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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Multidimensional Multiphysics Simulation of Nuclear Fuel Behavior

Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non- axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our multidimensional, multiphysics approach to analyze a missing pellet surface problem. The next example is the analysis of cesium diffusion in a TRISO fuel particle with defects. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Analysis of an MCU HEPA filter

A series of direct analyses on three portions (inlet, center, and outlet) of the High Efficiency Particulate Air (HEPA) filter material from the Modular Caustic-Side Solvent Extraction Unit (MCU) have been performed; this includes x-ray methods such as X-Ray Diffraction (XRD), Contained Scanning Electron Microscopy (CSEM) and X-Ray Fluorescence (XRF), as well as Fourier Transform InfraRed spectroscopy (FTIR). Additionally, two leaching studies (one with water, one with dichloromethane) have been performed on three portions (inlet, center, and outlet) of the HEPA filter material, with the leachates being analyzed by Inductively-coupled plasma emission spectroscopy (ICPES), Semi-Volatile Organic Analysis (SVOA) and gammascan. From the results of the analyses, SRNL feels that cesium-depleted solvent is being introduced into the HEPA filter. The most likely avenue for this is mechanical aerosolization of solvent, where the aerosol is then carried along an airstream into the HEPA filter. Once introduced into the HEPA filter media, the solvent wicks throughout the material, and migrates towards the outlet end. Once on the outlet end, continual drying could cause particulate flakes to exit the filter and travel farther down the airstream path.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Innovative Process for Comprehensive Treatment of Liquid Radioactive Waste - 12551

This paper presents the results of research activities aimed at creation of a principally new LRW distilling treatment method. The new process is based on the instantaneous evaporation method widely used in distillation units. The main difference of the proposed process is that the vapor condensation is conducted without using heat exchangers in practically ideal mode by way of direct contacting in a vapor-liquid system. This process is conducted in a specially designed ejector unit in supersonic mode. Further recuperation of excess heat of vaporization is carried out in a standard heat exchanger. Such an arrangement of the process, together with use of the barometric height principle, allows to carry out LRW evaporation under low temperatures, which enables to use excess heat from NPS for heating initial LRW. Thermal calculations and model experiments have revealed that, in this case, the expenditure of energy for LRW treatment by distilling will not exceed 3 kilowatt-hour/m{sup 3}, which is comparable with the reverse-osmosis desalination method. Besides, the proposed devices are 4 to 5 times less metal-intensive than standard evaporation units. These devices are also characterized by versatility. Experiments have revealed that the new method can be used for evaporation of practically any types of LRW, including those containing a considerable amount of oil products. Owing to arrangement of the evaporation process at low temperatures, the new devices are not sensitive to 'scale formation'. This is why, they can be used for concentrating brines of up to 500-600 g/l. New types of such evaporating devices can be required both for LRW treatment processes at nuclear-power plants under design and for treating 'non-standard' LRW with complex physicochemical and radionuclide composition resulting from the disaster at the Fukushima I Nuclear Power Plant.) As a result of accidents at nuclear energy objects, as it has recently happened at NPP 'Fukushima-1', personnel faces the necessity to take emergency measures and to use marine water for cooling of reactor zone in contravention of the technological regulations. In these cases significant amount of liquid radioactive wastes of complex physicochemical composition is being generated, the purification of which by traditional methods is close to impossible. According to the practice of elimination of the accident after-effects at NPP 'Fukushima' there are still no technical means for the efficient purification of liquid radioactive wastes of complex composition like marine water from radionuclides. Therefore development of state-of-the-art highly efficient facilities capable of fast and safe purification of big amounts of liquid radioactive wastes of complex physicochemical composition from radionuclides turns to be utterly topical problem. Cesium radionuclides, being extremely dangerous for the environment, present over 90% of total radioactivity contained in liquid radioactive wastes left as a result of accidents at nuclear power objects. For the purpose of radiation accidents aftereffects liquidation VNIIHT proposes to create a plant for LRW reprocessing, consisting of 4 major technological modules: Module of LRW pretreatment to remove mechanical and organic impurities including oil products; Module of sorption purification of LWR by means of selective inorganic sorbents; Module of reverse osmotic purification and desalination; Module of deep evaporation of LRW concentrates. The first free modules are based on completed technological and designing concepts implemented by VNIIHT in the framework of LLRW Project in the period of 2000-2001 in Russia for comprehensive treatment of LWR of atomic fleet. These industrial plants proved to be highly efficient and secure during their long operation life. Module of deep evaporation is a new technological development. It will ensure conduction of evaporation and purification of LRW of different physicochemical composition, including those containing hardness salts, resulted in generation of LRW concentrate 300-600 g/l. The method is based on utilization of supersonic ejector for intensification of thermal physic processes and performance of evaporation in brine recycling mode. All proposed technological solutions are totally based on patented Russian developments. Proposed work will allow to construct modular plants, which will be totally prepared for efficient purification of any types of liquid radioactive wastes from radionuclides in case of force majeure. According to proposed scheme concentration level of cesium radionuclides in safe-for-storage form will make up not less than 5000. With respect to purification from cesium radionuclides of liquid radioactive wastes stored at NPP 'Fukushima' about 10 t of inorganic sorbents, loaded in 160 protective filter-containers, will be required for solving this problem. The amount of secondary wastes will be reduced approximately in 5 times in comparison with traditional schemes, applied in purification of secondary LRW of Fukushima-1 by Areva (France) and Kurion (USA) companies. All units of modular plants will be constructed and manufactured as totally automated, providing their twenty-four-hour safe operation. Modular design will ensure efficiency and let optimize the costs of secondary LRW treatment. In order to ensure off-line operation in emergency conditions the plant should be equipped with auxiliary modules: energy and ventilation ones. Under normal conditions these modules can be stored in 'mothballed' condition at special warehouses under the authority of federal bodies. It will be reasonable to choose required transport facilities, the most suitable for transportation of modules to target destination beforehand, using vessel classification list.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Evaluation of Low-cost, Full HD, CMOS Cameras for Use in the Calcine Retrieval Project - 19530

This paper describes results from radiation testing conducted on low-cost, commercial off-the-shelf, full high-definition (HD) resolution (1920 x 1080 p), complementary metal-oxide-semiconductor (CMOS) camera sensors. The purpose of the testing was to evaluate the sensors' suitability for use in the Calcine Retrieval Project by determining their image quality and life expectancy while operating in an ionizing radiation field of 20.0 Gy/h. The two camera sensors tested were the Axis FA1105 and Axis FA1125. These commercial camera sensors were designed for retail security surveillance and have no radiation-hardened (rad-hard) components. In addition to radiation testing, the sensors' field-of-view, resolution, and low-light image quality were tested at several target distances to ensure they meet project requirements. Testing was divided into two phases. During Phase 1, four FA1105 sensors and four FA1125 sensors were tested simultaneously without any shielding in radiation fields of 25.5 Gy/h and 49.3 Gy/h. In Phase 2, three FA1105 and five FA1125 sensors were shielded by 22.2 mm of lead and tested simultaneously in radiation fields of 26.8 Gy/h, 52.3 Gy/h, 143.0 Gy/h, 200.7 Gy/h, 404.6 Gy/h, and 2,317.5 Gy/h. Dose rate values were established using film dosimetry readings. Apertures in the shielding allowed for a 40-degree field of view while shielding all parts of the sensor except for directly in front of the CMOS sensor array. All testing was conducted at the Idaho Accelerator Center in Pocatello, Idaho, using a 450 keV X-ray generator. Sensor images were automatically captured at regular intervals throughout the testing for qualitative evaluation by the test team. Film dosimetry was placed near the sensors to record actual dose totals. The 450 keV setting generates X-ray radiation with an energy level close to the primary gamma energy level produced by the Cesium-137 isotope in the calcine, which is 662 keV. Results from Phase 1 indicate the sensors failed at an average cumulative dose of 226.3 Gy. For a given 20.0 Gy/h field this results in a mean time to failure (MTTF) of 11.3 hours. Recorded images from all sensors showed minimal image degradation due to radiation-induced noise. When the sensors finally failed, the failure mode was an abrupt loss of video. The image went black suggesting some other component in the sensor head besides the CMOS sensor array was responsible for the failure. During a portion of Phase 1 testing, a commercial, rad-hard camera was irradiated in the 25.5 Gy/h field along with the eight Axis FA series sensors. Image quality from the FA series sensors significantly outperformed the image quality of the rad-hard camera in radiation-induced noise tolerance, image resolution, and dynamic range. During Phase 2 testing, all sensors produced excellent images in radiation fields as high as 404.6 Gy/h. One FA1125 sensor was tested at the highest dose rate of 2,317.5 Gy/h. At this dose rate, the image was color shifted towards red and contained a significant amount radiation-induced noise. The sensor did not fail and was producing usable imagery after one hour, at which time the test was ended due to allotted beam-time at the test facility. Eighteen hours after X-rays were shut off, the sensor image returned to normal color and all eight sensors continued to operate without any visible, permanent image degradation. For Phase 2 testing, it was not possible to establish a MTTF since none of the shielded sensors failed during the allotted test time. However, given the highest cumulative dose of 4,142.1 Gy received by a shielded FA1125 sensor, a minimum life expectancy of 207 hours (8.6 days) is predicted while operating in a 20.0 Gy/h field. Test results indicate the radiation tolerance of the FA series sensor's CMOS array, along with the small form factor, full HD image quality, wide dynamic range, and Ethernet standard makes it suitable for short-term use in monitoring and visually documenting waste remediation efforts for the Calcine Retrieval Project at the INL. The sensor's low-cost of about $140 US, compared to commercial rad-hard cameras costing between $5,000 and $7,000 US each, allow them to be used as an affordable, disposable resource for many short-duration, radioactive waste remediation efforts where decontamination of an expensive, commercial rad-hard camera is undesirable or not possible. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Multiscale, mechanistic modeling of cesium transport in silicon carbide for TRISO fuel performance prediction

Understanding cesium (Cs) transport in TRistructural ISOtropic (TRISO) particle fuel is crucial for predicting fission product release in high-temperature reactors. However, current challenges include significant scatter in diffusivity data and unexplained temperature-dependent diffusion regimes in the silicon carbide layer. This study addresses these challenges by developing a multiscale, mechanistic Cs transport model integrating atomistic simulations and phase field modeling. Our model quantifies temperature and grain size effects on Cs diffusivity, attributing experimentally observed regimes to a transition from bulk-dominated diffusivity at high temperatures to grain boundary-dominated diffusivity at lower temperatures. The model, validated against diffusion measurements and advanced gas reactor (AGR)-1 and AGR-2 post-irradiation fission product release data, enhances the predictive capability of the BISON fuel performance code. This study advances our understanding of Cs release from TRISO particles and its dependence on temperature and silicon carbide grain size, with implications for the safety and efficiency of high-temperature nuclear reactors.

BISON↗
DOE OSTI2022

Silicon Photomultipliers Coupled to Scintillators With the Emission Maximum at 550 nm

A majority of the silicon photomultipliers (SiPMs) are sensitive to blue and near-ultraviolet (NUV) photons that are not optimized for scintillators with the emission maximum at wavelengths longer than 500 nm. The red–green–blue (RGB) SiPM is developed for the maximum photon detection efficiency (PDE) at 550 nm, which is a good match for some high-light-yield scintillators, such as CsI:Tl (CsI) and Gd 1.5 Y 1.5 Ga 2 Al 3 O 12 :Ce (GYGAG). Comparisons are made for the performance of these scintillators coupled to the RGB SiPM and a blue-sensitive SiPM. Because it takes tens of nanoseconds for the microcells to recharge after registering a photon hit, the linearity of these scintillation detectors was studied for high-energy gammas where numerous scintillation photons are generated. In addition, the energy resolution of the 662-keV gamma emitted by 137 Cs was measured for temperatures between –20 °C and 50 °C. The nonlinearity was observed above 1 MeV for all measurements, and however, it can be corrected by energy calibration using a third-degree polynomial. For CsI, the energy resolution is better with the blue-sensitive SiPM because of the lower dark count rate (DCR). In contrast, GYGAG coupled to the RGB SiPM has a better energy resolution for temperatures below 30 °C because of the well-matched emission spectrum and PDE distribution. Nevertheless, the advantage disappears for temperatures above 30 °C due to the higher DCR. It would be useful to further develop the RGB SiPM with a lower DCR and higher operating temperatures.

42 ENGINEERING↗
DOE OSTI2018

Electronics Upgrades to the Green is Clean Phoswich Detector Systems and Programmatic Implementation at LANL - Phase II Completion - 18403

Los Alamos National Laboratory (LANL) radiological facilities produce low-density room trash that, in many cases, is not contaminated with radioactivity. It has been estimated that 50 to 90% of low-density room trash is free of radioactive contamination and eligible for inclusion in LANL's Green is Clean (GIC) program. The GIC program is a verification program for non-regulated waste from radiation controlled areas that has been actively segregated as nonradioactive through the use of the waste generator's acceptable knowledge. The GIC program has employed two large area Phoswich gamma ray detector systems (HERCULES and ZEUS) to verify that GIC candidate low-density waste contains no radioactivity above background. The large area (127 cm{sup 2}) Phoswich detectors consist of a thin NaI front crystal (3 mm thick) that is optically coupled to a thick CsI back crystal (50 mm thick). The scintillation properties of the two crystals are different enough to allow the system's electronic components to identify the origin of any pulse. The original systems utilize Nuclear Instrumentation Module (NIM) electronics to process the detector signals. The electronics currently in use by the systems are out of date and some components are no longer manufactured. In 2015, phase I upgrades to the GIC detector systems were completed. Phase I objectives included identifying more modern commercially available electronics that could replace the NIM electronics. A pulse shape discriminator (PSD) module, the Digital Gamma Finder (DGF) Pixie-4 made by XIA LLC, that fits in a compact National Instruments{sup TM} PXI chassis was identified to replace the majority of the original NIM bin electronics - a single Pixie-4 module replaces four NIM amplifiers and four NIM pulse shape analyzers (PSAs). Another phase I objective was to show that that new PSD could separate out the NaI and CsI detector signals in a manner equivalent to or better than the NIM electronics. Proof of concept was achieved in 2015 when a prototype upgrade system on ZEUS was built and tested and gating parameter settings were determined that successfully separated the NaI and CsI signals. Phase II required establishing a LabVIEW software interface with Pixie-4 electronics in lieu of the manufacturer provided Pixie-4 software called the Pixie Viewer. In addition, phase II included 1) final optimization of the gating parameters on the prototype Pixie-4 ZEUS system, 2) finalizing Requirements, Design and Test Plan documents, 3) LabVIEW software interface quality assurance (QA) testing, 4) setting up the HERCULES system with the Pixie-4 upgrade electronics, 5) check source and background calibration of both systems, 6) absolute efficiency determination on both systems and 7) procedure revisions and implementation of both systems into routine operations. In 2017 phase II was completed including the preparation of detailed Requirements, Design and Test Plan documents. The LabVIEW interface software that communicates and controls the Pixie-4 has been completed, merged with the preexisting LabVIEW operations software and fully tested. Gating parameter optimization, configuration of HERCULES with the Pixie-4 electronics, check source and background calibrations and absolute efficiency determinations have all been performed - quantitative results are presented in the full paper. For performance validation, comparison measurements of actual GIC bags on both the HERCULES system using the old NIM Bin electronics and the ZEUS system using the upgraded Pixie-4 electronics have been done and evaluated. Finally, procedure revisions and implementation of both systems into routine operations is currently in process and planned to be complete by the end of calendar year 2017. (authors)

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

The control room radiological habitability assessment of high temperature reactor pebble bed module (HTR-PM) - 18465

The control room radiological habitability is important for staff safety in nuclear power plant accident, which is also a licensing requirement of the High Temperature Reactor Pebble Bed Module (HTR-PM) in China. In this study, a comprehensive evaluation of the control room radiological habitability is performed for three design basis accident scenario of HTR-PM. For each scenario, three exposure pathways are taken into account, which includes the immersion external exposure, the inhalation exposure and the direct exposure from radioactive cloud in the environment. The atmospheric concentration is calculated using the NRC-recommended ARCON96 code, based on the two-year meteorological observation data at HTR-PM s site and the final building layout. The evaluation result demonstrates that the individual dose level of staffs in the control room is far below the requirement of the regulatory, which guarantees the radiological habitability of control room of HTR-PM. The contribution of each radionuclide has been analyzed, which suggests that iodine and metal radionuclides such as Cs-137 and Co-60 are primary contributor of the total dose, which are worthy of particular attention in the control room habitability design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Economic Feasibility Analysis and Engineering System Design Development of Novel Cesium Adsorbents - 18680

Potassium copper hexacyanoferrate-cellulose hydrogel composites (HCF-gels) and potassium copper hexacyanoferrate-embedded poly-citric acid hydrogel film (HPC) were synthesized. In Cs{sup +} removal experiments, HCF-gels exhibited exceptional Cs{sup +} adsorption capacities 2.06-2.32 mmol g{sup -1}. HPC showed a very high Cs{sup +} uptake of 667 mg/g KCuHCF. The adsorbents revealed stable and high Cs{sup +} removal efficiency across a wide pH range. Notably, the adsorbents could adsorb Cs{sup +} selectively in seawater containing Cs{sup +}. The combination of testing and radiation stability results indicated that the performance of the materials are maintained whether irradiated or not. The cost of the HCF-gels and HPC were estimated approximately 13.0055 $/kg and 38.2490 $/kg respectively. This is reasonable cost compared to commercialized Cs{sup +} adsorbents. New pool scrubbing system is designed for engineering system design to apply novel Cs{sup +} adsorbents under severe accident condition. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Cesium Removal Performance Comparisons of Crystalline Silicotitanate Media Batches with Savannah River Site Waste Simulant

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium salt cake solids) within the tank followed by at-tank ion exchange column treatment. Plans are underway to prepare and install a second TCCR unit at SRS. Capacity and particle size differences exist between archived (IE-911) and more recently-prepared CST media batches (9120- B and 9140-B). Side-by-side comparison testing was performed to evaluate the cesium removal performance of each batch to aid in selecting the preferred CST batch and media characteristics to load into the second TCCR unit. Batch contact equilibrium and flow-through column tests have been conducted with three CST batches using an SRS Average Simulant. Simulant batch contact equilibrium cesium loading results for the three CST batches (including two lots of one batch) are provided in Table ES-1. The 35 °C data indicates that the archived IE-911 batch has a higher cesium capacity than recently-prepared CST media and that the minor TCCR CST 9120-B media lot (2099000035) has similar cesium removal performance to the major TCCR lot (2099000034). Tests conducted at 25 ºC for the archived IE-911 CST batch indicated lower cesium removal performance with this simulant batch than was observed recently with a different SRS Average Simulant batch. A dilution factor (DF) is typically utilized when modeling engineered CST media cesium loading performance to account for mass contributions from the binder material. In cases where CST performance is lower than expected, this factor includes corrections for the binder and for low performance. The ZAM (Zheng, Anthony, and Miller) Isotherm Model DF values are provided in Table ES-1 for each simulant batch contact result. DF values near 0.5 were determined for the 9120-B and 9140-B CST batch contact tests while DF values near 0.6 (20% higher) were determined for the tests with IE-911. These dilution factors are lower than recently observed with a different SRS Average Simulant batch (9140-B DF = 0.68; IE-911 DF = 1.0). Based on these results, it appears that some component in the simulant solution used for equilibrium testing may have resulted in reduced cesium loading on the CST media.

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

Extraction-Scrub-Strip test results from the interim Salt Disposition Program Macrobatch 9 Tank 21H qualification samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 9 for the Interim Salt Disposition Program (ISDP). The Salt Batch 9 characterization results were previously reported. An Extraction-Scrub-Strip (ESS) test was performed to determine cesium distribution ratios (D (Cs) ) and cesium concentration in the strip effluent and decontaminated salt solution (DSS) streams; this data will be used by Tank Farm Engineering to project a cesium decontamination factor (DF). This test used actual Tank 21H material, and a blend solvent prepared by SRNL that mimics the solvent composition currently being used at the Modular Caustic-Side Solvent Extraction Unit (MCU). The ESS test showed acceptable performance with an extraction D (Cs) value of 52.4. This value is consistent with results from previous salt batch ESS tests using similar solvent formulations. This compares well against the predicted value of 56.5 from a recently created D (Cs) model

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

Extraction, -scrub, -strip test results from the interim salt disposition program macrobatch 10 tank 21H qualification samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 10 for the Interim Salt Disposition Program (ISDP). The Salt Batch 10 characterization results were previously reported.ii,iii An Extraction, -Scrub, -Strip (ESS) test was performed to determine cesium distribution ratios (D(Cs)) and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Tank Farm Engineering to project a cesium decontamination factor (DF). This test used actual Tank 21H material, and a sample of the NGS Blend solvent currently being used at the Modular Caustic-Side Solvent Extraction Unit (MCU). The ESS test showed acceptable performance with an extraction D(Cs) value of 110. This value is consistent with results from previous salt batch ESS tests using similar solvent formulations. This is better than the predicted value of 39.8 from a recently created D(Cs) model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Extraction, Scrub, and Strip Test Results from The Interim Salt Disposition Program Macrobatch 11 Tank 21H Qualification Samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 11 for the Interim Salt Disposition Program (ISDP). The Salt Batch 11 characterization results were previously reported. An Extraction, Scrub, and Strip (ESS) test was performed to determine cesium distribution ratios (D(Cs)) and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Tank Farm Engineering to project a cesium decontamination factor (DF). This test used actual Tank 21H material, and a sample of the Next Generation Solvent (NGS) Blend solvent currently being used at the Modular Caustic-Side Solvent Extraction Unit (MCU). The ESS test showed acceptable performance with an extraction D(Cs) value of 117. This value is consistent with results from previous salt batch ESS tests using similar solvent formulations. This cesium removal efficiency is better than the predicted value of 50.8 from a recently created D (Cs) model.

Peters, T.↗
DOE OSTI2015

Cs{sub 7}Sm{sub 11}[TeO{sub 3}]{sub 12}Cl{sub 16} and Rb{sub 7}Nd{sub 11}[TeO{sub 3}]{sub 12}Br{sub 16}, the new tellurite halides of the tetragonal Rb{sub 6}LiNd{sub 11}[SeO{sub 3}]{sub 12}Cl{sub 16} structure type

Two new rare-earth – alkali – tellurium oxide halides were synthesized by a salt flux technique and characterized by single-crystal X-ray diffraction. The structures of the new compounds Cs{sub 7}Sm{sub 11}[TeO{sub 3}]{sub 12}Cl{sub 16} (I) and Rb{sub 7}Nd{sub 11}[TeO{sub 3}]{sub 12}Br{sub 16} (II) (both tetragonal, space group I4/mcm) correspond to the sequence of [MLn{sub 11}(TeO{sub 3}){sub 12}] and [M{sub 6}X{sub 16}] layers and bear very strong similarities to those of known selenite analogs. We discuss the trends in similarities and differences in compositions and structural details between the Se and Te compounds; more members of the family are predicted. - Graphical abstract: Two new rare-earth – alkali – tellurium oxide halides were predicted and synthesized. - Highlights: • Two new rare-earth – alkali – tellurium oxide halides were synthesized. • They adopt slab structure of rare earth-tellurium-oxygen and CsCl-like slabs. • The Br-based CsCl-like slabs have been observed first in this layered family.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Fukushima Nuclear Crisis Recovery: A Modular Water Treatment System Deployed in Seven Weeks - 12489

On March 11, 2011, the magnitude 9.0 Great East Japan earthquake, Tohoku, hit off the Fukushima coast of Japan. This was one of the most powerful earthquakes in recorded history and the most powerful one known to have hit Japan. The ensuing tsunami devastated a huge area resulting in some 25,000 persons confirmed dead or missing. The perfect storm was complete when the tsunami then found the four reactor, Fukushima-Daiichi Nuclear Station directly in its destructive path. While recovery systems admirably survived the powerful earthquake, the seawater from the tsunami knocked the emergency cooling systems out and did extensive damage to the plant and site. Subsequent hydrogen generation caused explosions which extended this damage to a new level and further flooded the buildings with highly contaminated water. Some 2 million people were evacuated from a fifty mile radius of the area and evaluation and cleanup began. Teams were assembled in Tokyo the first week of April to lay out potential plans for the immediate treatment of some 63 million gallons (a number which later exceeded 110 million gallons) of highly contaminated water to avoid overflow from the buildings as well as supply the desperately needed clean cooling water for the reactors. A system had to be deployed with a very brief cold shake down and hot startup before the rainy season started in early June. Joined by team members Toshiba (oil removal system), AREVA (chemical precipitation system) and Hitachi-GE (RO system), Kurion (cesium removal system following the oil separator) proposed, designed, fabricated, delivered and started up a one of a kind treatment skid and over 100 metric tons of specially engineered and modified Ion Specific Media (ISM) customized for this very challenging seawater/oil application, all in seven weeks. After a very short cold shake down, the system went into operation on June 17, 2011 on actual waste waters far exceeding 1 million Bq/mL in cesium and many other isotopes. One must remember that, in addition to attempting to do isotope removal in the competition of seawater (as high as 18,000 ppm sodium due to concentration), some 350,000 gallons of turbine oil was dispersed into the flooded buildings as well. The proposed system consisted of a 4 guard vessel skid for the oil and debris, 4 skids containing 16 cesium towers in a lead-lag layout with removable vessels (sent to an interim storage facility), and a 4 polishing vessel skid for iodine removal and trace cesium levels. At a flow rate of at least 220 gallons per minute, the system has routinely removed over 99% of the cesium, the main component of the activity, since going on line. To date, some 50% of the original activity has been removed and stabilized and cold shutdown of the plant was announced on December 10, 2011. In March and April alone, 10 cubic feet of Engineered Herschelite was shipped to Seabrook Nuclear Power Plant, NPP, to support the April 1, 2011 outage cleanup; 400 cubic feet was shipped to Oak Ridge National Laboratory (ORNL) for strontium (Sr-90) ground water remediation; and 6000 cubic feet (100 metric tons, MT, or 220,400 pounds) was readied for the Fukushima Nuclear Power Station with an additional 100 MT on standby for replacement vessels. This experience and accelerated media production in the U.S. bore direct application to what was to soon be used in Fukushima. How such a sophisticated and totally unique system and huge amount of media could be deployable in such a challenging and changing matrix, and in only seven weeks, is outlined in this paper as well as the system and operation itself. As demonstrated herein, all ten major steps leading up to the readiness and acceptance of a modular emergency technology recovery system were met and in a very short period of time, thus utilizing three decades of experience to produce and deliver such a system literally in seven weeks: - EPRI - U.S. Testing and Experience Leading to Introduction to EPRI - Japan and Subsequently TEPCO Emergency Meetings - Three Mile Island (TMI) Media and Vitrification Experience by PNNL - Commercial Nuclear Power Plant Media Experience (including long term Cs removal) - DOE Low Active Waste (LAW) and High Level Waste (HLW) in High Salt and pH Conditions Media and Vitrification Experience - National Laboratory (e.g. Oak Ridge National Laboratory, ORNL) Ground Water Media Experience - Gulf Oil Spill Media Experience in Seawater - All Media Had to be Fully Tested at High Rad Levels in Seawater and Oil Before Arriving in Japan - Final Waste Form and Disposal Experience (e.g., vitrification) - 100 Metric Tons (6000 cubic feet or 220,400 pounds) of Media had to be Immediately Available with the same amount in production as replacement media. [To date, for 2011, 400 MT of media have been prepared for Japan alone.] - Remote Operation, Modular Water Treatment Equipment Design and Fabrication in both Commercial NPP and DOE Canyon Operations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Design and characterization of hierarchical aluminosilicate composite materials for Cs entrapment: Adsorption efficiency tied to microstructure

The growing quantity of nuclear waste and the serious threats to the environment challenge researchers to innovate and target new waste form technologies. In the past decades, considerable efforts have been devoted to developing highly selective sorbents followed by safe disposal with the assurance of chemical stability and robust retention performance. Zeolite-containing geopolymers are regarded as a possible 2-in-1 material able to both capture and sequester elements such as Cs in bed fixed column application perspective. Here, these composites show promise for combining extraction properties of zeolite powder due to its crystalline structure (high capacity and selective adsorption), with the tunable microstructure and the shaping feasibility of the geopolymer binder. For the development of materials devoted to Cs immobilization, porous zeolite/geopolymer composites were prepared by dispersing NaY zeolite particles in a geopolymer binder. The influence of the structural properties of such composites on their ability to entrap a large amount of Cs by an ionic exchange process was notably studied. Composites' compositions, porosities, morphologies and crystallinity were analyzed by scanning electron microscopy coupled with energy dispersive x-ray spectroscopy (SEM-EDX), x-ray diffraction analysis (XRD) and nitrogen adsorption/desorption studies. Experimental Cs sorption in batch mode was used to follow the ionic exchange phenomenon in these materials. Along with 5 wt% amount of zeolite in geopolymer improves the Cs adsorption performance offering multiple new adsorption sites. Additionally, the geopolymer mesopores are beneficial facilitating the access of Cs and its role as a binder is advantageous to tailor granular hierarchical structure for safer industrial application.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Engineering Evaluation of Barium Buildup in a Decayed CsCl Sealed Source and Potential Impact for Cesium Release from a Breached Source

For more than 50 years, radioactive 137 Cs has been a major source material for radioactive sealed sources, usually constructed as cesium chloride (CsCl) salt loaded into double-walled, stainless-steel capsules. A complication develops as 137 Cs decays to 137 Ba since this process creates a strongly reducing environment inside the capsule. A potential hazard exists if the capsule is breached and air ingress induces rapid exothermic oxidation, and this mechanism is suspected to be responsible for the well-known contamination incident at the Harbor View facility in Washington state. For this study, many thermodynamic evaluations were performed to assess the internal state of capsules after several decades of decay and to describe the potential oxidation if the capsule contents were to suddenly be exposed to air. Results suggest that reduction of impurities such as Cu, Fe, Pb, and Cr to metal will occur. If these impurities are lacking, it is possible that even Ba metal will form. In most cases, rapid oxidation can occur, and the exothermic reactions are sufficient to vaporize portions of the contents, which would include the remaining 137 Cs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗