Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Iterative”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Current generation by helicons and lower hybrid waves in modern tokamaks and reactors ITER and DEMO. Scenarios, modeling and antennae

The innovative concept and 3D full-wave code modeling the off-axis current drive by radio-frequency (RF) waves in large-scale tokamaks, ITER and DEMO, for steady-state operation with high efficiency is proposed. The scheme uses the helicon radiation (fast magnetosonic waves at high (20-40) ion cyclotron frequency harmonics) at frequencies of 500-700 MHz propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by helicons, in conjunction with the bootstrap current, ensure the maintenance of a given value of the total current in the stability margin q(0) {>=} 2 and q(a) {>=} 4, and will help to have regimes with a negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure {beta}{sub N} > 3 (the so-called advanced scenarios) of interest for the commercial reactor. Modeling with full-wave three-dimensional codes PSTELION and STELEC showed flexible control of the current profile in the reactor plasmas of ITER and DEMO, using multiple frequencies, the positions of the antennae and toroidal wave slow down. Also presented are the results of simulations of current generation by helicons in the DIII-D, T-15MD, and JT-60AS tokamaks. Commercially available continuous-wave klystrons of the MW/tube range are promising for commercial stationary fusion reactors. The compact antennae of the waveguide type are proposed, and an example of a possible RF system for today's tokamaks is given. The advantages of the scheme (partially tested at lower frequencies in tokamaks) are a significant decline in the role of parametric instabilities in the plasma periphery, the use of electrically strong resonator-waveguide type antennae, and substantially greater antenna-plasma coupling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Linear MHD stability analysis of post-disruption plasmas in ITER

Most of the plasma current can be replaced by a runaway electron (RE) current during plasma disruptions in ITER. In this case the post-disruption plasma current profile is likely to be more peaked than the pre-disruption profile. The MHD activity of such plasma will affect the runaway electron generation and confinement and the dynamics of the plasma position evolution (Vertical Displacement Event), limiting the timeframe for runaway electrons and disruption mitigation. In the present paper, we evaluate the influence of the possible RE seed current parameters on the onset of the MHD instabilities. By varying the RE seed current profile, we search for subsequent plasma evolutions with the highest and the lowest MHD activity. This information can be applied to a development of desirable ITER disruption scenario.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Influence of plasma pedestal profiles on access to ELM-free regimes in ITER

The influence of current density and pressure gradient profiles in the pedestal on the access to the regimes free from edge localized modes (ELMs) like quiescent H-mode in ITER is investigated. Using the simulator of MHD modes localized near plasma boundary based on the KINX code, calculations of the ELM stability were performed for the ITER plasma in scenarios 2 and 4 under variations of density and temperature profiles with the self-consistent bootstrap current in the pedestal. Low pressure gradient values at the separatrix, the same position of the density and temperature pedestals and high poloidal beta values facilitate reaching high current density in the pedestal and a potential transition into the regime with saturated large scale kink modes. New version of the localized MHD mode simulator allows one to compute the growth rates of ideal peeling-ballooning modes with different toroidal mode numbers and to determine the stability region taking into account diamagnetic stabilization. The edge stability diagrams computations and sensitivity studies of the stability limits to the value of diamagnetic frequency show that diamagnetic stabilization of the modes with high toroidal mode numbers can help to access the quiescent H-mode even with high plasma density but only with low pressure gradient values at the separatrix. The limiting pressure at the top of the pedestal increases for higher plasma density. With flat density profile the access to the quiescent H-mode is closed even with diamagnetic stabilization taken into account, while toroidal mode numbers of the most unstable peeling-ballooning mode decrease from n = 10−40 to n = 3−20.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Iterative algorithms for processing experimental data

The need to solve linear and nonlinear integral equations arise, e.g., in recovering plasma parameters from the data of multichannel diagnostics. The paper presents an iterative method for solving integral equations with a singularity at the upper limit of integration. The method consists in constructing successive approximations and calculating the integral by quadrature formulas in each integration interval. An example of application of the iterative algorithm to numerically solve an integral equation similar to those arising in recovering the plasma density profile from reflectometry data is presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Alfvén eigenmode evolution computed with the VENUS and KINX codes for the ITER baseline scenario

A new application of the VENUS code is described, which computes alpha particle orbits in the perturbed electromagnetic fields and its resonant interaction with the toroidal Alfvén eigenmodes (TAEs) for the ITER device. The ITER baseline scenario with Q = 10 and the plasma toroidal current of 15 MA is considered as the most important and relevant for the International Tokamak Physics Activity group on energetic particles (ITPA-EP). For this scenario, typical unstable TAE-modes with the toroidal index n = 20 have been predicted that are localized in the plasma core near the surface with safety factor q = 1. The spatial structure of ballooning and antiballooning modes has been computed with the ideal MHD code KINX. The linear growth rates and the saturation levels taking into account the damping effects and the different mode frequencies have been calculated with the VENUS code for both ballooning and antiballooning TAE-modes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Machine learning model inputs, outputs, and scripts associated with “Artificial intelligence-guided iterations between observations and modeling significantly improve environmental predictions”

NOTE: The manuscript associated with this data package is currently in review. The data may be revised based on reviewer feedback. Upon manuscript acceptance, this data package will be updated with the final dataset and additional metadata. This data package is associated with the manuscript “Artificial intelligence-guided iterations between observations and modeling significantly improve environmental predictions” (Malhotra et al., in prep). This effort was designed following ICON (integrated, coordinated, open, and networked) principles to facilitate a model-experiment (ModEx) iteration approach, leveraging crowdsourced sampling across the contiguous United States (CONUS). New machine learning models were created every month to guide sampling locations. Data from the resulting samples were used to test and rebuild the machine learning models for the next round of sampling guidance. Associated sediment and water geochemistry and in situ sensor data can be found at https://data.ess-dive.lbl.gov/datasets/doi:10.15485/1923689, https://data.ess-dive.lbl.gov/datasets/doi:10.15485/1729719, and https://data.ess-dive.lbl.gov/datasets/doi:10.15485/1603775. This data package is associated with two GitHub repositories found at https://github.com/parallelworks/dynamic-learning-rivers and https://github.com/WHONDRS-Hub/ICON-ModEx_Open_Manuscript. In addition to this readme, this data package also includes two file-level metadata (FLMD) files that describes each file and two data dictionaries (DD) that describe all column/row headers and variable definitions. This data package consists of two main folders (1) dynamic-learning-rivers and (2) ICON-ModEx_Open_Manuscript which contain snapshots of the associated GitHub repositories. The input data, output data, and machine learning models used to guide sampling locations are within dynamic-learning-rivers. The folder is organized into five top-level directories: (1) “input_data” holds the training data for the ML models; (2) “ml_models” holds machine learning (ML) models trained on the data in “input_data”; (3) “examples” contains files for direct experimentation with the machine learning model, including scripts for setting up “hindcast” run; (4) “scripts” contains data preprocessing and postprocessing scripts and intermediate results specific to this data set that bookend the ML workflow; and (5) “output_data” holds the overall results of the ML model on that branch. Each trained ML model resides on its own branch in the repository; this means that inputs and outputs can be different branch-to-branch. There is also one hidden directory “.github/workflows”. This hidden directory contains information for how to run the ML workflow as an end-to-end automated GitHub Action but it is not needed for reusing the ML models archived here. Please see the top-level README.md in the GitHub repository for more details on the automation. The scripts and data used to create figures in the manuscript are within ICON-ModEx_Open_Manuscript. The folder is organized into four folders which contain the scripts, data, and pdf for each figure. Within the “fig-model-score-evolution” folder, there is a folder called “intermediate_branch_data” which contains some intermediate files pulled from dynamic-learning-rivers and reorganized to easily integrate into the workflows. NOTE: THIS FOLDER INCLUDES THE FILES AT THE POINT OF PAPER SUBMISSION. IT WILL BE UPDATED ONCE THE PAPER IS ACCEPTED WITH ANY REVISIONS AND WILL INCLUDE A DD/FLMD AT THAT POINT. We thank the United States Forest Service, Washington Department of Fish and Wildlife, Washington Department of Natural Resources, Cowiche Canyon Conservatory, Washington State Parks and Recreation Commission (Scientific Research Permit #210901), and the Confederated Tribes and Bands of the Yakama Nation for access to field locations where the samples labeled “SSS” were collected. We also thank the Yakama Nation Tribal Council and Yakama Nation Fisheries for working with us to facilitate sample collection and optimization of data usage according to their values and worldview. WHONDRS consortium members were asked to provide any acknowledgments for the collection of samples labeled “CM” and the following is a list of acknowledgments that were submitted with their corresponding Site IDs: (MART) Research activities were conducted in part on the Wind River Experimental Forest within the Gifford Pinchot National Forest; (MP- 100379) Philadelphia is part of Lenapehoking, the ancestral homelands of the Lenape peoples; (MP-102398) Land surveyed is the ancestral homelands of the Nookhose'iinenno (Arapaho), Tsis tsis'tas (Cheyenne), and Nuuchu (Ute); (MP-100749 and MP- 100747) Georgia Coastal Ecosystem LTER, OCE-1832178; (SP-70 and SP-72) Eastern Shoshone, Shoshone-Bannock; (MP- 102944) Funded by Oregon Watershed Enhancement Board. On the traditional lands of the Confederated Tribes of the Siletz, Confederated Tribes of the Grand Rhonde, and the Clatsop-Nehalem Confederated Tribe; (MP- 100607) Holiday Creek is located on the traditional territory of the Monacan Indian Nation; (SP-45) Lafayette Blue Springs State Park; (MP-102420) NSF DEB-2016749; (MP-100019) New Hampshire Agriculture Experiment Station; (SP-35) Rayonier (land owner; https://www.rayonier.com/); (MP- 101276) US Department of Energy, Office of Science, Biological and Environmental Research, Subsurface Biogeochemical Research, Watershed Dynamics and Evolution SFA at ORNL; (MP- 103224) Watershed Dynamics and Evolution SFA at ORNL; (MP- 101584) Traditional lands of the Oceti Sakowin (Dakota, Lakota, Nakoda) and Anishinaabe Peoples.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Electromagnetic Analysis For The Design Of ITER Diagnostic Port Plugs During Plasma Disruptions

ITER diagnostic port plugs perform many functions including structural support of diagnostic systems under high electromagnetic loads while allowing for diagnostic access to plasma. The design of diagnotic equatorial port plugs (EPP) are largely driven by electromagnetic loads and associate response of EPP structure during plasma disruptions and VDEs. This paper summarizes results of transient electromagnetic analysis using Opera 3d in support of the design activities for ITER diagnostic EPP. A complete distribution of disruption loads on the Diagnostic First Walls (DFWs). Diagnostic Shield Modules (DSMs) and the EPP structure, as well as impact on the system design integration due to electrical contact among various EPP structural components are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

1.5 MW RF Load for ITER

Calabazas Creek Research, Inc. developed a 1.5 MW RF load for the ITER fusion research facility currently under construction in France. This program leveraged technology developed in two previous SBIR programs that successfully developed high power RF loads for fusion research applications. This program specifically focused on modifications required by revised technical performance, materials, and assembly specification for ITER. This program implemented an innovative approach to actively distribute the RF power inside the load to avoid excessive heating or arcing associated with constructive interference. The new design implemented materials and assembly changes required to meet specifications. Critical components were built and successfully tested during the program.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

US ITER Moving Forward

US ITER Project Manager Ned Sauthoff, joined by Wayne Reiersen, Team Leader Magnet Systems, and Jan Berry, Team Leader Tokamak Cooling System, discuss the U.S.'s role in the ITER international collaboration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Pre-Commercial Demonstration of Direct Non-Iterative State Estimator (DNSE+)

The objective of this DOE co-funded project is to demonstrate functionality and performance of a production-grade Direct Non-iterative State Estimator (DNSE) through its integration with NYPA’s Energy Management System (EMS) and an enhanced Real Time Dynamics Monitoring System (RTDMS) synchrophasor platform from Electric Power Group (EPG). A production-grade DNSE aims to overcome a major obstacle to the operational use of synchrophasor data in Synchro-Phasor Management Systems (SPMS) by providing SPMS applications with a consistent and a complete synchrophasor data foundation in the same way that a traditional EMS State Estimator (SE) provides to EMS applications. Specifically, the DNSE uses synchrophasor measurements and Supervisory Control and Data Acquisition (SCADA) measurements, together with the complete power system model, to simultaneously obtain the complete state of the entire utility’s operating model at rates that are close to the synchrophasor data rates. The full system coverage property is what distinguishes the DNSE from the linear synchrophasor SEs (which cover only parts of the operating model that are visible through synchrophasor measurements), and DNSE differs from traditional EMS SE in that the solution is obtained without iterations, without initial values (hence the “direct” in the name), and at a much higher execution rate. This presentation provides an overview of the project objectives, a brief description of the main tasks including the development work to productize and optimize DNSE from its current MATLAB code base, RTDMS enhancement for interfacing with the DNSE, factory acceptance testing, field installation, testing and demonstration, as well as the current status of the project execution.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2016

ITER Port Interspace Pressure Calculations

The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

Interspace ITER Vacuum vessel↗
DOE OSTI2013

MICADO: Parallel implementation of a 2D-1D iterative algorithm for the 3D neutron transport problem in prismatic geometries

The large increase in computing power over the past few years now makes it possible to consider developing 3D full-core heterogeneous deterministic neutron transport solvers for reference calculations. Among all approaches presented in the literature, the method first introduced in [1] seems very promising. It consists in iterating over resolutions of 2D and ID MOC problems by taking advantage of prismatic geometries without introducing approximations of a low order operator such as diffusion. However, before developing a solver with all industrial options at EDF, several points needed to be clarified. In this work, we first prove the convergence of this iterative process, under some assumptions. We then present our high-performance, parallel implementation of this algorithm in the MICADO solver. Benchmarking the solver against the Takeda case shows that the 2D-1D coupling algorithm does not seem to affect the spatial convergence order of the MOC solver. As for performance issues, our study shows that even though the data distribution is suited to the 2D solver part, the efficiency of the ID part is sufficient to ensure a good parallel efficiency of the global algorithm. After this study, the main remaining difficulty implementation-wise is about the memory requirement of a vector used for initialization. An efficient acceleration operator will also need to be developed. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

A method for reducing the largest relative errors in Monte Carlo iterated-fission-source calculations

In Monte Carlo iterated-fission-source calculations relative uncertainties on local tallies tend to be larger in lower-power regions and smaller in higher-power regions. Reducing the largest uncertainties to an acceptable level simply by running a larger number of neutron histories is often prohibitively expensive. The uniform fission site method has been developed to yield a more spatially-uniform distribution of relative uncertainties. This is accomplished by biasing the density of fission neutron source sites while not biasing the solution. The method is integrated into the source iteration process, and does not require any auxiliary forward or adjoint calculations. For a given amount of computational effort, the use of the method results in a reduction of the largest uncertainties relative to the standard algorithm. Two variants of the method have been implemented and tested. Both have been shown to be effective. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

A Decision-Based Iterative Modeling Framework for Probabilistic Performance Assessments - 18441

Radiological probabilistic performance assessments (PPAs) employ highly complex computer models to simulate the release of radionuclides from wastes in an engineered disposal facility, migration of these radionuclides over time, and resulting radiological dose to potentially exposed individuals. An iterative approach to PPA modeling is described to help define the appropriate complexity of the computer model. An initial model is developed that addresses all relevant processes and media, but that employs preliminary statistical parameter distributions to minimize the initial effort expended. Where appropriate, transport and exposure submodels with simplifying assumptions are also used in the initial model. Different assumptions and boundaries may be explored with the initial model to assess their importance in constraining model results. Exploration of the results of the initial model are used to determine the benefits of model refinements, and to guide such refinements to focus on specific areas of the model that most influence the model outputs of interest. This approach promotes efficiency and model parsimony, but does not compromise defensibility when performed in a transparent manner. Key model boundaries for a PPA include the time frame of the modeling (temporal boundaries), where future human receptors may be located (spatial boundaries), and what types of land uses and activities may exist in the future (exposure scenarios). Simplified transport and exposure models can be useful in helping decision makers and stakeholders establish such boundaries by analyzing the consequences of selecting different bounds and understanding which assumptions and boundaries have the greatest influence on the magnitude and distribution of model results. Pathway analysis and parameter sensitivity analysis performed on PPA model results are key components of the iterative modeling approach. The objective of pathway analysis is to identify the critical radionuclides and pathways contributing to radiation dose during the modeling performance period. An environmental 'pathway' is described by the connections from a waste package or cell in the disposal facility to a future human exposure event. These connections are commonly visualized using graphical means. A simple example of a potentially complete pathway for radon-222 is: radium-containing waste → radon emanation → upwards diffusion in cap material → diffusion/advection into a future building → inhalation exposure to a building occupant Parameter sensitivity analysis is conducted on the results of a PPA to identify those specific model input parameters whose distribution of values contributes the most to the distribution of target model output, such as annual radiation dose. Pathway and sensitivity analyses on initial modeling results provide information for guiding subsequent model development, and possibly interim decisions, by informing questions tied to the modeling objectives and performance criteria. For example, questions related to radiological doses from existing disposals may include: 1. Is radiation dose from radiological exposures either obviously negligible or obviously unacceptable based on initial modeling results? 2. If dose is not obviously negligible or unacceptable relative to applicable performance criteria, what model refinements would reduce uncertainty in the results to better inform decisions? (authors)

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

Human factors case study: An iterative approach to verifying and validating a new suite of reactor inspection and maintenance tooling - 177

An iterative approach was taken to perform the HF Verification and Validation (V and V) activities for a complex set of reactor inspection and maintenance tooling equipment known as the Bruce Reactor Inspection and Maintenance System (BRIMS). This paper is divided into the verification activities (of Human System Information Task Support and Design verifications) and the validation activities. The validation method used a graded approach by breaking validation activities into early usability testing and high fidelity validations. Early usability testing, allowed for issues and design concepts to be identified and explored. High fidelity validation activities tested that the final design requirements and operational goals of the system were met in the production equipment. The validation method has been described in detail by depicting the following: task selection criteria, tools, performance measures and data collection. Validation results demonstrate the benefits of the graded approach as issues were to be identified early in the design process, improvements were made during detailed design and additional testing could confirm that issues were resolved prior to commissioning. To capture and share the success of this iterative V and V approach lessons learned have been created for use in future V and V projects. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Overview of Nuclear Integration Activities for Mitigation of Radiation Loads to ITER Superconducting Magnets

The Nuclear Integration Unit (NIU) is a trans-departmental team from the ITER Organization and Domestic Agencies, responsible for all nuclear integration aspects of the facility and for ensuring that systems and components are suitable for their radiation environment and satisfy all related performance requirements. This is a very challenging task namely due to the large number of systems, their still evolving designs, the intricate and far-reaching impacts of these in nuclear responses everywhere, and the complex world-wide distribution of procurement responsibilities. A sound systems engineering approach that involves all relevant experts and stakeholders, and trades-off between different performance requirements in an optimum way, needs to be taken and sustained during design evolution and construction, always with the objective of final overall project success. One of the most important current occupations of the NIU is the evaluation of, and protection from, radiation conditions to superconducting magnets during plasma operations, and particularly nuclear heating. This requires neutronic analyses in massive computer resources, with state-of-the-art methods and complex, extensive computer models. The main ITER superconducting magnet systems subject to significant nuclear loads are the toroidal field coils (TFCs) and the poloidal field coils (PFCs). These are illustrated in Fig. 1. Nuclear responses in these depend mostly on the design of all in-vessel (blankets, manifolds, divertor, in-vessel coils, port plugs), vessel and thermal shield components. Sources to consider are the 500 MW DT burning plasma (14.1 MeV neutrons) and the activated water in the in-vessel cooling system (photons up to 7 MeV from N-16 and neutrons up to 3.5 MeV from N-17). This paper discusses a snapshot of the present understanding of nuclear heat conditions in both the TFC and PFC magnets. This includes integral values and distributions, as well as best estimates from computations and conservative upper estimates by consideration of systematic effects and random uncertainties. The computer models, methods and assumptions used for evaluation of these responses are described. Nuclear integration aspects and ongoing efforts to study and implement additional shielding and other mitigating measures are also discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

How to reduce Shutdown Dose Rates in ITER Diagnostics Equatorial Ports

The ITER vacuum vessel presents 54 access ports, arranged in three levels, where port plugs are inserted hosting different types of systems: test blanket modules, antennas, vacuum pumps, neutral beam injectors and diagnostics. At the equatorial level, nine out of the eighteen access ports present diagnostics port plugs, in which the diagnostics systems are deployed into three DSMs (Diagnostics Shield Modules). The DSMs, being a major constituent in volume of the port plugs, must help to fulfill the design target of Shutdown Dose Rates (SDDR) corresponding to 100 μSv.h{sup -1} after 10{sup 6} s of cooling time for planned in-situ maintenance operations in the port interspace. Additionally, the DSMs are largely responsible for the port plug drained weight limit of 45 T, which imposes limitations to the shielding technology. Up to now, all the proposed DSM designs combine differently stainless steel, water and/or B{sub 4}C as shielding to accommodate a maximum of 1.1 m{sup 3} of DSM bulk shield, contained into a L 1845 x W 518 x H 2011 mm stainless steel box weighting 5.8 T for each DSM. These proposals show insufficient shielding properties with respect to SDDR, while none of them satisfies currently the weight limit. However, during 2016, a new DSM concept has been developed, named 'modular DSM', with significant engineering advances and changes. The DSM length has been increased by 500 mm, up to 2345 mm. The SS box has been reduced to the minimum structure required, presenting the same weight while being larger and hosting a larger volume of bulk shielding inside. Thus, the volume of bulk shielding has been enlarged up to 1.5 m{sup 3} per DSM. The shielding technology is based on B{sub 4}C bricks deployed in SS trays that are attached to the vertical plates. This results in a bulk shielding made of 65% of B{sub 4}C, 10% SS and 25% void, with effective density of 2.4 g.cm{sup -3}. In summary, the modular DSM, without a weight increase, hosts extra 0.5 m3 of bulk shielding along an additional 500 mm of length, to be used with a light but highly neutron absorber, what allows fitting the weight limit of 45 T for the port plug once drained. From now, the non-modular DSMs will be referred as 'short DSMs'. The modular DSM's advantages with respect to the short DSM versions have shown that the DSMs design can be improved in generic approach, i.e. when the penetrations of the systems are not taken into consideration. A mention apart is required when port specific penetrations are considered. Provided that they will be an important radiation transmission path, their presence could significantly diminish the improvement. To clarify this issue, in this work we have carried out a nuclear analysis of the ITER equatorial port no. 11 integration considering the modular DSM. It is a first plasma port hosting the following 6 major diagnostics systems: Neutral Particle Analyzer (NPA), H-α monitor, Low Field Side (LFS) reflectometry, Visible - Ultraviolet divertor (VUV div), Visible - Ultraviolet core survey (VUV CS) and X-rays Crystal System (XRCS). For the baseline case, a SDDR matrix has been derived assigning responsibilities to the different EP no. 11 individual systems in terms of neutron flux transmission through them, and the related activation due to that neutron flux (activation of itself, but also of the rest of systems taken one by one). Different EP no. 11 configurations have been studied in addition to the baseline case. For example, the backfill of the port plug penetrations to protect all the empty space not required for systems sight. Another issue is the determination of the role of the Co impurity content minimization from 0.05 wt.% to 0.03 wt.% recently proposed. In this summary, only some results related to the baseline are given.

61 RADIATION PROTECTION AND DOSIMETRY↗