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DOE OSTI2019

Light Water Reactor Sustainability Program — A Summary of Collaborative Research and Development Activities

This report describes the collaborative research and development projects in which the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Program is engaged with industry, regulatory agencies, and other organizations to achieve outcomes that are directed at sustaining the U.S. nuclear power industryduring Fiscal Year 2018. One way that this Program accomplishes its objectives—to develop technologies and other solutions that can improve the reliability, sustain the safety, enhance the economic performance, and extend the life of current reactors—is by conducting research together with industry and key organizations through cooperative arrangements and plans, through coordinated planning and management of resources and activities, through periodic technical exchanges of data and information, or by other means of interaction (i.e., contracts, memoranda of understanding, non-disclosure agreements, etc.). Broadly, these interactions represent different types of collaborations—formal or informal agreements between two or more organizations to work on something together that may be for a short or extended duration and may involve a single purpose or broad set of goals. DOE’s research, development, and demonstration role focuses on enhancing the safe, efficient, and economical performance of the nation’s nuclear fleet while studying and mitigating aging phenomena and issues that are applicable to the service environments of operating reactors and require unique DOE laboratory expertise. Often, research and development (R&D) and demonstration activities are cost-shared, coordinated, or otherwise jointly conducted with industry, regulatory agencies, or other organizations through the aforementioned or other collaborative mechanisms. Operation of the existing fleet of plants to 60 years, extending the operating lifetimes of those plants beyond 60 years and, where practical, making further improvements in their productivity are essential to support the nation’s energy needs. Recently, several utilities have submitted or announced plans to submit applications to the U.S. NRC to begin the subsequent license-renewal process, extending the operating license period beyond 60 years from the date of their initial licensing. This marks an important planned milestone in the history of commercial nuclear power operations in the U.S.—one that underscores the long-term dependability of these plant designs and the commitment to their long-term performance by the organizations that operate them. The LWRS Program will continue to work with owner-operators to address the key issues needed to support the technical bases for continued safe long-term operation of our nation’s nuclear power assets.

42 ENGINEERING↗
DOE OSTI2019

SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) facility. The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Originally aiming at supporting sodium fast reactor (SFR) design, it is now extended to cover other non-light water reactor (non-LWR) concepts, such as high-temperature gas-cooled reactor (HTGR) designs, molten salt-fueled, and molten salt-cooled reactor designs. To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments relevant to these reactor concepts. The Compact Integral Effects Test (CIET) facility, which was designed to reproduce the thermal-hydraulics response of fluoride salt-cooled high-temperature reactors (FHR) under both forced- and natural-circulation conditions, has been identified and selected as one of benchmark test facilities for SAM code validation. Among available CIET experiments, three sets of them with distinctive characteristics were selected for SAM code validation purpose. The three sets experiments are: 1) Power step change transient tests; 2) DHX-DRACS natural circulation tests; and 3) heater frequency response tests. For all tests, steady-state or transient, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Distributed Temperature Sensing within a 19-pin Wire-Wrapped Fuel Bundle Model

It has always been difficult to measure local temperature and flow velocity within fuel bundle geometries because of their complex and compact layouts. High-resolution mapping of the flow field is especially challenging. High-resolution data is typically obtained using optical instruments but optical access to the interior regions of a bundle is severely limited. Velocity fields within fuel bundle models have been mapped with particle image velocimetry (PIV), an optical technique that employs lasers and cameras. Optical access to the bundle interior is obtained using transparent materials and matching the refractive index of the physical structures with that of the fluid. Testing is necessarily isothermal because refractive indices vary with temperature. However, the ultimate aim of such studies is understanding fluid dynamics and heat transfer within real reactors and so flow field mapping under nonisothermal conditions is of particular interest. Temperature fields can be mapped at high-resolution using laser-induced fluorescence (LIF). However, as an optical technique it is unsuitable for nonisothermal liquid flows. This is due to the same sort of optical distortion that restricts PIV. It is also unsuitable for opaque fluids and setups with restricted optical access. Fiber optic distributed temperature sensing (DTS) offers a high-resolution alternative that does not rely upon imaging. A single sensor can map temperature fields at resolutions far exceeding that achievable with conventional sensors such as thermocouples. Though fiber optic sensors cannot match the spatial resolution and bandwidth of LIF, they are suitable for nonisothermal and opaque flows. And though fibers are physical probes that perturb the flow, they can be strung across complex geometries to provide data in regions lacking the optical access necessary for techniques like LIF. Distributed temperature sensing has the potential to provide fuel bundle temperature data of a type that cannot be obtained with optical methods or conventional sensors. This paper reports on preliminary results of efforts to use fiber optic distributed temperature sensors in a fuel bundle model. We are conducting a feasibility study exploring implementation issues such as strain control in the mounting configuration and signal noise associated with flow-induced vibration. We describe here the setup and present preliminary data from our first phase testing with stagnant flow. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4

Using proton irradiation, this study investigates the individual influence of several factors on the corrosion kinetics of Zircaloy-4 in a hydrogenated water environment simulating a Pressurized Water Reactor (PWR). Using both simultaneous irradiation-corrosion and autoclave corrosion, we separately examine: (i) the effect of pre-irradiation on modifying the structure of the material, (ii) the impact of irradiation on creating defects in the growing oxide layer during corrosion, and (iii) the influence of irradiation on increasing the corrosion potential through radiolysis during corrosion. To replicate neutron-irradiated microstructure, two proton pre-irradiation schedules were employed: Schedule 1 (isothermal irradiation at 350 °C to 5 dpa) to simulate high-temperature PWR conditions, and Schedule 2 (two-step process: irradiation to 2.5 dpa at -10 °C followed by 2.5 dpa at 350 °C) to simulate lower temperature PWR and Boiling Water Reactor (BWR) conditions. Long-term autoclave corrosion testing over 360 days at 320 °C revealed no significant difference between unirradiated samples and those pre-irradiated according to either schedule, with all samples exhibiting sub-cubic kinetics within the pre-transition regime. Irradiated samples underwent Simultaneous Irradiation Corrosion (SIC) tests, corroding in 320 °C water while being irradiated with protons. Corrosion was found to accelerate in all SIC-tested samples relative to autoclave conditions, with the greatest increase observed in non-pre-irradiated samples. Pre-irradiation with either schedule resulted in a slower corrosion rate compared to non-pre-irradiated regions under SIC conditions. The degree of radiolysis observed in the SIC tests surpassed typical PWR conditions, approaching levels found in BWRs. Radiolysis products were identified as the primary contributors to accelerated corrosion, corroborated by radiolysis bar tests. Furthermore, these findings underscore the intricate interactions between irradiation, corrosion, and water chemistry in determining Zircaloy-4 corrosion kinetics within nuclear reactor environments.

Pressurized Water Reactor↗
DOE OSTI2025

Ultrasonic Transducer Irradiation Test Results

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 . A multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET ASI) program also provided initial support for this effort. This irradiation, which started in February 2014, is an instrumented lead test and real-time transducer performance data are collected along with temperature and neutron and gamma flux data. The irradiation is ongoing and will continue to approximately mid-2015. To date, very encouraging results have been attained as several transducers continue to operate under irradiation.

Daw, Joshua↗
DOE OSTI2019

Micro Reactor Instrumentation and Control FY2019 Report

Micro-reactors (i.e., very small transportable or mobile nuclear reactors with a capacity less than 20 MWt) are being developed to supply heat and power for various applications in remote areas, military installations, emergency operations, humanitarian missions, and disaster relief zones. A wide variety of reactor types are under consideration, including sodium-cooled fast reactors, molten salt reactors, light water reactors, very-high-temperature gas reactors, and heat pipe reactors. These miniaturized transportable reactor designs remain largely untested and unproven. System and component testing are needed to demonstrate design safety and system robustness, reliability and efficiency.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Water-cooled Reactor Cavity Cooling System Flow Analysis

A research team at University of Wisconsin - Madison designed and constructed a 1/4 scaled experimental facility to study natural circulation cooling in a water based reactor cavity cooling system (WRCCS) for decay heat removal in an advanced high temperature reactor. The facility is capable of natural circulation operation scaled for simulated decay heat removal (up to 45 kW input power, which is equivalent to 1.70 MW at full scale) and pressurized up to 2 bar pressure. The WRCCS facility has been used to study natural circulation flow behavior from transients due to evaporation of the water inventory from decay heat removal. The natural circulation flow is observed to be oscillatory with the mean flow increasing as the pressure head in the system drops due to evaporation of the water inventory during two-phase operation. A first principles model was developed to understand the behavioral limits and trends. Through the use of void fraction measurements and the model, we find that the mass flow rate in the system increases due to small increases in the time-averaged void fraction over the course of a four-hour boil-off experiment. This increase in void fraction causes an increase in the buoyancy head and mass flow. The model also determines the operational limits on the mass flow rate under two-phase operation that can be verified with extended tests in the WRCCS facility. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Development of a Griffin model of the advanced test reactor

In the pursuit of a higher fidelity deterministic simulation capability of the Advanced Test Reactor, it is important to have a fast yet accurate deterministic neutronics model. Here, to achieve this, we employed an advanced two-step method. The first step involves generating homogenized cross sections using OpenMC, a cutting-edge Monte Carlo neutron transport code. OpenMC offers excellent modular capabilities, allowing for easy component integration and flexibility in incorporating new designs into the model. The second step involves deterministic transport calculations, which are performed using Griffin, a reactor physics application based on the Multiphysics Object-Oriented Simulation Environment (MOOSE). To ensure the accurate spatial resolution and assignment of material cross sections, a Cubit-generated mesh for the Advanced Test Reactor is utilized as an intermediate step between the OpenMC and Griffin models; Griffin utilizes the mesh for its finite element solution, while OpenMC material identifications are written to the mesh file to be used in Griffin material assignments. Additionally, a Python-based script converts the cross sections generated by OpenMC into the ISOXML format required by Griffin. Initial comparisons using the Griffin diffusion solver indicated good agreement between the neutron multiplication factors obtained from the standalone OpenMC model and the Griffin model, with differences of less than 10 pcm in the 2D geometry configuration; it was later determined that this agreement was likely due to compensating effect and was more likely on the order of –700 pcm relative to the OpenMC solution. However, in three-dimensional calculations, an unacceptably large error (almost 8,000 pcm) was found in the Griffin solution with the diffusion solver. Subsequent calculations using Griffin’s discrete ordinates solver demonstrated substantially improved agreement, within 116 pcm of the OpenMC solution used to generate the cross sections for Griffin. Building on this capability, future work will seek to perform more detailed validation calculations. The ultimate goal is to evaluate both transient and multiphysics simulations of the reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3 Si 2 ) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3 Si 2 has been optimized and high phase purity U 3 Si 2 has been successfully produced. Results are presented from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3 . Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Delayed Fission Gamma-ray Characteristics of 232 Th, 233 U, 235 U, 238 U, and 239 Pu

Delayed fission gamma-rays play an important role in determining the time dependent ionizing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from activation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is necessary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution to the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray characteristics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 and experimental data and other published literature, including ENDF/B-VII.1. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Physics Validation Measurements for Highly-Reactive Experiment Packages in the INL Advanced Test Reactor

The Korea Atomic Energy Research Institute is currently in the process of qualifying the Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element will be irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

MEDE Project

Presentation on the Melt-Drain-Evaporation Testing associated with a non-irradiated Fermi-1 reactor assembly

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

NDMAS

Overview of Current ART-GCR Data: Fuel Fabrication, Irradiation Monitoring (Fuel & Graphite – near real-time for HDG-1), Post-Irradiation Examination (Fuel & Graphite), Graphite Characterization (Baseline and Irradiated), High Temperature Metals Mechanical Tests, Design, Methods, and Validation Data, Japan Atomic Energy Agency’s High Temperature Test Reactor (HTTR), Argonne National Laboratory’s Natural convection Shutdown heat removal Test Facility (NSTF), Oregon State University’s High Temperature Test Facility (HTTF), Generation IV International VHTR Materials Handbook, Additional related data, and Advanced Test Reactor operations (near real-time).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Pressure drop measurements across a helical coil steam generator configuration - 15997

The primary heat transport loop heat exchanger within a Very-High Temperature Reactor (VHTR) is most commonly a Helical Coil Steam Generator (HCSG). This design has been regarded as having increased heat transfer within a smaller design in comparison to other heat exchangers. The fluidic heat transfer properties of the HCSG is still under evaluation in tangent with HTRs due to its complex design. Heat transfer between these layers relies heavily on their relative location and fluidic properties. A model was constructed to study the pressure field on a rod s surface as well as across multiple layers of helically facing tubes. Experiments focused on evaluating the impact relative geometry played on the overall pressure drop of the test section with inlet fluid velocity ranging from 0 to 3 m/s. A study to compare the experimental results with the Gaddis-Gnielinski Correlative theoretical values of pressure drop in tube bundles shows that the complex helical geometry results in pressure drops similar to that of fully developed rod bundle although the test facility is comprised of two half rod helical coils. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Versatile Test Reactor Conceptual Core Design

The VTR is a 300-MW(thermal) sodium-cooled fast reactor (SFR) designed for the specific purpose of delivering unique testing capabilities to enable the advancement of all reactor technologies. With its flux level, irradiation volume, and operational flexibility, the VTR will enable accelerated testing of materials, fuels, and various components needing irradiation testing. Proven SFR technologies and design approaches have been leveraged in designing the VTR core, ensuring the highest possible readiness level. This resulted in the VTR using ternary metallic fuel and delivering fast flux levels in excess of 4 x 10 15 n/cm 2 ∙ s over large useful volumes, corresponding to about 60 dpa/year in steel. As part of the design efforts, the VTR core performance has been determined for a representative configuration, ensuring that the reactivity control systems offer sufficient shutdown margins, that the core can be safely cooled in all situations, and that reactivity feedback coefficients are conducive to a favorable safety behavior. Furthermore, the incorporation of features such as fuel assembly storage in the shield region supports the flexible and reliable operation of the VTR. Additional design work has been ongoing as well. This includes thorough shielding performance evaluations to ensure safe operation of the VTR, verification and validation of the design tools used to achieve compliance with Nuclear Quality Assurance (NQA-1) requirements, early assessment of the impact of irradiation experiments on the core performance envelope and associated margins, and in-depth uncertainty quantification efforts to quantify the anticipated range of performance characteristics. An experimental program supporting the VTR core design has been set up, with the current focus being on thermal-hydraulic experiments. The purpose of this experimental program is to obtain confirmatory measurements to serve directly as part of the core design basis or as part of the validation cases supporting the simulation tools used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

High Temperature Gas-Cooled Test Reactor Options Status Report

Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

VeraShift Usability Updates for Production Releases of VERA for Ex-core Calculations

The recently developed capability to perform ex-core calculations in the Virtual Environment for Reactor Applications (VERA) with Shift is now available in production releases of VERA. To support these production releases, updates to the VeraShift interface are needed continually. The updates completed under this milestone are documented in this report, including improvements to memory usage and code robustness, as well as the resolution of issues with unexpected behavior that arose during testing.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Comparison of Calculated and Experimental Flows and Pressures in the Orifice Flow Test Facility

A new experimental facility for materials irradiation and testing at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is being developed. Details of this facility have been presented in previous work. A prototype of this facility, the Thermosyphon Test Loop (TSTL), has been built and experimental data have been obtained and analyzed. Comparison of TSTL experimental data and calculations has been presented previously. The TSTL did not measure any flows (water or vapor) inside the test section because flowmeters could not be installed inside. Instead, two pressure sensors (P1 and P2) were installed at the bottom of the test section to measure the pressure differential ({Delta}P = P1-P2) across the orifices between the bottom of the downcomer and the bottom of the boiler. This {Delta}P is proportional to the water flow, but actual flows were not measured. Thus, another facility, the Orifice Flow Test Facility (OFTF), was built to measure water flow vs {Delta}P. The OFTF duplicates the bottom of the test section of the TSTL. The OFTF models the downcomer and the boiler including three dummy heaters (without power) inside the boiler. This paper presents the experimental data obtained and comparison to computer calculations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗