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The development and initial testing of the Ice Pig cleaning method for nuclear reprocessing plants - 15363

There are numerous situations throughout the nuclear industry where it would be advantageous to clean out enclosed systems, e.g. ducts and pipes, to remove sediments and radioactive contamination. However, conventional flushing methods result in the production of significant volumes of contaminated liquor. The wall Shear achieved with water flushing is insufficient to remove and carry dense deposits often found in nuclear facilities. Conventional mechanical pigs can be propelled through the duct by application of a driving medium such as compressed air or water. These are able to achieve very high wall shears, but can also get physically stuck where there is variation in the pipe / duct diameter or small radius bends. The field of application is best suited to straight constant diameter systems with no discontinuities such as Tee pieces, valves and in-pipe instrumentation. Driven by the desire to clean and displace radioactive materials from complex topology ducts and pipes, without the risk of having conventional pigs becoming lodged in a plant and coupled with the need to minimize effluent volumes, the UK Nuclear Industry led by Sellafield Limited (on behalf of the UK Nuclear Decommissioning Authority), has worked to demonstrate and underpin the use of the ice pigging technology for the nuclear industry. It is found that high ice fraction slurries flow as viscous liquids that exhibit a number of mechanical properties that offer much higher shear rates. The 'Ice Pig' exerts shear rates on pipe walls which are typically 1000 times greater than those achieved with water flowing at the same velocity. The 'Ice Pig' is able to negotiate almost any topology, flow through pipe networks of vastly differing diameter, through heat exchangers and in-line instrumentation etc. Experimental data and operational experience from other industries demonstrates that whilst there is high shear, the 'Ice Pig' does not damage the surface of the pipe wall. A further attractive characteristic of the 'Ice Pig' is that it eventually melts rendering it a simple effluent for processing by downstream processes. Due to the special nature of ice slurry, its unique rheological and physical characteristics, together with the complementary underpinning experimental work undertaken by Sellafield Limited enables the Ice Pigging process to move from an 'interesting' technology to one suitable for use by the nuclear industry for operations, Post Operational Clean Out (POCO) and decommissioning. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Materials and Methods for Streamlined Laboratory Analysis of Environmental Samples, FY 2016 Report

The International Atomic Energy Agency (IAEA) relies upon laboratory analysis of environmental samples (typically referred to as “swipes”) collected during on-site inspections of safeguarded facilities to support the detection and deterrence of undeclared activities. Unfortunately, chemical processing and assay of the samples is slow and expensive. A rapid, effective, and simple extraction process and analysis method is needed to provide certified results with improved timeliness at reduced costs (principally in the form of reduced labor), while maintaining or improving sensitivity and efficacy. To address these safeguard needs the Pacific Northwest National Laboratory (PNNL) explored and demonstrated improved methods for environmental sample (ES) analysis. Improvements for both bulk and particle analysis were explored. To facilitate continuity and adoption, the new sampling materials and processing methods will be compatible with existing IAEA protocols for ES analysis. PNNL collaborated with Oak Ridge National Laboratory (ORNL), which performed independent validation of the new bulk analysis methods and compared performance to traditional IAEA’s Network of Analytical Laboratories (NWAL) protocol. ORNL efforts are reported separately. This report describes PNNL’s FY 2016 progress, which was focused on analytical application supporting environmental monitoring of uranium enrichment plants and nuclear fuel processing. In the future the technology could be applied to other safeguard applications and analytes related to fuel manufacturing, reprocessing, etc. PNNL’s FY 2016 efforts were broken into two tasks and a summary of progress, accomplishments and highlights are provided below. Principal progress and accomplishments on Task 1, Optimize Materials and Methods for ICP-MS Environmental Sample Analysis, are listed below. • Completed initial procedure for rapid uranium extraction from ES swipes based upon carbonate-peroxide chemistry (delivered to ORNL for evaluation). • Explored improvements to carbonate-peroxide rapid uranium extraction chemistry. • Evaluated new sampling materials and methods (in collaboration with ORNL). • Demonstrated successful ES extractions from standard and novel swipes for a wide range uranium compounds of interest including UO 2 F 2 and UO 2 (NO 3 ) 2 , U 3 O 8 and uranium ore concentrate. • Completed initial discussions with commercial suppliers of PTFE swipe materials. • Submitted one manuscript for publication. Two additional drafts are being prepared. Principal progress and accomplishments on Task 2, Optimize Materials and Methods for Direct SIMS Environmental Sample Analysis, are listed below. • Designed a SIMS swipe sample holder that retrofits into existing equipment and provides simple, effective, and rapid mounting of ES samples for direct assay while enabling automation and laboratory integration. • Identified preferred conductive sampling materials with better performance characteristics. • Ran samples on the new PNNL NWAL equivalent Cameca 1280 SIMS system. • Obtained excellent agreement between isotopic ratios for certified materials and direct SIMS assay of very low levels of LEU and HEU UO 2 F 2 particles on carbon fiber sampling material. Sample activities range from 1 to 500 CPM (uranium mass on sample is dependent upon specific isotope ratio but is frequently in the subnanogram range). • Found that the presence of the UF molecular ions, as measured by SIMS, provides chemical information about the particle that is separate from the uranium isotopics and strongly suggests that those particles originated from an UF6 enrichment activity. • Submitted one manuscript for publication. Another manuscript is in preparation.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Investigation of the initial and volume recombination losses in gamma versatile cylindrical ionization chamber VGIC developed for gamma ray dosimetry

A versatile Gamma ionization chambers are used for flow control in systems with gamma nuclear reactors and reprocessing plants in and monitoring atmosphere around these facilities, this in order to protect staff. In the Laboratory Detection and Measures (LDM) Division for Study and Development of Nuclear Instrumentation (DSDNI) of CRNB, we designed, developed and characterized a versatile gamma ionization chamber (VGIC) to study experimentally its characteristics according to the geometry of the electrodes, the volume and pressure of the filler gas for the design of a gamma sealed chamber. The tests were conducted under the IEC (International Electro-technical Commission). In this paper, we present the results obtained in the various nuclear tests for characterization and calibration that we have made on the ionization chamber gamma VGIC prototype developed at our Department. To do this, three irradiators were operated at the Laboratory Calibration (SSDL) of the Department of Medical Physics Nuclear Research Center of Algiers (CRNA). Irradiator intensive gamma ({sup 60}Co: 1.25 MeV), one medium intensity gamma ({sup 137}Cs: 0.662 MeV) and 3rd low intensity ({sup 60}Co). Saturation curves and linearity were identified and the operating range and the sensitivity of the chamber have been deducted. The (I,V) characteristics of the chamber filled, with argon gas at 3 bar (0.3 M pa) pressure, for gamma ray irradiator sources were studied. To do so, the chamber was irradiated with gamma rays using different numbers of gamma sources (i.e. Up to 5). The plateau region is reached above 200 V and the detector operating voltage is found to be 600 V. It is observed that in the plateau region the slope is constant with an increase in the exposure rate. The (1/I, 1/V) and (I, l/V{sup 2}) characteristic curves reveal the presence of the initial and volume recombination losses. The volume recombination losses are found to be smaller than the initial recombination losses. Finally, in order to strengthen our results, we performed in the same conditions, a comparative characterization with another ionization gamma chamber type reference LND 504 (USA). (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

LaserSnake2: An Innovative Approach to Nuclear Decommissioning - 17080

Nuclear decommissioning challenges exist the world over and require innovative approaches to ensure safe and cost effective dismantling of high hazard nuclear facilities. LaserSnake technology makes use of the flexibility and dexterity of snake arm robots to deploy high-powered laser cutting, lifting and manipulation tools. This paper describes the development and use of the system to achieve the successful size reduction and removal of a substantial stainless steel vessel within the First Generation Reprocessing Plant (FGRP) at Sellafield in the UK. Due to the environment, nuclear decommissioning is an extremely challenging activity. High ambient radiation levels often preclude entry to the areas involved, and as industry tries to minimise dose uptake, even moderate radiation levels now require remote solutions. 'In-situ' size reduction of process plant, including vessels and pipework, will allow the removal of the nuclear inventory from a building prior to further decontamination and demolition. The term 'in situ' is in this case used to mean work done on plant items within their existing cells as opposed to moving items from their original location to a dedicated size reduction facility. Traditional size reduction cutting methods, such as reciprocating saws or grinders, pose a significant challenge for remote deployment platforms. This includes the loss of haptic feedback and the length of time taken to perform the task, and difficulty dealing with the reaction forces. Although remote engineered solutions for these tools are possible, they are are typically complicated, unreliable and expensive. Robot deployed laser cutting is widely used in non-nuclear industrial sectors, and offers a simpler method for size reduction. LaserSnake2 is a UK collaborative project, co-funded by the Nuclear Decommissioning Authority, Innovate UK and the Department of Energy and Climate Change. The project combined the skills of robotic experts and laser and optic specialists, to develop a remote handling technology suited for deployment within active nuclear environments. As well as meeting strict safety case requirements, the system provides a safe and cost effective method for decommissioning of complex nuclear spaces. An overview of the system development and deployment is presented in this paper. In July 2016, a successful decommissioning campaign was conducted in the FGRP at Sellafield. The LaserSnake2 system was used to size reduce a double-skinned dissolver vessel within a radiologically active nuclear cell. Planning of the size reduction method and processes, as well as results of the campaign itself are described in this paper. The containment procedures and radiological considerations for the system are presented along with the effects of operations on the cell and ventilation systems. LaserSnake, developed as a modular arm system with up to 4.5 m of articulation, carries an integrated tool comprising a high power, decommissioning specific, laser cutting head, high definition cameras and high powered illumination LEDs. It is designed for precise dismantling work within confined and hazardous environments. The tool can be interchanged, allowing the snake-arm robot also to carry grippers, environmental mapping sensors and other tools which may prove advantageous during a decommissioning campaign. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Cryogenic gamma detectors enable direct detection of 236 U and minor actinides for non-destructive assay [Cryogenic gamma detectors enable direct detection of minor actinides for non-destructive assay]

Here, we demonstrate the utility of a superconducting transition edge sensor (TES) γ-ray detector with high energy resolution and low Compton background for non-destructive assay (NDA) of a uranium sample from reprocessed nuclear fuel. We show that TES γ-detectors can separate low energy actinide γ-emissions from the background and nearby lines, even from minor isotopes whose signals are often obscured in NDA with conventional Ge detectors. Superconducting γ detectors may therefore bridge the gap between high-accuracy destructive assay (DA) and easier to-use NDA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Neptunium redox speciation and determination of its total concentration in dissolved fuel simulant solutions by spectrophotometry

Here, two new approaches to measure Np concentration in dissolved used nuclear fuel simulant (aqueous feed for PUREX process) by spectrophotometry are developed. The first approach is based on chemical reduction of Np in the feed to its tetravalent state using ascorbic acid with simultaneous conversion of Pu(IV) to Pu(III). Interfering effects from light absorbing fission and corrosion products are accounted for by measuring optical absorbance spectrum of aqueous raffinate after extraction of U, Np, and Pu by tributyl phosphate in dodecane. The second approach uses no chemical treatment at all and relies on spontaneous valency adjustment of Np to Np(V) by dilution of the feed with water to reduce its acidity to low decimolar range of nitric acid concentration. Results of Np determination in the feed by spectrophotometry are in good agreement with its concentration measured by ICP-MS.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Group hexavalent actinide separation from lanthanides using sodium bismuthate chromatography

Advanced used nuclear fuel (UNF) reprocessing strategies are limited by the complex radiochemical separations and engineering required to achieve the separation of actinides (An) from neutron scavenging lanthanides (Ln). The accessibility of the hexavalent oxidation state for the actinides (U – Am) provides a pathway to achieving a group hexavalent actinide separation from the trivalent lanthanides and Cm. The solid oxidant and ion exchanger, sodium bismuthate (NaBiO 3 ), has been demonstrated to quantitatively oxidize and separate Am from trivalent Cm in a column chromatographic system. This work expands on the use of NaBiO 3 chromatography to characterize the adsorption, kinetic, and elution behavior of U, Pu, and Eu. Separation factors over 200 with rapid kinetics were observed at dilute nitric acid concentrations with a complete An/Ln separation achieved in under an hour. In conclusion, the adsorption and chromatographic behavior of key fission products present in various reprocessing raffinates was characterized which demonstrated potential application of a NaBiO 3 -based separation following a TRUEX process.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Preparation of a uranium monocarbide anode and electrochemical characterization in molten LiCl-KCl-UCl 3

Porous uranium carbide (UC) pellets possessing moderate electrical conductivity were synthesized by reaction of UO 2 with graphite at temperatures up to 1550°C under rough vacuum. Conversions as high as 98% were achieved at soak times of 2-4 hours. The electrochemistry of the UC pellets in molten LiCl-KCl-6.5 wt% UCl 3 was explored using a variety of techniques including DC polarization methods, cyclic voltammetry, chronopotentiometry and bulk electrolysis. Here, the electrode reaction for anodic dissolution was found to be kinetically controlled by dissociation of UC to a transition state complex that was hypothesized to consist of a uranium atom partially complexed by chloride ions. Precise measurements of current efficiencies using chronopotentiometry indicated upper limits of 90.9 ± 3.4% and 98.3 +1.7/-3.7% for anode and cathode, respectively, when operating at anodic overpotentials near +300 mV. Bulk electrolysis of a UC pellet performed by passing 98% of the theoretical charge resulted in nearly complete recovery of its uranium content as highly pure metal at the cathode.

36 MATERIALS SCIENCE↗

Analysis and modeling of the equilibrium behaviors of U and Pu in molten LiCl-KCl/Cd system at 500 °C

In the field of pyroprocessing of used nuclear fuels, liquid cadmium is applied in different ways to recover transuranics from molten salt, while effecting some degree of separations between transuranics and lanthanides. In this work, a model of the separation performance between uranium and plutonium across the phase boundary between molten salt and liquid cadmium is presented. The salt is the LiCl-KCl eutectic with solute concentrations of UCl 3 and PuCl 3 . The cadmium is the liquid cadmium phase field from the Cd-U-Pu ternary alloy system. Lastly, the model used to describe the behavior of uranium and plutonium in the salt/cadmium system is applied to describe the behavior of plutonium and lanthanide in the salt/cadmium system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Metal–Organic Framework–Polyacrylonitrile Composite Beads for Xenon Capture

Mechanically robust forms of HKUST-1 metal-organic frameworks (MOFs) were fabricated by embedding the MOF in a passive polyacrylonitrile (PAN) matrix at different MOF loadings of 10–90 mass%. Here, the PAN is highly porous and acts as a scaffold that holds the active MOF adsorbent in place for Xe capture. Data presented herein show that the PAN matrix does not notably interfere with the Xe capture process where the Xe capacities scale somewhat linearly with the increase in MOF loading in the composites. Radiation testing of the composites under a gamma flux revealed that they are highly tolerant to these types of radiation fields.

36 MATERIALS SCIENCE↗

Iodine Adsorption in Metal Organic Frameworks in the Presence of Humidity

Used nuclear fuel (UNF) reprocessing represents a unique challenge when dealing with radionuclides such as isotopes of 85 Kr and 129 I 2 , due to their volatility and long half-life. However, efficient capture of 129 I 2 (t 1/2 = 15.7 x 10 6 years) from the nuclear waste stream can help to reduce the risk of releasing I 2 radionuclide into the environment and mitigate concerns about human health problems. Metal organic frameworks (MOFs) have been reported to be potential I 2 adsorbents: but the effect of water vapor, generally present in the reprocessing off-gas stream is rarely taken into account. Moisture stable porous MOFs, which can selectively adsorb I 2 in presence of water vapor is thus of great interest. Herein, the I 2 adsorption performance of two microporous MOFs is reported in presence of different humidity. I…π phenyl ring interactions are mainly responsible for the adsorption as revealed by single crystal XRD

Iodine↗

Actinide Lanthanide Separation Process – ALSEP

Separation of the minor actinides (Am, Cm) from the lanthanides at an industrial scale remains a significant technical challenge for closing the nuclear fuel cycle. To increase the safety of used nuclear fuel (UNF) reprocessing, as well as reduce associated costs, a novel solvent extraction process has been developed. The process allows for partitioning minor actinides, lanthanides and fission products following uranium/plutonium/neptunium removal; minimizing the number of separation steps, flowsheets, chemical consumption, and waste. This new process, Actinide Lanthanide SEParation (ALSEP), uses an organic solvent consisting of a neutral diglycolamide extractant, either N,N,N',N'-tetra(2 ethylhexyl)diglycolamide (T2EHDGA) or N,N,N',N'-tetraoctyldiglycolamide (TODGA), and an acidic extractant 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]), dissolved in an aliphatic diluent (e.g. n-dodecane). The An/Ln co-extraction is conducted from moderate-to-strong nitric acid, while the selective stripping of the minor actinides from the lanthanides is carried out using a polyaminocarboxylic acid/citrate buffered solution at pH anywhere between 3 and 4.5. The extraction and separation of the actinides from the fission products is very effective in a wide range of HNO3 concentrations and the minimum separation factors for lanthanide/Am exceed 30 for Nd/Am, reaching > 60 for Eu/Am under some conditions. The experimental results presented here demonstrate the great potential for a combined system, consisting of a neutral extractant such as T2EHDGA or TODGA, and an acidic extractant such as HEH[EHP], for separating the minor actinides from the lanthanides.

actinide separation↗

Simulant Testing of a Co-Decontamination (CoDCon) Flowsheet for a Product with a Controlled Uranium-to-Plutonium Ratio

Multiple tests of a co-decontamination (CoDCon) tributyl phosphate (TBP) based solvent extraction flowsheet are being performed to determine the uncertainties associated with maintaining a target U/Pu ratio in the Pu-containing product. The first two flowsheet tests used simple dissolved fuel simulants containing only U and Pu, but increasingly more complex simulants will but used in subsequent tests, with the ultimate goal of performing tests with irradiated fuel. The flowsheet is designed so as not to create a purified Pu stream, but rather a mixture of U and Pu with a U/Pu mass ratio of 7/3. Uranium and Pu are first co-extracted into the TBP solvent, and then U(IV) is used to strip the Pu from the loaded solvent. A fresh TBP solvent stream is introduced at the same stage the Pu-containing aqueous stream exits. The flowrate of the fresh TBP solvent stream is adjusted so that sufficient U bleeds over into the Pu product to produce the desired U/Pu ratio. The system is monitored in real time using optical spectroscopy, which allows for control of the U/Pu ratio through adjustment of the fresh TBP solvent flowrate. This paper will describe the results of the first two CoDCon flowsheet tests.

CoDCon, fuel recycle, fuel reprocessing, on-line m↗

A Review of Online Monitoring within Used Nuclear Fuel Recycling Processes

The processing of used nuclear fuels and related materials is often complex and variable. The ability to quickly optimize conditions to the material being processed can aid in increasing efficiency and safety, but requires very quick determination of the conditions present in the feedstock, the process, and the product. Furthermore, accurate quantification of materials such as enriched uranium and plutonium aids in maintaining material accountancy and avoiding nuclear proliferation risks. Traditional analytical methods require process samples to be collected and analyzed in a laboratory, which often takes days to weeks. Online monitoring is suitable for collecting this information nearly instantaneously, enabling much faster optimization of the process or detection of material diversion. Online monitoring is also beneficial as it is typically based on robust and nondestructive analytical methods, so no material is removed as samples. This review examines online monitoring relevant to used nuclear fuel processing for the determination of both chemical and physical parameters. The chemical parameters include quantities such as concentration, isotopic composition, and speciation. These values are often well suited to spectroscopic or spectrometric measurements as they are fast, nondestructive, and easily implemented in an online manner. Physical quantities are often more varied and include temperature, pressure, tank fill levels, and others. Due to the specificity of these quantities, specialized instrumentation is often used. However, this instrumentation is often amendable to online monitoring.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Bulk Handling Facility Modeling and Simulation for Safeguards Analysis

The Separation and Safeguards Performance Model (SSPM) uses MATLAB/Simulink to provide a tool for safeguards analysis of bulk handling nuclear processing facilities. Models of aqueous and electrochemical reprocessing, enrichment, fuel fabrication, and molten salt reactor facilities have been developed to date. These models are used for designing the overall safeguards system, examining new safeguards approaches, virtually testing new measurement instrumentation, and analyzing diversion scenarios. The key metrics generated by the models include overall measurement uncertainty and detection probability for various material diversion or facility misuse scenarios. Safeguards modeling allows for rapid and cost-effective analysis for Safeguards by Design. The models are currently being used to explore alternative safeguards approaches, including more reliance on process monitoring data to reduce the need for destructive analysis that adds considerable burden to international safeguards. Machine learning techniques are being applied, but these techniques need large amounts of data for training and testing the algorithms. The SSPM can provide that training data. This paper will describe the SSPM and its use for applying both traditional nuclear material accountancy and newer machine learning options.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Fukushima Daiichi Information Repository FY13 Status

The accident at the Fukushima Daiichi nuclear power station in Japan is one of the most serious in commercial nuclear power plant operating history. Much will be learned that may be applicable to the U.S. reactor fleet, nuclear fuel cycle facilities, and supporting systems, and the international reactor fleet. For example, lessons from Fukushima Daiichi may be applied to emergency response planning, reactor operator training, accident scenario modeling, human factors engineering, radiation protection, and accident mitigation; as well as influence U.S. policies towards the nuclear fuel cycle including power generation, and spent fuel storage, reprocessing, and disposal. This document describes the database used to establish a centralized information repository to store and manage the Fukushima data that has been gathered. The data is stored in a secured (password protected and encrypted) repository that is searchable and available to researchers at diverse locations.

97 MATHEMATICS AND COMPUTING↗

Design, Fabrication, and Shakeout Testing of ATALANTE Dissolver Off-Gas Sorbent-Based Capture System

A sorbent-based capture system designed for integration into the existing dissolver off-gas (DOG) treatment system at the ATelier Alpha et Laboratoires pour ANalyses, Transuraniens et Etudes de retraitement (ATALANTE) facility has been successfully designed and fabricated and has undergone shakeout testing. Discussions with personnel from the ATALANTE facility provided guidance that was used for the design. All components for this system were specified, procured, and received on site at Oak Ridge National Laboratory (ORNL). The system was then fabricated and tested at ORNL to verify operation. Shakeout testing resulted in a simplified system. This system should be easily installed into the existing facility and should be straightforward to operate during future experimental testing. All parts were selected to be compatible with ATALANTE power supplies, space requirements, and the existing DOG treatment system. Additionally, the system was demonstrated to meet all of four design requirements. These include (1) a dissolver off-gas flow rate of ≤100 L/h (1.67 L/min), (2) an external temperature of ≤50°C for all system components placed in the hot cell, (3) a sorbent bed temperature of ~150°C, and (4) a gas temperature of ~150°C upon entry into the sorbent bed. The system will be ready for shipment and installation in the existing DOG treatment system at ATALANTE in FY 2016.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Stability of Tritium and Iodine Sorbents in TPOG Conditions

US regulations could require the removal of both iodine and tritium from the off-gas stream of a used nuclear fuel (UNF) reprocessing facility. Advanced tritium pretreatment is a pretreatment step that uses high concentrations of NOR2R in a gas stream to volatilize tritium and iodine from UNF prior to traditional dissolution. The gaseous effluent from this process would then require abatement to remove tritium and iodine, but high levels of NOR2R could have a detrimental effect on the ability of various solid sorbents to remove the volatile radionuclides. For tritium and iodine, the sorbents of interest are 3Å molecular sieve (3AMS) for tritium and reduced silver mordenite (AgP 0 PZ), silver-functionalized silica-aerogel (AgAerogel), and silver-nitrate-impregnated alumina (AgA) for iodine. Prior research has demonstrated that exposure to high concentrations of NOR2R can reduce the iodine loading capacity of AgP 0 PZ by > 90% when exposed for 1 week. Research in Japan has demonstrated that AgA is more robust to NOR2R exposure than AgZ. The testing described here was intended to assess the effects of high concentrations of NOR2R on the iodine capture capacity of AgA and the water adsorption capacity of 3AMS. To determine the effect of extended exposure of the sorbents to NOR2R, both 3AMS and AgA were aged in a 75% NOR2R environment prior to loading. The 3AMS samples were aged for 1, 4, and 5.5 weeks at 40°C. They were then loaded with water in a 10°C dew point stream (corresponding to a water concentration of ~12,000 ppmv) at 40°C. There was no significant change in the water adsorption capacity of the 3AMS upon exposure to 75% NOR2R. The AgA samples were aged for 1, 2, and 4 weeks at 150°C and were loaded with 50 ppmv IR2R at 150°C. The results show that the iodine capture capacity of AgA is reduced by exposure to high concentrations of NOR2R. The iodine capacity reductions were 16%, 36%, and 76% for 1, 2, and 4 week exposures, respectively. This is less of a capacity loss than that seen in similar testing with the AgP 0 PZ sorbent.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗