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At least 19 records

Designing Lightweight Neutron Absorbing Composites Using a Comprehensive Absorber Areal Density Metric

Efforts to lightweight neutron absorbing composites are limited by incomplete understandings of the interaction between absorbing particles and their matrices. In this study, analytical models and a more physically representative simulation evaluated the penalty to neutron absorbing performance due to neutron channeling between large absorbing particles. Models and simulation agreed that B 4 C particles smaller than 100μm and especially those smaller than 10μm did not cause excessive neutron channeling. A more comprehensive neutron absorbing composite design metric – boron-10 equivalent areal density, which considers the particle size penalty and the matrix contribution to absorptivity – was introduced and used to estimate lightweighting via matrix substitution. Calculations using this new metric showed that a non-absorbing Mg matrix reduced mass by up to 35% over Al, constrained by the difference in mass density, while an absorbing Mg–Li matrix reduced mass by up to 60%, exceeding the difference in mass densities alone. Measurement of apparent absorber areal density through two experimental techniques – foil activation and direct counting – validated estimated absorber areal density as a neutron absorbing composite design metric. This updated understanding of the particle size penalty, newly introduced design metric, and experimental validation demonstrate a path to lightweight neutron absorbing composites.

Radiation shielding↗

Thermomechanical Properties of Neutron Irradiated Al 3 Hf-Al Thermal Neutron Absorber Materials

A thermal neutron absorber material composed of Al 3 Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al 3 Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4T m . Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al 3 Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al 3 Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al 3 Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of –6.3% for the annealed irradiated 36.5 vol% specimen.

36 MATERIALS SCIENCE↗

Microstructure of Neutron-Irradiated Al 3 Hf-Al Thermal Neutron Absorber Materials

A thermal neutron-absorbing metal matrix composite (MMC) comprised of Al 3 Hf particles in an aluminum matrix was developed to filter out thermal neutrons and create a fast flux environment for material testing in a mixed-spectrum nuclear reactor. Intermetallic Al 3 Hf particles capture thermal neutrons and are embedded in a highly conductive aluminum matrix that provides conductive cooling of the heat generated due to thermal neutron capture by the hafnium. These Al 3 Hf-Al MMCs were fabricated using powder metallurgy via hot pressing. The specimens were neutron-irradiated to between 1.12 and 5.38 dpa and temperatures ranging from 286 °C to 400 °C. The post-irradiation examination included microstructure characterization using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy. This study reports the microstructural observations of four irradiated samples and one unirradiated control sample. All the samples showed the presence of oxide at the particle–matrix interface. The irradiated specimens revealed needle-like structures that extended from the surface of the Al 3 Hf particles into the Al matrix. An automated segmentation tool was implemented based on a YOLO11 computer vision-based approach to identify dislocation lines and loops in TEM images of the irradiated Al-Al 3 Hf MMCs. This work provides insight into the microstructural stability of Al 3 Hf-Al MMCs under irradiation, supporting their consideration as a novel neutron absorber that enables advanced spectral tailoring.

36 MATERIALS SCIENCE↗

Neutron Absorber Plate Characterization Plan for Criticality Experiments Design

After being used in nuclear installations, depleted fuel can still be highly reactive and must be handled securely to prevent any radiological or criticality concerns. In particular, spent fuel from use in nuclear power reactors must be stored and transported in specifically designed containers using neutron absorber materials to prevent criticality. Various neutron absorber material types exist and are manufactured by various entities, as thoroughly described in the Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications written by EPRI. Presently, one of the most modern and most widely used types of neutron absorber material contains particles of boron carbide, or B 4 C, embedded in aluminum matrix: Boralcan, manufactured by Rio Tinto. It is very important for the community to know as much as possible about such neutron absorber materials. Therefore, in the recent years, a US Department of Energy National Nuclear Security Administration–Nuclear Criticality Safety Program funded project initiated design of an experiment that places Boralcan neutron-absorbing plates in an established critical assembly using low-enriched uranium fuel at the Sandia Pulsed Reactor Facility/Critical Experiments (SPRF/CX) apparatus at Sandia National Laboratories. The goal of the experiment is to produce high-quality benchmark data to submit to the International Criticality Safety Benchmark Evaluation Project (ICSBEP), for use in validating calculational tools and nuclear data by criticality safety analysts. The project, named IER-554, is currently in its final design stage, following a successful preliminary design. In the work documented in the design study, ten critical configurations using Boralcan neutron absorber plates were designed, and the experiment was proven to be feasible, with a predicted low k eff uncertainty around 100 pcm. An overview of the modeled cutout of the critical assembly with a Boralcan plate is shown in Figure 1, representing one of the configurations planned for the critical experiments. Before the plates are inserted in the critical assembly, it is necessary to know more about their composition and uniformity. This summary focuses on the plate characterization plans. Each plate will undergo (1) neutron transmission measurements at different locations to determine the 10 B areal density and (2) an in-depth x-ray computed tomography (XCT) examination to obtain the exact Sizes and distribution of the B4C powder particles inside the plates. In parallel, plate modeling studies are performed with a goal to determine the validity of the currently used approximation of modeling the neutron absorber plates as a homogeneous mixture of Aluminum 1100 alloy and B4C— instead of explicitly modeling the B4C particles. By using the experimental 10 B areal density measurements, and the exact size and location of the B4C particles obtained by XCT, a plate model can theoretically be built that reproduces the plate with extremely high fidelity. The results of this modeling study could increase the confidence of the criticality safety community in its modeling methods when using this type of neutron absorber material, and the industry could use these validations to change the boron loading credit limits from the U.S. Nuclear Regulatory Commission standard review plan for dry cask storage of spent nuclear fuel. The modeling calculations are performed with SCALE 6.3.0 using the KENO V.a sequence for criticality calculations with the ENDF/B-VIII.0 continuous-energy cross section library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluation of Neutron Absorber Panels from Zion Spent Fuel Pool

In spent fuel pools (SFPs), fixed neutron absorber materials are used to increase storage capacity while maintaining criticality safety margins. BORAL{sup R} is a commonly used neutron absorber material for both wet (SFPs) and dry storage (casks). The specific questions surrounding fixed neutron absorber materials include the following: 1. What are the conditions of the neutron absorber panels in SFPs? Is there any gross degradation that could cause potential concern for criticality safety of the pools? 2. Is the coupon monitoring approach adequate for monitoring the conditions of the panels as part of an aging management program? 3. For plants that do not have coupons, do current in situ measurement approaches provide accurate results? Could such in situ approaches be used as an alternative monitoring approach? In order to shed light on these technical questions using actual plant data, EPRI initiated the Zion comparative analysis project at the decommissioning Zion Nuclear Power Plant in late 2014. An overview of the Zion plant, SFP, and project was described in a previous publication. (authors)

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Evaluation of Neutron Absorbers in the DOE Standardized SNF Canister

The goal of this evaluation was to evaluate different options for a neutron absorber in the U.S. Department of Energy (DOE) Standardized spent nuclear fuel (SNF) Canister. This was accomplished by taking the two most relevant criticality evaluations and replacing the advanced neutron absorber (ANA) basket material with borated stainless steel. This work was initiated, because alternative storage options for DOE SNF has become a priority for DOE. Idaho National Laboratory (INL) is initiating activities to ensure the continued operation of the Advanced Test Reactor (ATR) by providing alternative and redundant storage options for spent ATR fuel. In addition, the DOE Office of Environmental Management (EM) has recently initiated studies to understand and improve the technical basis for long-term dry storage of aluminum clad SNF (e.g., ATR, High Flux Isotope Reactor [HFIR]) (Connolly 2018). While currently in the lab-scale phase, the next step involves validation and verification (V&V) of the lab-scale results. One method to perform V&V is to load aluminum-clad spent fuel in an instrumented DOE Standardized SNF Canister as a demonstration. In order to minimize the need for repackaging before disposal and to provide a representative environment in a demonstration, the fuel is planned to be loaded with an appropriate criticality control mechanism, such as neutron absorbing basket material. While ANA was originally selected as the basket material for ATR fuel in the Yucca Mountain Repository, it has never been produced on a large scale. Therefore, incorporating this material in the DOE Standardized SNF Canister as part of a near-term demonstration could be more challenging than moving forward with a commercially available material, such as borated stainless steel. Past evaluations eliminated borated stainless steel as a long-term neutron absorbing material, because it corroded too quickly and may not always remain in the waste package after degradation. This led to the invention and selection of ANA for use as the basket material in the DOE Standardized SNF Canister. After selecting ANA, new corrosion tests were performed on borated stainless steel using a different method for fabrication. This method of fabrication significantly improved the corrosion resistance properties of borated stainless steel, so much so that it out-performed ANA in corrosion testing. Researchers used the information obtained from these tests to select borated stainless steel as the neutron absorber in the Transportation, Aging, and Disposal Canister (TAD) designed for commercial SNF. This evaluation compares the criticality control of borated stainless steel to ANA. In every case and scenario, the calculated effective neutron multiplication factor (k eff ) using a borated stainless steel basket was lower than that using an ANA basket, though the borated stainless steel must be thicker due to corrosion effects over the regulatory time period. Although the borated stainless steel performed better than the ANA, it still required additional neutron-absorbing material, gadolinium shot, for the calculated k eff to fall below the upper critical limit of 0.93 in a few cases evaluated. More research using thicker baskets or inserts could be performed in an attempt to lower the k eff without the use of gadolinium shot. In addition, a reevaluation of the degraded ATR material used in the previous evaluations may prove that the original assumptions were over-conservative.

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The Gd based Neutron Absorber for Spent Fuel Storage and Transport Cask - 17636

The criticality safety calculations were performed for integral model which has single rack wall of stainless steel added neutron absorber material. The integral model was verified by reference model composed of 3 layer; basket cell, neutron absorber and sheath. For the neutron absorber material, Gd was selected due to its high neutron absorption cross section in thermal region compared to B. The mathematical model based on validated code, KENO-Va in SCALE was used and the data library is ENDF/B-VII 238-Group. The k{sub eff} was calculated for changes in Gd and B contents in flooded condition. The Gd added stainless steel shows higher performance than B at same atomic percent and reactivity reduction decreases as absorber content increases. Therefore, the Gd 0.47 at% was selected for 'optimum Gd content' for effective use of neutron absorber. With Gd 0.47 at%, B, Sm, Er and Eu was considered for additional neutron absorber and B shows the best performance among them. The Gd 0.47 at% with B 2.05 at% was selected for (optimum neutron absorber content'. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hydrothermal reactivity of neutron absorber composites

The stability of neutron absorber composite materials at hydrothermal conditions was tested in a series of two-week experiments to mimic spent nuclear fuel disposal. Coupons, composed of boron carbide (B 4 C) sintered with and encased in aluminum, were increasingly altered in experiments at 150, 230, and 300 °C and pressures of 150 bar. Alteration of aluminum to boehmite (γ-AlO(OH)) and hydrogen gas generation occurred over the range of investigated temperatures, but is most significant at 300 °C. We found that the formation of boron-bearing mineral phases was not detected; however, aqueous boron was present in the reaction fluids.

36 MATERIALS SCIENCE↗

Overview of neutron absorber materials used in spent fuel pools - 14622

Spent fuel pools (SFPs) were originally designed with low density storage racks that maintained subcriticality based on geometric separation. With the increasing need for storage, due to elimination of reprocessing as a viable alternative, SFPs with high-density racks required neutron absorber materials to control reactivity and maintain criticality safety margins. The Nuclear Energy Institute (NEI) conducted an industry wide survey in 2013 and the survey was updated recently based on additional utility data provided by Electric Power Research Institute (EPRI). The survey results show that about one-quarter of the pools in the United States either do not have neutron absorber material in the SFP or do not credit it even when it is present. Also, about half of the plants in the U.S. use BORAL{sup R} as the neutron absorber material followed by Metamic, Carborundum, Alcan, and borated stainless steel. For neutron absorber monitoring, the majority of the plants rely on a coupon testing program. However, for those plants that never installed coupons or have exhausted their supply of coupons from coupon sampling, in situ measurements at periodic intervals can be performed to determine the status of the neutron absorber. In this paper, an overview of the types of neutron absorber materials used in SFPs, the distribution of absorber material use by material type, types of monitoring programs as part of aging management programs, operational experience to date with different absorber materials based on monitoring programs, and industry initiatives toward addressing long-term performance issues related to neutron absorber material degradation and monitoring are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Overview of zion comparative analysis project for assessment of Boral{sup R} neutron absorber material performance and monitoring in spent fuel pools - 14618

Neutron absorber materials are used in spent fuel pool (SFP) storage racks to control reactivity. In the United States, about half of the operating plants currently use BORAL{sup R} neutron absorber material in SFP racks. Although operational experience spanning more than 25 years has not shown any loss of the BORAL{sup R} core material, corrosion effects, including blisters and pits, have been observed and reported. To improve understanding for the behavior of BORAL{sup R} in SFPs and evaluate the performance of monitoring approaches, EPRI initiated the Zion comparative analysis project in 2014. The Zion project involves three phases. During the first phase of the project, in situ measurements were performed on selected SFP rack panels in December 2014. Panels for in situ measurements were selected after an extensive analysis of each panel in the pool to ensure that the selected panels represented the pool in terms of variation in neutron and gamma dose values. As part of the second phase of the project, the remaining 10 coupons from the Zion SFP were shipped to a Penn State University (PSU) laboratory for analysis in January 2015. During the final phase of the project, a subset of the panels for which in situ measurements were performed will be harvested from the Zion SFP, cut into samples, and analyzed at PSU. The overall scope of this project will allow comparison of in situ measurements to real panel data as well as data obtained from coupon analysis. In this paper, an overview of the Zion comparative analysis project, the criteria and procedure for panel selection for in situ measurements and harvesting, preliminary coupon analysis results and in situ measurement results are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Applications of neutron-absorbing structural amorphous metal (SAM) coatings for criticality safety controls of used fuel storage, transportation, and disposal - 14534

Used nuclear fuel contains fissionable materials ({sup 235}U, {sup 239}Pu, {sup 241}Pu, etc.). To prevent nuclear criticality in used fuel storage, transportation, and disposal, neutron-absorbing materials (or neutron poisons, such as borated stainless steel, Boral{sup TM}, Metamic{sup TM}, Ni-Gd, etc.) would have to be applied. Corrosion-resistant, iron-based structural-amorphous metals (SAMs) have been tested to determine their relative corrosion resistance. Many of these materials can be applied as coatings with advanced High Velocity Oxy-Fuel (HVOF) thermal-spray technology. SAM2X5, SAM1651, and other SAMs are amorphous-metal composite alloys that have been identified as having outstanding corrosion resistance. Because of its high boron content, SAM2X5 can be applied as the neutron-absorbing coatings to the metallic support structure for criticality-safety controls of used fuel in racks in wet storage pool and baskets inside the storage containers, the transportation cask, and eventually the disposal containers in repository disposal. Research and experiment conducted at Lawrence Livermore National Laboratory indicated that the high boron-containing SAM2X5 coating could be an effective criticality control material for used fuel management. The neutron irradiation experiments and the neutron transmission measurements conducted at McClellan Nuclear Radiation Center indicated that extensive fast neutron irradiation did not change the structure of the amorphous SAM2X5 melt-spun ribbons, and SAM2X5 exhibited effective neutron absorbing capability, similar to Boral{sup TM} and Metamic{sup TM}. (authors)

36 MATERIALS SCIENCE↗

Accident analysis of spent fuel storage and transportation cask with Gd based neutron absorber

The shortage of nuclear spent fuel storage pool has been issued. The spent fuel storage and transportation cask is one of the major alternatives for storing spent fuel. The use of advanced neutron absorber is important for effective storage of spent fuel in spent fuel cask. Gadolinium based neutron absorber is chosen since it has high neutron absorption cross section in thermal neutron region compared to boron and it does not produce any gas element when it reacts with a neutron. The major requirement for neutron absorbing materials in the spent fuel cask system is to maintain subcriticality in normal, off-normal and accident conditions. In this study, the criticality calculations under accident conditions are conducted to ensure the safety of spent fuel cask. Major accident conditions are flood and cask tip-over. The full and partial flooded condition is analyzed and 100% flooded condition has the highest reactivity. The water density is selected as 1 g/cm{sup 3} because k{sub eff} is highest at this condition. When the cask tip-over, the flux trap reduces. In this case, a huge increase in reactivity of 800 pcm takes place. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Neutron Absorber Plate Characterization Plan for Criticality Experiments Design [Slides]

The goal of this presentation is to show the feasibility of performing a critical experiment with a B 4 C/Aluminum alloy neutron absorber plate. The motivation of this experiment is to provide integral data on modern neutron absorber product “Boralcan” by Rio Tinto. In 2020, used in about 20% of spent fuel pools in the US, and in casks for storage and transportation in the US and in Europe. The added value of this experiment: Neutron absorber plate modeling method verification: homogeneous or not, a lot of potential savings by relaxing boron loading credit limits. No available critical benchmark involving Boralcan, other boron-related experiments are relatively old. Nuclear data validation for B, C, and Al in the form of B 4 C in Al matrix.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Neutron Absorber Considerations for the DOE Standardized Canister

The Department of Energy (DOE) standardized canister is being re-evaluated to support the management of non-commercial spent nuclear fuel. The standardized canister would allow a broad range of DOE-managed fuels to be loaded, stored, transported, and disposed of while avoiding the need to re-open the canister after initial loading, thus preventing repackaging of the material. A key safety function of this package is to ensure the material maintains a sub-critical configuration for storage, transportation, and disposal. DOE-managed fuels come from a wide range of reactor types, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. To adequately address the diversity of DOE spent nuclear fuel (SNF), a number of basket were designed. The basket material needed to have the ability to ensure subcriticality, be structurally sound, and be corrosion resistant. One such option that was investigated in the late-1990s through mid-2000s was a corrosion-resistant, nickel-chromium-molybdenum alloy containing gadolinium, which was termed the advanced neutron absorber (ANA). A number of analyses were performed to verify the structural, corrosive properties, and neutron absorption performances of this material as well as to compare other potential materials to be used for the basket. Gadolinium-containing carbon steel and borated stainless steel were also considered and evaluated. Boral was briefly considered, but due to its corrosive properties. Each of the evaluated materials have strengths and weaknesses. ANA is the best thermal neutron absorber but lacks confirmatory research for corrosion and weldability. Gadolinium containing carbon steel provides excellent moderator exclusion capabilities but does not contain a sufficient supply of phosphate to ensure the gadolinium precipitates. To minimize the probability of a critical configuration, it is highly desirable for the gadolinium to precipate the probability that the materials will remain within the waste packages over the postclosure time periods of interest. Borated stainless steel has structural integrity and impressive corrosion resistance in most environments, but can undergo galvanic corrosion when in close proximity/contact with aluminum and may experience less than desirable corrosion resistance. This report provides an extensive list and comparison of the materials evaluated to ensure sub-criticality of DOE standardized canister in all credible scenarios. More analyses need to be performed using new and updated corrosion information to re-evaluate the neutron absorber that best meets the objectives of the DOE standardized canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Update on the use of Neutron Absorbers for Storage Applications at Y-12

The Y-12 National Security Complex has a high density enriched uranium storage system at the Highly Enriched Uranium Materials Facility (HEUMF) that has been deployed for nearly 10 years. A robust neutron absorber system is provided through the design of Rackable Can Storage Boxes (RCSBs) and their configuration into storage racks. The requirements from ANSI/ANS-8.21, Use of Fixed Neutron Absorbers in Nuclear Facilities outside Reactors, are met with the RCSB and rack system at HEUMF through various design aspects, verification criteria, functional testing, and simulation modeling. Previous publications on a neutron absorber storage system at Y-12 were present at ANS in 2001 and 2002, when the RCSBs were still in a conceptual phase. The publications focused on ongoing studies at the time of what materials worked best for storage applications and how to test those materials, as well the practical design of the unit. The RCSBs have proven to provide a safe, static, and low maintenance storage system, with some minor design issues and several opportunities for future improvement.

42 ENGINEERING↗

Hot deformation behaviors and processing maps of B{sub 4}C/Al6061 neutron absorber composites

In this study, the hot deformation behaviors of 30 wt.% B{sub 4}C/Al6061 neutron absorber composites (NACs) have been investigated by conducting isothermal compression tests at temperatures ranging from 653 K to 803 K and strain rates from 0.01 to 10 s{sup −1}. It was found that, during hot compression, the B{sub 4}C/Al6061 NACs exhibited a steady flow characteristic which can be expressed by the Zener-Hollomon parameter as a hyperbolic-sine function of flow stress. High average activation energy (185.62 kJ/mol) of B{sub 4}C/Al6061 NACs is noted in current study owing to the high content of B{sub 4}C particle. The optimum hot working conditions for B{sub 4}C/Al6061 NACs are found to be 760–803 K/0.01–0.05 s{sup −1} based on processing map and microstructure evolution. Typical material instabilities are thought to be attributed to void formation, adiabatic shear bands (ASB), particle debonding, and matrix cracking. Finally, the effect of the plastic deformation zones (PDZs) on the microstructure evolution in this 30 wt.% B{sub 4}C/Al6061 composite is found to be very important. - Highlights: •The hot deformation behavior of the 30 wt.% B{sub 4}C/Al6061 NACs was first analyzed. •The 3D efficiency map and the instability map are developed. •The optimum hot working conditions were identified and validated by SEM and TEM. •The hot deformation schematic diagram of 30 wt.% B{sub 4}C/Al6061 NACs is developed.

36 MATERIALS SCIENCE↗

Preliminary Design of Critical Experiments Involving Commercially Available B 4 C Neutron Absorber Plates with Low-Enriched UO 2 Fuel [Slides]

Conclusions of this presentation conclude that the experiment is feasible with at least 10 critical configurations, and low expected benchmark uncertainty. Potential high impact for the understanding of commercially used neutron absorber plates. Next steps are to think about other configurations: rod pitch, other rods – Finalize plate support, uncertainty study. Additionally, the potential use of IER-441 central test region to safely place the plate in the core. Next steps include the Determination of the neutron absorber plate characterization path and the Potential SINBAD benchmark of neutron transmission measurements.

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