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DOE OSTI2017

Which parameters might help to predict an impact on reactivity for mixed fissile units in a storage array?

Several studies have already been performed on mixed fissile units in storage arrays and concluded that the neutron spectrum of the fissile material, the shape of fissile units, the interstitial moderation and structural materials may have an impact on reactivity. For this paper, calculations are performed on a simplified infinite planar array with one level of fissile material units axially reflected by concrete. Fissile material is modelled by a cylinder shape centered in the array mesh selected (120 cm height and 60 cm pitch). Either air or interstitial moderation by water is modelled between the fissile units. Three parameters are studied: the fissile materials (with various moderation conditions), the cylinder dimension and the nature of the structural material. The main following conclusions can be drawn from these calculations. First, for several mixed-array cases, air-spaced mixed arrays present a high increase on reactivity whereas interstitial moderation between fissile units leads to no or little impact. Secondly, fissile unit shapes do not explain directly the mixing impact on reactivity. Thirdly, structural materials have an important impact on reactivity in mixed array (either an increase or a decrease) due to their absorbing and/or moderating properties either when air or interstitial moderation are modelled between fissile units. Fourthly, no simple parameter was found to predict the impact on reactivity of mixed array in all studied configurations. However, a comparison of different neutron spectrum parameters can help to predict the absence or potential presence of a 'mixing effect'. Indeed, if all parameters are similar, no 'mixing effect' is observed. But, if these parameters present differences, there might be a 'mixing effect'. (authors)

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DOE OSTI2017

An evaluation of in-line component process modules

The purpose of this paper is to discuss the analysis of abnormal configurations of in-line components, tanks and pumps in support of the integration of Criticality Safety into the Uranium Processing Facility (UPF) at Y-12 National Security Complex (Y-12). In-line components are components that are used in solution piping systems, including instrumentation, valves, pumps, and other general-use equipment. Mounted on self-contained skids/modules, in-line components (instruments, seal pots, level switches, etc.), tanks, and pumps are associated with a wide variety of the UPF processes. Using computer generated models expected to bound abnormal conditions for process module configurations, this report evaluates the impact of loss of spacing between in-line components in in-line component arrays and the effects of abnormal events. All fissile materials are considered to have a {sup 235}U enrichment of 100% and full theoretical density of 18.81 g/cm{sup 3} to conservatively represent the maximum credible enrichment available. Water and concrete were utilized to account for the construction materials associated with the skids and other reflective surfaces. Uranyl nitrate, was evaluated over the range of 0 to 550 g U/L for various normal and abnormal configurations of components mounted on skids. Abnormal events included fissile spills, sprinkler activation, and loss of spacing due to a seismic event. The results from this analysis provide a basis that supports development of nuclear safety design criteria, which generates standardized volume limits expected to bound the actual internal volume of the planned components. The results also provide a basis for consistent Nuclear Criticality Safety (NCS) control of in-line components, pumps, and tanks, for the ultimate development of Criticality Safety Process Studies in the UPF. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Testing and Deployment of Remote Systems to Remove Highly Radioactive Soil at Hanford's 324 Building - 18043

U.S. Department of Energy (DOE) contractor CH2M HILL Plateau Remediation Company (CHPRC) is in the process of procuring, testing and deploying remotely operated tools to safely remove highly contaminated soil from beneath the 324 Building on the Hanford Site in southeastern Washington State. The 324 Building operated for more than 40 years, completing a variety of high-dose engineering development and demonstration activities inside six hot cells that shielded the workers from the highly radioactive materials. Activities to decommission, decontaminate and demolish the facility began in 2008. In 2010, soil contamination was found in the soil under one of the hot cells, and building deactivation was paused to evaluate the soil contamination. Sampling efforts, using borehole technology, revealed significant under-building contamination with dose rates measuring has high as 130 Sv/hr (13,000 R/hr). The highly-contaminated soil has to be removed before traditional open air deactivation activities can continue. CHPRC completed proof of concept testing and built a full-scale test facility to demonstrate the effectiveness of three primary tools that will accomplish the floor and soil removal. The tools include a commercially available concrete saw, and a commercially available excavator arm that have been modified to operate remotely in the high-radiation environment within the hot-cell. In order to reduce the wear on the hot-cell overhead cranes, a remote operated rail and cart system has been designed to transport soil and debris out of the hot-cell to an airlock area where the waste will be separated, according to its level of radioactivity. A camera and light system has been designed to provide visual and audio coverage of the hot-cell area that will allow the operations to proceed from remote control stations. This paper will describe the primary remote operated equipment designed for the project and the challenges and processes for developing, testing and deploying the equipment necessary to excavate the soil from below the 324 Building. (authors)

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DOE OSTI2018

H-Canyon Ventilation System Exhaust Tunnel Inspection at the Savannah River Site - 18052

The Department of Energy (DOE) H-Canyon Facility at the Savannah River Site (SRS) is operated by Savannah River Nuclear Solutions (SRNS). The Facility Structural Integrity Program (SIP) performs periodic visual inspections to confirm the component is capable of performing its safety function. This Program provides the process to document the conditions of safety related structural systems and components, their damage mechanisms, and recommend repairs if required before there is impact on the safety envelope. The H-Canyon Exhaust Tunnel is part of the H-Canyon Exhaust Ventilation System and is periodically inspected under the Structural Integrity Program. The Tunnel performs a Safety Class passive design feature function that is available 100% of the time. The Tunnel directs Canyon process air from the Canyon to the Sand Filter System. The radiological and chemical airborne activity coupled with the physical configuration of the tunnel precludes a manned entry into the tunnel. Tunnel inspections have been performed using cameras on a stick, referred to as pole cameras prior to CY2003. Since CY2003, five camera equipped crawlers have been used with varying levels of inspection success. With each successive crawler inspection, expectations have increased to perform tasks beyond visual examinations. The 2017 crawler performed the visual inspection but also included improvements on crawler performance, such as testing the use of wireless capability and remediation of crawler travel route, to the addition of stationary cameras and lighting at key locations. The latest 2017 crawler successfully traveled the entire length of the tunnel between the Canyon and Sand Filter System for the second time since development, thus providing key visual data for the integrity program [2]. In addition, several access ports were installed in 2017 following the removal of concrete core samples that allow for new visual inspection of the Canyon Crossover tunnel, thus providing information for an area that was generally inaccessible for monitoring. Planning for the 2019 crawler inspection is underway with the expectations to improve upon inspection cameras and also introduce mapping technologies such as LIDAR. The challenge to obtain meaningful data from the remote inspections of a 60+ year operating Canyon Exhaust tunnel continues to be a complex endeavor requiring close coordination of multiple organizations, adherence to Conduct of Operations, Radiological Safety, Worker Safety, etc. principles as well as alignment with Facility Safety Basis requirements. This paper discusses the continuing improvements in inspection techniques and the challenges associated these inspections. (authors)

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DOE OSTI2018

Waste Encapsulation and Storage Facility Stabilization and Ventilation Project-18121

Department of Energy contractor CH2M HILL Plateau Remediation Company (CH2M) is responsible for the safe and compliant operation of the Waste Encapsulation and Storage Facility (WESF), a nuclear materials storage facility at Hanford containing 1,936 highly radioactive capsules containing cesium and strontium. The capsules are currently stored under approximately 4 m (13 ft) of water. The water shields workers from radiation and helps cool the capsules. WESF is a large, concrete structure nominally 48 m (160 ft) long, 30 m (98 ft) wide and 12 m (40 ft) high with the pool cells being approximately 4 m (13 ft) below grade level. It is a robust facility with walls ranging from 0.3 m to 1 m (1 ft to 3 ft) thick. While the capsules are currently in a safe configuration, WESF is an aging facility, and DOE and CH2M are modifying the facility to ensure continued safe and compliant operations and to support the eventual transfer of capsules to dry storage. Dry storage would eliminate the possibility of a release of radioactive material in the unlikely event of a major earthquake that might result in loss of pool storage water and overheating and breach of one or more capsules. The project was divided into two major deliverables: 1) Ventilation: replacement of the existing ventilation system which included a below-grade high efficiency particulate air (HEPA) filter pit with an above grade HEPA filter skid and new fans and ducting. 2) Stabilization: grouting deactivated facility hot cells and other void spaces to the maximum extent practical to reduce material at risk from legacy contamination. The aging ventilation system relied on HEPA filters that exceeded their operational life. Due to the high levels of radionuclide contamination, replacing the filters was impractical and highly hazardous. CH2M installed a new ventilation system to provide safe and compliant ventilation for future operations. The project used an existing, unused ventilation skid that was designed for another project. Repurposing the unused skid presented additional technical challenges to modify the system for the project specific requirements. The new system is above ground, easier to maintain and is built to current standards. New connections inside the facility were installed to allow the new ventilation to operate effectively. The stabilization project is unique in that core drilling and grouting was performed in an operating Hazard Category 2 nuclear facility, which brought forth safety basis and environmental regulatory challenges. The hot cells were isolated from all existing support utilities including electrical, process chemicals, manipulator ports, instrument air, etc. to seal the hot cells and prevent highly contaminated supernate or grout from leaking into adjacent galleries. A high flow grout was used to maximize the distance between injection ports. (authors)

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DOE OSTI2018

NORM in Building Materials: A Case Study in Releasing and Disposing of Building Debris within the DOE System - 18123

In 2013, the Office of Management and Budget issued Memorandum 2013-02, clarifying its 'Freeze the Footprint' policy. Memorandum 2013-02 requires that all federal agencies maintain or reduce their total square footage, within each subsequent year, from a FY2012 baseline. Memorandum 2013-02 also requires that total square footage remain constant with respect to the type/intended use of a building. At Los Alamos National Laboratory (LANL), this means that new construction cannot occur without demolition of equivalent (both in type and square footage) building(s). Since 2013, decommissioning and demolition of aging, unused facilities at LANL has increased. In 2016, the TA16-280 building complex was scheduled for demolition. TA16-280 is a former high explosives (HE) processing facility that consists of six buildings connected to each other via weatherproof walkways. It was constructed in the 1950's and has been subsequently used for HE shipping, receiving, and storage. Interviews with radiation protection staff suggest that some of the buildings within the complex may have been used to store depleted uranium metal parts in a non-dispersible form. It is LANL policy to conduct decommissioning surveys for radioactive materials when building history cannot confirm that a structure is radiologically non-impacted. LANL follows the Multi-Agency Radiation Survey and Assessment of Materials and Equipment (MARSAME) process to develop its surveys and clear structures from radiological control. After clearance, a structure may be dispositioned as non-regulated waste, at a significant cost savings compared to regulated disposition pathways such as low-level waste. MARSAME measurements conducted in the 16-280 complex were elevated compared to reference measurements of surrogate materials, and a majority of the survey units failed to pass indistinguishable from background (IFB), which is LANL's unregulated clearance criteria. Further measurements conducted both in situ and in the laboratory indicated the presence of naturally-occurring thorium in several building materials including: poured concrete, the glaze atop ceramic blast tiles, and the porcelain enamel on the exterior of blast-proof ceiling lamps. Other than naturally-occurring thorium, there were no elevated radionuclide concentrations detected in building materials. It should be noted that there was no history of thorium use in the 16-280 complex, and the material in question was in all cases fixed to and within the building materials in question. Except in certain specific cases, such material is exempt from regulation under the Atomic Energy Act of 1954 [3], as amended, and is therefore exempt from regulation under the Nuclear Regulatory Commission, the Department of Energy, and agreement state regulatory frameworks. Despite this regulatory exemption, disposition of material from the 16-280 complex proved to be challenging, principally because local industrial landfills stipulate 'No radioactive material' in their Waste Acceptance Criteria. This disconnect between what is regulated and what is disposable creates a situation where some DOE sites find it easier to disposition property as Low Level Waste in lieu of pursuing a non-regulated disposition pathway. This presentation will explore the specifics of the 16-280 decommissioning project as well as the implications of the release/disposition disconnect for naturally occurring radioactive material and project planning at LANL. (authors)

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DOE OSTI2018

Closing the Book on the WVDP Vitrification Facility - 18222

Demolition of the West Valley Demonstration Project's Vitrification Facility was initiated on September 11, 2017, approximately 21 years after the facility produced its first canister of vitrified high-level waste. In the center of the structure, the heavily-reinforced concrete Vitrification Cell operated for six years, safely immobilizing about 24 million curies of high-level radioactive waste in borosilicate glass. When the facility was shut down in 2002, it had achieved the highest operational availability of any vitrification facility in the world. It was the first facility in the United States to complete its vitrification mission and is now the first in the nation to undergo demolition. This paper summarizes the selection, design, operation, deactivation and demolition of this facility. (authors)

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DOE OSTI2018

Measurement of Gas Bubbles Encased in Bitumen Nuclear Waste Drums with Muon Scattering Tomography - 18237

In some countries low and intermediate level nuclear waste is stored in steel containers with either pure bitumen added to fill the free volume, or after mixing of the waste with bitumen. Oxidation of the bitumen by interacting with the nuclear waste results in the production of hydrogen. Since pure bitumen is impermeable to water and gases, this hydrogen can assemble in bubbles, possibly resulting in bitumen overflowing out of the waste containers. It is necessary to develop ways of measuring the gas bubbles present in nuclear waste without intervening and compromising the container. We show that this can be done using muon scattering tomography, by applying an analysis method, first developed to find high-Z materials in a lower Z background of concrete, to data from a low-Z material in a higher Z background. The performance of this method is shown, by implementing it on Monte Carlo simulation data. A resolution on the total gas volume of 1.6±0.8% has been achieved for bubbles of 2 liters or more. It is also shown that the method can determine if the gas is located in separate bubbles or concentrated in a bigger one. (authors)

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DOE OSTI2018

The DISTINCTIVE University Consortium: Structural Integrity - 18272

The Engineering and Physical Sciences Research Council (EPSRC) sponsored DISTINCTIVE consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) is developing technologies for civil infrastructure repair, in-situ subsurface waste immobilisation, and groundwater protection during construction and decommissioning. The consortium has contributed to the development of skilled cross-disciplinary civil engineers and scientists, that have the knowledge and experience required to develop engineering solutions tailored for application within radiologically contaminated sites. The Structural Integrity Theme focuses on challenges ranging from site-scale infrastructure preservation and restoration, through injectable ground barriers for risk mitigation, to the remote characterisation and handling of individual waste packages. The main aim of the theme is to develop novel engineering solutions, tailored for use on radiologically contaminated sites, for: ground protection; infrastructure characterisation; concrete restoration and waste characterisation. Technologies should minimise current, and future, radiation exposure of the workforce whilst providing economically viable engineering solutions. (authors)

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DOE OSTI2018

Hanford Double Shell Tank Integrity Inspection - Integration of Customized Robotics - 18400

28 steel, double-shell, million gallon waste storage tanks located at the Hanford site in Washington State store both radioactive and hazardous waste. To monitor their condition and ensure continued viability, these tanks are inspected periodically through a comprehensive integrity program. In 2012, the first leak in a double-shell tank at Hanford was discovered in the 241-AY tank farm (Tank AY-102) and the failure was observed to be from the primary tank bottom. However, the exact failure location and damage mechanism have not been determined. A decision was made to pursue the capability to inspect the primary tank bottom region in order to provide additional tank integrity information. This work is particularly challenging due to the confined space in the double-shell tank annulus region (i.e. the 30 inch space between the primary and secondary tank shells), the limited dimensions of the tank risers that provide access to the annulus space (risers as small as 10 inches in diameter), and also the very small dimensions of the air channels in the concrete refractory pad that the primary waste storage tank rests on (channels are approximately two inches in diameter). Veolia Nuclear Solutions, along with Key Subcontractor Inuktun, were contracted to provide the capability to conduct these tank inspections in support of Hanford's double-shell tank integrity program. The Veolia Nuclear Solutions /Inuktun team designed, built, tested and delivered a tethered miniature robotic crawler system that utilizes the marsupial approach - a larger deployment crawler positions a smaller inspection crawler. The deployment crawler is installed via the tank riser, is magnetically coupled to the side of the tank shell, and drives to the selected deployment location at the end of one of the refractory channels where the small inspection robot is subsequently deployed. Operation of the equipment is controlled from a trailer outside the tank farm (a distance of several hundred feet). The system design includes a custom mechanism that provides for positioning and cable handling as well as the controls interface (laptop connected via control cables). The solution leveraged the Veolia Nuclear Solutions/Inuktun team's experience with delivering similar modular systems for other US and international sites. This simple but innovative solution delivered on an accelerated project schedule underlines the ability to provide effective robotic solutions to the nuclear industry. (authors)

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DOE OSTI2018

Spectroscopic and Computational Analysis of Chemical Pollutants Interactions on Steel - 18409

There is limited understanding of the scientific interaction of the chemical contaminants to DOE facility materials which can alter the stability of the materials, affecting the general safety. The resulting studies will have a direct impact towards developing an enhanced cost-effective decontamination method and understanding the negative effects caused by the chemical waste on the structural integrity of the solid surface in contact. First order kinematic models provide a better fit to the experimental data, implying that one layer of Hg adsorbs to the steel surface. Energy dispersive X-ray analysis confirmed that stainless steel samples are mainly composed of Fe, Ni, and Cr, and have a low carbon content. Objective: To determine the physiochemical interactions of Mercury (Hg) and Cesium (Cs) to steel and concrete using spectroscopic, adsorption and computational studies. To develop a multiscale simulation that shows the effects of adsorption and diffusion of alkali metals on the mechanical performance of steel alloys used in nuclear power plants. Approach: Molecular Dynamics (MD) simulations were used to compute the stress gradient generated in steel due to Hg adsorption. Finite element models were developed for simulating the effects of temperature and pressure in a diffusion analysis. Based on experimental data, one layer of Hg was used for the MD simulations. Material properties at the atomic level were determined from the MD simulations The stress gradient from the MD simulations is used as an input parameter in the FE simulation. Conclusions and Future Work: Computational methods such as MD and FE simulations were used to determine the effects of mercury adsorption in stainless steel samples. The adsorption process induces a stress gradient in the facility materials, which can affect the mechanical performance of facility materials under service. Tensile stresses help the diffusion process while compressive have the opposite effect. Temperature has a positive effect in the mass diffusion process. Future work will evaluate the effects of temperature on the material properties at the molecular level using both experimental and MD methods. Future work will evaluate the mechanical performance of facility materials at a wide range of environmental variables.

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DOE OSTI2018

Aging Management Considerations for Dry Cask Storage Canisters - 18472

Dry cask storage systems (DCSSs) are used to store used nuclear fuel (UNF) at independent spent fuel storage installations (ISFSIs) prior to final geologic storage or reprocessing. As storage durations increase beyond the twenty-year period commonly implemented for licensing the storage sites, it is important to consider aging related degradation issues for DCSS components. A large portion of the DCSS designs used throughout the world employ a ventilated concrete overpack with a welded stainless steel canister that confines the UNF. This paper explores key issues related to aging management for these canister components, including: (1) canister environment, (2) inspection limitations, and (3) aging management program guidance and requirements. While the austenitic stainless steel used for the canisters are generally resistant to most forms of environmental attack, it is possible for conditions to arise that lead to pitting and even stress corrosion cracking of the canister systems. A large number of studies related to canister degradation have been performed by industry groups and government researchers, both in the US and worldwide. These advances in the collective knowledge provide insight into risk factors for DCSS. Such information is beneficial to improve ISFSI siting decisions and ensure successful relicensing of existing ISFSIs. This paper highlights relevant research and also discusses applicability to degradation of other equipment and containers used to process and isolate UNF and other nuclear waste streams from the public and the environment. Inspection of the canister surface is an important tool to investigate and evaluate the potential for degradation. Existing DCSS designs typically do not offer ready access to the canister surface in order to perform the inspection. Further complicating inspection efforts are the surface conditions present on the sheltered canisters after extended storage times, including atmospheric debris such as dust and pollen. Given ongoing development efforts, inspection capabilities and associated limitations should be assessed at multiple points in time over the course of an aging management program to incorporate new advances in the field. This paper explores the limitations of currently available capabilities in inspection and delivery, and provides insight into the gaps remaining to be filled to permit an effective inspection program. Recommendations and requirements related to programs for aging management of canisters have been published or are in progress from a variety of groups, including industry groups, regulatory authorities, and Codes and Standards. An effective aging management program should consider the technical bases of these various inputs and maintain compliance with requirements. Given multiple sources, decisions will be required when considering conflicting guidance, and this paper will include discussion of the relative merits of the various approaches. (author)

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DOE OSTI2018

Dynamic Leaching Method for Intact Salt-stone Samples - 18495

At the Savannah River Site (SRS) low activity salt solution is stabilized via encapsulation within a grout termed salt-stone. Salt-stone is emplaced into large (multi-million liter) concrete storage containers referred to as Salt-stone Disposal Units (SDUs). Technetium-99 (Tc-99) and iodine-129 (I-129) are long-lived radionuclides contained in the low activity salt waste and subsequently incorporated into the grout waste form: both radionuclides are contributors to the potential long-term radiation exposure of the public and environment surrounding the SDUs. With respect to Tc-99, in the reducing, high pH environment within the grout, Tc-99 is expected to be relatively immobile since it exists in a reduced Tc(IV) oxidation state in the form of sparingly soluble hydrated oxides (TcO{sub 2}.xH{sub 2}O) or sulfides (TcS{sub x}). However, in the presence of O{sub 2} (associated with the future infiltration of air or oxygenated ground waters into the salt-stone monolith) it is possible for redox-sensitive Tc(IV) to transition into highly soluble (and mobile) Tc(VII) species, such as pertechnetate (TcO{sub 4}{sup -}), which are more readily transported to the surrounding environment. In contrast, I-129 is highly soluble and its leaching behavior is not anticipated to be significantly impacted by potential changes in redox environment or pH within salt-stone. Traditional approaches to quantifying the leaching behavior of Tc-99 and I-29 from cementitious matrices have involved partitioning experiments using size-reduced (crushed/ground) samples, and determination of the radionuclide fraction immobilized by the cementitious solids. Such experiments create artificially high solid-liquid contact areas that likely result in higher Tc-99 and I-129 leachate concentrations than would be expected for intact, monolithic samples. In the current study a new technique, termed the Dynamic Leaching Method (DLM), is being used to initially investigate the Tc-99 leaching behavior from monolithic salt-stone samples; the system will be used at a later date to evaluate the leaching characteristics of I-129. The data derived using this technique is intended to inform the SRS Salt-stone Disposal Facility (SDF) Performance Assessment (PA) which models the long-term transport of radionuclides from the SDUs to the environment. The DLM utilizes a flexible-wall permeameter to achieve saturated leaching under an elevated hydraulic gradient to simulate and accelerate the future transport of groundwater through salt-stone. Leachates are regularly analyzed with respect to radionuclide concentrations, key constituent concentrations (e.g., Al, Ca, Mg, Na, Si, CO{sub 3}{sup 2-}, NO{sub 2}{sup -}, NO{sub 3}{sup -}, SO{sub 4}{sup 2-}), pH, and reduction potential (E{sub h}) in order to develop an understanding of how chemical changes associated with the pore solution influence grout weathering and radionuclide leaching behavior. This paper provides a brief update with respect to data derived from preliminary Tc-99 leaching evaluations as the DLM was being developed, but the paper is predominantly focused on new sample data in which the DLM technique, pre- and post-DLM salt-stone sample characterization, and leachate analysis protocols have been optimized. (authors)

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DOE OSTI2018

Salt Reconsolidation and Repository Engineered Barriers - 18651

Salt repositories for the permanent disposal of hazardous waste provide extraordinary advantage over other geo-media because geotechnical barriers can be constructed with confidence using crushed salt, which is a perfectly compatible material that is readily available by mining practice. Other excellent barrier materials that consist of specialty mixes of salt-based concrete, bentonite, and bitumen bolster confidence that a repository in geologic salt formations will remain impermeable forever. Practically speaking, achieving high-performance engineered barriers assumes an isolation role as important as the host formation itself. Long-term performance assessment intrinsically includes evaluating seal-system design and demonstrated operational characteristics. Consistent with global interests, specific research on salt-repository seal systems not only established significant technical bases but identified possible improvements through straightforward research and laboratory and field investigations. Engineered barriers of crushed salt with possible additives, such as clay and water, evolve to create hydrological and mechanical characteristics that approach those of the host salt formation. Demonstrating early functionality of seal systems reduces uncertainty, which is introduced when calculating or predicting when performance properties become effective in the future. (authors)

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DOE OSTI2016

Incredibility of a Postulated Criticality Event for Enriched Uranyl Nitrate Storage

At the H-Canyon outside facilities, highly enriched uranyl nitrate solutions collected from leaks or during deliberate maintenance activities is collected into 5-gallon polyethylene bottles. The current criticality safety evaluation for the facility postulates that a criticality event is possible if the solutions are allowed to concentrate through evaporation, freezing, storing multiple bottles in a geometrically unfavorable array, or any combination of these conditions. Using chemical data on uranyl nitrate freezing from Savannah River National Laboratory, protected limits on concentration and enrichment, and normal operational restrictions, criticality was shown to be incredible under all normal and credible abnormal conditions. A bounding normal solution is analyzed under normal conditions and shown to be substantially subcritical. A bounding credible abnormal solution is defined as 9.2 g U{sup 235}/L and 73 wt. % enriched and 0 M excess nitric acid. The concentration and enrichment are limitations already protected by the facility for other processes. Because acidity is not specifically restricted, the most reactive lower bound is selected. The credible abnormal worst placement is in a concrete basin reflected on three sides. There the bottle(s) would be subject in the worst cases to flooding (full water reflection) and postulated concentration through evaporation or freezing. Computational modeling was performed using KENOVI in the SCALE 6.1 code package. (authors)

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DOE OSTI2016

The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Materials Challenges and Experimental Approaches to Understanding Long-Term Materials Behavior for Extended Nuclear Plant Operations

Commercial nuclear reactors present a very challenging environment for materials to operate within. Materials undergo combinations of environmental stressors that include high temperatures, contact with water, high stress loading, fatigue, and intense radiation fields. Over twenty five different metal alloys can be found within the primary and secondary systems along with numerous types of materials present in the containment vessel, concrete structures, instrumentation and control equipment, cabling, buried piping, and other support facilities. Dominant forms of degradation may vary greatly between the different systems, structures and components. Understanding materials degradation is necessary for informed long-term operation decisions that establish materials limits, support focused planning of materials inspections, provide economic and safety based decisions on component replacement, and allow for more appropriate regulatory guidelines. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Optimization of Backfill Material Transport for the Future Belgian Geological Disposal Facility - 19071

In Belgium, long-lived, low- and intermediate level waste (B waste) will be conditioned in concrete monoliths B and high-level waste (C waste) in so-called Supercontainers (SC). It is planned to dispose of the waste packages in a Geological Disposal Facility (GDF). The B and C waste shall be disposed of each in one emplacement field at a depth between 200 m and 600 m below the surface. Two access galleries run through each field and connect the disposal galleries with a central shaft and support zone. After the emplacement of a specified number of waste packages, a stepwise backfilling of the voids is planned in order to stabilize the galleries, to limit the amount of voids, and to ensure a favourable chemical environment. Taking into account the objectives of backfilling and the specific material properties of the waste packages two mortars were developed to backfill the disposal galleries. In particular, for reasons of safety at work, the mortars should be hydraulically pumped through pipelines. Accordingly, in any case, it is necessary to operate storage facilities of the raw materials, a mix plant and a pump. The raw materials are foreseen to be supplied by trucks from external producers. Due to this fact, the better accessibility of the plants and with the aim of minimizing noise and dust emissions and the space for technical systems in the GDF, it is advisable to install the technical equipment above ground outside the GDF control area. However, the feasibility of this basic concept depends on the processing time of the backfill materials and the capacity of the transport system. Thus, although short transport times can theoretically be realized if pipelines with a small cross sections are chosen, in this case the required power of the pump increases significantly. If no suitable configuration can be found, the distance of the mixing and pumping system to the backfill segment has to be shortened. In this case a pneumatic conveyor system would have to be installed to transport the solids to a mix plant and pumping unit located in the underground galleries to be able to meet the desired backfill rates. In order to identify the optimum solution, the decision criteria and the advantages and disadvantages of existing and planned transport systems were evaluated. The investigations showed that it is technically feasible to pump the backfill material from above ground into the disposal galleries. As failures of the backfill system would result in interruptions of the general disposal operation, a redundant pump and pipeline system was developed. The two lines can be operated separately or simultaneously in order to realize high flow rates needed for the backfilling of segments far from the shaft. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗