Search NASA⌕ Search

SEARCH · Search NASA

Results for “component testing”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Design of Modular Test Facility for Particle/sCO2 Heat Exchanger Evaluation

Heliogen has designed a 1.3 MW particle and supercritical carbon dioxide (sCO2) test loop to retire technical and manufacturing risks. The data from this facility will be used to validate near commercial-scale particle heat exchanger modules. The heat exchanger is scaled to capture all features of the full-scale modules. Testing will be conducted with CARBOBEAD HSP 16/30, a larger particle size than other recent testing. The system was designed to ASME codes with the constraint that sCO2 piping and other major components are stainless steels.

14 SOLAR ENERGY↗
DOE OSTI2021

Draft Rules for Alloy 617 Creep-Fatigue Design Using an EPP+SMT Approach

The report describes an improved new creep-fatigue design method for structural components in high temperature nuclear service. The new method uses an integrated elastic-perfectly plastic (EPP) analysis and Simplified Model Test (SMT) approach that greatly simplifies the design process by avoiding the separate evaluation of creep and fatigue damages and eliminating the requirement of stress classification. The report includes draft design rules, presented in a format compatible with an ASME nuclear Code Case, as well as a commentary on the rules and the validation of the rules using pressurized SMT (p-SMT) test data. Areas of potential improvement have been identified for further evaluation. The report also develops EPP+SMT design charts for Alloy 617 at temperatures between 800°C and 950°C by extrapolating high strain range, short hold test data to low strain range and long hold times, typical for structural components in high temperature nuclear service. The design charts are currently limited to 10,000 cycles to avoid excessive extrapolation outside the test database. More robust alternate extrapolation procedures will be considered in future work. The report includes several verification problems comparing the new EPP+SMT design method with current ASME creep-fatigue design methods. The comparison demonstrates that the EPP+SMT method significantly reduces the over-conservatism in current methods. Finally, the report includes a fully-documented sample problem with multiple load cases detailing the application of proposed EPP+SMT design rules.

36 MATERIALS SCIENCE↗
DOE OSTI2016

FY 2016 Status Report: CIRFT Testing Data Analyses and Updated Curvature Measurements

This report provides a detailed description of FY15 test result corrections/analysis based on the FY16 Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) test program methodology update used to evaluate the vibration integrity of spent nuclear fuel (SNF) under normal transportation conditions. The CIRFT consists of a U-frame testing setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages to a universal testing machine. The curvature of rod bending is obtained through a three-point deflection measurement method. Three linear variable differential transformers (LVDTs) are used and clamped to the side connecting plates of the U-frame to capture the deformation of the rod. The contact-based measurement, or three-LVDT-based curvature measurement system, on SNF rods has been proven to be quite reliable in CIRFT testing. However, how the LVDT head contacts the SNF rod may have a significant effect on the curvature measurement, depending on the magnitude and direction of rod curvature. It has been demonstrated that the contact/curvature issues can be corrected by using a correction on the sensor spacing. The sensor spacing defines the separation of the three LVDT probes and is a critical quantity in calculating the rod curvature once the deflections are obtained. The sensor spacing correction can be determined by using chisel-type probes. The method has been critically examined this year and has been shown to be difficult to implement in a hot cell environment, and thus cannot be implemented effectively. A correction based on the proposed equivalent gauge-length has the required flexibility and accuracy and can be appropriately used as a correction factor. The correction method based on the equivalent gauge length has been successfully demonstrated in CIRFT data analysis for the dynamic tests conducted on Limerick (LMK) (17 tests), North Anna (NA) (6 tests), and Catawba mixed oxide (MOX) (10 tests) SNF samples. These CIRFT tests were completed in FY14 and FY15. Specifically, the data sets obtained from measurement and monitoring were processed and analyzed. The fatigue life of rods has been characterized in terms of moment, curvature, and equivalent stress and strain..

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Summary of Methodology for Mitigating Risks Associated with Licensing and Qualifying AM Nuclear Materials

The US Department of Energy’s Advanced Materials and Manufacturing Technologies (AMMT) program focuses on accelerating the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy. Laser powder bed fusion (LPBF) is one of the most popular additive manufacturing (AM) processes for fabricating components with intrinsically complex geometries. LPBF was extensively explored for nuclear applications under the previous Transformational Challenge Reactor program. Additionally, Oak Ridge National Laboratory developed and licensed the Peregrine software and larger digital platform that couples machine learning and in situ data collection during AM to detect anomalies and any evolved defects. The digital platform will be critical to (1) the qualification of AM components for nuclear applications that link location-specific data to macroscopic properties and (2) predict final component performance. Current in situ process monitoring tools are valuable for observing the formation of stochastic flaws, but additional data are needed to predict the resulting microstructures and associated material performance. Rapid cooling rates and large thermal gradients have caused large heterogeneities in the microstructure, which cause anisotropy in mechanical performance. The AMMT program is evaluating the best approaches for addressing these heterogeneities and their effect on component performance using a combination of multiscale modeling, enhanced in situ process monitoring, and high throughput experimental testing. This report summarizes strategies for mitigating the risks associated with qualifying AM components, including developing new sensing capabilities for in situ process monitoring and characterizing melt pool solidification and residual stresses to inform multiscale modeling efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Spent nuclear fuel dynamic performance under normal conditions of transport

The objective of this project is to evaluate the spent nuclear fuel (SNF) vibration integrity under normal condition of transport. The SNF fatigue strength data has been collected from various pressurized water reactors. The CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester) has been used to simulate transportation environments. CIRFT is an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The CIRFT consists mainly of a U-frame testing setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages to a universal testing machine. The curvature of rod bending is obtained through a three-point deflection measurement method. Three linear variable differential transformers (LVDTs) are used and clamped to the side connecting plates of the U-frame to capture the deformation of the rod. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response. The major findings of CIRFT test on the high burn-up SNF are as follows: -) SNF system interface bonding plays an important role in SNF vibration performance; -) Fuel structure contributes to the SNF system stiffness; -) There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and -) SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

42 ENGINEERING↗
DOE OSTI2022

High-Fidelity Mock Development for the Insensitive High Explosive TATB

Due to the violent reaction potential of high explosives, in tests where a reaction is not desired, it can be safe and convenient to use an inert mock that can mimic relevant properties of the associated explosive. Use of mock materials is well-established in explosives work, but most existing mocks are generally meant to mimic only one or two properties of a given explosive, potentially failing to match the explosive in other critical ways. Recently, there has been interest in developing robust thermomechanical mocks for many common explosives. Here, eleven materials are systematically chosen and investigated as potential thermomechanical mocks for TATB, the explosive component of PBX 9502. DSC, nanoindentation, and compression testing are used to narrow down to most appropriate mocks and identify their advantages and limitations. Hardness and elastic modulus are reported for the first time for the molecular crystals iodoanthranilic acid, hexamine, melamine, and trithiolane dioxide, and the crystal structure is reported for the first time for the molecular crystals iodoanthranilic acid and iodosalicylic acid.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Designing Cladded Components for High Temperature Nuclear Service: Part I—Analysis Methods

Use of corrosion-resistant cladding can greatly extend the design life of structural components in many advanced reactor systems. However, there are currently no ASME design rules for cladded components to guard against creep-fatigue failure and ratcheting strain accumulation in elevated temperature nuclear service. This paper, presented in two parts, addresses this gap by proposing a design strategy for cladded components that do not require long-term testing of clad materials. The proposed approach relies on approximate design analysis methods for two types of clad materials—a soft clad that creeps much faster than and has lower yield stress than the class A base material and a hard clad that creeps much slower than and has higher yield stress than the class A base material. Part I discusses the approximate analysis strategies for the clad materials—treat a soft clad as perfectly compliant and a hard clad as linear elastic—and Part II develops a complete set of design rules for each of the two types of cladded components. In conclusion, finite element analyses of representative high temperature reactor components show that the proposed design analysis methods can bound the design quantities in soft cladded components and approximate the design quantities in hard cladded components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Robust foil regenerator flow loss and heat transfer tests under oscillating flow condition

A regenerator works as an energy storage component in a Stirling engine. Thus, it is one of the most critical components that makes the Stirling engine highly efficient. Traditionally, the regenerators are made of wire media such as woven screens and random fibers. These wire regenerators experience considerably high flow losses through the regenerator due to the flow separation and the resulting vortices. The high flow losses associated with the flow separation and vortices result in a lower engine efficiency. This research is to test a new robust foil regenerator to investigate its feasibility for free-piston Stirling engine applications. The robust foil regenerator is configured with parallel flow channels and stiffening ribs. Comparing to complex flows patterns in woven screen and random fiber regenerators, the flows in the parallel plates are simple and the flow losses are considerably small, and the heat transfer coefficient is also relatively small. To test the regenerator experimentally, a regenerator test rig was manufactured. To verify the accuracy of the test rig, a woven screen regenerator was made and tested prior to the robust foil regenerator test. The friction factor correlation of the woven screen regenerator shows a good agreement with the published regenerator correlations. The robust foil regenerator tests were conducted under the oscillating flow condition with different Reynolds numbers. The flow losses were measured with dynamic pressure transducers and the friction coefficient was derived from measured pressure drop. The measured temperature difference over the regenerator was used to calculate the heat transfer coefficient and Nusselt number. The figure of merit with two methods of NPH/NTU ratio and j/f ratio was calculated to evaluate the suitability of the robust foil regenerator. The results show that the robust foil regenerator has considerably higher figure of merit and therefore, better performance than other types of regenerator.

42 ENGINEERING↗
DOE OSTI2017

Tracking the Key Constituents of Concern of the WTP LAW Stream

The testing results presented in the present report were also obtained on a DM10 melter system operated with the primary WTP LAW offgas system components with recycle, as specified in the statement of work (SOW) [6] and detailed in the Test Plan for this work [7]. The primary offgas system components include the SBS, the WESP, and a recycle system that allows recycle of liquid effluents back to the melter, as in the present baseline for the WTP LAW vitrification. The partitioning of technetium and other key constituents between the glass waste form, the offgas system liquid effluents, the offgas stream that exits the WESP, and the liquid condensate from the vacuum evaporator were quantified in this work. The tests employed three different LAW streams spanning a range of waste compositions anticipated for WTP. Modifications to the offgas system and operational strategy were made to expedite the approach to steady state concentrations of key constituents in the glass and offgas effluent solutions during each test.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Extraction, Scrub, and Strip Test Results for the IBC BOBCalix Solvent Samples

A second vendor for solvent extraction components has been chosen for use at the Salt Waste Processing Facility (SWPF). IBC Advanced Technologies (IBC) has provided samples of their BOBCalix, as well as complete solvent. These samples were tested for quality assurance purposes to ensure their use at SWPF is acceptable. Two Extraction, Scrub, and Strip (ESS) tests were performed on solvent samples from IBC. The purpose of testing these samples was to determine if the solvents are displaying the correct extraction and stripping behavior with cesium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

RELAP-7 Pressurizer Component Development Updates

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory (INL). RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). As part of the efforts to expand the capability for PWR simulation, an equilibrium single-region pressurizer model has been implemented in RELAP-7. The pressurizer component can simulate pressure and water level change through insurge, spray, and heating processes. Two simple tests – one for insurge process and another for outsurge process – have been reported to demonstrate and verify the functions of the pressurizer model. The typical single-phase PWR system model presented in the first RELAP-7 milestone report has been updated, as part of system level test for the new pressurizer model. The updated PWR system model with the pressurizer component can be used for more realistic transient simulations. The addition of the equilibrium single-region pressurizer model represents the first step of developing a suite of pressurizer models with different levels of fidelities. Non-equilibrium two-region two-zone model and non-equilibrium five-region three zone model will be implemented in the near future. Cross comparison and validation will be pursued when higher fidelity pressurizer models are completed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

VERA 3.6 Release

CASL has developed a formalized process to release the Virtual Environment for Reactor Applications (VERA) to non-CASL users through the Radiation Shielding Information Computational Center (RSICC). This milestone represents a major release of VERA with updates that significantly reduce computational resources required and introduce new capabilities related to fuel performance integration. This release will also be used by CASL Test Stands and training activities. VERA 3.6 represents an incremental release, with improvements to core component capabilities, significant performance improvements, and updated documentation. This release followed a significantly revised release process, with formal documentation, testing, readiness reviews, and release authorization. The major tasks that were completed as part of this milestone were: 1) Revision of the VERA RSICC Release Process and Procedures; 2) Identification of the VERA release components; 3) Creation of a Release Candidate; 4) Independent evaluation of the Release Candidate by the Advanced Modeling Applications (AMA) Focus Area; 5) A Release Candidate Readiness Review; 6) Development and execution of a System Test Plan; 7) Updates of manuals and other required documentation; 8) Export Control reviews; 9) A Release Readiness Review; and 10) Delivery of the release package to RSICC. Synopses of the work done to complete each of these major tasks are summarized in Section 2 below. Supporting details are provided in the PHI Kanban ticket #4424 Support VERA 3.6 Release Activities and its child tickets.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

An observational constraint on stomatal function in forests: evaluating coupled carbon and water vapor exchange with carbon isotopes in the Community Land Model (CLM4.5)

Abstract. Land surface models are useful tools to quantify contemporary and future climate impact on terrestrial carbon cycle processes, provided they can be appropriately constrained and tested with observations. Stable carbon isotopes of CO2 offer the potential to improve model representation of the coupled carbon and water cycles because they are strongly influenced by stomatal function. Recently, a representation of stable carbon isotope discrimination was incorporated into the Community Land Model component of the Community Earth System Model. Here, we tested the model's capability to simulate whole-forest isotope discrimination in a subalpine conifer forest at Niwot Ridge, Colorado, USA. We distinguished between isotopic behavior in response to a decrease of δ13C within atmospheric CO2 (Suess effect) vs. photosynthetic discrimination (Δcanopy), by creating a site-customized atmospheric CO2 and δ13C of CO2 time series. We implemented a seasonally varying Vcmax model calibration that best matched site observations of net CO2 carbon exchange, latent heat exchange, and biomass. The model accurately simulated observed δ13C of needle and stem tissue, but underestimated the δ13C of bulk soil carbon by 1–2 ‰. The model overestimated the multiyear (2006–2012) average Δcanopy relative to prior data-based estimates by 2–4 ‰. The amplitude of the average seasonal cycle of Δcanopy (i.e., higher in spring/fall as compared to summer) was correctly modeled but only when using a revised, fully coupled An − gs (net assimilation rate, stomatal conductance) version of the model in contrast to the partially coupled An − gs version used in the default model. The model attributed most of the seasonal variation in discrimination to An, whereas interannual variation in simulated Δcanopy during the summer months was driven by stomatal response to vapor pressure deficit (VPD). The model simulated a 10 % increase in both photosynthetic discrimination and water-use efficiency (WUE) since 1850 which is counter to established relationships between discrimination and WUE. The isotope observations used here to constrain CLM suggest (1) the model overestimated stomatal conductance and (2) the default CLM approach to representing nitrogen limitation (partially coupled model) was not capable of reproducing observed trends in discrimination. These findings demonstrate that isotope observations can provide important information related to stomatal function driven by environmental stress from VPD and nitrogen limitation. Future versions of CLM that incorporate carbon isotope discrimination are likely to benefit from explicit inclusion of mesophyll conductance.

58 GEOSCIENCES↗
DOE OSTI2021

ASME Code Rules and ASTM Standards Integration for Ceramic Composite Core Materials and Components 1

Fiber-reinforced ceramic matrix composites have many desirable properties for high-temperature nuclear applications, including excellent thermal and mechanical properties and reasonable to outstanding radiation resistance. Over the last 20 years, the use of ceramic composite materials has already expanded in many commercial nonnuclear industries as fabrication and application technologies mature. The new ASME design and construction rules under Section III, Subsection HH, Subpart B lay out the requirements and criteria for materials, design, machining and installation, inspection, examination, testing, and the marking procedure for ceramic composite core components, which is similar to the established graphite code under Section III, Subsection HH, Subpart A. Moreover, the general requirements listed in Section III, Subsection HA, Subpart B are also expanded to include ceramic composite materials. The code rules rely heavily on the development and publication of standards for composite specification, classification, and testing of mechanical, thermal, and other properties. These test methods are developed in the American Society for Testing and Materials Committee C28 on Advanced Ceramics with a current focus on ceramic composite tubes. Details of the composites code, design methodology, and similarities to the graphite code, as well as guidance for the development of specifications for ceramic composites for nuclear application and recent standard developments, are discussed. The next step is to "close the gap" to support licensing aspects by validating the code with benchmarking data.

Geringer, Josina↗
DOE OSTI2016

In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening

Argonne National Laboratory (ANL), under the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316SS) material which is widely used in the US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In a second paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.

Cyclic plasticity material modeling↗
DOE OSTI2016

FY 2016 Status Report: Documentation of All CIRFT Data including Hydride Reorientation Tests (Draft M2)

The first portion of this report provides a detailed description of fiscal year (FY) 2015 test result corrections and analysis updates based on FY 2016 updates to the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) program methodology, which is used to evaluate the vibration integrity of spent nuclear fuel (SNF) under normal conditions of transport (NCT). The CIRFT consists of a U-frame test setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages connecting to a universal testing machine. The curvature SNF rod bending is obtained through a three-point deflection measurement method. Three linear variable differential transformers (LVDTs) are clamped to the side connecting plates of the U-frame and used to capture deformation of the rod. The second portion of this report provides the latest CIRFT data, including data for the hydride reorientation test. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNF samples into a single zone. A detailed examination revealed that, at the same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), LMK, and mixed uranium-plutonium oxide (MOX). Just looking at the strain, LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The third portion of this report provides finite element analysis (FEA) dynamic deformation simulation of SNF assemblies . In a horizontal layout under NCT, the fuel assembly’s skeleton, which is formed by guide tubes and spacer grids, is the primary load bearing apparatus carrying and transferring vibration loads within an SNF assembly. These vibration loads include interaction forces between the SNF assembly and the canister basket walls. Therefore, the integrity of the guide tubes and spacer grids critically affects the vibration intensity of the fuel assembly during transport and must be considered when developing the multipurpose purpose canister (MPC) design for safe SNF transport.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

SU-F-BRA-13: Knowledge-Based Treatment Planning for Prostate LDR Brachytherapy Based On Principle Component Analysis

Purpose: To create a knowledge-based algorithm for prostate LDR brachytherapy treatment planning that standardizes plan quality using seed arrangements tailored to individual physician preferences while being fast enough for real-time planning. Methods: A dataset of 130 prior cases was compiled for a physician with an active prostate seed implant practice. Ten cases were randomly selected to test the algorithm. Contours from the 120 library cases were registered to a common reference frame. Contour variations were characterized on a point by point basis using principle component analysis (PCA). A test case was converted to PCA vectors using the same process and then compared with each library case using a Mahalanobis distance to evaluate similarity. Rank order PCA scores were used to select the best-matched library case. The seed arrangement was extracted from the best-matched case and used as a starting point for planning the test case. Computational time was recorded. Any subsequent modifications were recorded that required input from a treatment planner to achieve an acceptable plan. Results: The computational time required to register contours from a test case and evaluate PCA similarity across the library was approximately 10s. Five of the ten test cases did not require any seed additions, deletions, or moves to obtain an acceptable plan. The remaining five test cases required on average 4.2 seed modifications. The time to complete manual plan modifications was less than 30s in all cases. Conclusion: A knowledge-based treatment planning algorithm was developed for prostate LDR brachytherapy based on principle component analysis. Initial results suggest that this approach can be used to quickly create treatment plans that require few if any modifications by the treatment planner. In general, test case plans have seed arrangements which are very similar to prior cases, and thus are inherently tailored to physician preferences.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

The SNS Moderator Demonstration Facility

We describe a moderator demonstration facility at the Spallation Neutron Source (SNS). We will use the Beam Test Facility (BTF) at SNS to provide a moderator neutronics test stand with which we will verify the performance gains expected from the innovative moderator concepts central to the SNS Second Target Station. These concepts include large-volume para-hydrogen moderators and high-brightness moderators, as well as new moderator materials. The BTF provides a test facility for the new Radio-Frequency Quadrupole accelerator (RFQ) to be installed at SNS and produce a pulsed H− beam of 2.5 MeV energy. We will add to the BTF a beam chopper similar to that already used in the SNS at the RFQ exit, various beam transport components, a neutron-producing target, a cryogenic moderator test stand, a reflector-shielding assembly, and a performance assessment neutron beamline.

Iverson, Erik B.↗