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Fukushima two year processing update

On March 11, 2011, now two years ago, the magnitude 9.0 Great East Japan earth quake, Tohoku, hit off the Fukushima coast of Japan. This was one of the most powerful earthquakes in recorded history and the most powerful one known to have hit Japan. The ensuing tsunami devastated a huge area resulting in some 25,000 persons confirmed dead or missing. The perfect storm was complete when the tsunami then found the four-reactor, Fukushima- Daiichi Nuclear Station directly in its destructive path. Some 2 million people were evacuated from a fifty mile radius of the area and evaluation and cleanup began. A tremendous effort has been made, by many nationalities, since this time to restore this damaged plant and surrounding area and to return a great deal of the residents to their homes and farm lands. While most of the outcome of this unprecedented natural and manmade disaster was negative, both in Japan and worldwide, there have been some extremely valuable lessons learned and new emergency recovery technologies and systems developed to cope with the aftermath of this disaster. This paper describes new technology developed to selectively remove radioactive materials dangerous to workers, local citizens, and the natural environment from seawater used to cool the damaged reactors at Fukushima. As always, the mother of invention is necessity. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

2015 Annual Summary Report for the Area 3 and Area 5 RWMs

This annual summary report presents results and conclusions of the Fiscal Year (FY) 2015 review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs) in support of the Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site (NNSS). The report presents results and conclusions of the FY 2015 review and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (R&D) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the NNSS relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and R&D activities were reviewed to determine the adequacy of the CAs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Support of SRS salt-stone regulatory compliance - 15586

The SRS salt-stone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. The salt-stone Processing Facility (SPF) is permitted as a wastewater treatment facility per the South Carolina Department of Health and Environmental Control Regulations (SCDHEC) South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-67. The salt-stone Disposal Facility is permitted as a solid waste landfill site, as defined by SCDHEC Regulations R61-107.19. The salt-stone Sampling and Analysis Plan provides the SCDHEC with the chemical and physical characterization strategy for the salt solution and salt-stone waste form. During operation, samples are collected from Tank 50H on a quarterly basis and are analyzed by the Savannah River National Laboratory (SRNL) for over chemical and radioactive components. The analytical results are compared to the salt-stone waste acceptance criteria in a technical report. SRNL also prepares salt-stone samples from the quarterly sample and premix binder materials provided by the SPF to determine the non-hazardous nature of the grout to meet the requirements of the SCDHEC regulation for the 'Characteristic of Toxicity', R.61-79.261.24(b) and the 'Universal Treatment Standards', R.61-79.268.48.4. salt-stone samples are collected, size reduced, packaged, and shipped to an independent United States EPA certified laboratory, or a laboratory that holds an equivalent certification such as one from the National Environmental Laboratory Accreditation Program to perform the EPA Toxicity Characteristic Leaching Procedure (TCLP) extraction of the salt-stone sample. SRNL reviews the results of the TCLP analysis and prepares a report detailing the sample preparation and analysis that is used by the salt-stone Facility to demonstrate the nonhazardous nature of the salt-stone waste form. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Joint US/Russian Studies of Population Exposures Resulting from Nuclear Production Activities in the Southern Urals

Beginning in 1948, the Soviet Union initiated a program for production of nuclear materials for a weapons program. The first facility for production of plutonium was constructed in the central portion of the country east of the southern Ural Mountains, about halfway between the major industrial cities of Ekaterinburg and Chelyabinsk. The facility now known as the Mayak Production Association and its associated town, now known as Ozersk, were built to irradiate uranium in reactors, separate the resulting plutonium in reprocessing plants, and prepare plutonium metal. The rush to production, coupled with inexperience in handling radioactive materials, lead to large radiation exposures, not only to the workers in the facilities, but also to the surrounding public. Fuel processing started with no controls on releases, and fuel dissolution and accidents in reactors resulted in release of about 37 PBq (1015 Bq) of 131I between 1948 and 1967. Designed disposals of low- and intermediate-level liquid radioactive wastes, and accidental releases via cooling water from tank farms of high-level liquid radioactive wastes, into the small Techa River caused significant contamination and exposures to residents of numerous small riverside villages downstream of the site. Discovery of the magnitude of the aquatic contamination in late 1951 caused revisions to the waste handling regimes, but not before over 200 PBq of radionuclides (with large contributions of 90Sr and 137Cs) were released. Liquid wastes were diverted to tiny Lake Karachay (which today holds over 4 EBq); cooling water was stopped in the tank farms. In 1957, one of the tanks in the tank farm overheated and exploded; over 70 PBq, disproportionately 90Sr, was blown over a large area to the northeast of the site; a large area was contaminated and many villages evacuated. This area today is known as the East Urals Radioactive Trace (EURT). Each of these releases was significant; together they have created a group of cohorts unrivaled in the world for their chronic, low-dose-rate radiation exposure. The 26,000 workers at Mayak were highly exposed to external gamma and inhaled plutonium. A cohort of individuals raised as children in Ozersk is under evaluation for their exposures to radioiodine. The Techa River Cohort consists of over 30,000 people who were born before the start of exposure in 1949 and lived along the Techa River. The Techa River Offspring Cohort consists of about 21,000 persons born to one or more exposed parents of this group - many of whom also lived along the contaminated river. The EURT Cohort consists of about 18,000 people who were evacuated from the EURT soon after the 1957 explosion and another 8000 who remained. These groups together are the focus of dose reconstruction and epidemiological studies funded by the US, Russia, and the European Union to address the question “Are doses delivered at low dose rates as effective in producing health effects as the same doses delivered at high dose rates?”

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗

FABRICATION AND DEPLOYMENT OF THE 9979 TYPE AF RADIOACTIVE WASTE PACKAGING FOR THE DEPARTMENT OF ENERGY

This paper summarizes the development, testing, and certification of the 9979 Type A Fissile Packaging that replaces the UN1A2 Specification Shipping Package eliminated from Department of Transportation (DOT) 49 CFR 173. The DOT Specification Package was used for many decades by the U.S. nuclear industry as a fissile waste container until its removal as an authorized container by DOT. This paper will discuss stream lining procurement of high volume radioactive material packaging manufacturing, such as the 9979, to minimize packaging production costs without sacrificing Quality Assurance. The authorized content envelope (combustible and non-combustible) as well as planned content envelope expansion will be discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up into small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and also foreign decommissioning projects can provide solutions with were unknown before. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Role of Electron and Atom Probe Tomography in Characterization of Nuclear Fuels

Dramatic advances in materials characterization capabilities over the last few years enabled microstructural and chemical 3D characterization of specimens at the atomic-scale with unprecedented quantitative precision and high data collection rates. These improvements allow for the transition of electron tomography (ET) and atom probe tomography (APT) techniques to a much broader research field and their implementation in the characterization of nuclear fuels and highly radioactive materials. Nanoscale characterization has been widely developed for non-nuclear applications, yet in the nuclear arena, particularly for fuels, these modern characterization techniques have not been applied in concert to elucidate fundamental material behavior in metal fuel. In this work, we present cross-correlated APT and ET data and provide atomic-scale representations of microstructure and chemistry of nuclear fuels, which is essential in the validation of current and future atomistic and meso-scale modeling efforts. A firm understanding of microstructural evolution in reactor materials is imperative and will lead to deeper understanding of in-reactor degradation behavior, creating a new generation of radiation-resistant materials and fuels with the potential for significant increases in burn-up. In summary, two complementary techniques were used to characterize phases formed as a result of interdiffusion between U-Pu-Zr metal fuel and cladding and fission gas bubbles produced in U-Mo fuel upon irradiation. Understanding in-reactor degradation behavior of these materials is critical to advancing nuclear energy technology. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Developing Surrogate Far-Field Nuclear Fallout and its Rapid Decontamination from Aircraft Surfaces

Naval aircraft and vehicles deployed in response to the tsunami of 2011 and subsequent nuclear meltdown at the Fukushima Daiichi nuclear power station were exposed to and contaminated with radioactive materials released to the atmosphere. In response, the Navy and Argonne National Laboratory are testing various decontamination methods to clean aircraft surfaces exposed to far-field nuclear fallout. Far-field fallout, which can travel hundreds of miles, includes dissolved radionuclides and fine particulate matter. To simulate contamination with such materials, dissolved cesium-137 ( 137 Cs) and radiolabeled particles were deposited on coupons (1 x 1 in. sections) of tire, acrylic, and painted metal surfaces. These coupons were subjected to six different decontamination methods, including ones involving four strippable products, a soap solution, and tap water. Removals of dissolved 137 Cs were generally high: over 90% of the deposited radionuclide was removed after three decontaminations. To simulate exposure to inert, radioactive fallout or other inert, airborne particulate, silica (SiO 2 ) particles were used as surrogate fine particulate matter. The silica particles were radio-labeled by first depositing europium-152 ( 152 Eu) or americium-241 ( 241 Am) on the particle surface as sesquioxides (Eu 2 O 3 or Am 2 O 3 ), then coating it with sodium silicate. Scanning electron microscopy confirmed that particle size was not altered by the treatment, while leaching studies into numerous solutions and decontamination agents confirmed that minimal amounts of the radiolabel dissolved into solution. Three different methods to apply the radioactive particles to the silica were evaluated. Pressing the coupons into a dish containing the particles resulted in particle clumping on the surface, while applying the particles within water also resulted in heterogeneous surface coverage. Suspending the particles in ethanol and then applying them to the surface supplied homogeneous surface coverage that provided acute and precise removal percentages. Finally, particle-surface interactions seemingly play an important role in radiological decontamination as the removal percentages were lower than those of soluble 137 Cs for aqueous decontamination methods. These decreased removal percentages indicate that more complex methods such as strippable decontamination agents are required to remove particulate far-field fallout from aircraft and similar vehicle surfaces.

36 MATERIALS SCIENCE↗

Review of Concrete Biodeterioration in Relation to Buried Nuclear Waste

Long-term storage of low level radioactive material in below ground concrete disposal units (DUs) (Saltstone Disposal Facility) is a means of depositing wastes generated from nuclear operations of the U.S. Department of Energy. Based on the currently modeled degradation mechanisms, possible microbial induced effects on the structural integrity of buried low level wastes must be addressed. Previous international efforts related to microbial impacts on concrete structures that house low level radioactive waste showed that microbial activity can play a significant role in the process of concrete degradation and ultimately structural deterioration.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Survey Report on Commercial Off-The-Shelf (COTS) Tracking Tools

Radioactive/nuclear (R/N) materials are integral parts of many business sectors such as medicine, industry, and research. In the United States alone, approximately 3 million packages containing R/N materials are shipped annually via highways, rail, air, and water. These materials are utilized in medical diagnostics and treatments, industrial radiography, and scientific research. The secure and efficient transportation of R/N materials is paramount to public safety and national security. Recent incidents involving the loss of R/N materials during transportation highlight significant security and safety concerns. For instance, in December 2024, a package containing four encapsulated Selenium-75 sources was reported lost in transit from the Czech Republic to Madrid, raising alarms about the security of such shipments. Similarly, a Germanium-68 pin source was reported lost in transit in New Jersey in December 2024, underscoring the risks associated with transporting radioactive materials.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Characterization of gamma-ray background outside of the High Flux Isotope Reactor

Static gamma-ray measurements collected between September 2017 and March 2018 in the vicinity of the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory are presented. Temporal events are identified and tracked, including radioactive material transfers, rain events, and 41 Ar emissions. The remaining data is used to determine the average background and its variation. Finally, these results will be used to benchmark Monte Carlo transport models of the HFIR/REDC vicinity and provide training data for material transfer tracking algorithms and data fusion algorithms to determine HFIR operational status under the Multi-Informatics for Nuclear Operations Scenarios program.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Fracture Toughness Evaluation for Sandia Mock-up Stainless Steel Canister Weldment Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials that have no current outlet than storage. These materials are often held in long-term storage as an interim stage within their lifecycle. Lifecycles can include reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. Stress corrosion cracking of interim storage containers has been identified as a high priority data gap by the Department of Energy. However, little has been done with regards to the canister weld material properties, including the fracture toughness, and their impact on stress corrosion, until recently from Sandia National Laboratory canister mockup program. Furthermore, because no post-weld heat treatment was required, the associated high tensile residual stress within these canister welds can drive the initiation and growth of stress corrosion cracking (SCC) cracks. Thus, the fracture mechanical property, such as fracture toughness, is an essential information to understand the canister weldment fracture resistance behavior and to support mitigation protocol development to counter SCC. In this work, we carried out fracture toughness evaluation on the received mock-up canister weldments from Sandia, using spiral notch torsion fracture test technology. The details of the weldment fracture toughness evaluation program development at ORNL are presented in this progress report. The objective of this research is to collect fracture toughness data on the mock-up canister weldment from Sandia National Laboratory. Data will be collected under quasi-static fracture loading using the Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support the ongoing SNF dry storage canister modeling activities and to address regulatory issues associated with canister container integrity and the follow-on SNF system transport reliability. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre-cracks are needed. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient crack length to increase the sample fracture tip localized constraint. Fatigue pre-crack growth techniques are employed to introduce sharp crack front into the sample. Previously, only rough calculations were applied to estimate the compliance evolution in the SNTT crack growth process; while accurate quantitative descriptions have never been attempted. This generates an urgent need to understand the crack evolution during the SNTT fracture testing process of ductile materials. The newly developed governing equations for SNTT crack growth estimate are used to control the fatigue crack growth to effectively reaching the target crack length. The detailed SNTT approach and its estimated fracture toughness for the as-received Sandia canister weldment and the associated baseline 304 stainless steel are presented in this report. In the past SNTT technology has also been successfully applied to investigate the fracture behavior of X52 and X80 steels and the X52 welded materials used for hydrogen infrastructures. The current study in supporting Sandia SNF Dry Storage Mock-up Canister Program is the first attempt of applying SNTT to the highly ductile stainless steel and the associated weld materials. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to SS304 steel and SS304/308 weld material. The estimate JQ upon fracture for the baseline SS304 steel is at 1618 lb/in. The estimate JQ upon fracture for the SS304/308 weld from HAZ and weld specimens are 720 lb/in and 850 lb/in, respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Technical Report for Calculations of Atmospheric Dispersion at Onsite Locations for Department of Energy Nuclear Facilities

The Department of Energy (DOE) has performed an evaluation of the technical bases for the default value for the atmospheric dispersion parameter χ/Q. This parameter appears in the calculation of radiological dose at the onsite receptor location (co-located worker at 100 meters) in safety analysis of DOE nuclear facilities. The results of the calculation are then used to determine whether safety significant engineered controls should be established to prevent and/or mitigate the event causing the release of hazardous material. An evaluation of methods for calculation of the dispersion of potential chemical releases for the purpose of estimating the chemical exposure at the co-located worker location was also performed. DOE’s evaluation consisted of: (a) a review of the regulatory basis for the default χ/Q dispersion parameter; (b) an analysis of this parameter’s sensitivity to various factors that affect the dispersion of radioactive material; and (c) performance of additional independent calculations to assess the appropriate use of the default χ/Q value.

61 RADIATION PROTECTION AND DOSIMETRY↗

Los Alamos National Laboratory 2022 Annual Site Environmental Report (Rev. 2)

Los Alamos National Laboratory (Laboratory) annual site environmental reports are prepared each year by the Laboratory’s environmental organizations as required by U.S. Department of Energy Order 231.1B, Administrative Change 1, Environment, Safety, and Health Reporting, and Order 458.1, Administrative Change 4, Radiation Protection of the Public and the Environment. The chapters in this report discuss our compliance with environmental laws, regulations, and orders (Chapter 2, Compliance Summary); how we manage the Laboratory’s environmental performance and assure the quality of data from analysis of environmental samples (Chapter 3, Environmental Programs and Analytical Data Quality); how we monitor for air emissions of radioactive materials and for weather conditions (Chapter 4, Air Quality); how we monitor for effects of Laboratory operations on groundwater quality (Chapter 5, Groundwater Protection); how we monitor the levels of chemicals and radionuclides in storm water runoff and sediment (Chapter 6, Watershed Quality); how we monitor for the levels and effects of chemicals and radionuclides in plants, animals, soil, and vegetation (Chapter 7, Ecosystem Health); and finally, what radioactive dose or risk from chemical exposure that members of the public could experience as a result of Laboratory operations (Chapter 8, Public Dose and Risk Assessment).

54 ENVIRONMENTAL SCIENCES↗

Development of a method for high LET irradiation of liquid systems

A variety of processes, from material sterilization and cancer treatment to used nuclear fuel recycling, benefit from quantifying the sensitivity of the system to radiation. Determining the effects of alpha irradiation on a system may be of complementary interest to the effects of gamma irradiation, as alpha radiation has higher linear energy transfer (LET) and will likely result in different chemical damage effects. This becomes important in advanced nuclear fuel cycle processes where the radioactive materials to be handled in solutions contain significant amounts of alpha emitters. In this work, we describe a method for studying high LET radiation in a liquid system using a TRIGA® reactor and the 10 B(n,α) 7 Li reaction. By fitting a model based on neutron diffusion and absorption to experimentally obtained Fricke dosimetry data, the high LET dose to a sample was predictable over the full range of reactor power available and varying 10 B concentration. This method may be applied to study the effects of high LET radiation on any liquid system as long as a suitable molecule containing boron is used and appropriate neutron diffusion coefficients are known. A wide range of high LET dose rates from <10 Gy/h to >1,000 kGy/h may be obtained with this method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

How to Handle the Conflict between Safety and Security - 18564

The purpose of safety and security for nuclear activities are the same. They are both necessary measures to protect people and the environment from radioactive material. There are many measures that enhance both the safety and the security but there are also measures serving one area that are in direct conflict with the other. How information is handled is maybe the best example to demonstrate this conflict between interests. In the security field, information is handled on a need-to-know basis, preventing valuable details to fall into hands of antagonists, terrorists etc. The safety culture, on the contrary, is based on knowledge and transparency and it encourages the sharing of information and experience to improve the safety. There is little guidance on how to manage the conflicts between safety and security for a transport system moving nuclear fuel and radioactive waste and especially when there are conflicts between different legislation areas, for example the Act on Nuclear Activities and Safety of Life at Sea. The Swedish Nuclear Fuel and Waste Management Co, SKB, transports all Swedish spent fuel and operational waste from the Nuclear Power Plants to the central storage facilities in Sweden. To manage the transport system, SKB owns a fleet of packages, road vehicles and an INF 3 ship. For the shipments of nuclear material, the security focus has increased rapidly over the years after the 9/11 attack, followed by regulatory requirements on enhanced security. SKB is currently running a project to further strengthen the transport security, an activity that requires a careful balance between safety and security. This paper describes the methods used and the challenges experienced trying to manage the conflicts between safety and security for the Swedish nuclear transport system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Technical Assessment of Proposed LAW Simulant for Cold Commissioning

Prior to radioactive operations, the Hanford Tank Waste Treatment and Immobilization Plant (WTP) will implement a cold-commissioning plan or strategy that will include water runs and other melter feed simulants to not only test the various engineering systems, unit operations, and processing strategies with respect to safety and performance, but to also allow facility operations to gain operational experience before introducing radioactive material into the vitrification facilities. Based on direction from Bechtel National, Inc. to eliminate nitrogen oxide (NOx) generation, EnergySolutions has recommended the use of a simple NaOH solution with glass formers to support this stage of testing during cold commissioning.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

EVIDENCE FOR ASYMMETRIC DISTRIBUTION OF CIRCUMSTELLAR MATERIAL AROUND TYPE Ia SUPERNOVAE

We study the properties of low-velocity material in the line of sight toward nearby Type Ia supernovae (SNe Ia) that have measured late phase nebular velocity shifts (v{sub neb}), thought to be an environment-independent observable. We have found that the distribution of equivalent widths of narrow blended Na I D1 and D2 and Ca II H and K absorption lines differs significantly between those SNe Ia with negative and positive v{sub neb}, with generally stronger absorption for SNe Ia with v{sub neb} {>=} 0. A similar result had been found previously for the distribution of colors of SNe Ia, which was interpreted as a dependence of the temperature of the ejecta with viewing angle. Our work suggests that (1) a significant part of these differences in color should be attributed to extinction, (2) this extinction is caused by an asymmetric distribution of circumstellar material (CSM), and (3) the CSM absorption is generally stronger on the side of the ejecta opposite to where the ignition occurs. Since it is difficult to explain (3) via any known physical processes that occur before explosion, we argue that the asymmetry of the CSM is originated after explosion by a stronger ionizing flux on the side of the ejecta where ignition occurs, probably due to a stronger shock breakout and/or more exposed radioactive material on one side of the ejecta. This result has important implications for both progenitor and explosion models.

79 ASTRONOMY AND ASTROPHYSICS↗