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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Consolidating Data on Spent Nuclear Fuel into a Unified Database

This paper describes the comprehensive commercial spent nuclear fuel (SNF) database, the Unified Database (UDB), and discusses the UDB's (1) content and design with limited data examples and potential applications, (2) external interfaces, and (3) planned enhancements to streamline national SNF data collection and preservation. The UDB is the nucleus of the Used Nuclear Fuel Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST and DARDS). The UDB is being developed for the US Department of Energy Office of Nuclear Energy (DOE-NE) to provide a credible, controlled data source for key information for: various analytical purposes, including assessing issues and uncertainties related to the extended storage and transportability of SNF after extended storage; supporting research and development (R and D) prioritization; and preserving SNF related information for future generations. The UDB will serve as a national resource for use by DOE-NE or a future management and disposal organization in planning for future storage, transport, and disposal of SNF. The UDB is also being designed with the flexibility to provide information on greater-than-class C (GTCC) waste, government-owned SNF, and high-level radioactive waste (HLW). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Evaluation of Neutron Absorber Panels from Zion Spent Fuel Pool

In spent fuel pools (SFPs), fixed neutron absorber materials are used to increase storage capacity while maintaining criticality safety margins. BORAL{sup R} is a commonly used neutron absorber material for both wet (SFPs) and dry storage (casks). The specific questions surrounding fixed neutron absorber materials include the following: 1. What are the conditions of the neutron absorber panels in SFPs? Is there any gross degradation that could cause potential concern for criticality safety of the pools? 2. Is the coupon monitoring approach adequate for monitoring the conditions of the panels as part of an aging management program? 3. For plants that do not have coupons, do current in situ measurement approaches provide accurate results? Could such in situ approaches be used as an alternative monitoring approach? In order to shed light on these technical questions using actual plant data, EPRI initiated the Zion comparative analysis project at the decommissioning Zion Nuclear Power Plant in late 2014. An overview of the Zion plant, SFP, and project was described in a previous publication. (authors)

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

Probabilistic Risk Assessment of Spent Fuel Pool Cooling System Failure Considering Safety and Security Initiating Events

In order to quantify the probable risk associated with a safety system, nuclear safety risk evaluation process is applied which is generally performed in three steps. The first step is to calculate the damage frequency of a critical system. This step is followed by the radioactive source term estimation due to the potential system damage. The final step is the assessment of off-site consequences. The safety risk formula is shown by Eq. (1). P{sub F} is the probability of safety system failure (the probability used describes a random, independent event), and C is the consequence value. R{sub SAF} = P{sub F} * C (1) The security risk associated with an attack on critical infrastructure and/or special nuclear materials theft is usually evaluated using three- term risk equation (probability of attack, probability of adversary success, and consequence). A security threat and the related initiating events are not considered in a pure safety risk analysis. The vulnerability of the physical protection system is performed independently even though the consequence resulting from a security initiating event can be similar to a safety related initiating event and thus gets double counted. The security risk can be calculated using Eq. (2) R{sub SEC} = P{sub A} * P{sub S} * C (2) PA is the probability of attack, P{sub S} is the probability of adversary success (P{sub S} = 1 - P{sub I} P{sub N}). P{sub I} being the probability of interruption which accounts for the probabilities of detection of all timely detection elements along an adversary path. PN being the probability of neutralization: the result of engagement of the response force with the adversaries after the interruption has occurred. C is the consequence magnitude from a successful attack. The two security probabilities do not describe random, independent events. As a result, because of the differences in the randomness and independence of the respective safety and security probabilities, the safety and the security risks are currently calculated separately. The consequence term in nuclear safety and security risk evaluation have the same definition. An effort is made here to study the effect on the overall risk by considering security initiating events along with the subsequent failure of the intermediate safety element. The target selected is a spent fuel storage pool for analyzing the public risk arising from the interplay between nuclear security and safety initiating events. EASI (Estimate of Adversary Sequence Interruption) model is used to calculate the probability of interruption related to a security system that leads to the calculation of the security initiating event frequency. Further progression of this initiating security event to a top event is analyzed using a Probabilistic Safety Assessment (PSA) tool, SAPHIRE (version 8) to obtain a joint pool damage frequency. For the progression of events leading to a top event, the failure of the intermediate safety elements is assumed to be external random events such as heavy load drop and fire events. The objectives of this study is to (1) determine the double accounting of frequency of failure, risk and allocated resources in the scenario of a non-random security initiating event leading to a safety random event chain (2) compare the damage frequency from pure security with the frequency of combined security-safety failure event and pure safety failure events. (authors)

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

Shielding analysis of TN-32 spent fuel dry cask with SCALE

The objective of the University of Florida Non Destructive Evaluation (NDE) team is to use Emission Source Tomography to determine and verify the spent fuel location and condition in TN-32 dry casks. In order to investigate this technique, a radiation analysis of TN-32 cask will be performed with MAVRIC control module of SCALE 6.1 and MCNP6 code. The radiation source term from active fuel regions and fuel assembly structure components is calculated, and the MAVRIC geometry model is demonstrated in this paper. Finally, the procedure and future work is listed in this paper. And the shielding analysis result will be presented in ANS winter conference. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Estimating Public Dose Along Hypothetical Spent Nuclear Fuel Transportation Routes

In the future, it will be necessary to transport large quantities of spent nuclear fuel (SNF) from nuclear power plant sites to an interim storage facility (ISF) or a repository when such a facility is available. Because it is impossible to shield all of the radiation from this material, the general public will receive a very small, but quantifiable, dose from this transportation campaign. This paper describes the estimated collective effective radiation dose to the public due to SNF shipments under routine conditions and compares the estimated dose figures to the natural background radiation dose. To determine the estimated collective effective dose to the public, this analysis uses the US Department of Energy's (DOE) Stakeholder Tool for Assessing Radioactive Transportation (START), a geographic information system (GIS) routing tool, and a set of unit dose factors (UDFs) which were developed to calculate the estimated dose to the public along transportation routes. The estimated doses of radiation to which the public would be exposed during shipments of SNF is expected to be a consideration, along with many other factors, when making routing decisions. In all cases studied in this paper, estimated doses to the public will be orders of magnitude lower than the natural background radiation dose and will meet all US federal regulatory requirements for radiation dose/exposure to members of the public. The sample routes analyzed in this paper are expected to be representative of future SNF shipment routes in terms of length and mode of transportation, although they do not represent actual routes or destinations. This paper considers four routes from two shutdown reactor sites to two destinations. It must be noted that the specified origin sites are not necessarily the first sites under consideration for a de-inventory process. Further, the destination sites reviewed in this report are only to illustrate estimated dose figures along different routes and should not be taken as potential sites for an ISF or repository. For illustrative purposes only, the shortest distance rail route was found between each origin and destination point, and the collective effective dose to the public from the SNF cask was estimated along each route. This paper is not intended to make recommendations for routing decisions, but rather is intended to inform further discussion about the transportation of SNF and other radioactive materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Thermal Evaluations for Hypothetically Drain-Down Spent Fuel Storage Facility at SRS

The L-Area Spent Nuclear Fuel (SNF) underwater storage facility at Savannah River Site (SRS), referred to as L-Basin SNF facility, has an array of Vertical Tube Storage (VTS) racks and a group of High Flux Irradiation Reactor (HFIR) fuel cores. The underwater facility is integrally designed as a wet SNF storage system with a chemically-controlled water pool. The SNF storage racks are seismically qualified. The L-Basin pool holds about 3.4 million gallons of water with concrete wall thickness of about 5 ft and water depths of 17 ft to 30 ft. The facility has mainly two different elevation levels of 30 ft and 17 ft below the ground level. At the level 30 ft below the ground, a maximum of 110 racks of VTS and 5 racks of Dry Cave can be stored. At the level of 17 ft below the ground, 120 HFIR cores can be stored, considering that each rack contains two HFIR cores. In case of a hypothetical accident, all cooling water in the storage pool may be drained away, resulting in dry SNF storage condition. The primary objective of the work is to estimate the maximum fuel temperature in a conservative manner when the L-Basin facility is completely dry due to the hypothetical drain-down of the pool water. A three-dimensional Computational Fluid Dynamics (CFD) modeling approach was taken for the thermal evaluations of the water-drained SNF storage facility. The modeling calculations were performed, assuming that all of the storage racks in L-Basin are fully loaded with SNF assemblies containing decay heat sources. In the case of full loading, the L-Basin facility stores 14,600 SNF assemblies in VTS racks and 120 cores in HFIR racks. The thermal analysis results of the L-Basin model yielded maximum temperatures of VTS and HFIR fuel assemblies for different fuel loading scenarios and boundary conditions of the SNF storage racks.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Use of High-Frequency Fields for Centrifugal Separation of Spent Nuclear Fuel

The possibility of using a rotating magnetic field in a low-frequency direct-flow plasma centrifuge for separating spent nuclear fuel is studied. Estimates of the separative effect in a model binary mixture with atomic mass 120 and 240 were made for concrete values of the parameters of the setup and the properties of rf-discharge plasma. The hydrodynamic parameters of the rotating and expanding plasma column, the separative effect, and the capacity of the setup are obtained.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2016

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Development, feasibility, and uncertainty of radioactive 22 Na tracer dilution and gamma spectroscopy for mass determination of molten salt for pyroprocessing spent nuclear fuels

Accurate knowledge of the total mass of molten salts both for pyroprocessing spent nuclear fuels and for molten salt reactors is necessary for safeguards purposes. However, it is challenging to know or measure the total mass of molten salt due to the complicated shapes as well as the compositional—and thus density—changes that occur over time during operation. Here, we investigated radioactive tracer dilution (RTD) as a potential safeguards technique for application to the total mass measurement of LiCl–KCl–UCl 3 salt used for uranium electrorefining, with a focus on the feasibility, uncertainty, and fission product effect of RTD for salt mass measurement. To this end, experimental-scale RTD tests (600–700 g of total mass) were initially carried out by adding 22 Na tracer salt into LiCl–KCl salt containing 0, 5, and 28wt% radioactive salts from an electrorefiner (ER) for refining uranium. Upon the completion of the experimental-scale RTD tests, we performed a scale-up RTD test by directly adding 22 Na tracer salts to the ER, which uses about 12 kg of LiCl–KCl–UCl 3 salt during normal operation. This paper reports the main features of RTD for total mass measurement of molten salt and summarizes the results of the experimental-scale and scale-up RTD tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235 U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via high performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Adsorption of molecular iodine and alkyl iodides from spent-nuclear-fuel-reprocessing off-gas using reduced silver mordenite

Radioactive iodine is a hazardous byproduct of spent-nuclear-fuel reprocessing that must be removed from the off-gas stream before it can be discharged. Reduced silver mordenite (Ag 0 Z) is currently the baseline material for iodine capture in the U.S. Although the performance characteristics and capture mechanisms of I 2 and CH 3 I have been established, similar investigations into long-chain organic iodides have yet to be performed. In this study, thin beds of Ag 0 Z were loaded with I 2 , CH 3 I, C 4 H 9 I, and C 12 H 25 I (5 ppm to 50 ppm) carried in a dry air stream at 150°C. The maximum iodine capacity was 105 ± 5 mg I/g Ag 0 Z for all species. Saturated Ag 0 Z samples were characterized using scanning electron microscopy, X-ray fluorescence, X-ray photoelectron spectroscopy, diffuse reflectance UV–visible spectroscopy, pair distribution function analysis, and thermogravimetric analysis. Further, near-complete Ag utilization and similar physical/chemical properties were observed for all samples. Through a comparison with previous studies and an investigation of aged Ag 0 Z, we propose that iodine species react with Ag + at exchange sites, forming α-AgI within the mordenite channels, and with surface Ag 0 nanoparticles, yielding β-/γ-AgI. The available Ag sites in the interior (Ag + ) and exterior (Ag 0 ) of mordenite determine the adsorption capacity since α-AgI formation is limited by the total pore volume. Potential iodine uptake routes were summarized for aging and non-aging environments. A scalable predictive model was implemented for deep-bed iodine removal, and predictions were in good agreement with experimental data. Sensitivity analysis suggests that iodine uptake kinetics is governed by pore diffusion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Polyester-based epoxy vitrimer integrating spent coffee ground as a natural filler

Coffee is a widely consumed beverage, but approximately 80% mass of coffee cherry is discarded as waste. Here, in this study, spent coffee grounds (SCG) were recycled and used as a natural filler in a polyester-based epoxy matrix made from reactions of diglycidyl ethers of Bisphenol A and polyethylene glycol with methyl nadic anhydride. The properties of SCG/epoxy composites with various SCG contents up to 40% w/w were evaluated using various techniques such as FT-IR spectroscopy, tensile testing, optical microscopy, differential scanning calorimetry, and rheology. The SCG filled polyester-based epoxy matrix demonstrated excellent dimensional stability and retained a permanently crosslinked network, but re-shaping was possible at ~150 °C through transesterification, which occurred due to a dynamic chain exchange reaction between the ester group of the epoxy matrix and the hydroxyl group of the SCG fillers. The transesterification also improved the interfacial adhesion between the matrix and SCG fillers, resulting in tensile strength of over 20 MPa and a modulus of around 2.5 GPa, even with high SCG contents up to 40% w/w. While neat, cured epoxy shows thermoset behavior, filling the matrix with SCG enables reforming and origami-inspired spontaneous change in shape.

42 ENGINEERING↗
DOE OSTI2025

Selective leaching and solvent extraction of Lithium from spent batteries

Pre-extracting Li from spent lithium-ion batteries (LIBs) is crucial because the recovery efficiency of Li is low after other critical metals have been extracted. Traditional methods involving black mass roasting followed by water leaching have resulted in the extraction of 76 % Li and 61 % Al. However, pre-leaching Li from pretreated black mass using an oxalic acid solution at both ambient and elevated temperatures significantly improved results, achieving 98.1 % Li, 99.5 % Al, and 100 % Fe extraction while leaving Ni, Co, Mn, and Cu behind under optimal conditions. Oxalic acid crystals were recovered by refrigerating the leach solution at temperatures below 5 °C and were reused with nearly identical leaching efficiency. Selective extraction of Li from the oxalate leach solution was achieved using Cyanex® 936P under optimal conditions. Here, due to its extremely low Li extraction efficiency, Dichloromethane proved unsuitable as a diluent. Comparative extraction tests using Cyanex® 936P, Cyanex® 272, and DEHPA in kerosene demonstrated that Cyanex® 936P is an excellent extractant for Li, effectively separating it from other impurities. Under optimal conditions, 98.8 % of available Li was extracted using 20 % Cyanex® 936P, compared to 51.1 % with Cyanex® 272 and 39.9 % with DEHPA in kerosene. Additionally, stripping Li from Cyanex® 936P using H 2 SO 4 and HCl was explored, with HCl yielding the best performance.

Black mass↗
DOE OSTI2025

In-situ ion irradiation of fission products in a spent UO 2 fuel

This study investigates the behavior of fission gas bubbles and five metal precipitates (5MPs) (Mo, Ru, Rh, Tc, Pd) in spent uranium dioxide (UO 2 ) fuel under various ion irradiation doses, temperatures, and flux conditions. Utilizing in-situ ion irradiation and advanced transmission electron microscopy, we analyzed the evolution of fission gas bubbles and 5MPs in UO 2 samples from the Belgium Reactor 3 (BR-3). Our findings reveal significant shrinkage of fission gas bubbles and 5MPs with increasing irradiation dose, accompanied by a decrease in pair density. We demonstrate that ion irradiation induces a homogeneous re-solution process where individual atoms are ejected from bubbles and precipitates, leading to their dissolution and subsequent re-precipitation in the matrix. In conclusion, this study provides critical insights into the dynamic behavior of fission products under irradiation, facilitating the development of predictive models and contributing to the optimization of nuclear fuel performance and safety.

Fission products↗
DOE OSTI2016

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets to the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Spent nuclear fuel system dynamic stability under normal conditions of transportation

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Monitoring of helium gas leakage from canister storing spent nuclear fuel: Radiological consequences and management

Chloride-induced stress corrosion cracking (CISCC) of canisters has been identified as a potential degradation mechanism requiring ageing management — especially in a marine air environment. Methods for monitoring helium gas leakage from canisters have been developed on the basis of a phenomenon that the bottom temperature of the canister increases as the top temperature of the canister decreases when helium gas leaks during storage. In this paper, we consider the radiological consequences of CISCC based on two scenarios of helium gas leakage and propose effective management that includes a timeline for monitoring and mitigation. Monitoring helium gas leakage by measuring the difference in temperature between canister surfaces in the dry cask storage of spent fuel (SF) can detect leakage before SF undergoes oxidation as a result of air ingress and before country-specific regulatory limits of off-site doses are exceeded, thereby enabling mitigatory action.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗