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All sourcesNASA NTRSDOE OSTI
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 OSTI2017

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Pellet Cladding Interaction In-Reactor Ramp Testing in a World without the Halden Boiling Water Reactor

One of the most crucial performance areas for fuel rods in water cooled nuclear power plants is interaction between cladding tubes and fuel pellets. Experimental programs in test reactors have provided key data to help fuel developers and plant operators understand Pellet Cladding Interaction (PCI) phenomena and optimize their strategies for reliable fuel performance. Approximately 50 years of “ramp” testing programs, where the fission heating rate is deliberately manipulated in test rods, have been performed in a handful of test reactors to reveal and understand PCI behaviors such as iodine-assisted stress corrosion cracking. Unfortunately, the test reactors most engaged in this type of work have all been retired over the years up to the recent and unexpected closure of the Halden Boiling Water Reactor which effectively caused a hiatus in PCI ramp testing programs. The need for ramp testing is crucial at this time to enable refined PCI understanding as more plants consider implementing flexible operations, increased fuel rod burnup limits, and new fuel technologies with enhanced accident tolerance. This paper reviews some of the key PCI phenomena that must be addressed with in-pile testing and surveys past test reactor’s methods for achieving the needed conditions. A strategic approach is then presented using test reactors and complimentary facilities which are still available today. Near term data opportunities are put forth along with capability development strategies that will ensure future longevity in this field of research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Sensitivity Analysis of OECD ATLAS A5.1 Test using MARS-KS code

The ATLAS was designed to model a reduced-height primary system of a typical APR1400 which was developed by the Korean Nuclear Industry. The ATLAS has the 1/2-height, 1/144-area and 1/288-volume scales for the APR 1400 and it is scaled for full pressure and temperature conditions. A benchmark test, test A5.1, was performed to analyze and quantify the extent of scaling distortion in the ATLAS, as a part of a OECD-ATLAS project. The SB-CL-32 test which was conducted using the Large Scale Test Facility (LSTF) of the Rig-of-Safety Assessment-V (ROSA-V) Program was selected as the counterpart test to address the scaling issue. The detailed description of the LSTF test facility and the SB-CL-32 test can be found in the literature. The objective of the A5.1 test is to investigate the thermal hydraulic phenomena during cold leg small break LOCA with steam generator secondary side depressurization as the Accident Management action. In addition to that, the scaling effect analysis of ATLAS was performed in the several viewpoints. In this study, sensitivity analysis of A5.1 test was performed using MARS-KS code. Several experimental conditions of A5.1 test were varied to study effects on the reactor coolant system during the SBLOCA transient. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Thermal Testing of Waste Packages for a Storage Environment

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard TRU waste to the Waste Isolation Pilot Plant (WIPP) for disposal. The POC consists of an inner Pipe Container (PC) surrounded by organic fiber dunnage inside of a 7A drum. The PC was design to maintain separation of fissile material and to provide shielding from radiation. The POC has many design features that are similar to Type B packages that do provide protection from engulfing fires. However, it was not specifically designed to prevent release of stored material during a long-duration fire. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites (e.g., Waste Isolation Pilot Plant). Radioactive waste storage operators would like to be able to claim some level of protection is provided by the POCs for thermal assaults that could occur within storage facilities when these POCs contain PCs with combustibles. In 1997 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provide during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drums generating sufficient internal pressure to pop off its lid and expose the top of the PC to the fire environment. However, these tests are not bounding for POC/PC containing combustibles. With the addition of combustible materials in the PC, the possibility exists for generation of higher quantities of gases and much higher pressures inside the PC, which can lead to early rupture of the PC filter and higher radioactive aerosol release fractions. Therefore, a new series of tests was required to determine the temperature response of the exposed PC and determine the amount of material that could be released. DOE contacted SNL in 2015 to discuss the 1997 SNL tests and came to the conclusion that those tests were not bounding for POC/PC containing combustibles Therefore, DOE commission a new set of tests at SNL. Starting in 2015, SNL conducted fire tests to examine whether the contents of the PC would reach a temperature that would result in generation of sufficient gas to cause overpressurization of the PC, and subsequent release of its content. This paper will describe the various test conducted thus far, and present preliminary results from these tests.

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

FORCE Regression Testing

Via programs including the Light Water Reactor Sustainability and Integrated Energy Systems, the U.S. Department of Energy has invested in the Framework for Optimization of ResourCes and Economics (FORCE) software framework (Idaho National Laboratory 2024a) for the technical and economic analysis of nuclear-integrated energy systems (IES). Nuclear IES expand the use of nuclear from traditional baseload electricity generation to a flexible and adaptive source of combined heat and power. Nuclear heat can be used in the production of a variety of energy currencies such as hydrogen and ammonia as well as other heat applications including water desalination and district heating. FORCE is designed with the intent to provide interconnected analysis tools that enable the accurate technical and economic assessment of specific nuclear IES configurations for individual energy markets. FORCE consists of three main analysis pathways: HYBRID (Idaho National Laboratory 2024b), which contains high-resolution physical models for IES; Holistic Energy Resource Optimization Network (HERON) (Idaho National Laboratory 2024c), which analyzes IES long-term economic viability; and Optimization of Real-time Capacity Allocation (ORCA) (Idaho National Laboratory 2024d), designed for real-time control of IES via digital twins and optimal decision making, including autonomous and remote operation research. Development of the FORCE ecosystem is guided by three pillars: capability, which assures that the computational requirements of IES analysis are met by the software tools; reliability, which provides for consistent code performance and expected behaviors; and accessibility, which lowers the barrier to entry for using the software and accelerates analysis by users beyond the FORCE primary developers. Reliability of the FORCE ecosystem is established according to the American Nuclear Society?s Nuclear Quality Assurance (NQA-1) program [American Society of Mechanical Engineers 1982], with specific levels of software quality assurance (SQA) within NQA-1 applied to each software tool in FORCE. As the tools within FORCE have matured, some integration algorithms to accurately connect the software tools for holistic analysis have been developed and deployed within the FORCE software repository. In accordance with NQA-1 standards, regression tests are required to guarantee the software performs consistently even when new capabilities are added to the software. In this report, we document the deployment of both unit tests, which test the consistent behavior of small pieces of the FORCE code base, as well as integration tests, which test the consistent performance of full use cases for the FORCE integration algorithms. We further document the encapsulation of these tests within a test harness, which collectively checks for each successful test completion on demand. Finally, we document the automation of the test harness using GitHub Actions [GitHub 2024], which require all tests succeed before any new capability or other changes can be added to the FORCE integration software

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

Mobility of heavy metals through granitic soils using mini column infiltration test

This study is about the mobility of cadmium through compacted granitic soils. Two granitic soils namely the Broga (BGR) and Kajang (KGR) granitic soils were collected in Selangor, Malaysia. Physical and chemical tests were applied for both granitic soils to determine the physical and chemical properties of soil materials. Physical test results shows granitic soils (BGR and KGR) have high percentage of sand ranging between 54%–63% and 46%–54% respectively, an intermediate and intermediate to high plasticity index as well as high specific gravity ie; 2.50–2.59 and 2.45–2.66 respectively. For chemical test, granitic soils shows acidic pH values ranged from 5.35–5.85 for BGR and pH 5.32–5.54 for KGR. For organic matter, SSA and CEC test, it shows low values ranged from 0.22%–0.34% and 0.39%– 0.50% respectively for organic matter test, 17.96 m{sup 2}/g–21.93 m{sup 2}/g and 25.76 m{sup 2}/g–26.83 m{sup 2}/g respectively for SSA test and 0.79 meq/100g–1.35 meq/100g and 1.31 meq/100g–1.35 meq/100g respectively for CEC test. Mini column infiltration test was conducted to determine the retention of cadmium while flowing through granite soils. This test conducted based on the falling head permeability concepts. Different G-force ranging from 231G to 1442G was used in this test. The breakthrough curves show the concentration of Cd becomes higher with the increasing of G-force for both granitic samples (BGR and KGR). The selectivity sorption for both granites ranked in the following decreasing order of; 231G>519G>923G>1442G. Results demonstrated that granitic soils also have low buffering capacity due to low resist of pH changes.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Parameter estimation with Sandage-Loeb test

The Sandage-Loeb (SL) test directly measures the expansion rate of the universe in the redshift range of 2 ∼< z ∼< 5 by detecting redshift drift in the spectra of Lyman-α forest of distant quasars. We discuss the impact of the future SL test data on parameter estimation for the ΛCDM, the wCDM, and the w{sub 0}w{sub a}CDM models. To avoid the potential inconsistency with other observational data, we take the best-fitting dark energy model constrained by the current observations as the fiducial model to produce 30 mock SL test data. The SL test data provide an important supplement to the other dark energy probes, since they are extremely helpful in breaking the existing parameter degeneracies. We show that the strong degeneracy between Ω{sub m} and H{sub 0} in all the three dark energy models is well broken by the SL test. Compared to the current combined data of type Ia supernovae, baryon acoustic oscillation, cosmic microwave background, and Hubble constant, the 30-yr observation of SL test could improve the constraints on Ω{sub m} and H{sub 0} by more than 60% for all the three models. But the SL test can only moderately improve the constraint on the equation of state of dark energy. We show that a 30-yr observation of SL test could help improve the constraint on constant w by about 25%, and improve the constraints on w{sub 0} and w{sub a} by about 20% and 15%, respectively. We also quantify the constraining power of the SL test in the future high-precision joint geometric constraints on dark energy. The mock future supernova and baryon acoustic oscillation data are simulated based on the space-based project JDEM. We find that the 30-yr observation of SL test would help improve the measurement precision of Ω{sub m}, H{sub 0}, and w{sub a} by more than 70%, 20%, and 60%, respectively, for the w{sub 0}w{sub a}CDM model.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2022

ADAPTIVE GROUP TESTING WITH MISMATCHED MODELS

Accurate detection of infected individuals is one of the critical steps in stopping any pandemic. When the underlying infection rate of the disease is low, testing people in groups, instead of testing each individual in the population, can be more efficient. In this work, we consider noisy adaptive group testing design with specific test sensitivity and specificity that select the optimal group given previous test results based on pre-selected utility function. As in prior studies on group testing, we model this problem as a sequential Bayesian Optimal Experimental Design (BOED) to adaptively design the groups for each test. We analyze the required number of group tests when using the updated posterior on the infection status and the corresponding Mutual Information (MI) as our utility function for selecting new groups. More importantly, we study how the potential bias on the ground-truth noise of group tests may affect the group testing sample complexity.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Fabrication, Welding, and Inspection Techniques in Preparation for In-Reactor Testing of Accident Tolerant Fuels

After the Fukushima events in 2011, DOE-NE in collaboration with nuclear industry shifted R&D emphasis to accident performance of LWR fuels under extended loss of active cooling and steam exposure. DOE-NE has created a roadmap for the “Development of Light Water Reactor Fuels with Enhanced Accident Tolerance.” The mission of the Accident Tolerant Fuel (ATF) Roadmap is to develop the next generation of LWR fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. The ultimate goal of the ATF roadmap is to support the insertion of lead fuel rods (LFRs) or lead fuel assemblies (LFAs) of an Accident Tolerant Fuel into a commercial LWR within 10 years (i.e., by the end of FY-2022). As a step toward this goal, an irradiation test series has been developed to assess the performance of proposed ATF concepts under normal LWR operating conditions. Data generated by this test program will be used to establish the feasibility of certain aspects of proposed ATF concepts, as well as provide information to support screening among concepts; as such, it is an integral part of Phase I: Feasibility Assessment and Down-Selection outlined in the ATF Roadmap. This irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL), and it has been designated as the ATF-1 test series. Current fission reactors use zirconium-based fuel cladding because of its extremely low macroscopic thermal neutron absorption cross-section, good high temperature strength, and decent corrosion resistance. However, advanced, innovative materials may provide these same benefits while increasing reactor safety margin, core power density, and fuel utilization. These advanced fuel cladding systems will allow revolutionary cladding performance and enhanced fuel mechanical designs, however, challenges exist in design, analysis and fabrication of innovative, never before tested, fuel cladding systems for in-reactor testing. This paper highlights the challenges associated with design, fabrication and welding, and inspection of innovative materials and actions taken to address those challenges in preparation for the Phase I ATR irradiation testing. The lessons learned from Phase I of this experiment can be used to guide researchers for design and analysis of future in-reactor testing of advanced fuel cladding systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

SU-F-T-249: Application of Human Factors Methods: Usability Testing in the Radiation Oncology Environment

Purpose: As a quality improvement measure, we undertook this work to incorporate usability testing into the implementation procedures for new electronic documents and forms used by four affiliated radiation therapy centers. Methods: A human factors specialist provided training in usability testing for a team of medical physicists, radiation therapists, and radiation oncologists from four radiotherapy centers. A usability testing plan was then developed that included controlled scenarios and standardized forms for qualitative and quantitative feedback from participants, including patients. Usability tests were performed by end users using the same hardware and viewing conditions that are found in the clinical environment. A pilot test of a form used during radiotherapy CT simulation was performed in a single department; feedback informed adaptive improvements to the electronic form, hardware requirements, resource accessibility and the usability testing plan. Following refinements to the testing plan, usability testing was performed at three affiliated cancer centers with different vault layouts and hardware. Results: Feedback from the testing resulted in the detection of 6 critical errors (omissions and inability to complete task without assistance), 6 non-critical errors (recoverable), and multiple suggestions for improvement. Usability problems with room layout were detected at one center and problems with hardware were detected at one center. Upon amalgamation and summary of the results, three key recommendations were presented to the document’s authors for incorporation into the electronic form. Documented inefficiencies and patient safety concerns related to the room layout and hardware were presented to administration along with a request for funding to purchase upgraded hardware and accessories to allow a more efficient workflow within the simulator vault. Conclusion: By including usability testing as part of the process when introducing any new document or procedure into clinical use, associated risks can be identified and mitigated before patient care and clinical workflow are impacted.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2023

Evaluation of Serial Testing After Exposure to COVID-19 in Early Care and Education Facilities, Illinois, March–May 2022

Objective: To understand SARS-CoV-2 transmission in early care and education (ECE) settings, we implemented a Test to Stay (TTS) strategy, which allowed children and staff who were close contacts to COVID-19 to remain in person if they agreed to test twice after exposure. We describe SARS-CoV-2 transmission, testing preferences, and the number of in-person days saved among participating ECE facilities. Methods: From March 21 through May 27, 2022, 32 ECE facilities in Illinois implemented TTS. Unvaccinated children and staff who were not up to date with COVID-19 vaccination could participate if exposed to COVID-19. Participants received 2 tests within 7 days after exposure and were given the option to test at home or at the ECE facility. Results: During the study period, 331 TTS participants were exposed to index cases (defined as people attending the ECE facility with a positive SARS-CoV-2 test result during the infectious period); 14 participants tested positive, resulting in a secondary attack rate of 4.2%. No tertiary cases (defined as a person with a positive SARS-CoV-2 test result within 10 days after exposure to a secondary case) occurred in the ECE facilities. Most participants (366 of 383; 95.6%) chose to test at home. Remaining in-person after an exposure to COVID-19 saved approximately 1915 in-person days among children and staff and approximately 1870 parent workdays. Conclusion: SARS-CoV-2 transmission rates were low in ECE facilities during the study period. Serial testing after COVID-19 exposure among children and staff at ECE facilities is a valuable strategy to allow children to remain in person and parents to avoid missing workdays.

Public, Environmental & Occupational Health↗
DOE OSTI2018

Absolute Locations of the North Korean Nuclear Tests Based on Differential Seismic Arrival Times and InSAR

Absolute location of the announced Democratic People’s Republic of Korea (DPRK) nuclear test on 6 January 2016 is constrained by fitting associated, Interferometric Synthetic Aperture Radar (InSAR)-based ground displacement with elastic finite-element modeling of an underground explosion source, including the effects of topography. The other five announced nuclear tests are located using arrival times and differential arrival times of regional and teleseismic body waves and constraints on the 6 January 2016 event location. Nuclear tests on 6 January 2016, 9 September 2016, and 3 September 2017 are under the summit ridge of Mt. Mantap, and tests on 25 May 2009 and 12 February 2013 are south of the topographic crest under the steep southern face of the mountain. The first test on 9 October 2006 was located near the crest of a separate topographic ridge approximately 2.87 km east of the 6 January 2016 event. Several unannounced events have occurred in the vicinity of the test site. The location uncertainty ellipse of an event approximately 8 min 30 s after the 2017 announced test covers the DPRK test site and is likely to have occurred there. Finally, additional events on 27 September 2017 and 12 October 2017 are 4–8 km northwest of the test site, and location probability regions do not overlap the test site.

58 GEOSCIENCES↗
DOE OSTI2022

LCLS-II-HE Cryomodule Testing at Fermilab

22 Linac Coherent Light Source II (LCLS-II) cryomodules were successfully tested at the Cryomodule Test Facility (CMTF) at Fermilab. Following the completion of the LCLS-II testing program, CMTF has shifted to testing cryomodules for the LCLS-II High Energy upgrade (LCLS-II-HE). The first LCLS-II-HE cryomodule, the verification cryomodule (vCM), was successfully tested and verified the readiness of LCLS-II-HE cryomodule testing at CMTF, and production cryomodule testing has begun. Presented here are the production cryomodule test acceptance criteria, testing plan, and cryomodule test results so far.

43 PARTICLE ACCELERATORS↗
DOE OSTI2013

Letter Report: LAW Simulant Development for Cast Stone Screening Test

More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second facility will be needed for the expected volume of additional LAW requiring immobilization. A cementitious waste form known as Cast Stone is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with waste acceptance criteria for the IDF disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. Further, the waste form must be tested to provide the technical basis for understanding the long term performance of the waste form in the IDF disposal environment. These waste form performance data are needed to support risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF. A testing program was developed in fiscal year (FY) 2012 describing in some detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW (Westsik et al. 2012). Included within Westsik et al. (2012) is a section on the near-term needs to address Tri-Party Agreement Milestone M-062-40ZZ. The objectives of the testing program to be conducted in FY 2013 and FY 2014 are to: • Determine an acceptable formulation for the LAW Cast Stone waste form. • Evaluate sources of dry materials for preparing the LAW Cast Stone. • Demonstrate the robustness of the Cast Stone waste form for a range of LAW compositions. • Demonstrate the robustness of the formulation for variability in the Cast Stone process. • Provide Cast Stone contaminant release data for PA and risk assessment evaluations. The first step in determining an acceptable formulation for the LAW Cast Stone waste form is to conduct screening tests to examine expected ranges in pretreated LAW composition, waste stream concentrations, dry-materials sources, and mix ratios of waste feed to dry blend. A statistically designed test matrix will be used to evaluate the effects of these key parameters on the properties of the Cast Stone as it is initially prepared and after curing. The second phase of testing will focus on selection of a baseline Cast Stone formulation for LAW and demonstrating that Cast Stone can meet expected waste form requirements for disposal in the IDF. It is expected that this testing will use the results of the screening tests to define a smaller suite of tests to refine the composition of the baseline Cast Stone formulation (e.g. waste concentration, water to dry mix ratio, waste loading).

Russell, Renee L.↗
DOE OSTI2014

Ranchero Armature Test LA-43-CT-2: PBXN-110 Explosive with no smoothing layer. Lower Slobovia, 6/13/13

This test, documented as LA-43-CT-2, was performed to verify calculations which indicate that the smoothing layer employed for Ranchero Armatures is not required when the slapper point spacing is 18 mm (“On the efficacy of an acrylic layer in the Ranchero armature,” Bob Watt, published in “Ranchero Status Report” LA-14463). [Naming this test –CT-2 acknowledges the previous test performed during 1/11 as the initial test in the series, and we will refer to that test as LA-43-CT-1 in the future.] In addition, higher velocity is achieved when the volume occupied by the smoother is replaced by HE. Two framing cameras were used on the shot, and a 20 point PDV array was fielded to ascertain armature performance in specific conditions. The conclusions are that the armature is smooth enough, and the terminal velocity is increased from ~3.1 mm/μs to ~3.3 mm/μs. The armature tested had casting voids in the HE that were observed in pre-test inspections, and were expected to cause ruptures. (See Appendix 3 of the Ranchero Status Report.) Such “blowouts” were observed in all three locations identified in pre-shot inspection. In addition, ruptures occurred in locations having no pre-shot indication of a problem. These unpredicted blowouts have a difference in appearance from those typically seen. Further visual examination of the inspection radiographs do not reveal the cause, and that is a serious concern for future efforts with such castings. A path forward will include attempting to digitize the inspection x-rays for the charge tested, to see if the cause of the “new” ruptures can be found via computer enhancement. In addition, there are two additional armatures from the same casting lot, and these might also be tested to see if there is some problem with the HE or the armature material in this batch. For now, this plan is on hold, pending the outcome of tests and cost estimates related to using PBX-9501 in place of the PBXN-110. Casting issues would be eliminated in such configurations.

42 ENGINEERING↗
DOE OSTI2014

Alternate Reductant Cold Cap Evaluation Furnace Phase II Testing

Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further research and development of this flowsheet eliminated the formic acid, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric-glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12 th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the CEF cold cap and vapor space data to the benchmark melter flammability models; Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: Define new cold cap reactions and global kinetic parameters in support of the melter flammability model development; Quantify off-gas surging potential of the feed; Characterize off-gas condensate for complete organic and inorganic carbon species. After charging the CEF with cullet from Phase I CEF testing, the melter was slurry-fed with glycolic flowsheet based SB6-Frit 418 melter feed at 36% waste loading and was operated continuously for 25 days. Process data was collected throughout testing and included melter operation parameters and off-gas chemistry. In order to generate off-gas data in support of the flammability model development for the nitric-glycolic flowsheet, vapor space steady state testing in the range of ~300-750°C was conducted under the following conditions, (i) 100% (nominal and excess antifoam levels) and 125% stoichiometry feed and (ii) with and without argon bubbling. Adjustments to feed rate, heater outputs and purge air flow were necessary in order to achieve vapor space temperatures in this range. Surge testing was also completed under nominal conditions for four days with argon bubbling and one day without argon bubbling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

FY14 Status Report: CIRFT Testing Results on High Burnup UNF

The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4, and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.

fuel vibration integrity↗