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DOE OSTI2018

Design of, and Test Plan for, an Iodine and Tritium Capture System for a NO 2 -Based Tritium Pretreatment Process

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas of a processing plant, including 3 H, 14 C, 85 Kr, and 129 I. One potential simplification to the management of the off-gas streams could be achieved through an efficient tritium pretreatment (TPT) step in which the UO 2 fuel is oxidized either by air or by NO2 before dissolution. Oxidation of the UO2 fuel matrix results in the release of tritium contained within. Upfront removal of tritium from the fuel in a pretreatment step, followed by the abatement of tritium, can minimize or eliminate the distribution of tritium throughout the plant, also decreasing or eliminating the need for tritium capture on multiple off-gas streams. The use of NO 2 as the oxidant in an advanced tritium pretreatment (ATPT) allow the oxidation to be performed at lower temperatures and may result in the quantitative release of iodine from the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Savannah River National Laboratory – General Fusion 2023 INFUSE Report (Rev.1)

This report describes the results from an INFUSE research project, where Savannah River National Laboratory (SRNL) in collaboration with General Fusion (GF) used process modeling to understand and optimize commercial power plant (CPP) fuel cycle designs based on parameters provided by GF. The study primarily focused on two candidate fuel cycles with different blanket materials, one with a lead lithium eutectic (LLE) blanket and the other with a pure lithium (Li) blanket. LLE benefits from a low melting point, favorable neutronics, and lower reactivity, but liquid lithium has the potential for higher tritium breeding ratios (TBR) and does not poison the plasma as a high Z contaminant. It was found that the main differences between LLE and Li designs are the extraction technologies required to remove tritium from the blanket and the amount of tritium and its distribution within the facility. More than 80% of the in-process tritium inventory for the LLE design is contained in the isotope separation system, while for the Li design, over 60% of the in-process tritium inventory is contained within the blanket material. This is due to significant tritium retention by Li. For the Li blanket, the burden of tritium processing rests on the blanket extraction technology rather than the traditional exhaust processing route. Thus, the blanket extraction technology is a main driver of tritium inventory in the Li system and determines the subsequent interface with the tritium processing plant.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2018

Initial Assessment of Ruthenium Removal Systems for Tritium Pretreatment Off-Gas

The processing of used nuclear fuel releases volatile and semivolatile radionuclides into the off-gas streams associated with the various processing steps. Ongoing efforts to develop methods to control the release of gaseous radionuclides to the environment have identified ruthenium (specifically 106 Ru) as one of the semivolatile nuclides requiring a high degree of abatement. The tritium pretreatment process will release a fraction of the ruthenium as volatile ruthenium tetroxide (RuO 4 ). Scoping tests were conducted to characterize the adsorption/deposition of RuO 4 from dry gas streams by two different substrates: steel wool and silica gel. A simple test system was assembled to produce RuO 4 through oxidation of powdered ruthenium metal, followed by a temperature-controlled sorption column. Tests were conducted with no sorbent present in the column to obtain baseline data on the evolution of RuO 4 and deposition of ruthenium on the internal surfaces of the column. Silica gel and steel wool were tested as adsorbent media. Gaseous effluent from the column was scrubbed with caustic solution to remove any remaining ruthenium. Ultraviolet/visible spectroscopy (UV/Vis) and inductively coupled plasma–mass spectroscopy analyses were used to measure the amount of ruthenium in the scrub solutions. The UV/Vis analysis proved to be problematic because of precipitation of ruthenates, ruthenium oxides, and hydrated ruthenium oxides that occurred over time. An effective stabilization agent that does not adversely affect the desired spectra needs to be identified if this method is to be used in the future. There were indications that oxidation of powdered ruthenium metal did not generate a stream with steady concentration of RuO 4 , so improved techniques are needed for follow-on experiments designed to infer deposition rate. Both steel wool and silica gel are able to decrease the effluent Ru concentration by a factor of 10–100 during the 5-hour tests conducted. These tests were conducted using a dilute RuO 4 -bearing dry gas stream and the short test duration did not result in sorbent saturation or in an estimation of Ru breakthrough time. RuO 4 desorbs from silica gel at room temperature. As interactions between RuO 4 and silica gel are not irreversible, it cannot function as a disposal media without pretreatment to reduce RuO 4 to a less mobile species. Silica gel may be useful as a RuO 4 sorbent or delay media in some applications. Based on visual observations, steel wool was an effective deposition substrate with no evidence of RuO 4 release. The deposition form should be identified and methods to assay the ruthenium content need to be developed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Water Processing for Isotope Recovery Using Porous Zero Valent Iron

Reduction-oxidation cycles of metals can be harnessed to create a reusable tritiated water processing system. The concept is straightforward; a tritium-contaminated steam passes over a hot metal bed converting the metal to a metal oxide and liberating hydrogen isotopes for further processing and isotope separation. The bed is regenerated by converting the metal oxide back to a bare metal using protium gas, creating a clean water stream. Free oxygen is not produced during the cyclical process, making it safe for use in a hydrogen processing facility, and the only by-product is detritiated water. Porous zero valent iron (p-ZVI) has been identified as an ideal candidate material for this process due to its low cost, unique morphology, and favorable thermodynamics. Therefore, investigations of p-ZVI were conducted to better understand how a bed composed of such material would behave in a tritium processing facility. The thermal and physical properties were assessed, along with cycling and isotope effects. The results indicate that p-ZVI beds could serve as a low-cost, reusable system for the treatment of water in tritium processing facilities.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Tritiated Sulfur Hexafluoride Disposition Strategies

Tritium (T 2 ) is a radioactive isotope of hydrogen that is produced in nuclear fission reactions and is often used in nuclear fusion reactions and accelerator-based applications for medical isotope production. As a hydrogen isotope, tritium can readily bind to hydroxyl radicals (OH), forming tritiated water (HTO or T 2 0), and to carbon atoms. Tritium decays to helium-3 ( 3 He) via beta-decay with max decay energy of 18.6 keV. While it is not an external radiation hazard, it can be an internal radiation hazard if tritium is inhaled, ingested, or absorbed through the skin. In applications where tritium is handled, tritium confinement is performed using different barriers to minimize releases to the environment. For gaseous (elemental), liquid (oxide), and metal (hydride) tritium, process piping and components provide the primary confinement function. Secondary tritium confinement is typically provided by inert (i.e. non-flammable gases such as nitrogen, argon, or helium) gloveboxes which are connected to a tritium stripper system. Primary tritium confinement barriers typically produce low volumes of high tritium concentrations of tritiated methane from carbon in steels or organic materials along with tritiated water/tritium oxide (e.g. HTO) and tritiated ammonia from reactions with oxygen and nitrogen. Tritium escaping primary confinement into secondary confinement atmospheres (e.g. gloveboxes) produce higher volumes of lower activity contamination than found in process piping. Tritium contamination also occurs by leaks or tritium permeation/diffusion through confinement materials. Tritium from inside primary confinement barriers will diffuse or leak out of the primary confinement barrier and usually into the air, if the system is inside an air hood or ventilated hot cell, or into the secondary confinement (e.g. glovebox) atmosphere which is either exhausted or stripped based on the function of the secondary confinement (glovebox) system. Accelerator based processes for medical isotope production represent an atypical tritium contamination challenge. In medical isotope production, deuterium supply gas is ionized and accelerated to a tritium gas target to produce neutrons that are then used to produce the medical isotopes through additional nuclear fission reactions. To create large voltage differentials for accelerator operations, an electrical insulation medium is needed to prevent or rapidly quench electric discharges. A common electrical insulation medium utilized in accelerator applications is sulfur hexafluoride (SF 6 ) gas. SF 6 has a high dielectric strength and allows for the construction of smaller accelerator systems compared to other electrical insulation mediums such as air or dry nitrogen. Due to tritium permeation/diffusion through accelerator process and confinement materials, there is the possibility that tritium can contaminate the electrical insulation medium of the accelerator. Tritium contaminated SF 6 creates a material without any obvious processes for managing the contamination, reuse, or disposal of the used SF 6 . This document will discuss possible management strategies for tritium contaminated SF 6 for accelerator-based processes for Molybdenum-99 (Mo-99) production.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

Evaluation of Detritiation Strategies for Mitigating Tritium Releases

Tritium confinement is performed using different barriers to minimize releases to the environment. Primary confinement is usually performed by process piping and components, secondary confinement typically by inerted gloveboxes connected to a continuously operating tritium stripper system. Some facility designs employ tertiary tritium confinement by using process room confinement with a tritium stripper system activated after an accident to further mitigate tritium releases to the environment. Secondary confinement of some tritium systems such as radiation-hardened or shielded-cell enclosures are not easily achieved using designs for inerted glovebox. These enclosures can be challenging to seal for secondary tritium confinement due to heating, ventilation, and air conditioning (HVAC) needs for temperature control, as well as feedthroughs for equipment and instrumentation, are potential pathways for tritium leaks or loses. Additional tritium leak paths can be created in secondary confinement enclosures due to a Design Basis Accident (DBA) seismic event. These new, larger tritium leak rates from the DBA event are usually not part of the stripper system design basis so all tritium released to the secondary confinement enclosure is assumed released to the environment. Detritiation strategies for shielded enclosures and/or gloveboxes with significant leak rates are necessary to minimize off-site radiological dose consequences. To address this, the relationship between the confinement system leak rate and stripper system (recirculating) flow rate was evaluated to meet prescribed maximum allowable environmental tritium emissions from the event. The simple analysis described assumes a constant leak rate from the confinement system (i.e. shielded enclosure or glovebox) while a recirculating stripper system strips tritium from the system– a competition between tritium release and tritium recovery. At a high level, the resulting expression summarizes the relationship between system leak rate (F L ) and stripper flow rate (F S ) relative to allowable tritium release (Q A ) and initial tritium release (Q 0 ): $\frac{F_L}{F_s}$ ≈ $\frac{Q_A}{Q_0}$. This report derives the relationship between these parameters, examines the impact of finite stripping times followed by purging of the stripped volume. The analysis provides example results for up to 30 gram tritium releases: the maximum tritium inventory of a Hazard Category III (Department of Energy) Nuclear Facility. The detritiation model presented represents a high-level analysis of tritium recovery versus losses for “leaky” confinement enclosures such as shielded cells after large tritium releases. Key parameters for doing the analyses are the system leak rate, stripper flow rate, initial tritium release, and allowable tritium release values. The analyses show a proportional decrease in releases with reduction in initial tritium release but a non-linear increase in tritium recovery with increased stripper flow rate or reduced leak-rate. The analyses also apply to glovebox confinement systems which could have significant increases in leak rates after a DBA. The analysis recognizes but does not include tritium absorption followed by re-emission from the walls of the confinement volume – an analysis which could be pursued in future studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Deuterium Permeation Through a Self-Supported Palladium-Silver Membrane in Helium Gas Mixtures

Palladium membranes and membrane reactors can separate and purify tritium from impurities in the plasma exhaust processing system for the fusion energy fuel cycle. Membranes can also act as a continuous separation method to remove tritium from helium streams in the breeder blanket tritium extraction system, such as from the purge gas of solid breeders. To investigate the potential of these membranes for these applications, we performed a deuterium permeation campaign with a self-supported palladium-silver (Pd-25Ag wt%) tube of 0.15 m length, 76 µ m wall thickness, and 3.0 mm inner diameter. A gas mixture of 3.95% D 2 and a balance of He was delivered to the inside of the tube and D permeated radially outwards through the membrane into a vacuum chamber. Permeation experiments were conducted at 300 ° C, 350 ° C, 400° C, and 450°C, from 100 to 1000 sccm total flow rate, and with total pressures of 90, 150, 190, and 250 kPa. Further, deuterium permeation was consistently lower than predicted from diffusion-limited permeation models, thus we developed a transport model that included gas-phase mass transfer and surface reactions to model experimental results. The dissociation constant was optimized to fit the developed model to experimental data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Tritium glovebox stripper system seismic design evaluation

The use of glovebox confinement at US Department of Energy (DOE) tritium facilities has been discussed in numerous publications. Glovebox confinement protects the workers from radioactive material (especially tritium oxide), provides an inert atmosphere for prevention of flammable gas mixtures and deflagrations, and allows recovery of tritium released from the process into the glovebox when a glovebox stripper system (GBSS) is part of the design. Tritium recovery from the glovebox atmosphere reduces emissions from the facility and the radiological dose to the public. Location of US DOE defense programs facilities away from public boundaries also aids in reducing radiological doses to the public. This is a study based upon design concepts to identify issues and considerations for design of a Seismic GBSS. Safety requirements and analysis should be considered preliminary. Safety requirements for design of GBSS should be developed and finalized as a part of the final design process.

42 ENGINEERING↗

DOE OSTI2019

TRP Cutter Head Improvement

The current Target Rod Preparation (TRP) Cutter Head suffers from overheating problems during TPBAR breaching operations. The overheating leads to work stoppage due to the periodic need to allow the TRP Cutter Head motor to cool. This work stoppage and allowance for cooldown of the motor leads to a very inefficient TPBAR breaching process. Due to the increased need for Tritium gas, the system must be improved to increase the efficiency of the TPBAR breaching process. The project would include the following major steps: Determine likely causes of overheating of the TRP cutter head motor. Identify and research methods for cooling currently installed motor. Identify and research alternate motors for TRP cutter head. Procure a new motor for testing. Perform TPBAR breaching mockup testing with new motor. Write report on results and recommendations on potential path forward.

42 ENGINEERING↗

DOE OSTI2020

SR19036 - Improve TRP Cutter Head Performance

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Tritium cleanup system and method

Work area cleanup systems and methods are described for removing tritium from the atmosphere of a work area such as inert gas gloveboxes. Systems utilize a multi-column approach with parallel processing. Tritium of a tritium-contaminated stream is converted into tritiated water and adsorbed onto the separation phase of a first column as a second, parallel column can be simultaneously regenerated. The gaseous stream that exits the column during the regeneration phase can carry a high tritium concentration. The system can also include and a separation stage during which the tritium of the gaseous regeneration stream can be separated from the remainder of the regeneration product.

Xiao, Xin↗

DOE OSTI2017

Investigation of Sorption and Diffusion Mechanisms, and Preliminary Economic Analysis

This report describes the synthesis and evaluation of molecular sieve zeolite membranes to separate and concentrate tritiated water (HTO) from dilute HTO-bearing aqueous streams. Several monovalent and divalent cation exchanged silico alumino phosphate (SAPO-34) molecular sieve zeolite membranes were synthesized on disk supports and characterized with gas and vapor permeation measurements. The pervaporation process performance was evaluated for the separation and concentration of tritiated water. Experiments were performed using tritiated water feed solution containing tritium at the high end of the range (1 mCi/mL) anticipated in a nuclear fuel processing system that includes both acid and water streams recycling. The tritium concentration was about 0.1 ppm. The permeate was recovered under vacuum. The HTO/H2O selectivity and separation factor calculated from the measured tritium concentrations ranged from 0.99 to 1.23, and 0.83-0.98, respectively. Although the membrane performance for HTO separation was lower than expected, several encouraging observations including molecular sieving and high vapor permeance are reported. Additionally, several new approaches are proposed, such as tuning the sorption and diffusion properties offered by small pore LTA zeolite materials, and cation exchanged aluminosilicates with high metal loading. It is hypothesized that substantially improved preferential transport of tritium (HTO) resulting in a more concentrated permeate can be achieved. Preliminary economic analysis for the membrane-based process to concentrate tritiated water is also discussed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2015

TSTA piping and flame arrestor operating experience data

Not Available

Cadwallader, Lee C.↗

DOE OSTI2018

Design of Laboratory System to Recover Tritium from Zirconium Recycle and Plan to Test System

The objective of the tritium separations task is to develop a viable process to separate tritium from the zirconium recycle off-gas stream. In support of this long-term objective, the goal of the present work is to design a test system and prepare a test plan for an alternate tritium capture method suitable for, and compatible with, recycle of the reactive gas (i.e., chlorine) in the zirconium recycle process. The zirconium recycle process is under development to recover zirconium from the Zircaloy® cladding component of used nuclear fuel. Small quantities of all isotopes of hydrogen (all H) are trapped in the Zircaloy as zirconium hydride. Processing releases and converts this hydrogen to hydrogen chloride, including tritium-substitu ted hydrogen chloride, 3 HCl (a.k.a., TCl). This compound of tritium must be removed from the zirconium recycle process off-gas and sequestered to satisfy regulations governing emissions of tritium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2015

Production and characterization of thin film group IIIB, IVB and rare earth hydrides by reactive evaporation

A recent short history of reactive evaporation by D. M. Mattox [History Corner—A Short History of Reactive Evaporation, SVC Bulletin (Society of Vacuum Coaters, Spring 2014), p. 50–51] describes various methods for producing oxides, nitrides, carbides, and some compounds, but hydrides were not mentioned. A study was performed in the mid-1970s at the General Electric Company Neutron Devices Department in Largo, FL, by the author to study preparation of thin film hydrides using reactive evaporation and to determine their unique characteristics and properties. Films were produced of scandium (Sc), yttrium (Y), titanium (Ti), zirconium (Zr), and the rare earth praseodymium (Pr), neodymium (Nd), gadolinium (Gd), dysprosium (Dy), and erbium (Er) hydrides by hot crucible filament and electron beam evaporation in atmospheres of deuterium and tritium gases. All-metal vacuum systems were used and those used with tritium were dedicated for this processing. Thin film test samples 1000 nm thick were prepared on 1.27 cm diameter molybdenum disk substrates for each occluder (i.e., an element that can react with hydrogen to form a hydride) material. Loading characteristics as determined by gas-to-metal atomic ratios, oxidation characteristics as determined by argon–sputter Auger analysis, film structure as determined by scanning electron microscope analysis, and film stress properties as determined by a double resonator technique were used to define properties of interest. Results showed hydrogen-to-metal atomic ratios varied from 1.5 to 2.0 with near maximum loading for all but Pr and Nd occluders which correlated with the oxidation levels observed, with all occluder oxidation levels being variable due to vacuum system internal processing conditions and the materials used. Surface oxide levels varied from ∼80 Å to over 1000 Å. For most films studied, results showed that a maximum loading ratio of near 2.0 and a minimum surface oxide level of ∼80 Å could be obtained with a bulk film oxygen level of ∼0.54 oxygen as determined by microprobe analysis when an evaporation rate of ∼0.313 mg/cm{sup 2} min was used in an atmosphere of D{sub 2} or T{sub 2} gas at a system deposition pressure of 1 × 10{sup −3 }Torr (1.33 × 10{sup −1 }Pa) in an evaporation time of ∼2 min. Platelet type (i.e., a film microstructure showing an overlay of flat plates with large grain sizes) film structures were observed for most films with some film mechanical properties determined (i.e., grain size and Vickers μ-hardness), and reduced stress levels were seen with initial normalized differential (tensile) stress levels being (1.0–4.0) × 10{sup 8 }dyne/cm{sup 2} for tritium loaded samples and (1.5 ± 0.5) × 10{sup 9 }dyne/cm{sup 2} for deuterium loaded samples. Also, stress aging characteristics were determined for some hydride films prepared in a radioactive tritium gas atmosphere. Tritium loading, however, had the undesirable characteristic of having to dispose of the internal processing system fixtures, which can be minimized, but the reactive evaporation technique produced desirable thin films.

36 MATERIALS SCIENCE↗

DOE OSTI2013

Forest Irrigation of Tritiated Water: A Proven Tritiated Water Management Tool - 13357

Tritium releases from the Old Radioactive Waste Burial Ground (ORWBG) at the SRS in South Carolina has impacted groundwater and surface water. Tritiated groundwater plumes discharge into Fourmile Branch which is a small tributary of the Savannah River, a regional water resource. Taking advantage of the groundwater flow paths and the local topography a water collection and irrigation system was constructed and has been used at the SRS for over a decade to reduce these tritiated water releases to Fourmile Branch. The tritiated water is transferred to the atmosphere by evaporation from the pond surface, and after irrigation, wetted surface evaporation and evapotranspiration through the forest vegetation. Over the last decade SRS has irrigated over 120,000,000 gallons of tritiated water, which diverted over 6000 curies away from Fourmile Branch and the Savannah River. The system has been effective in reducing the flux of tritiated groundwater by approximately 70%. Mass balance studies of tritium in the forest soils before operations and over the last decade indicate that approximately 90% of the tritiated water that is irrigated is transferred to the atmosphere. Dose studies indicate that exposure to site workers and offsite maximally exposed individual is very low, approximately 6 mrem/year and 0.004 mrem/year, respectively. To consistently meet the flux reduction goal of tritium into Fourmile Branch optimization activities are proposed. These efforts will increase irrigation capacity and area. An additional 17 acres are proposed for an expansion of the area to be irrigated and a planting of approximately 40 acres of pine forest plantations is underway to expand irrigation capacity. Co-mingled with the tritiated groundwater are low concentrations of chlorinated volatile organic compounds (cVOCs), and 1,4-dioxane. Research studies and SRS field data indicate the forest irrigation system may have an added benefit of reducing the mass of these co-contaminants via degradation. This semi-passive system makes use of natural processes of hydrology and evapotranspiration to manage tritium-contaminated water by reducing its entrance into site streams and the Savannah River, as well as treating low levels of co-mingled VOCs. SRS expects to operate the system until the tritium decays to levels that represent a minimal impact to Fourmile Branch and the Savannah River, and meets the stakeholder expectations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Cryogenic system with GM cryocooler for krypton, xenon separation from hydrogen-helium purge gas

In the thorium molten salt reactor (TMSR), fission products such as krypton, xenon and tritium will be produced continuously in the process of nuclear fission reaction. A cryogenic system with a two stage GM cryocooler was designed to separate Kr, Xe, and H{sub 2} from helium purge gas. The temperatures of two stage heat exchanger condensation tanks were maintained at about 38 K and 4.5 K, respectively. The main fluid parameters of heat transfer were confirmed, and the structural heat exchanger equipment and cold box were designed. Designed concentrations after cryogenic separation of Kr, Xe and H{sub 2} in helium recycle gas are less than 1 ppb.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Assembly and Testing of a Tritium and Iodine Removal System for Use with Advanced Tritium Pretreatment

US regulations will require the removal of iodine and tritium, along with other volatile and semivolatile radionuclides, from the off-gas streams of nuclear fuel reprocessing plants. Tritium pretreatment is a head-end operation that could be incorporated within nuclear fuel reprocessing plants for the upfront removal and trapping of these volatiles species. Quantitative, up-front removal in a single compact system is of significant advantage because otherwise tritium and iodine distributes to other unit operations, each of which have associated off-gas streams also requiring treatment. Oak Ridge National Laboratory is presently developing an advanced tritium pretreatment process based on the oxidation of used nuclear fuel at low temperature using NO 2 /O 2 mixtures, which produces a fine powder. This process can result in the quantitative volatilization of both tritium and iodine. The powder product can be selected to be U 3 O 8 , UO 3 , or a nitrate by adjusting the processing conditions and the process can be readily interfaced to either conventional aqueous-based or advanced pyrochemical–based reprocessing plants. In preparation for a planned kilogram-scale hot cell demonstration of the advanced tritium pretreatment process with used nuclear fuel (UNF), an integrated iodine and tritium capture system was designed and assembled. Three experimental runs of this system were completed successfully. All the samples were submitted for analysis. The results and data analysis will be covered in a subsequent report to be issued by the Off-Gas Sigma Team. Several additional tests should be undertaken to provide additional information on the system’s performance. The objectives for the additional tests and test conditions are presented in this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗