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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Next Generation System Analysis Model Recently Added Features and Future Plans - Abstract

The Nuclear Waste Policy Act of 1982, as amended (NWPA 1982), established the federal government’s responsibility to accept spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from waste owners and generators for ultimate disposition. SNF generated by the current fleet of commercial nuclear reactors is being stored at the reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The US Department of Energy Office of Nuclear Energy (DOE-NE) is developing an Integrated Waste Management Program (IWMP) comprising a suite of options and supporting analyses to enable future informed choices. The IWMP is applying integrated waste management system architecture analysis, system engineering, and decision analysis principles to inform potential future decisions regarding potential nuclear waste management system architectures. Architecture analyses of the IWM system are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all major aspects of the back end of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool designed for the express purpose of modeling the IWM system. NGSAM imports data from the Oak Ridge National Laboratory (ORNL) Unified Database (e.g., historic assembly information, thermal profiles for assembly heat, at-reactor dry storage loadings) to ensure that the simulation initializes with a realistic representation of the state of commercial SNF in the United States. Recent major enhancements that have been implemented into NGSAM since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort and buffer car acquisition. • Addition of heavy haul and barge routes for some sites, as well as support for user-defined inter-modal routes. • Updates to the logic that checks the thermal maps prior to package transport. • Addition of an allocation method that predicts when reactor sites will pack assemblies from their pools for dry storage and allocates packages to those reactor sites in the preceding periods, favoring direct transport packages and reducing the number of packages that reactor sites pack for dry storage at their ISFSIs. • Addition of reactor site family operational limits, which are used to limit the number of loads from the pool and from dry storage at a given reactor site per year. • Support has been added for multiple canister loading maps and packages having multiple compatible transportation overpacks. • Updates in the handling of non-commercial fuel, including a new database containing data to support the updates. • Support for repackaging at reactor sites. • Implementing additional output reports or modifying existing reports. • User edits can now be created and edited via the NGSAM website. • Ability to load packages for dry storage at ISF pools. • Same-type package blending at DOE sites. • Support for multi-mode transloading at reactor sites. These new features have improved NGSAM capabilities and/or improve the user experience with the model and will be discussed in more detail. The initial NGSAM requirements for advanced reactor fuels, reprocessing, treatment, and conditioning are preliminary and are described at a high level in this paper: analysts will provide more specific requirements to the NGSAM team in the future. Additionally, there are many data needs associated with modeling advanced reactors in NGSAM, but many of the data or plans are still in progress and/or yet to be fully defined. However, this document describes an initial exploration of the data relevant to this program. Advanced reactor data will likely require revision as concepts evolve and new considerations are made. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by DOE. No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Two-color high speed thermal imaging system for laser-based additive manufacturing process monitoring

Monitoring melt pool temperature in laser powder bed fusion by providing a build laser that produces a laser beam that is directed onto the melt pool and produces an incandescence that emanates from the melt pool, receiving the incandescence and producing a first image having a first spectral band and a second image having a second spectral band, and determining the ratio of said first image having a first spectral band and said second image having a second spectral band to monitor the melt pool temperature.

Calta, Nicholas P.↗
DOE OSTI2015

Preliminary Design of Critical Function Monitoring System of PGSFR

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation control and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A study of transfer of UNF from non-disposable canisters - 15388

A study was performed by AREVA and URS for the Department of Energy (DOE) [1] to establish optimal methods for the opening of dry storage canisters (DSC) and the subsequent repackaging of used nuclear fuel (UNF) from the DSCs into generic standard transportation, aging, and disposal canisters (STADs) in wet or dry environments at reactor, consolidated storage, or repository sites. Eight dry and wet cutting technologies were evaluated: (1) skiving; (2) milling; (3) arc gouging/plasma torch; (4) cutting wheel; (5) diamond cable cutting; (6) laser cutting; (7) sawing; and (8) electrical discharge machining. The evaluated UNF repackaging technologies options included: (1) wet transfer in a pool; (2) dry transfer in a dry transfer system (DTS); (3) dry transfer in a fixed hot cell (FHC); (4) dry transfer in a mobile hot cell (MHC), and (5) a hybrid approach utilizing wet transfer in a pool and dry transfer in a FHC. A systems engineering evaluation was utilized for ranking the eight viable cutting options and the five different repackaging approaches for each of the sites. The systems engineering evaluation took into account six criteria (i.e., life cycle costs, worker dose, operational efficiency, technical maturity, ease of licensing, and waste generation) and performed a qualitative pairwise ranking of their relative importance to one another. Then quantitative or qualitative factors associated with each criterion (e.g., cost in millions of dollars, relative ease of licensing, etc.) for each option at each location (reactor, consolidated storage facility [CSF], repository) were developed based on historical experience, existing design studies, current 'state-of-the art', and/or engineering judgment. Using this input, the systems engineering evaluation identified dry skiving as the preferred cutting technology, with the wheel cutting technology in either a dry or wet environment identified as a promising technology that, with some RD and D, could mature to a better option (depending also on the disposition of the used DSC). For a reactor and a CSF/repository site, the systems engineering evaluation ranked the repackaging option using the skiving dry cutting technique and wet transfer of UNF in a pool as the highest option. However for a CSF and a repository site, the team recommended the hybrid option using the skiving dry cutting technique and both dry and wet transfer of UNF in a pool and FHC as a potentially better option (providing specific benefits such as higher throughput and more flexibility in operation). (authors)

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

Large Cask, Small Building, AREVA's Safe MLLW Disposal Solution Despite Facility Limitations - 16295

In order to begin decommissioning of a 1950's research reactor, Class C Mixed Low Level Waste (MLLW) had to be removed from the facility and disposed of. The AREVA TN RAM cask, designed and licensed for shipping Type B waste packages, was chosen as the most appropriate shipping solution. Due to the building layout, crane capacity and pool size, the standard TN RAM cask loading process could not be utilized. Unique solutions and approaches had to be developed to complete the project. AREVA designed a smaller liner insert that would house the four MLLW components in a single waste package to solve the problems related to crane capacity and physical pool dimensions. AREVA also designed and fabricated a shield bell to provide radiation shielding while moving the liner insert from the reactor pool to the cask that was staged adjacent to the reactor pool. The AREVA team worked with the disposal facility to cover requirements for disposal of Class B MLLW. This included designing spacers filled with inert material to fulfill the requirement for maximum 10% void space in the waste package. This disposal method for Class C waste met the requirements of macro-encapsulation. Therefore, no additional treatment action was required when disposing of Class C MLLW. AREVA's solution provides a model for disposing MLLW requiring macro-encapsulation in a single waste package. This solution also provides a model for how to adapt large shipping casks and liners to smaller aging commercial, DOE or research facilities. (authors)

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

Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental Study of Molten Metallic Fuel Relocation in Sodium-Filled Fuel Structures

The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) is being developed by Korea Atomic Energy Research Institute (KAERI). While their design incorporates inherent safety features demonstrated in the landmark Shutdown Heat Removal Test (SHRT) conducted at EBR-II including a large pool configuration and uranium alloy fuel, it is still necessary to evaluate metal fuel relocation and theirs consequences in case of postulated core disruptive accidents. Extensive data regarding fuel relocation in sodium from both in-pile and out-of-pile tests were obtained for oxide fuels in the 1960's - 1970's, while few studies of metallic fuel behavior were performed because of concerns that metallic fuel could not achieve high burnup due to the irradiation induced swelling problem. Following new metal fuel design development that mitigated the swelling problem, experiments regarding interaction of molten metal fuels and liquid sodium have been conducted by several researchers. Most studies that were conducted using metal simulants in various liquids concluded that the molten metals were heavily fragmented after liquid contact and the dominant factors governing the fragmentation were superheating and latent heat of fusion of the molten metal fuel. Gabor et al. conducted experiments using kilogram quantities of various molten uranium alloys in an open sodium pool configuration. They found that the metal fuel fragments were basically in forms of filaments and sheets with a high bed voidage in the order of 0.9. Their calculation indicated that the debris beds formed by relocated metal fuels in an open sodium pool would be largely coolable by conduction heat transfer; and even if deep beds were to form, convective heat transfer and boiling heat transfer could preclude further melt penetration. Although Gabor's tests are useful in evaluating the coolability of relocated metal fuels during a postulated accident in a SFR, they focused on late phase behavior in which the core material relocates downward through the core support plate into the inlet plenum where the structures are more open and thus consistent with the open pool configuration. There is still a knowledge gap related to nature of metal fuel fragmentation and relocation within the fuel assembly structures. Argonne National Laboratory (ANL) is planning to conduct experiments regarding metal fuel relocation in the core structures which mock up the PGSFR's fuel assembly design. In the present study, preliminary tests for demonstration of melting and injecting the metal fuel were conducted using copper. Metal fuel relocation experiments with single fuel pin test section are planned to be conducted for understanding the flow and freezing behavior of the molten metal fuel in a narrow sodium channel and for evaluating the coolability of the relocated fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2026

Improving Experience Beyond the Ride: Joint Assignment and Repositioning for Rideshare

In this study, a novel joint assignment and repositioning strategy is designed, implemented, and tested in a simulated traffic environment. The strategy addresses limitations of previously constructed rideshare repositioning algorithms by integrating the consideration of rider pooling into the optimization. By considering pooling, the strategy can minimize excess vehicle movement and increase riders' use of pooled services over a zonal-flow-based repositioning strategy and a fleet operating with no repositioning at all. The new strategy can increase pooling by up to 2.3% compared to no repositioning in a large region, and up to 0.1% in a small region, without compromising overall fleet revenue. Additionally, compared to a baseline, flow-based repositioning strategy, the new strategy can reduce empty vehicle travel by up to 30% in the larger region, and 2.1% in the smaller simulated area.

Paul, Joseph↗
DOE OSTI2020

In-situ Observations of Directed Energy Deposition Additive Manufacturing Using High-Speed X-ray Imaging

In laser-based directed energy deposition (DED) additive manufacturing, interactions among the laser beam, particle flow, and melt pool influence the properties of the solidified final part. Two separate DED systems, one with high powder flow rates to represent industrial-scale DED processing and the other with low powder flow rates for individual particle tracking, were synchronized with the high-speed imaging setup at the Advanced Photon Source in Argonne National Laboratory. In-situ x-ray imaging of the DED process using both systems highlighted the influence of powder flow rates. Increased powder flow rates resulted in less laser absorption into the melt pool, leading to a transition from a keyhole mode to a melt pool without a keyhole but with surface fluctuations due to powder flow. Increased velocities of particles during powder flow resulted in a decrease in particle melting times and a greater propensity for porosity formation. Altogether, better understanding of the interactions that occur during various scales of the DED process will enable flexibility, control, and new materials development in DED-based additive manufacturing.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Role of scan strategies on thermal gradient and solidification rate in electron beam powder bed fusion

Local microstructure control in electron beam powder bed fusion (EB-PBF) is of great interest to the additive manufacturing community to realize complex part geometry with targeted performance. The local microstructure control relies on having a detailed understanding of local melt pool physics (e.g., 3-D melt pool shape as well as spatial and temporal variations of thermal gradient (G) and solidification rate (R)). In this research, a new scan strategy referred to as ghost beam is numerically evaluated as a candidate to achieve the targeted G and R of IN718 alloy. The boundary conditions for simulations, including the speed (490 mm/s) and spatial locations of the beam within a given layer, are obtained by using series of snapshot images, recorded at 12,000 frames per second, using a high-speed camera. The heat transfer simulations were performed using TRUCHAS an open-source software deployed within a high-performance computational infrastructure. The simulation results showed that reheating at short beam on-time and time delay decreases both G and R. Local variation of R at the center of the melt pool trailing edge showed periodic temporal fluctuations. Finally, the ghost beam scan strategy was compared to other existing raster and spot scan strategies.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Predictive process mapping for laser powder bed fusion: A review of existing analytical solutions

One of the main challenges in the laser powder bed fusion (LPBF) process is making dense and defect-free components. These porosity defects are dependent upon the melt pool geometry and the processing conditions. Power-velocity (PV) processing maps can aid in visualizing the effects of LPBF processing variables and mapping different defect regimes such as lack-of-fusion, under-melting, balling, and keyholing. This work presents an assessment of existing analytical equations and models that provide an estimate of the melt pool geometry as a function of material properties. The melt pool equations are then combined with defect criteria to provide a quick approximation of the PV processing maps for a variety of materials. Finally, the predictions of these processing maps are compared with experimental data from the literature. Here, the predictive processing maps can be computed quickly and can be coupled with dimensionless numbers and high-throughput (HT) experiments for validation. The present work provides a boundary framework for designing the optimal processing parameters for new metals and alloys based on existing analytical solutions.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Performance estimation of resonance-enhanced dual-buoy wave energy converter using coupled time-domain simulation

This article presents the modeling methodology and performance evaluation of the resonance-enhanced dual-buoy WEC (Wave Energy Converter) by HEM (hydrodynamic & electro-magnetic) fully-coupled-dynamics time-domain-simulation program. The numerical results are systematically compared with the authors’ 1/6-scale experiment. With a direct-drive linear generator, the WEC consists of dual floating cylinders and a moon-pool between the cylinders, which can utilize three resonance phenomena from moon-pool dynamics as well as heave motions of inner and outer buoys. The contact and friction between the two buoys observed in the experiment are also properly modeled in the time-domain simulation by the Coulomb-friction model. Moon-pool resonance peaks significantly exaggerated in linear potential theory are empirically adjusted through comparisons with measured values. A systematic comparative study between the simulations and experiments with and without PTO (power-take-off) is conducted, and the relative heave displacements/velocities and power outputs are well matched. Then, parametric studies are carried out with the simulation program to determine optimum generator parameters. The performance with various wave conditions is also assessed.

16 TIDAL AND WAVE POWER↗
DOE OSTI2013

Application of underwater spectrometric system for survey of ponds of the MR reactor (NRC Kurchatov institute)

The underwater spectrometric system for survey the bottom of material science multi-loop reactor MR ponds was elaborated. This system uses CdZnTe (CZT) detectors that allow for spectrometric measurements in high radiation fields. The underwater system was used in the spectrometric survey of the bottom of the MR reactor pool, as well as in the survey located in the MR storage pool of highly radioactive containers and parts of the reactor construction. As a result of these works irradiated nuclear fuel was detected on the bottom of pools, and obtained estimates of the effective surface activity detected radionuclides and created by them the dose rate. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

A life cycle and product type based estimator for quantifying the carbon stored in wood products

Background: Timber harvesting and industrial wood processing laterally transfer the carbon stored in forest sectors to wood products creating a wood products carbon pool. The carbon stored in wood products is allocated to end-use wood products (e.g., paper, furniture), landfill, and charcoal. Wood products can store substantial amounts of carbon and contribute to the mitigation of greenhouse effects. Therefore, accurate accounts for the size of wood products carbon pools for different regions are essential to estimating the land-atmosphere carbon exchange by using the bottom-up approach of carbon stock change. Results: To quantify the carbon stored in wood products, we developed a state-of-the-art estimator (Wood Products Carbon Storage Estimator, WPsCS Estimator) that includes the wood products disposal, recycling, and waste wood decomposition processes. The wood products carbon pool in this estimator has three subpools: (1) end-use wood products, (2) landfill, and (3) charcoal carbon. In addition, it has a user-friendly interface, which can be used to easily parameterize and calibrate an estimation. To evaluate its performance, we applied this estimator to account for the carbon stored in wood products made from the timber harvested in Maine, USA, and the carbon storage of wood products consumed in the United States. Conclusion: The WPsCS Estimator can efficiently and easily quantify the carbon stored in harvested wood products for a given region over a specific period, which was demonstrated with two illustrative examples. In addition, WPsCS Estimator has a user-friendly interface, and all parameters can be easily modified.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Vibro-acoustic Imaging at the Breazeale Reactor

The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spent TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Deep Convection: Onset, Diurnal Cycle, and Role in Climate Variability and Sensitivity (Final Report)

The ASR SGP testbed for continuous modeling offers new opportunities to improve GCM parameterizations. We conducted an ASR investigation of convective and cloud microphysical processes that used existing ARM data as a testing ground for meaningful approaches to model evaluation and set the stage for later strategic use of the testbed to improve cumulus parameterizations and better understand deep convective cloud feedback. Our research consisted of the following tasks: Diurnal cycle of continental convection Deep convection at the SGP occurs in two forms: Organized mesoscale convection that originates upstream and produces peak precipitation at night; and local shallow convection that develops into deep convection that peaks in late afternoon. Local transitions are a logical next step from the shallow convection that will be the early focus of the testbed. This transition constrains entrainment and cold pools, since GCMs produce a continental rain peak near noon. We studied ARSCL-observed transition cases using the WRF, determined the sources of SCM diurnal cycle errors and made parameterization changes as appropriate. Factors controlling convective detrainment and anvil evolution Convective detrainment is a major source of high cloud ice water content and cloud forcing. The GISS GCM diagnoses a convective updraft speed profile, assumes a particle size distribution (PSD), and uses size-fallspeed relations to determine the fraction of condensate carried upward vs. precipitating. We used SCM case studies for TWP-ICE and MC3E for which vertical velocity has been retrieved to test whether entrainment improvements produce realistic updraft speeds and then improved the updraft microphysics parameterization. We compared detrained ice water contents and fall speeds to ARM and other field experiment data to refine the GCM’s fall speed and PSD assumptions. Role of convective cloud radiative processes in the Madden-Julian Oscillation (MJO) Changes in entrainment and rain evaporation and a cold pool parameterization have produced an MJO in the GISS GCM for the first time and allowed us to reproduce the KAZR dependence of convection depth on humidity during AMIE-Gan. We determined whether recent model changes including cold pools improve or degrade the MJO and made adjustments accordingly.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Corrosion test of LWR cladding tubes in high temperature seawater

In the case of cooling system failure in spent fuel pools (SFP) due to station black out and/or loss of coolant accidents, a pool water level must be kept high enough to avoid overheat of spent fuels. Seawater is one of the candidate coolant materials in such emergencies. In the present work, we have studied the corrosion behavior of zirconium alloys in seawater. Tube samples of Zircaloy-2 and -4 claddings were immersed in artificial seawater (3, 10 and 20 wt.%) at high temperature (80 - 85 Celsius degrees). Boric acid was added to the seawater at a corrosion test of Zircaloy-4. Two types of specimens were prepared according to the simulation of circumferential tensile stress. One was a plugged tube specimen, which was pressurized to around fuel rod pressure in a spent fuel pool by argon gas. Another specimen was of C-ring shape, where large circumferential tensile stress was partially induced at the surface. A notch was longitudinally grooved to increase the local tensile stress on some of the C-ring specimens. The specimen surface was observed by a scanning electron microscopy (SEM) after the corrosion test. After the 200 day long corrosion test for the plugged tube specimens without boric acid, Mg(OH){sub 2} deposits were observed on the surface of Zircaloy-2 and -4 specimens whereas Mg(OH){sub 2} was not observed with boric acid because of probably a lower pH. Neither pitting corrosion nor stress corrosion cracking (SCC) was observed on the surface of all plugged tube specimens. Only in the C-ring specimens in 20 wt.% seawater without boric acid and in 3 and 10 wt.% seawater with boric acid, small cracks were observed at the notch. The results of the present work show that Zircaloy-2 and -4 have a high tolerance towards corrosion and SCC in high temperature seawater.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗