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DOE OSTI2015

Status of HLW disposal in the USA and rational, progressive, paths forward based upon lessons learned in the USA and abroad since 1973 - 15103

The development of the USA's only candidate (since 1987) HLW-repository at the Yucca Mountain (YM) site, Nevada, was halted by the Obama administration in 2009. Four years later, the Obama administration proposed a new strategy for the back end of the nuclear fuel cycle that would open a consent-based HLW repository 'by 2048'. At the end of 2014, both the YM HLW repository and the new HLW-disposition (storage and disposal) strategy remain on hold due to lack of enabling legislation. With this as background, and based upon our active involvement in, monitoring of, and reporting on HLW management programs in the USA and abroad since 1973, we submit that: 1. Politics, political will, and legal actions and outcomes will continue to govern progress. 2. Public acceptance and sustained support (PASS) may influence politics and political will. 3. PASS is time-dependent and governed by: a. Ideologies; b. Perceived benefits and risks; c. Level and equity of radiation protection provided by applicable regulations; d. Perceived robustness/safety of the proposed disposition concept; and e. Perceived credibility and competency vested in the involved organizations and their respective leaders (= TRUST). We therefore recommend that the USA concurrently pursues at least two, 'centralized' HLW-storage and -disposal solutions based upon best-available relevant knowledge and majority-PASS at facility-host locations. Furthermore, safe(r) and more secure disposal of defense/government-generated HLW could probably commence at least 15 years earlier than 'by 2048' in/at: 1. The YM repository/site. 2. The WIPP repository/site. 3. 'Local' deep boreholes. (authors)

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DOE OSTI2018

Determining Unabated Airborne Radionuclide Emissions Monitoring Requirements Using Inventory-Based Methods

Compliance with the airborne radionuclide emission monitoring requirements in the National Emission Standards for Hazardous Air Pollutants (NESHAP; Title 40 of the U.S. Code of Federal Regulations Part 61, Subpart H) and State requirements in Washington Administrative Code 246–247: Radiation Protection – Air Emissions and 173-480: Ambient Air Quality Standards and Emission Limits for Radionuclides were evaluated for Pacific Northwest National Laboratory (PNNL) operations. Additional guidance may be found in the U.S. Department of Energy Handbook, Environmental Radiological Effluent Monitoring and Environmental Surveillance. To meet regulatory requirements, reviews of planned or proposed projects and activities provide the basis for implementing necessary monitoring adjustments or for implementing changes to projects and activities in a timely manner. Potential unabated offsite doses were evaluated for emission locations managed by PNNL and licensed to the Department of Energy. These locations were at facilities in Richland, Washington (i.e., the Hanford Site and PNNL Richland Campus) and in Sequim, Washington, (PNNL Marine Sciences Laboratory). This report describes the inventory-based methods and provides the results for the NESHAP assessment performed in 2017.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2015

On effective dose for radiotherapy based on doses to nontarget organs and tissues

Purpose: The National Council for Radiation Protection and Measurement (NCRP) published estimates for the collective population dose and the mean effective dose to the population of the United States from medical imaging procedures for 1980/1982 and for 2006. The earlier report ignored the effective dose from radiotherapy and the latter gave a cursory discussion of the topic but again did not include it in the population exposure for various reasons. This paper explains the methodology used to calculate the effective dose in due to radiotherapy procedures in the latter NCRP report and revises the values based on more detailed modeling. Methods: This study calculated the dose to nontarget organs from radiotherapy for reference populations using CT images and published peripheral dose data. Results: Using International Commission on Radiological Protection (ICRP) 60 weighting factors, the total effective dose to nontarget organs in radiotherapy patients is estimated as 298 ± 194 mSv per patient, while the U.S. population effective dose is 0.939 ± 0.610 mSv per person, with a collective dose of 283 000 ± 184 000 person Sv per year. Using ICRP 103 weighting factors, the effective dose is 281 ± 183 mSv per patient, 0.887 ± 0.577 mSv per person in the U.S., and 268 000 ± 174 000 person Sv per year. The uncertainty in the calculations is largely governed by variations in patient size, which was accounted for by considering a range of patient sizes and taking the average treatment site to nontarget organ distance. Conclusions: The methods used to estimate the effective doses from radiotherapy used in NCRP Report No. 160 have been explained and the values updated.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2017

Risks of Hazardous Substances in Radioactive Waste for the Konrad Repository in Germany - 17522

When disposing of radioactive waste with negligible heat generation, the activity of the radionuclides is limited. However, the waste including radioactive material comprises a certain proportion of non-radioactive material. The risk of hazardous contamination of the groundwater by non-radioactive substances must be taken into account. It must be ascertained that the non-radioactive material cannot affect near-surface groundwater, which is a resource for drinking water supply. To achieve this, limit values on the amount of hazardous substances in the total waste volume shall be determined. In 2002 a regulatory approval was granted to the 'Konrad' final storage facility for radioactive waste with negligible heat generation according to the German Water Resources Act. This approval is embedded in the license (plan-approval decision) based on the German Atomic Energy Act. The approval is based on the model of a possible discharge of contaminants from the final storage facility. In a long-term simulation spanning 300,000 years, contact of the waste with the groundwater in the surrounding rock formations over the postoperational phase was conservatively assumed. The maximum amount of allowable disposal of substances - such that no concern of affecting the near-surface groundwater arises - can be derived from the solubility of the respective substances. Implementing the permit according to the Water Resources Act requires: - classification of the waste through delivery declaration obligations - monitoring the substance inspection in the waste to assess its suitability for the disposal, and - stock keeping procedures for the disposal waste by the operator. While the repository is operated by the Federal Office for Radiation Protection (BfS) (http://www.bfs.de./EN/home/home{sub n}ode;), reporting to the Federal Environment Ministry, the statutory water permit is issued and applied by the State of Lower Saxony as responsible state authority. A waste package/container list supplements a comprehensive list of substances. From this - depending on the respectively conditioned waste - material vectors can be derived. For this purpose, BfS provides a database with a substances and container list to the waste owners. In the meantime, first experiences have been collected in applying the approval. More than 540 waste containers with a total mass of almost 6,000 Mg have been controlled and are ready for disposal. In our presentation we will show the recent development in supervising the preparation for the disposal of mixed waste. (authors)

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DOE OSTI2015

Outdoor characterization of radio frequency electromagnetic fields in a Spanish birth cohort

There is considerable public concern in many countries about the possible adverse effects of exposure to non-ionizing radiation electromagnetic fields, especially in vulnerable populations such as children. The aim of this study was to characterize environmental exposure profiles within the frequency range 100 kHz–6 GHz in the immediate surrounds of the dwellings of 123 families from the INMA-Granada birth cohort in Southern Spain, using spot measurements. The arithmetic mean root mean-square electric field (E{sub RMS}) and power density (S{sub RMS}) values were, respectively, 195.79 mV/m (42.3% of data were above this mean) and 799.01 µW/m{sup 2} (30% of values were above this mean); median values were 148.80 mV/m and 285.94 µW/m{sup 2}, respectively. Exposure levels below the quantification limit were assigned a value of 0.01 V/m. Incident field strength levels varied widely among different areas or towns/villages, demonstrating spatial variability in the distribution of exposure values related to the surface area population size and also among seasons. Although recorded values were well below International Commission for Non-Ionizing Radiation Protection reference levels, there is a particular need to characterize incident field strength levels in vulnerable populations (e.g., children) because of their chronic and ever-increasing exposure. The effects of incident field strength have not been fully elucidated; however, it may be appropriate to apply the precautionary principle in order to reduce exposure in susceptible groups. - Highlights: • Spot measurements were performed in the immediate surrounds of children's dwellings. • Mean root mean-square electric field and power density values were calculated. • Most recorded values were far below international standard guideline limits. • Data demonstrate spatial variability in the distribution of exposure levels. • While adverse effects are proven, application of the precautionary principle may be appropriate.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2018

Long-term effects of ionizing radiation after the Chernobyl accident: Possible contribution of historic dose

Highlights: • Paper considers the role of delayed non-targeted effects (DNTE) in environmental radiation protection (ERP). • Comparison of the DNTE dose response relationship (DRR) in Chernobyl birds 9–11 generations after accident. • Dose reconstruction suggests the pattern of DNTE induced by the historical dose is similar datasets. • Conclude the historic dose contribution is important to consider. The impact of the Chernobyl NPP accident on the environment is documented to be greater than expected, with higher mutation rates than expected at the current, chronic low dose rate. In this paper we suggest that the historic acute exposure and resulting non-targeted effects (NTE) such as delayed mutations and genomic instability could account at least in part for currently measured mutation rates and provide an initial test of this concept. Data from Møller and Mousseau on the phenotypic mutation rates of Chernobyl birds 9–11 generations post the Chernobyl accident were used and the reconstructed dose response for mutations was compared with delayed reproductive death dose responses (as a measure of genomic instability) in cell cultures exposed to a similar range of doses. The dose to birds present during the Chernobyl NPP accident was reconstructed through the external pathway due to Cs-137 with an estimate of the uncertainty associated with such reconstruction. The percentage of Chernobyl birds several generations after the accident without mutations followed the general shape of the clonogenic survival percentage of the progeny of irradiated cells, and it plateaued at low doses. This is the expected result if NTE of radiation are involved. We suggest therefore, that NTE induced by the historic dose may play a role in generating mutations in progeny many generations following the initial disaster.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2017

Preliminary Development of a Nuclide Inventory Code for HTGR

Nuclide inventory calculation for a reactor core is the basis of radiation protection, shielding design and radioactive environmental impact assessment of the power plant. HTR-PM (high-temperature gas-cooled reactor pebble-bed module) under construction in China is one of the candidates for Gen-IV advanced nuclear power system.1 In the design stage of HTR-PM, KORIGEN has been utilized in the inventory calculation.2 However, the current version of KORIGEN was developed several decades ago. In order to improve the design capability of the analysis code package for HTR-PM, a new computer code for nuclide inventory calculation called NUIT (Nuclide Inventory Tool) is being developed based on the DEPTH3 code. The NUIT code tries to utilize accurate and efficient burnup solvers, incorporate well-validated and most recent nuclear data libraries, and provide user-friendly IO interface. This paper presents the preliminary work in the development of NUIT.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRSDate not supplied

Development of the Advanced Regolith Ground Operations (ARGO) Test Bed: A Robotic Excavation and Construction Test Facility with Simulated Lunar Environments

NASA’s Artemis Program is working towards developing a sustained presence on the Moon and eventually Mars. To achieve this goal, robotic excavation, site preparation and construction technologies are under development to establish the capability to construct infrastructure such as launch/landing pads and radiation protection shelters. Technologies must be proven in simulated Lunar conditions prior to surface demonstration missions. To that end, the Relevant Additive Construction Technology (REACT) Announcement of Collaboration Opportunities (ACO) project with AI Space Factory and the NASA Kennedy Space Center’s (KSC) Granular Mechanics and Regolith Operations Laboratory (a.k.a. Swamp Works) has developed the Advanced Regolith Ground Operations (ARGO) Test Bed. ARGO includes a ~1.5m x 1.5m x 1.2m (~5x5x4ft) vacuum chamber, cryogenically cooled thermal shroud, 3-axis robotic positioning system, and regolith bin. For the REACT project, a pellet extruder, feed hopper, and heated 600mm x 600mm (23.6x23.6in) build plate have been installed on ARGO to advance the Technology Readiness Level (TRL) of regolith-polymer composite Fused Deposition Modeling (FDM) additive construction systems, processes, and materials. This paper will focus on the design and operational characteristics of the ARGO Test Bed with pellet extruder.

regolith↗

NASA NTRSDate not supplied

Development of the Advanced Regolith Ground Operations (ARGO) Test Bed – A Robotic Excavation and Construction Test Facility with Simulated Lunar Environments

regolith↗

DOE OSTI2015

Encapsulation of ILW metals by melting technology - 15360

This paper outlines how encapsulation using melting technology can be performed and describe the positive effects in a handling and repository perspective. Based on a feasibility study of a conditioning method for reactor core components and other ILW metals with a high content of long-lived radionuclides, the paper gives an overview of the concept and how the method can be facilitated. The paper covers identification and analysis of alternative techniques that supports the aim from a repository perspective, which is to strengthen the defense in depth for long-lived waste to be disposed. Embedding of radioactive metallic ILW components can isolate the activated component from corrosion and leakage of radionuclides into the repository barrier systems. By implementing a system that delays the mobility of nuclides the leakage to the biosphere can be delayed and by then decreased significantly. Conditioning by embedding has potential to strengthen a repository concept significantly. It may in certain cases also reduce the requirements on the repository barriers. In addition to the above, embedding also serves as an efficient radiation protection, which simplifies post treatment handling of the waste. The concept can be applied on a wide spectrum of metallic components. Positive effects in a handling and repository perspective can be achieved if low level metal not subject to free release is used as embedding material, since conditioning of metallic ILW waste can be carried out at significant lower dose rates and without increasing the total amount to be disposed. The paper discuss why there are embedding alternatives for ILW metals that are considered to be both technically feasible and reasonable in cost both with regards to investment and operational cost. A cost-benefit analysis as well as a risk analysis is discussed in order to support the feasibility assessment of a facility for encapsulation of ILW metals by melting technology. (authors)

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DOE OSTI2025

Archival records housed at USTUR support radium dial worker dosimetry

The American radium dial worker (RDW) cohort of over 3200 persons is being revisited as part of the Million Person Study (MPS) to include a modern approach to RDW dosimetry. An exceptional source of data and contextualization in this project is an extensive collection of electronic records (digitized from existing microfilm and microfiche) housed at the United States Transuranium and Uranium Registries (USTUR). Although the type, extent, and quality (e.g. legibility) of record(s) varies between individuals, the remarkable occupational, medical and demographic data include in vivo radiation measurements (e.g. radon breath, whole body counts), autopsy results, medical records (including copies of radiographs), interviews over the years, and correspondence. Of particular dosimetric interest are the details of radiation measurements. For example, there are some instances where hand-written and transcribed values are both available, along with notes providing context for why a particular measurement in a series of measurements was chosen to assign an intake, or if there were concerns about a particular measurement. Born prior to 1935, RDW have nearly all passed away. Thus, the updated dosimetry, especially for the skeletal tissues, will allow the correlation of lifetime cumulative dose with radiation risk. Here we review typical information available in this collection of historical records and highlight some interesting finds. Additionally, we discuss the relevance to current and ongoing work related to updating the dosimetry of the RDW in the MPS, including providing an example of the usefulness of information contained in these records. The RDW cohort provides a unique historical perspective on occupational exposure to radium, making it a valuable dataset for understanding long-term health effects and improving current radiation protection standards.

Million Person Study↗

DOE OSTI2015

Evaluation of the response of tritium-in-air instrumentation to HT in dry and humid conditions and to HTO vapor

Nuclear plant operators (power generation, decommissioning and reprocessing operations) are required to monitor releases of tritium species for regulatory compliance and radiation protection purposes. Tritium monitoring is performed using tritium-in-air gas monitoring instrumentation based either on flow-through ion chambers or proportional counting systems. Tritium-in-air monitors are typically calibrated in dry conditions but in service may operate at elevated levels of relative humidity. The NPL (National Physical Laboratory) radioactive gas-in-air calibration system has been used to study the effect of humidity on the response to tritium of two tritium-in-air ion chamber based monitors and one proportional counting system which uses a P10/air gas mixture. The response of these instruments to HTO vapour has also been evaluated. In each case, instrument responses were obtained for HT in dry conditions (relative humidity (RH) about 2%), HT in 45% RH, and finally HTO at 45% RH. Instrumentation response to HT in humid conditions has been found to slightly exceed that in dry conditions. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

A Dual Alpha/Gamma Imaging Tool to Manage Decommissioning Operations

The knowledge of the radiological state of a nuclear plant under decommissioning is a key factor at any stage of the project: from the initial characterization (to define the best scenario and the main dismantling steps), during the monitoring of the decontamination operations, for the waste management and, last but not least, for the final controls. During a facility's operation, a clear view of the process radiological level is equally necessary, in order to plan maintenance scheduling and optimize interventions by personnel. Radiation protection teams generally ensure worker radiological safety, and also supply dose mappings for each of the areas associated with typical spectra. In most cases, this mapping does not give enough information to be used as input data when preparing maintenance or clean-up work. With advances in activity modeling calculation codes (using 3D), the way measurements are made to recover the best input radiological data on site is a key factor. Nuclear instrumentation is still in progress and the evolution of the technologies and software allow us to get imaging information in real time. The gamma camera is going to evolve and to become more compact and more sensitive. New portable gamma cameras have to be tested under high level dose rate. The alpha camera is also evolving: from the first prototype tested on site to solar blind alpha camera (to reduce the influence of light) to the new dual alpha/gamma prototype. (authors)

42 ENGINEERING↗

DOE OSTI2019

A Comprehensive Radiological Characterization of the ACRR Facility

Radiation Protection (628) in conjunction with Nuclear Facility Operations (1381) completed a comprehensive radiological conditions characterization of the ACRR facility between August 2015 and January 2017 to better understand the radiological environment in and around the ACRR High-Bay. The overall goals of the characterization, as identified by 1381 and 628, were to determine potential dose to workers during routine reactor operations, areas of potential elevated dose in and around the ACRR High-Bay, the relationship between reactor power and dose during steady-state operations, and the dose per pulse for pulsed operations. To accomplish this, eight configurations were identified of interest for characterization based on increased dose potential for which field surveys were completed to include assessment of the neutron spectra for each, determination of neutron dose conversion factors (Neutron Codes) based on defined neutron spectra, mapped dose surveys for each configuration, as well as determination of the relationship between dose and reactor power level during steady-state reactor operations. Steady-state survey data was also used to calculate dose per pulse approximations during pulsed reactor operations.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

Experience with the Quality Control of Radioactive Waste Packages for the German Repository Konrad - 16493

The German repository Konrad for radioactive waste with negligible heat generation is designed to accept 303,000 m{sup 3} of waste packages in total. Based on the Konrad licensing conditions, any treatment and conditioning of radioactive waste should be planned and carried out in accordance with the waste acceptance requirements of the repository Konrad in order to produce waste packages suitable for disposal. The qualification of the waste treatment processes and the resulting products are achieved by interaction of different parties involved. On behalf of the Federal Office for Radiation Protection - the competent German authority for disposal of radioactive waste - TUV NORD EnSys Hannover GmbH and Co. KG acts as one of the independent expert organizations for more than twenty years in the field of qualification of waste treatment campaigns and review of waste packages documentations. The permitted amount of nuclide activity in the radioactive waste packages depends on the characteristics of the radioactive waste products. Projecting a suitable strategy for loading the waste containers contributes to determine in advance, what requirements the radioactive waste products have to fulfill as well as what properties of the waste products need to be verified and documented during conditioning or treatment process. On the basis of various examples, it is being demonstrated how appropriate advance process planning can ensure that radioactive waste packages are generated to comply with the waste acceptance requirements. In consequence, technical measures due to requalification can be avoided even in case of unpredictable difficulties. (authors)

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DOE OSTI2017

Activation Calculation for the Dismantling and Decommissioning of a Light Water Reactor Using MCNP{sup TM} with ADVANTG and ORIGEN-S

The boiling water reactor (BWR) Philippsburg 1 (KKP1) and the pressurized water reactor (PWR) Neckarwestheim 1 (GKN I), with a thermal power of 2575 MW{sub th} resp. 2497 MW{sub th}, were shut down and entered directly the post-operational phase which is covered by the operating license held by the nuclear power plant owner. The decommissioning of a nuclear power plant (NPP) which is licensed under chap. 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning requires a decommissioning license. This will include essential changes in requirements for the systems and components and will mainly lead to the immediate dismantling of the nuclear power plant which is now the only permitted decommissioning strategy in Germany. In this context, neutron activation calculations for the structural components of KKP1 and GKN I have to be carried out to predict occurring activities and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP6) combined with automated variance reduction tools (ADAVANTG) are used and are coupled with the activation and decay module ORIGEN-S. This paper presents the calculation procedure and the obtained results with a validation by measured activations and gamma dose rates.

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DOE OSTI2015

The conduct of the united states department of energy nuclear criticality safety program hands-on training and education courses - 14621

The United States Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP), funded and managed by the National Nuclear Security Administration (NNSA), offers courses to educate nuclear criticality safety (NCS) engineers, process supervisors, and others, and to provide experimental, hands-on training addressing important characteristics of neutron multiplying systems. This training satisfies the American National Standard Institute/American Nuclear Society (ANSI/ANS)-8.26-2007 requirement for participating in hands-on experiments with fissionable materials and provides awareness and understanding of DOE Orders, Guides and Rules, ANS standards, and hazard identification techniques, as well as control design, implementation, and maintenance in criticality safety evaluations. Since the pilot course was developed in 2011, a variety of courses have been developed for NCS practitioners and process supervisors. Attendees have come from several US and international sites. Through NCSP international collaborations and relationships, students from the Atomic Weapons Establishment (AWE) in the United Kingdom, Atomic Energy of Canada, and the Institute for Radiation Protection and Nuclear Safety in France have participated in these courses in the US. Manager courses are available for process supervisors and operations managers with NCS responsibilities in their daily jobs. Manager courses are taught at Sandia National Laboratory (SNL) and the National Criticality Experiments Research Center (NCERC); they offer a combination of hands-on experiments and classroom discussions. A two-week course is also offered for NCS practitioners, with a week of classroom instruction at Los Alamos National Laboratory (LANL) focused on generation of NCS evaluations, and a week of hands-on training utilizing the critical assemblies at either SNL or NCERC. Cleared students can attend the hands-on training at either SNL or NCERC, and uncleared students can attend at SNL. Courses continue to evolve and continuously improve as important elements of NCS evaluations (e.g., human factors, non-destructive assay (NDA) techniques, process criticality accident lessons-learned, and hand calculation methods) are integrated with criticality safety evaluation exercises. Courses include tours of process operations at SNL, NCERC, and LANL, and interaction with instructors experienced in conducting critical experiments and working in NCS. The NNSA NCSP provides management, funding, and technical direction for these courses and makes them available with no tuition cost to NCS practitioners. This paper provides a status report to the NCS community on the conduct of these courses and how they will likely evolve in the future. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Pressurizing Hagan and SAVY containers to 30-psig (air) to measure the release of analytical cerium oxide test powder

In response to an ESS surrounding the storage of sealed sources at TA-55, personnel from RP-SVS (Radiation Protection Services) and ORI-2 (Operational Readiness & Execution) were approached by ES-55 (Facility System Engineering) to provide technical testing of known nuclear material storage containers. Testing criteria were determined by SB-PF (Safety Basis for Plutonium Facilities). Tests were conducted in conjunction with multiple related projects at the Aerosol Sciences Laboratory (TA-03-0130-0103). An experimental system was quickly developed to deliver a pressure pulse (30-psig) that mimics a sealed source burst scenario. A series of twelve tests was conducted. Six tests were done with two different 5QT SAVY-4000 containers, where each SAVY filter-lid combination was subjected to three successive test insults. Three tests were done with a (0.375" diameter filter) 8Q.T Hagan container, and three tests with a (0.625" diameter filter) 8QT Hagan container. An unused, fresh container filter was used for each tested Hagan container, each receiving only one test insult per filter. Cerium oxide (CeO 2 ) powder was loaded (100 grams per test) into a nozzle in the tested Hagan and SAVY containers, and the nozzle was hard-plumbed to a ball valve and a pressure source. This system was installed into the Los Alamos RRFMC (Respirable Release Measurement Chamber), which is an integrated multipurpose aerosol wind tunnel that satisfies NQA-1 subpart 2.4 for R&D work. The ball valve was fitted with a mechanical linkage for operation from outside the wind tunnel. An aerodynamic particle sizer counted the particle concentrations and size distributions of released aerosol. Respirable aerosol released during the tests was measured and a correction factor for wind tunnel flowrate and internal duct deposition was applied.

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