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DOE OSTI2022

Effect of Initial Water Saturation on The Performance of Fracturing Fluids With and Without Polyallylamine under Simulated EGS Conditions

Objectives/Scope: StimuFrac (US Patents 9,873,828 B2 and 9,447,315 B2), a CO2-reactive polymer aqueous solution [polyallylamine (PAA) 1wt% in water] combined with CO2, can be used as a less water-intensive fracturing fluid for enhanced geothermal systems (EGS). Our previous results show that in hot dry rock (HDR), PAA/CO2 fracturing fluids outperformed other fluids such as water, CO2, and CO2/water in generating large fractures with less fluid consumed. The objective of this work is to study the effect of initial water saturation of rock on the performance of StimuFrac fluid in ½ foot cubic rock samples and under representative EGS pressure/temperature conditions using cyclic and constant flow rate injection strategies. The fracturing results are compared with results using different fracturing fluids in terms of controlling fracture propagation rates, fracture hydraulic conductivity, breakdown pressures and volumes of fluids required. Methods/Procedures/Process: In all tests, water was initially injected into the rock to increase the water saturation before the fracturing processes to simulate actual geothermal reservoir conditions. For the cyclic injection, one complete cycle consisted of (1) a PAA slug (or water slug) injection followed by (2) CO2 injection to initiate the fracture. In the second step of the first cycle, the pressure of CO2 is increased until a maximum pressure is reached (fracture is initiated at this moment), and then the injection of CO2 is allowed for another 30 seconds to propagate the fracture. Then, the two-step cycle of PAA followed by CO2 injection (up to 2-4 mL/min) was continued. For the constant flow rate injection strategies, the initial water saturation was increased by injecting water at 1000 psi and 200°C for three days. After that, an initial slug of water, CO2, or PAA was injected and then fracturing was initiated and propagated by injecting CO2 at a constant flow rate. Applications/Significance/Novelty: The results of this study suggest that water saturation, especially near the wellbore region, will significantly affect the fracturing fluid transmission into the rock porous media by changing the relative permeability of CO2 or water, thus affecting the fracture initiation and propagation. In this study, fracturing with cyclic injection or constant flow rate injection methods were performed using three different kinds of fluids systems. These fluids are water, CO2, or CO2 with PAA. Splitting the rock samples in half after fracturing reveals that the fracture propagation is significantly limited under these high water saturated conditions compared to dry initial conditions: The fractures propagate less than 1/3 length of the distance from the wellbore to rock surface, and in some cases no fracture is generated. This may be caused by the fact that leak-off is dominating the fracturing process and the injected fluid flow rate is not high enough to overcome the leak-off even under high flow rate injection conditions. Additionally, CO2 could be leaking off into the wellbore annulus and this may be making it more difficult to generate pressure gradients away from the near-wellbore region.

Jian, Guoqing↗
DOE OSTI2022

Integrated Risk-Informed Condition Based Maintenance Capability and Automated Platform: Technical Report 3

This project is a collaborative research effort between PKMJ Technical Services LLC, Idaho National Laboratory, and Public Service Enterprise Group (PSEG) Nuclear, LLC. The collaboration, led by PKMJ Technical Services LLC, is part of the industry Funding Opportunity Announcement (FOA) award under Advanced Nuclear Technology Development FOA #DE-FOA-0001817. The pilot demonstration focuses on the Circulating Water System (CWS), an important non-safety-related system that impacts the power generation capability of the plant site. Achieving riskinformed condition-based Predictive Maintenance (PdM) on the CWS will result in significant economic benefits, and the developed methodologies can also be applied to other plant systems. This approach supports an industry goal of ensuring that nuclear power generation remains a viable, economically competitive option in the energy market. Operation and Maintenance (O&M) costs include labor-intensive Preventive Maintenance (PM) programs that involve manually performed inspection, calibration, testing, and maintenance of plant assets at periodic frequencies as well as time-based replacement of assets, irrespective of condition. This project offers an alternative by focusing on riskinformed condition-based maintenance to reduce O&M costs while still maintaining plant health and safety. This report summarizes the progress made toward achieving a risk-informed condition-based maintenance approach. The research and development (R&D) activities presented in this report are associated with development of a nuclear digital platform application, integration of fault signature models, and automated work management processes. The fault signatures and Machine Learning (ML) models are key components in predictive analytics and are heavily leveraged to improve the insights received by existing plant process data sources. Availability of the analysis results within a centralized digital platform enhances efficiency by enabling automation of activities otherwise performed manually. Personnel are presented with enhanced information that can be used to evaluate plant status and risks. Utilizing the enhancements to data analytics supports automated responses, (i.e. issuance of work orders) to address developing equipment faults and thus preventing forced, unplanned shutdowns of components or systems. The R&D activities described within this report lay the foundation for developing and demonstrating a digital automated platform to centralize the implementation of condition monitoring and response to equipment faults. The digital automated platform is cloud-based and designed to enable improved efficiency of plant processes. The digital platform includes content related to maintenance optimization, fault signature analysis, and plant records, which can all be used to support efficiencies when located within a centralized digital platform. These efficiencies could be further enhanced when deployed through industry-wide deployment of the technology to improve insights and processes based upon economies of scale.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Conditioning and Repackaging of Spent Radioactive Cs-137 and Co-60 Sealed Sources in Egypt - 13490

Radioactive Sealed sources (RSSs) are widely use all over the world in medicine, agriculture, industry, research, etc. The accidental misuse and exposure to RSSs has caused significant environmental contamination, serious injuries and many deaths. The high specific activity of the materials in many RSSs means that the spread of as little as microgram quantities can generate significant risk to human health and inhibit the use of buildings and land. Conditioning of such sources is a must to protect humans and environment from the hazard of ionizing radiation and contamination. Conditioning is also increase the security of these sources by decreasing the probability of stolen and/or use in terrorist attacks. According to the law No.7/2010, Egyptian atomic energy authority represented in the hot laboratories and waste management center (centralized waste facility, HLWMC) has the responsibility of collecting, conditioning, storing and management of all types of radioactive waste from all Egyptian territory including spent radioactive sealed sources (SRSSs). This paper explains the conditioning procedures for two of the most common SRSSs, Cs{sup 137} and Co{sup 60} sources which make up more than 90% of the total spent radioactive sealed sources stored in our centralized waste facility as one of the major activities of hot laboratories and waste management center. Conditioning has to meet three main objectives, be acceptable for storage, enable their safe transport, and comply with disposal requirements. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Carbonation Behavior of Pure Cement Hydrates under Supercritical Carbon Dioxide Conditions - 12199

Carbonation of cement-based waste forms using a supercritical carbon dioxide (SCCO{sub 2}) is a developing technology for the waste immobilization of radioactive and non-radioactive wastes. However, the detail carbonation behaviors of cement matrices under the SCCO{sub 2} condition are unknown, since cement matrices forms very complex phases. In this study, in order to clarify the crystal phases, we synthesized pure cement hydrate phases as each single phases; portlandite (Ca(OH){sub 2}), ettringite (Ca{sub 6}Al{sub 2}(SO{sub 4}){sub 3}(OH){sub 12}.26H{sub 2}O), and calcium silicate hydrate (n CaO---m SiO{sub 2} ---x H{sub 2}O), using suspensions containing a stoichiometric mixture of chemical regents, and performed carbonation experiments using an autoclave under supercritical condition for carbon dioxide. The XRD results revealed both the carbonate phases and co-product phases depending on the initial hydrate phases; gypsum for Ettringite, amorphous or crystalline silica for calcium silicate hydroxide. Thermogravimetric analysis was also performed to understand carbonation behaviors quantitatively. According to the experimental results, it was found that the major reaction was formation of calcium carbonate (CaCO{sub 3}) in all cases. However, the behaviors of H{sub 2}O and CO{sub 2} content were quietly different: Portlandite was most reactive for carbonation under SCCO{sub 2} conditions, and the CO{sub 2} content per one molar CaO was ranged from 0.96 ∼ 0.98. In the case of Ettringite, the experiment indicates partial decomposition of ettringite phase during carbonation. Ettringite was comparatively stable even under the SCCO{sub 2} conditions. Therefore, a part of ettringite remained and formed similar phases after the ettringite carbonation. The CO{sub 2} content for ettringite showed almost constant values around 0.86 ∼ 0.87. In the case of calcium silicate hydrate, the carbonation behavior was significantly influenced by the condition of SCCO{sub 2}. The CO{sub 2} content for the calcium silicate hydrate had values that ranged from 0.51 ∼ 1.01. The co-products of the carbonation were gypsum (CaSO{sub 4}) for ettringite, silica gel (SiO{sub x}) and silica (SiO{sub 2}) for calcium silicate hydrate, which also contributed to the densification of the particles. The production of co-products enhanced the change to their morphology after the carbonation. (authors)

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

Modeling used fuel response to normal conditions of transportation package drops to access geometric sensitivities

The US Department of Energy Used Fuel Disposition program is performing research to determine the loading conditions that used nuclear fuel (UNF) experiences during normal conditions of transportation (NCT). This loading environment includes mechanical shock and vibration as the UNF is transported by rail or truck, and it also includes potential package free drop conditions, which are the topic of this study. NCT drop test requirements for UNF package designs are defined in 10 CFR 71.71. The free drop test height is 30 cm for the UNF packages considered in this study. Explicit dynamic finite element models of generic rail and truck UNF packages containing a highly resolved 17x17 PWR fuel assembly are used to estimate the potential range of loading experienced by UNF during NCT horizontal package drops onto an unyielding surface. There are many variables and uncertainties that can affect the loading of UNF during this type of free drop scenario. This study focuses on geometric variables, such as gap conditions and small adjustments to the angle of impact, in a number of analysis cases. The LS-DYNA finite element analyses calculate detailed results for all the components of the fuel assembly. The results show that the peak localized axial strain in the fuel cladding is estimated to be 0.006421 (nominal) with an upper bound of 0.014511, based on the peak instantaneous bending moment of 63.65 N-m and other uncertainty factors. For the assumed low burnup condition of the UNF, the cladding yield strain is 0.011027 and the failure strain is 0.015574, which means that the calculated upper bound strain approaches the cladding failure limit without exceeding it. The peak shear force on the cladding was calculated to be 7879 N. This could be significant to cladding integrity if applied to the UNF in a situation where the fuel pellets do not support the shear transfer. Further evaluation is needed to determine if this is a credible threat to cladding integrity for these 30 cm free drop scenarios.

42 ENGINEERING↗
DOE OSTI2016

Secondary creep behavior of Zr-4 claddings under LOCA conditions

The thermo-mechanical behavior of Zircaloy-4 fuel rods under Loss-Of-Coolant Accident (LOCA) conditions in water cooled reactors is investigated. A custom experimental setup is dedicated to the high-temperature creep ballooning study of 90 mm long cladding samples. Creep tests were performed under an inert environment (argon), for temperatures from 750 to 850 C. degrees and internal pressures ranging from 1 to 5 MPa. During its operating life, Zr-4 cladding is submitted to oxidation and hydriding. These parameters strongly influence the creep behavior of Zirconium alloys at high temperatures. A first campaign on as-received Zr-4 was performed. Creep-rates are computed using 2-Dimensional Digital Image Correlation (2D-DIC) and are correlated to local temperatures measured using Near Infra-Red thermography. As-received Zr-4 behavior laws are determined using Finite Element Model Updating (FEMU). Norton exponents and activation energies are determined for thermal-mechanical conditions of the performed tests. The results show that the effect of applied stress on the Norton exponent is evidenced in the study for both α-phase and (α + β)-phase domains. The determined stress exponent magnitudes are very consistent with literature data depending on the thermomechanical test conditions. The activation energy parameter is very sensitive to the thermo-mechanical conditions and remains valid within a limited temperature and stress ranges to the authors' knowledge. The beginning of the (α + β)-phase domain is clearly identified with the activation energy increase. Finally, steady-state creep is accurately modeled depending on ballooning thermo-mechanical conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Clad to coolant heat transfer during RIA in PWR conditions

The power pulse during an insertion of reactivity accident (RIA) leads to a transient heating of the fuel. Heat transfer toward the coolant through the clad and associated temperature transient of the clad is of safety significance since it may largely impact the risk of clad failure. For sufficiently large enthalpy deposited within the fuel, clad temperature increases rapidly and this can lead to coolant boiling. This paper concerns the critical review of the present understanding and of the models for the clad to coolant heat transfer in RIA for PWR conditions. The results of past experiments are reviewed and the reliability of current models dedicated to these conditions is analyzed. It appears that the use of classical models for wall to fluid heat transfer does not allow predicting the clad temperature transient as measured in RIA-related experiments. The deviation is all the more large when film boiling conditions are reached. To reproduce the experimental data, some correlations of heat transfer can be fitted but the reliability of such model out of the range of conditions of the experimental data is low. It is mainly due to a rather low level of understanding of the heat transfer in very transient conditions (i.e. when the time scale of variation of the main heat transfer parameters, like a wall temperature, is less than the typical internal time scale of the heat transfer process itself). It results in a very large scattering in the numerical simulation of clad temperature when several fuel codes are considered. To improve the reliability of these calculations, it is believed that the study of transient boiling heat transfer thanks to devoted experimental facilities is required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Flow Rate Dependency of Permeability Changes with the Deposition of CSH under the Highly Alkaline Ca-rich Condition - 17175

The formation of calcium silicate hydrate (CSH) around the geological repository is one of the notable factors for the assessment of the radionuclide migration. This study discussed the clogging effect of CSH in flow path with the change in the flow rate based on the flow experiments using micro flow cell. In the experiment, the change of the permeability with the alteration of granite under the highly alkaline and Ca-rich condition was examined. As a result, the permeability in flow path gradually decreased with the periodical increase and decrease of the permeability. In addition, under the condition of the lower flow rate (0.06 ml/h), the permeability became much smaller than those for other conditions of the higher flow rates. These suggest that the clogging of flow path will progress with the local deposition/moving of CSH, in particular, under the condition of the lower flow rate. Retardation effect of radionuclide migration will be expected because it is prospected that flow rate of groundwater is lower than that of this study in the actual condition of repository. (authors)

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

The Effect of Bicarbonate on Autunite Dissolution in the Presence of Shewanella oneidensis under Oxygen Restricted Conditions - 16429

Uranium is a key contaminant at many U.S. Department of Energy sites that served a leading role in the nation's defense for over 50 years. Uranium contamination of soil and groundwater is of great environmental concern due to the toxicological properties of the uranyl species. In an effort to decrease the concentration of soluble uranium, tripolyphosphate injections were identified as a feasible remediation strategy for sequestering uranium in situ in contaminated groundwater at the Hanford Site (WA). The introduction of sodium tripolyphosphate into uranium-bearing porous media results in the formation of uranyl phosphate minerals (autunite) of general formula (X{sub 1-2}[(UO{sub 2})(PO{sub 4})]{sub 2-1}.nH{sub 2}O), where X is a monovalent or divalent cation. The stability of the uranyl phosphate minerals in the subsurface is a critical factor since it determines the long-term effectiveness of the sodium tripolyphosphate injection as a remediation strategy. The behavior of uranium and its mobility in the subsurface are affected by various factors such as the chemistry of pore water, groundwater and soil minerals, the presence of complex-forming ligands, and micro-organisms that thrive under these conditions. Bicarbonate presence in the aqueous phase facilitates uranium desorption from soil and sediments and promotes the release of U(VI) into the aqueous phase, thus increasing uranium mobility in natural waters. In neutral or basic pH conditions, uranium can readily complex with a wide variety of ligands such as carbonate and form soluble uranyl-carbonato complexes such as UO{sub 2}(CO{sub 3}){sub 2}{sup 2-} and UO{sub 2}(CO{sub 3}){sub 3}{sup 4-}. Bicarbonate complexes have been identified in contaminated pore water at the Hanford Site. Furthermore, the presence of bacteria can significantly affect uranium mobility since bacteria may bio-enhance the release of U (VI) from uranyl-phosphate minerals, thus liberating uranium in the aqueous phase in an effort to obtain phosphorous, a vital nutrient for their metabolism. The Columbia River at the site exhibits large stage variations, causing fluctuations in the water table. These water table fluctuations and multiple rise-and-fall cycles in the river created an oxic-anoxic interface in this region. Previous assessments of Hanford sediment samples collected from this area noted a decline in cultivable aerobic bacteria and suggested the presence of facultative anaerobic bacteria. Shewanella oneidensis MR1 is a common bacterial group in soils and Shewanella species have been found in the Hanford soil. Shewanella oneidensis is a facultative anaerobic species which also belongs to a group known as dissimilatory metal-reducing bacteria (DMRB), due to their ability to reduce metals as a part of their metabolic pathway. Therefore, understanding the role of Shewanella is a critical factor affecting the outcome of environmental remediation. The objective of this research was to demonstrate the significance of bacteria-uranium interactions and investigate the effect of bicarbonate in the aqueous phase on Ca-autunite dissolution in the oxygen-restricted conditions. Sterile 100-mL glass mixed bioreactors with weighed amounts of ground Ca-autunite particles were used for initial experimentation. These autunite-containing bioreactors were injected with bacterial cells after the autunite equilibrated with the media solution amended with 0 mM, 3 mM 5 mM and 10 mM concentrations of bicarbonate. Additional abiotic samples were prepared without bacterial inoculation to provide a control for the biotic samples. The concentration of U(VI) determined in the aqueous phase was in direct correlation to the concentration of bicarbonate present in the solution. It was noted that an increase in bicarbonate concentration led to higher U(VI) concentration in the aqueous phase. A sharp decrease in U(VI) concentration was observed in the bicarbonate-free samples, probably due to a microbe mediated reduction of U(VI) to U(IV). In the case of samples amended with 3 mM bicarbonate, a slight, yet significant, decrease in U(VI) was also observed compared to the abiotic control. However, compared to the abiotic controls, there was no change in U(VI) levels in the aqueous phase as a result of bacterial inoculation for the samples amended with 5 and 10 mM bicarbonate for a period of time as long as 50 days. These results imply that the uranyl-carbonate complexes (UO{sub 2}(CO{sub 3}){sub 2}{sup 2-} and UO{sub 2}(CO{sub 3}){sub 3}{sup 4-}) that are present in carbonate-rich environments might be interfering with the microbial reduction in the oxygen- restricted conditions; however, this conclusion requires further investigation. Finally, the data indicate that release of U, P and Ca from autunite was non-stoichiometric. In conclusion, the high stability of uranyl-phosphate complexes makes them a strong candidate for the remediation efforts to sequester uranium in the subsurface. Nevertheless, in a bicarbonate-rich environment, Ca-autunite has a high probability of dissolution and the presence of metal-reducing bacteria does not seem to impede this process in the presence of higher bicarbonate concentrations under oxygen restricted conditions. (authors)

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

Conditioning and Requalification of Legacy Waste Packages - 18648

In most German interim storage facilities, hundreds of waste packages are waiting for re-conditioning and qualification to comply with the waste acceptance criteria (WAC) for the Konrad repository. Especially the waste classification as well as some characteristics and the description of the material composition in its current state do not comply with the Konrad requirements since the conditioning of these waste packages took place up to 40 years ago. The German name for these waste packages is 'Altgebinde', hereafter referred to as 'legacy waste packages'. The present article will provide an overview of the measures for the qualification of these legacy waste packages to fulfill the Konrad requirements. The following measures, when necessary, are carried out for the qualification of legacy waste packages: - Collection and evaluation of all documents related to the primary conditioning process, e. g. the used equipment, documented waste and packaging properties, conditioning and radiology, if available (hereafter 'old documents'). - Preparation of a conceptual report and a process control quality plan based on existing data, with appropriate measures to qualify the waste packages. Assessment and approval of the process control quality plan by the competent authority. - Characterization of the waste and packaging properties by visual inspection or tomography, recording or updating of dose rate, contamination and mass, verification of product stability, for example by gas analysis of the interior atmosphere of the container. - Characterization of the nuclide inventory by gammas scans, sample taking to analyze nuclides which can not be measured by means of in-situ gamma spectrometry, and for material declaration. - Qualification or post-conditioning of the legacy waste packages if necessary, e. g. by drying the waste packages and packaging them in containers approved for disposal. - Compilation of all results of the above mentioned tests and steps to create a waste package documentation, including the calculations for the nuclide inventory and the description of the material composition of the waste packages. - Receiving approval by the competent authority after evaluation of the final documentation. The aim behind this structured approach is to prevent additional, unnecessary measures and treatment of the waste. Consequently, the additional characterization is preferably carried out on the basis of the available documents. (authors)

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

Control of Radioactive Methyl Iodide Gases by Mesoporous Hybrid Adsorbents at High Humid Condition - 19329

During the operation of nuclear facility as well as decommissioning stage, various radioactive gaseous wastes are generated. It is generally recognized that radioiodine is potentially one of the most hazardous gas on the human body and environment. Iodine-131 release in the gaseous effluents at the nuclear facilities had been a critical environmental concern. Reducing iodine release is a critical issue in radioactive waste system evaluation, especially in their adherence to the 'as low as reasonably achievable' (ALARA) criteria because radioiodine among these gaseous wastes has radiological harm. Therefore, numerous studies for the removal of radioiodine from the off-gas stream of nuclear facilities have been performed with formation processes and chemical structure. There has been some concern regarding the use of TEDA-impregnated activated carbon as an adsorbent for radioiodine removal, because of the low ignition point of TEDA and the possible release of TEDA from the charcoal in continuously flowing air streams. Therefore, the TEDA-impregnated activated carbon cannot use to trap radioiodine generated from high humid condition, due to the rapid drop of the removal efficiency. It is obvious that high humid can be expected during design basis event such as loss-of-coolant accident (LOCA). Therefore, investigation on the retention of organic radioiodine (CH3I131) by various activated carbons at high humid condition has extensively been performed so far. But there are no good results on the adsorption of methyl iodide at high humid condition up to now. As a result, the promising approach to the problem of iodine removal under high humid conditions is to use organic-inorganic hybrid adsorbents in which are enough retention characteristics from hydrophobic property. In order to develop the best removal efficiency for the control of radioiodine gas generated from a high humid process, a hydrophobic adsorbents was developed and the performance test of adsorbent using radioactive methyl iodide tracer was also performed. Demonstration test of adsorbent using radioactive methyl iodide tracer was performed at high humid conditions up to 97% relative humidity. (authors)

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

Performance Evaluation of AG-1 FC Filters Under Upset Conditions and the Effects of Aging on Filters - 19501

High efficiency particulate air (HEPA) filters are routinely used in nearly every operating Department of Energy (DOE) and National Nuclear Safety Administration (NNSA) nuclear facility. These nuclear grade HEPA filters installed in DOE and NNSA confinement ventilation systems are credited as the final barrier to protect the environment, the public, and the worker from accidental release of virtually all airborne radioactive materials, and must meet the American Society of Mechanical Engineers' (ASME) Code on Nuclear Air and Gas Treatment (CONAGT) AG-1 standard. This testing activity evaluates ASME AG-1 Section FC axial flow HEPA filters, both separator and separator-less, under upset conditions. Additionally, testing efforts have analyzed the effects of filter degradation due to aging. In 1997, the Defense Nuclear Facilities Safety Board (DNFSB) raised concerns of the potential vulnerability of HEPA filters in vital safety systems, both from a design stand point and the potential for degradation of HEPA filter performance over time. The DOE Office of Environmental Management (DOE-EM) has a Cooperative Agreement with Mississippi State University (MSU) Institute for Clean Energy Technology (ICET) to conduct testing of HEPA filters. DOE-EM has established a Technical Working Group (TWG) of over thirty subject matter experts from the nuclear industry to provide input and oversight of ICET projects and activities. This has included establishment/approval of testing parameters for evaluating these filters under ASME NQA-1 compliant Test Plans. The current scope of work involves evaluation of the performance envelope for new and aged separator and separator-less designs of AG-1 Section FC axial flow HEPA filters. A range of elevated test conditions (T and RH) are used to determine the operating envelope within which DOE nuclear safety experts can credit installed HEPA filter performance, thereby establishing a risk-informed DOE service life. This paper will present preliminary test results of new and aged axial flow filters tested at ICET. Results in this paper include data collected from new separator and separator-less type FC HEPA filters and equivalent test data for a limited group of aged separator and separator-less type FC HEPA filters. Testing for both scopes of work evaluate initial filter performance under ambient conditions of temperature (60 deg. F - 80 deg. F (∼16 - ∼27 deg. C)) and relative humidity (40-60%), loading of each filter with Aluminum Trihydroxide, Al(OH)3, to a specified differential pressure, and then exposing the filter to specified elevated temperature and relative humidity conditions. In addition to the upset conditions of elevated temperature and relative humidity, the performance of HEPA filters under fire scenarios is not well characterized. ICET is currently designing a burn box and testing apparatus to study the effects of fire from burning liquids and a wide range of blends of combustible solids. (authors)

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

Augmented Monitoring and Condition Assessment Program for SNF Wet Storage Life Extension - 20489

Approximately 27 MTHM of spent nuclear fuel (SNF) owned and managed by the U.S. Department of Energy, Office of Environmental Management is stored in the L Basin at the Savannah River Site (SRS). This 'DOE SNF' is comprised of approximately 12,000 aluminum-clad, aluminum-based fuel assemblies (∼7 MTHM), and approximately 2000 non-aluminum fuel assemblies (∼20 MTHM). A program is in progress to perform Non-Destructive Examination (NDE) of the fuel and their storage containers to characterize the materials' condition, and to evaluate the effects of service to enable continued safe wet storage of the SNF. The Augmented Monitoring and Condition Assessment Program (AMCAP), a two-part program to develop and implement remote underwater Non-Destructive Examination of the aluminum SNF, and of the containers of the non-aluminum SNF, respectively, is aimed at the characterization and evaluation of corrosion degradation of aluminum fuel and container materials. The predominant design of the aluminum SNF (ASNF) stored in L Basin is the Materials Test Reactor (MTR) equivalent design, a plate fuel design. These SNF are stored in 5'' diameter tubes or 5'' x 5'' squares (called bundles) that are nominally 12' or up to 14' long. The SNF is de-bundled and inspected using a custom-designed MTR Fuel Inspection Table. The inspection table provides for indexed fuel positioning for a video camera examination with controlled lighting. A total of 10 of the SNF originating from foreign research and test reactors were selected for examination based on burnup, enrichment, and prior damage caused by service/storage history. A special inspection campaign of these 10 assemblies is in progress. The observed corrosion damage included minor to moderate attack from general corrosion, pitting, crevice, end grain and galvanic corrosion. Example results from the completed inspections are shown and discussed. The focus is a comparison of the as-received condition versus the as-found current storage condition that will serve to validate Water Chemistry and Corrosion Monitoring Programs. The non-aluminum SNF (NASNF) stored in L Basin are of diverse design that includes various geometries with claddings of stainless steel, Zircaloy, and Hastelloy. The fuel core materials include uranium alloys, oxides/mixed oxides, and carbides. These SNF materials, originating from early experimental and test reactors, are in various physical forms including single fuel elements and cut pieces. The fuel is stored in L Basin in various configurations including in direct bundled storage in aluminum tubes and in isolation cans that are in the bundles or in a larger over-size storage container (OSC). Concern with inside-out corrosion and the potential loss of configuration control and ability to handle the storage containers prompted the development of remote NDE methods that include visual and UT technologies to assess the condition of the containers. Candidate stored materials were selected for a special inspection campaign; the development of the NDE methods for the inspection for galvanic, crevice and sediment-induced corrosion (inside-out) are discussed. A summary of the SNF storage in L Basin at the SRS, and an overview of the AMCAP to enable continued safe storage of DOE SNF are described. (authors)

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

Development and implementation of an in-situ cable condition monitoring method for nuclear power plants - 130

The sustainability of the existing fleet of nuclear power reactors depends on utilities' ability to manage the aging of key components that cannot easily or economically be replaced or upgraded. Cables, metals, and concrete structures are among the major and key components that the industry and government are working on to ensure the long term reliability and safety of nuclear plants. To ensure that these key components continue to meet their performance requirements, new testing and condition monitoring technologies must be developed that are capable of identifying and quantifying degradation in these materials. The work presented in this paper focuses on the development and implementation of an in-situ condition monitoring method for nuclear power plant cables. This cable condition monitoring method is based on the frequency domain reflectometry (FDR) technique, which is a non-destructive in-situ electrical test that uses the principle of transmission line theory to locate and quantify impedance changes in a cable circuit. Through extensive research, development, and validation, the FDR test technique has been adapted as a cable condition monitoring method that trends with increasing cable degradation. In addition, the FDR data has been correlated to the industry standard elongation at break (EAB) test to quantify the degree of localized degradation in nuclear plant cable insulation material caused by long term exposure to high temperatures and radiation. By correlating FDR to EAB, the aging condition of cables routed through harsh environments can be better understood in relation to its ability to perform its intended function. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Characteristics Analysis of SGTR accidents under the Shutdown Conditions

The Steam Generator Tube Rupture (SGTR) accident is one of design basis accidents that could result in undesirable radiation release as it bypasses the containment. According to Ref. [1], SGTRs have been experienced in several nuclear power plants and Ulchin 4 of Korea is one of them. Basically, SGTR analysis assumes that it occurs at full power, the SGTR at Ulchin 4 in 2002, was in the shutdown condition. Therefore, the transient behavior was different from the cases in full power conditions and especially, the effectiveness of Emergency Operating Procedure (EOP) needs to be checked if it is also applicable to the shutdown conditions, because EOP is prepared for full power conditions. In this paper, we set up a MARS-KS code model for Ulchin 4 unit and simulated the SGTR accident for benchmarking the model to the actual plant behavior during the accident. And then, several case studies were done to investigate the consequence of SGTR and the effectiveness of the EOP, focusing on the amount of the primary to secondary discharge flow which was influenced by operator actions. At the end, we'll discuss about the EOP for the shutdown conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

MECHANISTIC MULTIPHYSICS MODELING OF CLADDING RUPTURE IN NUCLEAR FUEL RODS DURING LOSS-OF-COOLANT ACCIDENT CONDITIONS

The Loss of Coolant Accident (LOCA) is a design basis accident that is included as part of the safety analysis of nuclear power plants. As the nuclear industry desires to increase the cycle length and discharge burnup of existing nuclear power plants they must demonstrate safe operation during a LOCA on high burnup fuel. During a LOCA transient on high burnup fuel rods, the rods may undergo a process known as fuel fragmentation, relocation, and dispersal (FFRD). To permit dispersal, the cladding encapsulating the fuel must rupture with an opening size large enough to allow the fragmented fuel particles to release. Current licensing tools used by industry and the United States Nuclear Regulatory Commission are limited in geometric fidelity and materials that can be analyzed. These simulation tools generally employ a quasi-two-dimensional (1.5D or Layered1D) or 2D-RZ axisymmetric geometric representations exclusively. While a valid approach under some instances, there are times when important physics have an asymmetric behavior in the fuel rod. Examples include fuel fragmentation, thermal-hydraulic boundary conditions, and cladding rupture, all of which are important for LOCA analysis. As industry pursues burnup extensions it must be demonstrated that fuel dispersal can be mitigated or eliminated. To do this, an understanding of the rupture opening after cladding failure is required. This work presents the development of a model for predicting the size and location of the rupture opening in failed fuel rods during LOCA conditions using advanced modeling and simulation tools. In order to supply the rupture model with appropriate boundary conditions, improvements to fuel fragmentation, axial relocation and oxidation modeling were required. First, the eXtended Finite Element Method (XFEM) is used to mechanistically predict the number of fuel fragments that form due to material strength randomization, criteria for strength randomization, mesh density, power ramping rates and irradiation effects. These predictions with associated uncertainty were compared to empirical correlations developed for UO2 verifying that they can be used with increased confidence in subsequent axial relocation analyses. Secondly, a new first-of-its-kind Layered2D computational framework was developed that provides the ability to apply azimuthally varying boundary conditions while providing discrete layers to track fuel movement during the LOCA. An existing fuel axial relocation model developed for Layered1D was extended to work within the Layered2D framework. A large sensitivity study was performed on the initial version of the model to identify modeling parameters of particular importance, with the emissivity used for radiation after blowdown being the primary source of uncertainty. Then, a simplistic approach to incorporate mechanical degradation of the cladding due to oxidation was also developed to investigate the impact of reduced cladding thickness on predictions of the time to failure of cladding tubes. It was found that the cladding will typically rupture prior to a reduction in thickness that is sufficient to impact the rupture behavior. A model was then developed for predicting cladding rupture that transfers the cladding surface temperatures, rod internal and external pressures, fast neutron fluence, and fast neutron flux from a more detailed Layered1D, Layered2D, or 2D-RZ analysis to a 3D cladding only analysis. Comparisons of the rupture model to a few experiments indicate reasonable predictions. The rupture model was then applied to two accident tolerant fuel concepts (FeCrAl and Cr-coated Zircaloy) where it predicted that both ATF concepts would have smaller rupture openings and delayed rupture times than the standard Zircaloy-4 cladding material under identical loading conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An Alternative Frictional Boundary Condition for Computational Fluid Dynamics Simulation of Friction Stir Welding

For better application of numerical simulation in optimization and design of friction stir welding (FSW), this paper presents a new frictional boundary condition at the tool/workpiece interface for computational fluid dynamics (CFD) modeling of FSW. The proposed boundary condition is based on an implementation of the Coulomb friction model. Using the new boundary condition, the CFD simulation yields non-uniform distribution of contact state over the tool/workpiece interface, as validated by the experimental weld macrostructure. It is found that interfacial sticking state is present over large area at the tool-workpiece interface, while significant interfacial sliding occurs at the shoulder periphery, the lower part of pin side, and the periphery of pin bottom. Due to the interfacial sticking, a rotating flow zone is found under the shoulder, in which fast circular motion occurs. The diameter of the rotating flow zone is smaller than the shoulder diameter, which is attributed to the presence of the interfacial sliding at the shoulder periphery. For the simulated welding condition, the heat generation due to friction and plastic deformation makes up 54.4 and 45.6% of the total heat generation rate, respectively. In conclusion, the simulated temperature field is validated by the good agreement to the experimental measurements.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Model validations for low-global warming potential refrigerants in mini-split air-conditioning units

To identify low GWP (global warming potential) refrigerants to replace R-22 and R-410A, extensive experimental evaluations were conducted for multiple candidates of refrigerant at the standard test conditions and at high-ambient conditions with outdoor temperature varying from 27.8 C to 55.0 C.. In the study, R-22 was compared to propane (R-290), DR-3, ARM-20B, N-20B and R-444B in a mini-split air conditioning unit originally designed for R-22; R-410A was compared to R-32, DR-55, ARM-71A, L41-2 (R-447A) in a mini-split unit designed for R-410A. To reveal physics behind the measured performance results, thermodynamic properties of the alternative refrigerants were analysed. In addition, the experimental data was used to calibrate a physics-based equipment model, i.e. ORNL Heat Pump Design Model (HPDM). The calibrated model translated the experimental results to key calculated parameters, i.e. compressor efficiencies, refrigerant side two-phase heat transfer coefficients, corresponding to each refrigerant. As a result, these calculated values provide scientific insights on the performance of the alternative refrigerants and are useful for other applications beyond mini-split air conditioning units.

54 ENVIRONMENTAL SCIENCES↗