Data indicate that stable iodine (I- 127) concentrations are up to three orders of magnitude greater than I-129, limiting the effectiveness of potential technologies because they are not isotope-specific. Ex situ approaches also lacked the effectiveness and maturity required to treat I-129. Hence, the technology evaluation did not identify any technologies that needed treatability testing, driven by site and contaminant properties that hinder effectiveness and/or implementability of the technologies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Full Length Assembly Testing in PELICAN (Final Report)

In support of the development of the U.S. Department of Energy (DOE) Versatile Test Reactor (VTR), a thermal hydraulics test facility was constructed to generate experimental measurement of the pressure drop across a single full-scale assembly containing prototypic axial reflectors, fuel, and plena components. Constructed and operated at Argonne National Laboratory, the Pressure drop Experimental Loop for Investigations of Core Assemblies in Nuclear reactors (PELICAN) facility was designed to achieve hydraulic conditions identical to those anticipated for a full-scale fuel assembly located in the VTR core in the region with the highest flow rate. Using water as surrogate for liquid sodium, the flow loop was operated at elevated temperatures and pressures to match the thermophysical properties of liquid sodium and ensure matching Reynolds and Euler numbers. The measurement objectives for data generated from this test facility was driven primarily by the validation needs for code calculations and simulations of the reference VTR core. These objectives focused on the need to validate pressure drop results across the various segments of the fuel assembly as they relate directly to the pumping power and safety of the reactor. Presented in this report are experimental results and analytical comparisons based on testing of a full-length assembly in PELICAN. Housed within a hexagonal test section extending 3.4 m in length, the tested assembly features a prototypic lower reflector, grid plates, wire-wrapped rod bundle, upper reflector, and exit region. The rod bundle extends over 1.5 m in length and contains 217 individual wire-wrapped rods with dimensions that best reflect the reference VTR design. The as-tested bundle assembly was fabricated using 316 stainless steel 0.25-inch (6.35-mm) diameter rods wrapped with 0.04-inch (1.016-mm) diameter wire at a helical pitch of 10.51 inch (26.6 cm). Details of the method for in-house wire-wrapping, assembly, and installation are provided later in this report. Experimental measurements of pressure drop at 19 positions along the test assembly were recorded for a range of flow conditions, with special attention paid to key locations within the assembly, including component inlet and outlet, transition, and wire-wrapped rod bundle regions. Testing conditions were based on 110°C water with flow rates ranging from 50 to 450 GPM (3 to 27 kg/s) at the inlet of the test assembly generating Reynolds numbers and velocities up to ~8.0×10 4 and ~7.8 m/s, respectively, within the rod bundle region. Non-dimensional values for the friction factor were then calculated based on these experimental measurements and compared against those predicted by various analytical correlations available from open literature. Predictions by the upgraded Cheng and Todreas, Rehme, and Novendstern correlations fell within 4% to those values measured experimentally.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Exclusion and Verification of Remote Nuclear Reactors with a 1-kiloton Gd -Doped Water Detector

To date, antineutrino experiments built for the purpose of demonstrating a nonproliferation capability have typically employed organic scintillators situated as close to the core as possible (typically at a distance of a few meters to tens of meters) and have not exceeded a few tons in size. One problem with this approach is that proximity to the reactor core requires accommodation by the host facility. Water Cherenkov detectors located offsite, at distances of a few kilometers or greater, may facilitate nonintrusive monitoring and verification of reactor activities over a large area. As the standoff distance increases, the detector target mass must scale accordingly. This paper quantifies the degree to which a kiloton-scale gadolinium-doped water Cherenkov detector can exclude the existence of undeclared reactors within a specified distance, and remotely detect the presence of a hidden reactor in the presence of declared reactors, by verifying the operational power and standoff distance using a Feldman-Cousins-based likelihood analysis. A 1-kton scale (fiducial) water Cherenkov detector can exclude gigawatt-scale nuclear reactors up to tens of kilometers within a year. In conclusion, when attempting to identify the specific range and power of a reactor, the detector energy resolution is not sufficient to delineate between the two.

42 ENGINEERING↗

DOE OSTI2022

Computed Tomography Scanning and Geophysical Measurements of the Enhanced Geothermal Systems (EGS) Collab SURF Core

The computed tomography (CT) facilities and the Multi-Sensor Core Logger (MSCL) at the National Energy Technology Laboratory (NETL) in Morgantown, West Virginia were used to characterize core from 8 wells from the Enhanced Geothermal Systems (EGS) Collab project located in the Sanford Underground Research Facility’s (SURF) 4850 level. The work in this report falls in Experiment 1 in a three-phase experiment, with the primary purpose to increase understanding of hydraulic fracturing. The primary impetus of this work is a collaboration between U.S. Department of Energy (DOE) and the EGS Collab. The resultant datasets are presented in this report and can be accessed from NETL's Energy Data eXchange (EDX) online system using the following link: https://edx.netl.doe.gov/dataset/surf-collab-core. All equipment and techniques used were non-destructive, enabling future examinations and analyses to be performed on these cores. None of the equipment used was suitable for direct visualization of the pore space in the core, although fractures and discontinuities were detectable with the methods tested. Low resolution CT imagery with the NETL medical CT scanner was performed on the entire core. Qualitative analysis of the medical CT images, coupled with X-ray fluorescence (XRF), P-wave, and magnetic susceptibility measurements from the MSCL were useful in identifying zones of interest for more detailed analysis as well as fractured zones. The ability to quickly identify key areas for more detailed study with higher resolution will save time and resources in future studies. The combination of methods used provides a multi-scale analysis of the core; the resulting macro and micro descriptions are relevant to many subsurface energy related examinations traditionally performed at NETL.

47 OTHER INSTRUMENTATION↗

DOE OSTI2016

Radiation Characterization Summary: ACRR Cadmium-Polyethylene (CdPoly) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline

This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the cadmium-polyethylene (CdPoly) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-CdPoly-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to Drew Tonigan for helping field the activation experiments in ACRR, David Samuel for helping to finalize the drawings and get the parts fabricated, and Elliot Pelfrey for preparing the active dosimetry plots.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

The design, implementation, and performance of the LZ calibration systems

LUX-ZEPLIN (LZ) is a tonne-scale experiment searching for direct dark matter interactions and other rare events. It is located at the Sanford Underground Research Facility (SURF) in Lead, South Dakota, USA. The core of the LZ detector is a dual-phase xenon time projection chamber (TPC), designed with the primary goal of detecting Weakly Interacting Massive Particles (WIMPs) via their induced low energy nuclear recoils. Surrounding the TPC, two veto detectors immersed in an ultra-pure water tank enable reducing background events to enhance the discovery potential. Intricate calibration systems are purposely designed to precisely understand the responses of these three detector volumes to various types of particle interactions and to demonstrate LZ's ability to discriminate between signals and backgrounds. In this paper, we present a comprehensive discussion of the key features, requirements, and performance of the LZ calibration systems, which play a crucial role in enabling LZ's WIMP-search and its broad science program. The thorough description of these calibration systems, with an emphasis on their novel aspects, is valuable for future calibration efforts in direct dark matter and other rare-event search experiments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

Influence of irrigation approaches and spatial geolocation on tritium speciation, uptake and depuration

Pine needles and tree cores from a tritium (T) contaminated phytoremediation forest at the Savannah River Site (SRS in Aiken, SC) Mixed Waste Management Facility (MWMF) were measured for total T and T speciation and compared to other locations at the SRS and the surrounding area. Tree core ages ranged from 9 to 14 years old, covering over half of the ~20-year on-going remediation efforts, while pine needles represent more recent time periods of 1-to-2-year increments. Remedial irrigation efforts at the MWMF are found to directly influence the pine needle T concentrations. The T content in the MWMF samples is higher than non-irrigated needle samples from other locations around the SRS. Furthermore, the different forms of organic bound T are preferentially stored in tree core tissue, compared to pine needles where tritiated water dominates.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2017

Resonance Self-shielding Impact on Neutron Spectrum Determination for Missouri S and T Reactor

Neutron Activation Analysis is a useful technique for characterizing the neutron flux at nuclear facilities. It involves irradiating samples to produce unstable isotopes, which enables qualitative and quantitative analyses of unknown samples. In the process, samples are bombarded with neutrons; new elements then become radioactive isotopes due to absorption of neutrons. These emit different amounts of gamma energy during the decay process, and gamma rays are counted using HPGe. Each reaction product has a known half-life and cross-section, and by measuring the amount of gamma radiation energy emitted from a radioactive isotope, it is possible to identify the isotope and activity. However, the neutron spectrum is not precise or representative of the actual neutron environment because of the self-shielding effect. This occurs with large samples that have large cross-sections or thicknesses in the resonance range of the neutron flux spectrum. Several methods have been used to mitigate this problem, such as thinning or diluting alloys samples, calculating the self-shielding factor for each material, and generating a corrected library for the unfolding procedure code. The Missouri University of Science and Technology Reactor (MSTR) is a pool-type research reactor that operates up to 200 kW. It uses light-water for both moderation and natural convective heat removal. The reactor core consists of 15 fuel elements positioned in a 9 x 6 aluminum grid plate, with four elements used for reactor power control. The fuel element is 19.75% low enriched uranium in 18 curved fuel plates. The reactor has seen various configurations starting with configuration designation 101W and subsequent reconfiguration increases the numeric designation by 1. The current configuration, 120W, was implemented in 2010. The new configuration provides more in-core locations for activation experiments and uses more fuel elements than previous ones. One of the irradiation locations in the MSTR core is the source holder. It is desired to improve the documentation of neutronic characteristics of the MSTR which are cardinal to experiments supported by the facility. One of these characteristics is the neutron spectrum at various locations in the reactor. Successful efforts yielded the neutron spectrum at a known location through the unfolding of the activity of irradiated foils. The purpose of this current study is to investigate the impact of resonance self-shielding of some of the activated foils on the determined spectrum.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Re-Analysis of the RRR/SEG Fast Thermal Coupled Facility

The Rossendorf Research Reactor (RRR), or Rossendorfer Ringzonenreacktor, was an Argonaut type zero power research reactor located at the Rossendorf research facility in Dresden Germany. The RRR consists of an annular core surrounded by a gr¬aphite reflector. Inside the core is an additional removable internal graphite reflector. This internal reflector was replaced in December of 1972 with a rapid deployment lattice known as Schnelles Einsatz-Gitter (SEG). The insertion of the SEG experimental lattice leads to a fast-thermal coupled system with the annular RRR reactor being the thermal driver and the SEG being the fast insertion lattice. The RRR/SEG integral experimental setup was used to perform small sample reactivity measurements using the pile oscillator method. By loading the SEG lattice with specially designed unit cells, a specific adjoint spectra could be obtained (only SEG 6 had a special radial arrangement). With a nearly energy-independent adjoint spectrum, the slowing-down effect becomes a constant and essentially disappears, such that any chan¬¬¬¬ge in reactivity can be attributed to the absorption component of the sample material. SEG 4, 5, and 7 exhibit this characteristic. If the adjoint function is strongly energy-dependent, then the scattering component will dominate. In the case of both SEG 6 lattices, where the adjoint is also monotonously rising, the scattering effect will always be negative. This analysis aims to re-evaluate the central reactivity worths of different samples for the SEG 4, 5, 6, and 7 experiments using modern continuous energy monte-carlo.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Investigation of Treat M2 and M3 calibration experiments for benchmark analysis

The Transient Reactor Test Facility (TREAT) located at Idaho National Laboratory is a graphite moderated and reflected, air cooled core. This reactor was originally designed and operated to experimentally test fuel under extreme transient conditions. To this end, TREAT successfully performed many series of steady-state and transient tests until operations were discontinued in 1994. With the restart of the reactor planned for 2018, accurate models of previous reactor experiments are necessary to verify the ability to predict conditions for planned upcoming tests. In this work, the methodology for modeling the M2 and M3 calibration experiments (M2/M3CAL) will be described along with results from the variation of key parameters including the boron concentration. Additionally, this model will be used to demonstrate the correlation between reactor behavior and the implementation of 'nuclear graphite' thermal scattering law (TSL) data to capture the behavior of thermal neutrons in the TREAT core.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Radiation Characterization Summary: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline (ACRR-PLG-CC-32-cl).

This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

Vehar, David W.↗