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This report describes various approaches and techniques associated with robotic spent fuel monitoring. The purpose of this description is to improve the quality of measured signatures, reduce the inspection burden on the IAEA, and to provide frequent verification.
To support more rigorous analysis on global security issues at Sandia National Laboratories (SNL), there is a need to develop realistic data sets without using "real" data or identifying "real" vulnerabilities, hazards or geopolitically embarrassing shortcomings. In response, an interdisciplinary team led by subject matter experts in SNL's Center for Global Security and Cooperation (CGSC) developed a hypothetical case description. This hypothetical case description assigns various attributes related to international SNF transportation that are representative, illustrative and indicative of "real" characteristics of "real" countries. There is no intent to identify any particular country and any similarity with specific real-world events is purely coincidental. To support the goal of this report to provide a case description (and set of scenarios of concern) for international SNF transportation inclusive of as much "real-world" complexity as possible -- without crossing over into politically sensitive or classified information -- this SAND report provides a subject matter expert-validated (and detailed) description of both technical and political influences on the international transportation of spent nuclear fuel.
The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.
BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, but baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.
Summary level reports from the Spent Fuel Database for use at DOE-EM headquarters and for use at IAEA.
This document presents the test plan for Task 2 of the Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan. This task’s overarching goal is the resolution of gas generation resulting from the radiolytic degradation of the oxide layers on aluminum. The uncertainty associated with the build-up of gaseous radiolytic products inside the ASNF storage canisters is one of the dominant technical challenges associated with long term dry storage of aluminum clad fuel. The objectives of the Task 2 test plan are to: 1. Understand how radiation chemistry inside the ASNF dry storage canister impacts extended dry storage performance. 2. Understand the radiation induced mechanisms governing the production of radiolytic gases and alteration of gas composition within a canister. 3. Correlate the observed changes in material properties of ASNF with the radiolytic degradation of aluminum oxide in fill environments of variable moisture and gas composition. 4. Compile a predictive kinetic modeling tool, to provide quantitative mechanistic insight into the in-cask radiation chemistry of ASNF. 5. Interface with the canister modeling activity of Task 3 to deliver degradation rates for package performance of both sealed and vented scenarios.
This report covers the experiments performed by Pacific Northwest National Laboratory and Oregon State University at their Testing, Research, Isotope, and General Atomic (TRIGA) test reactor. These experiments fall under a larger goal to maintain Continuity of Knowledge (CoK) over a spent fuel pool in the absence of light. During FY 17, PNNL was tasked with investigating the ex-pool solutions to this investigation. Preliminary analysis done in FY 16 resulted in the identification of IR and UV cameras as a viable option. One experiment conducted that year yield results showing that a pure IR camera would not function to image thermal objects through water due to the large absorption cross section of IR light within water. This experiment lead PNNL to investigate cameras that ranged from UV to IR in the hopes that the cameras would be able to image the objects more efficiently with cameras that ranged across these spectra. This constraint lead us to being conversations with Phoenix Engineering based out of Georgia as they have cameras that sample UV-IR and Vis – IR light spectra. The two cameras identified and eventually used for this experiment were the HAWK-216 and OWL 640 cameras. The specific QE curves of these cameras can be found in subsequent sections. The results discussed in this report pertain to the images captured at the OSU TRIGA reactor under varying power levels to see what thresholds and minimum detectable thermal output could be detected by these cameras.
At the Savannah River Site (SRS), approximately 7 MTHM of aluminum-clad spent nuclear fuel is currently in wet storage at L-basin. A significant portion of this fuel is slated to be dried and sealed in DOE standard canisters for long-term storage. This report will focus on a few fuel types, which should bound all other fuel stored at SRS. These consist of the reference fuel assembly, the MURR fuel and the HFIR fuel. These bound a hypothetical maximum scenario for MTR box fuel, a maximum for decay heat in an assembly, and a maximum for the aluminum surface area, respectively. The geometry for the packages were created from DOE specifications for storage of each of the fuels. As long as the minimum time from reactor discharge to sealed storage is 3 years, then the highest temperatures which occur within the sealed canisters do not exceed 100° C for all scenarios considered here. In the geometry considered, while the HFIR fuel has a large surface area, the packaging configuration leaves large void areas, such that the aluminum surface to free volume ratio is about 30% less than an ATR or MURR packaging configuration.
This presentation will be presented at the NNSS FY 2020 SDRD Annual (Webex) Meeting scheduled for September 23–24, 2020. Multi-modal, Multi-energy Approach for Neutron Interrogation of Spent Fuel RSLN-022-19, Year 2
Engineering-scale tests were performed to evaluate the relative effectiveness of forced gas and vacuum-drying processes on aluminum-clad spent nuclear fuel (ASNF) in preparation for extended dry storage. Drying models were developed to improve confidence in the likely range of conditions at the start of dry storage and to understand the factors most significant to each process. Testing was performed at the Holtec International Forced Helium Dehydration training facility in Camden, New Jersey, by students from the University of South Carolina in collaboration with Idaho National Laboratory. A full-size Type 1a basket, with a full-diameter vessel configuration (based on the Department of Energy standardized canister design), was used in combination with a 10 assembly surrogate ASNF arrangement to provide a proxy for geometry. A resistance heater incorporated in one assembly simulated the influence of 100 W decay heat. This instrumented chamber allowed testing of the removal of bulk water, physisorbed water, and chemisorbed water over a range of operating parameters with either drying process. Thick oxide films were grown on aluminum test plates to mimic the ASNF surface chemistry. These one-use test plates were housed within select assemblies during drying. Post-test analysis of these plates gave a measure of the moisture removed during each test, which was compared to a control. Instrumentation provided detailed thermal, humidity, and gas or pressure data relevant to each test. Model simulations generally compared favorably to instrument data for the tests. While both processes were able to remove bulk water, removal of chemisorbed water was shown to depend heavily on local temperature. To the disadvantage of vacuum drying, operations approaching or exceeding 220°C achieved consistently more dehydroxylation.
Several experimental techniques were utilized to understand the corrosion and SCC performance of possible techniques for repairing cracks and pits in SS304 utilized for dry cask storage of spent nuclear fuel. The repairs examined included cold spray, FSW, soldering, and VFAW. The corrosion and SCC assessment techniques utilized for comparing repairs against the base material were 4-pt bending SCC testing in boiling MgCl 2 at stress levels between 0 MPa and 350 MPa, immersion testing in 40% RH brine solution meant to mimic a service-like environment, SSRT SCC testing in 40% RH brine solution. For FSW alone, scanning electrochemical cell microscopy and microelectrochemical syringe cell were utilized to understand the corrosion performance across the weld, in the retreating/advancing sides, and base material. Based on the results of this study, the FSW and cold spray repair techniques proved most viable for SCC repair in SS304 dry cask storage applications and continuing research is highly recommended to understand and quantify in further detail their SCC performance compared to unrepaired SS304. Additionally, solder and VFAW (patch or incline) would not be recommended based on the corrosion and SCC findings in this study.
PNNL staff performed an independent technical review (ITR) of the aluminum-clad spent nuclear fuel (ASNF) program Task 2. The objective of Task 2 is to determine G-values for the radiolytic production of gaseous molecular hydrogen (H 2 ) from the aluminum oxyhydroxide layers present on ASNF cladding. The G-values need to be defensible, bounding, but not overly conservative. As part of the ITR, the team reviewed program documents, Task 2 work plans, and results to date. In addition, the team held three separate video conferences with the program research staff to ask clarifying questions and obtain additional detail. The PNNL ITR team concludes that if the program follows the methodology and rigor outlined in the test matrices, the program will develop G-values for a defensible technical basis of extended storage of ASNF. Recommendations for improvement are given.
A presentation covering the work done by the C430 Department at Idaho National Lab on thermal concerns for extended dry storage of aluminum clad spent nuclear fuel (ASNF). Computational fluid dynamics software Star CCM+ was used to model hypothetical accident conditions applied to a dry storage configuration containing ASNF. The hypothetical accident conditions being examined were a fire in which the dry storage configuration is fully engulfed in flames and a situation in which the vent ports of the overpack are completely blocked.
This research explored vitrification of U-Zr nuclear fuel that exhibits an explosion hazard when processed by common nitric acid dissolution. Glass frits were chosen based on properties that had been previously measured as well as being demonstrated successfully in production at the Defense Waste Process Facility (DWPF). Zirconium and Aluminum metal powders were chosen to represent the spent fuels after literature demonstrated that the potential phase that causes the issue for explosive reaction in the effluent stream is tied to the Zirconium-III phase field. DSC measurements were carried out on various simulants and waste forms to better understand thermal behavior of these materials. Zirconium (Zr) was tested with both nitric acid and the new effluent flowsheet that is under development by the SRNL Chemical Process Control (CPC) group. The nitric acid test did show the expected exothermic behavior as well as how the behavior is affected by particle size of the powders used. In contrast, the effluent samples did not show an exothermic peak. However, the concentration of metal in the samples was orders of magnitude lower than in the nitric acid testing and further testing of the effect of concentration on the exothermicity that could be measured confirmed a dampened signal. Tests were run that included frit, both DWPF Frit 510 and Iron Phosphate frit, exhibiting complex curves that included multiple endothermic and exothermic peaks during heating and cooling. The final crucible samples appear similarly to the bulk glasses pointing to these peaks being a part of the dissolution of the simulant into the base glasses.
This report is Supplement 22 to Nuclear Power Plant Infrastructure Evaluations for Removal of Spent Nuclear Fuel (PNNL-35730) and contains the infrastructure evaluation for Duane Arnold.
This report is Supplement 21 to Nuclear Power Plant Infrastructure Evaluations for Removal of Spent Nuclear Fuel (PNNL-35730) and contains the infrastructure evaluation for Three Mile Island.
One approach for mitigation of stainless steel (SS) spent fuel canisters (SFCs) is to coat the welds or even the entire SFC with a cold-sprayed nickel coating. The coating can serve as a magnetostrictive interface between an electromagnetic acoustic transducer (EMAT) and the SS plate material. The EMAT can send ultrasound into the canister and “listen” for reflections caused by chloride-initiated stress corrosion cracking (CISCC) or other damage mechanisms that may lead to leaks if undetected and unmitigated. Here, this project outlines an approach for monitoring the integrity of SFCs using magnetostrictive EMAT sensors, which emit and detect discontinuity-initiated reflections indicative of pits or cracks likely arising from CISCC. Flat 13 mm SS plates with welds similar to actual canister welds were prepared, with flat-bottom holes 3 mm and 6 mm deep and 6 mm in diameter placed within the weld heat-affected zone (HAZ). The most effective sensor configurations were able to detect critical discontinuities more than 2 m away. Although a single sensor could be robotically deployed, the large volume covered by each sensor could support a small number of permanently mounted sensors, equipped with an external connector for periodic interrogation. The HAZ of a hypothetical SFC cylindrical surface could be effectively monitored with just 16 sensors.
The confinement boundaries of spent nuclear fuel (SNF) canisters are typically fusion welded. Welded microstructures, strain hardening, and residual stresses combined with a chemically aggressive, chloride-rich environment led to concerns that the welded canister may be susceptible to chloride-induced stress corrosion cracking (CISCC). A comprehensive understanding of the modification of stainless steel (SS) metallurgical and mechanical properties by fusion welding could accelerate the predictive analysis of CISCC susceptibility. This paper describes a submerged arc welding (SAW) procedure that was developed and qualified on 12.7 mm (0.5 in.) thick AISI 304/304L SS to produce joints in a way similar to actual SNF canister manufacturing. This procedure has the potential to reduce the production cost and weld CISCC susceptibility by using fewer welding passes and lower heat input than current industrial applications. Global and local mechanical behaviors and properties, as well as residual stress distributions on the welded joint, were studied. The results indicate that hardness values in the fusion zone (FZ) and heat-affected zone (HAZ) are slightly higher than that of the base metal. Strain localization was presented in the HAZ before the tensile stress reached its maximum value, and then it shifted to the FZ. The specimen finally broke in the FZ. High tensile residual stresses exhibited in the FZ and the nearby HAZ suggest the highest CISCC-susceptible spots. The maximum tensile residual stresses were along the welding direction, indicating that if cracks occur, they would be perpendicular to the welding direction. This study involved developing and qualifying a SAW procedure for SNF canister production. The new procedure yielded cost savings (SAW working efficiency increased by about 80%), improved mechanical properties, and presented moderate residual stresses. Analysis revealed that the welded joint’s low-stress and high-stress damage assessments may be affected by shifts in the strain localization spot under loading.