Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “SHUTDOWN”

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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
NASA NTRS1984

Shutdown characteristics of the Mod-O wind turbine with aileron controls

Horizontal-axis wind turbines utilize partial or full variable blade pitch to regulate rotor speed. The weight and costs of these systems indicated a need for alternate methods of rotor control. Aileron control is an alternative which has potential to meet this need. The NASA Lewis Research Center has been experimentally testing aileron control rotors on the Mod-U wind turbine to determine their power regulation and shutdown characteristics. Experimental and analytical shutdown test results are presented for a 38 percent chord aileron-control rotor. These results indicated that the 38 percent chord ailerons provided overspeed protection over the entire Mod-O operational windspeed range, and had a no-load equilibrium tip speed ratio of 1.9. Thus, the 38 percent chord ailerons had much improved aerodynamic braking capability when compared with the first aileron-control rotor having 20 percent chord ailerons.

Miller, D. R.↗
NASA NTRS1987

Combustion stability analysis of preburners in liquid propellant rocket engines during shutdown

A linearized one-dimensional lumped-parameter model capable of predicting the occurrence of the low frequency combustion instability (chugging) experienced during preburner shutdown in the Space Shuttle Main Engines is discussed, and predictions are compared with NASA experimental results. Results from a parametric study of parameters including chamber pressure, fuel and oxygen temperatures, and the effective bulk modulus of the liquid oxidizer suggest that chugging is probably affected by conditions at shutdown through the fuel and oxidizer temperatures. It is suggested that chugging is initiated when the fuel, oxidizer, and helium temperature and flow rates pass into an unstable region, and that chugging may be terminated by decaying pressures.

Lim, Kair-Chuan↗
NASA NTRS1988

SSME technology test bed fast shutdown assessment

The fastest practical shutdown sequence for the Technology Test Bed Engine is developed. A sequence is defined and key issues addressed. The conclusion of the study is that a fastest shutdown sequence within the existing SSME design capability is practical and can cut oxidizer consumption by 50 percent. However, the revised sequence would not have prevented any of the prior experienced SSME incidents and would introduce development risk to the program.

Liebman, A.↗
DOE OSTI2015

Safety shutdown separators

The present invention pertains to electrochemical cells which comprise (a) an anode; (b) a cathode; (c) a solid porous separator, such as a polyolefin, xerogel, or inorganic oxide separator; and (d) a nonaqueous electrolyte, wherein the separator comprises a porous membrane having a microporous coating comprising polymer particles which have not coalesced to form a continuous film. This microporous coating on the separator acts as a safety shutdown layer that rapidly increases the internal resistivity and shuts the cell down upon heating to an elevated temperature, such as 110.degree. C. Also provided are methods for increasing the safety of an electrochemical cell by utilizing such separators with a safety shutdown layer.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2018

Preliminary Evaluation of Removing Spent Nuclear Fuel From Shutdown Sites - Fort Calhoun Site Visit - 18048

This is a technical paper that does not take into account contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. An updated evaluation of removing spent nuclear fuel (SNF) from 14 shutdown nuclear power plant sites was conducted during 2017. The major update since 2016 was the addition of the recently shutdown Fort Calhoun site to the evaluation. The Fort Calhoun site is located on the western shore of the Missouri River in Washington County in eastern Nebraska, about 31 km north of Omaha, Nebraska. Fort Calhoun has been shut down since October 24, 2016 and final removal of SNF from the reactor vessel was completed on November 13, 2016. The site visit to Fort Calhoun occurred in May 2017 and involved meetings with the Fort Calhoun staff and the Union Pacific Railroad. Participants in the site visit included the Federal Railroad Administration, Pahrump Paiute Tribe, Prairie Island Indian Community, Nebraska State Patrol, Nebraska Department of Health and Human Services, Iowa Department of Transportation Office of Rail Transportation, Iowa Department of Transportation Office of Motor Vehicle Enforcement, and the Council of State Governments-Midwest. The Fort Calhoun site was found to have two transportation mode options for the removal of SNF, rail and barge. In the case of rail, an onsite rail spur could be reinstalled from the existing Cargill industrial spur that runs through the Fort Calhoun site, or an onsite trans-load onto the Cargill industrial spur could be performed. In 2006, a barge was used to ship steam generators, a pressurizer, and a reactor vessel head to the Fort Calhoun site; consequently, barge transport of SNF offsite is potentially an option. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Assessment of Multi-Step CADIS Parameter Sensitivity to Neutron Flux Spectrum in Shutdown Dose Rate Calculations

Shutdown dose rate (SDDR) analysis requires (a) a neutron transport calculation to estimate neutron flux fields, (b) an activation calculation to compute radionuclide inventories and associated photon sources, and (c) a photon transport calculation to estimate final dose rate. In some applications, accurate full-scale Monte Carlo (MC) SDDR simulations are needed for very large systems with massive amounts of shielding materials. For example, SDDR assessments are required throughout the biological shield (bio-shield) of the ITER experimental facility to evaluate the required waiting period after the shutdown of ITER and to identify the locations where human accessibility should be prohibited. The bio-shield is a large cylindrical concrete structure (30 m tall and 34 m in diameter) surrounding the very complex tokamak machine. Determining the effects on SDDR of important factors, such as the cross talk (interactions) between the different ports of ITER, is possible only through full-scale simulations that involve all the complex inner details of the ITER tokamak machine. These simulations are impractical because calculation of space- and energy-dependent neutron fluxes throughout the structural materials is needed to estimate the distribution of radioisotopes causing the SDDR. Determining a good importance function that can be used in biasing a neutron MC calculation for SDDR is complicated because it is difficult to explicitly express the SDDR response function, which depends on subsequent computational steps. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) hybrid MC/deterministic method was developed to speed up the SDDR MC neutron transport calculation using an importance function that represents the neutron importance to the final SDDR. The MS-CADIS method uses the CADIS method, which has been Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC0500OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for the United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan). successfully used for more than a decade in radiation shielding applications to generate consistent source biasing and weight window (WW) variance reduction parameters to accelerate continuous-energy MC simulations. However, because MS-CADIS focuses on multi-step shielding calculations such as SDDR calculations, it develops an importance function for the initial radiation transport calculation (e.g., the neutron calculations in SDDR simulations) that represents the importance of particles to the final response of the overall simulation. To develop this importance function, the MS-CADIS method uses an adjoint calculation with a specific neutron adjoint source that represents the neutron's contribution to the final SDDR. Brute force calculation of the MS-CADIS neutron adjoint source requires a deterministic neutron transport calculation coupled with activation calculations in each element of the deterministic mesh. These calculations are needed to determine the MS-CADIS parameter for the neutron interactions that produce radioisotopes contributing to the SDDR. The MS-CADIS parameter is the microscopic cross section of the radioisotope-producing interaction, multiplied by the mass of this radioisotope existing at the end of the irradiation and decay scenario, and divided by the steady-state rate of the interaction. Because the activation calculations are performed for each radionuclide in each mesh element, implementation of the MS-CADIS method can be computationally expensive. However, these neutron transport and activation calculations are not needed if we assume that the radioisotopes are created due to single neutron interactions and their concentrations do not change as a result of burnup by neutron irradiation. The use of this approximation, which renders the MS-CADIS parameter independent of the neutron flux, seems reasonable in calculating the MS-CADIS importance function because this function is needed only to speed up the MC calculation, not to calculate the final SDDR. In this work, we investigate the validity of this approximation by investigating the behavior of the MS-CADIS parameter as a function of the neutron flux spectra in the Fusion Evaluated Nuclear Data Library (FENDL) Neutronics and Shielding Benchmark problem that contains representative isotopes and flux environments of fusion energy systems. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Automated power-following control for nuclear thermal propulsion startup and shutdown using MOOSE-based applications

This paper describes an investigation of various automated control strategies applied to a full-core multiphysics Griffin/Bison/RELAP-7 model of a prototypical nuclear thermal propulsion system. In all cases, control is achieved by actuating control drums based on the demanded power and predicted quantities from the numerical model. One key finding is that hybrid proportional integral derivative controllers – a novel type of controller that uses both power and reactivity predicted signals – can demonstrate a level of performance rivaling that of period-generated control. The former requires parameter tuning, while the latter mostly necessitates providing reactivity coefficients and temperature rates of change, which could be very challenging to accurately measure in real-time. In addition, decay heat plays an important role in determining cooling requirements during cooldown phases. A decay heat model, accounting for burn time and throttling, was derived, and is incorporated within the model to simulate the steady-state and shutdown phases and satisfyingly follow the power demand. However, temperature overshoots and non-constant specific impulse during throttling will necessitate further improvements.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2023

Comprehensive Analysis of Streaming and Shutdown Dose Rate Experiments at JET with ORNL Fusion Neutronics Workflows

Current experimental fusion systems and conceptual designs of fusion pilot plants (FPPs) are growing in complexity and size. Several radiation metrics are crucial to the safe operation of fusion machines, including neutron flux streaming through openings and the shutdown dose rate (SDDR). Most current designs of advanced experimental fusion systems—and the most probable candidates for FPPs—are based on the tokamak concept, which is prone to neutron streaming through the myriad openings needed for diagnostic and support systems. SDDR is caused by decay gamma rays from radionuclides that become activated by neutrons during the operation of a fusion system that use deuterium-deuterium (DD), tritium-tritium, or deuterium-tritium plasma. Because computational tools have become essential for determining these radiation metrics, they must be validated against reliable and applicable experimental data. Experiments at the Joint European Torus (JET) provide a unique source of experimental data for validating computational tools and nuclear data used to determine SDDR and neutron fluxes in streaming-dominated geometries. Here, this paper presents the comprehensive analysis of the high-performance DD JET SDDR, and streaming experiments performed using Oak Ridge National Laboratory (ORNL) fusion workflows. The computational results were compared with experimental results that consist of online SDDR measurements with ionization chambers and neutron fluence streaming measurements using thermoluminescent detectors. The ratio of calculated-to-experimental SDDR values ranges from 0.6 to 2.5, and the streaming results range from 0.5 to 8.0. Future work will include analyzing the JET 2021 DTE2 campaign alongside the integration of the Shift Monte Carlo transport code into all ORNL fusion neutronics workflows.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Implementation of the D1S Methodology for Shutdown Dose Rate Calculations in the OpenMC Monte Carlo Particle Transport Code

We present an implementation of the direct one-step (D1S) methodology for shutdown dose rate (SDR) calculations in the OpenMC Monte Carlo particle transport code. In addition to being the first fully open-source D1S implementation, it is also the first to require no ad hoc source code or nuclear data library modifications. The code can seamlessly switch between production of prompt and decay photons based on a user input parameter, and the decay data needed for decay photon generation are made available through a depletion chain file, which is already used for OpenMC’s built-in depletion/activation solver. A set of Python functions significantly eases the burden of computing and applying time correction factors needed to properly account for the time dependence of radionuclide activity. To assess the accuracy of the D1S implementation, SDR calculations have been carried out for three problems: a prism of iron irradiated by 14-MeV neutrons, the ITER port plug computational benchmark, and the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from the Shielding INtegral Benchmark Archive and Database (SINBAD). For each of these problems, comparisons were made to calculations using the rigorous two-step (R2S) method. The results on the iron prism problem illustrate how the D1S method achieves superior spatial resolution compared to the R2S method without the need for spatial discretization of the activation regions. The D1S and R2S results for the ITER port plug benchmark agree well with previously reported results in the literature. While the D1S results are 10% to 15% lower than the R2S results, this may be due to stochastic uncertainty and/or spatial discretization in the R2S calculations. On the FNG dose rate benchmark problem, the D1S method produces dose rate estimates that are within 4% of the dose rates predicted using a cell-based R2S workflow. The D1S estimates of the SDR are also in reasonable agreement with the experimental measurements and show the same basic trends that have been observed in previous works. A qualitative analysis of the execution time and uncertainty for the R2S and D1S workflows suggests that the D1S method would attain a higher figure of merit.

D1S method↗
DOE OSTI2026

A New Shutdown Dose Rate Benchmark Problem for Representative Fusion Applications

Here, this work introduces a new benchmark problem for calculating shutdown dose rates (SDDRs) aimed at fusion reactor applications. The model is designed to represent a simplified version of a typical ITER port plug. The responses of interest include neutron flux, gamma flux, and gamma SDDR at 12 different locations scattered throughout the port. This article outlines the geometry specifications of the problem, provides material definitions for the components, specifies the required responses to be calculated, and presents the source definition information. The need for this benchmark arises from the limited availability of publicly accessible references, with only one benchmark representing the typical dimensions and materials found in fusion systems. This existing benchmark has been cited extensively, reflecting the demand within the scientific community to test both established and novel workflows for SDDR calculations. However, since its presentation at a conference in 2011, the results have become increasingly well known. Moreover, the absence of formal publication and peer review has led to the details of this benchmark being extracted from secondary sources, such as subsequent studies that reference it. As a result, analysts are left with significant flexibility in interpreting the key parameters, which can be adjusted to account for unknown systematic errors, ultimately reproducing the already well-known responses. This new benchmark serves as an updated version of that earlier work, with the aim of providing a more reliable description of the materials and their impurities, which is crucial for assessing activation and subsequent gamma emission. Additionally, it seeks to provide a geometry that more closely represents an ITER port plug. The improvements in the problem definition will lead to a more reproducible benchmark problem, while also presenting the radiation transport community with a completely new challenge. The results will be published in a future article to allow analysts adequate time to analyze this problem independently.

Benchmark↗
DOE OSTI2013

Novel Hybrid Monte Carlo/Deterministic Technique for Shutdown Dose Rate Analyses of Fusion Energy Systems

The rigorous 2-step (R2S) method uses three-dimensional Monte Carlo transport simulations to calculate the shutdown dose rate (SDDR) in fusion reactors. Accurate full-scale R2S calculations are impractical in fusion reactors because they require calculating space- and energy-dependent neutron fluxes everywhere inside the reactor. The use of global Monte Carlo variance reduction techniques was suggested for accelerating the neutron transport calculation of the R2S method. The prohibitive computational costs of these approaches, which increase with the problem size and amount of shielding materials, inhibit their use in the accurate full-scale neutronics analyses of fusion reactors. This paper describes a novel hybrid Monte Carlo/deterministic technique that uses the Consistent Adjoint Driven Importance Sampling (CADIS) methodology but focuses on multi-step shielding calculations. The Multi-Step CADIS (MS-CADIS) method speeds up the Monte Carlo neutron calculation of the R2S method using an importance function that represents the importance of the neutrons to the final SDDR. Using a simplified example, preliminarily results showed that the use of MS-CADIS enhanced the efficiency of the neutron Monte Carlo simulation of an SDDR calculation by a factor of 550 compared to standard global variance reduction techniques, and that the increase over analog Monte Carlo is higher than 10,000.

hybrid Monte Carlo/deterministic↗
DOE OSTI2021

Benchmarking a Pronghorn Model of the Natural Convection Shutdown Heat Removal Test Facility Against STAR-CCM+ (Poster)

Summary Verification of Pronghorn’s Net Radiation Transfer via Code-to-Code comparison with STAR-CCM+. Comparison based on Natural Convection Shutdown Heat-Removal Test Facility (NSTF). Near-perfect agreement if conjugate heat transfer inside the riser channels is replaced by fixed temperature. Difference of up to 5K if conjugate heat transfer is modeled (correlation vs. full CFD).

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2012

Validation of the integration of CFD and SAS4A/SASSYS-1: Analysis of EBR-II shutdown heat removal test 17

Recent analyses have demonstrated the need to model multidimensional phenomena, particularly thermal stratification in outlet plena, during safety analyses of loss-of-flow transients of certain liquid-metal cooled reactor designs. Therefore, Argonne's reactor systems safety code SAS4A/SASSYS-1 is being enhanced by integrating 3D computational fluid dynamics models of the plena. A validation exercise of the new tool is being performed by analyzing the protected loss-of-flow event demonstrated by the EBR-II Shutdown Heat Removal Test 17. In this analysis, the behavior of the coolant in the cold pool is modeled using the CFD code STAR-CCM+, while the remainder of the cooling system and the reactor core are modeled with SAS4A/SASSYS-1. This paper summarizes the code integration strategy and provides the predicted 3D temperature and velocity distributions inside the cold pool during SHRT-17. The results of the coupled analysis should be considered preliminary at this stage, as the exercise pointed to the need to improve the CFD model of the cold pool tank. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Initial Site-Specific Studies for Removing SNF from Six Shutdown Sites - 18042

The US Department of Energy (DOE) is exploring options for developing a large-scale transportation system for the safe transport of spent nuclear fuel (SNF) and high-level radioactive waste to future storage or disposal facilities. This complex, large-scale transportation system will involve coordinated, integrated activities. Interfaces with various transportation and non-transportation activities must be defined, and systems must be developed to ensure successful integration when the system becomes operational. To provide an understanding of the tasks and interfaces required for the complete de-inventory of particular sites where SNF is stored, DOE solicited bids from integrating contractors to develop initial site-specific de-inventory reports. The reports were developed for several nuclear power plant sites selected to represent a range of geographic locations, existing on-site SNF storage systems, and expected transportation modal options. Each report is presented in a self-contained narrative describing a potential campaign for removing all SNF and greater-than-class-C low-level radioactive waste (GTCC waste) from each specified site. Each study narrative includes site information; an analysis of the near-site transportation infrastructure; a concept of operations for the de-inventory; an estimated budget and schedule for the duration of operations; and a safety and security plan with site-specific considerations, emergency response considerations, and recommendations for the next steps DOE could take to de-inventory the site. The reports provide cost estimates for analytical purposes only; they do not attempt to specify the party or parties responsible for the costs. These reports present a first look at how an integrating contractor might recommend removing SNF from particular sites. They are written from the perspective of the contractor and do not represent DOE's plans. At this writing, six sites have been studied: - Connecticut Yankee, - Big Rock Point, - Humboldt Bay, - Maine Yankee, - Trojan, - Kewaunee. DOE may develop de-inventory reports for all reactor sites, with an initial focus on shutdown reactor sites. The order in which sites have been selected for study should not be considered an indicator of the order used by DOE to remove SNF from the sites. Since the SNF's destination is not important for the purpose of this work, DOE tasked the contractor with planning transportation of the SNF from the site to a generic railroad location in the approximate geographical center of the United States. These reports represent an initial study of the specific sites. Per the terms of the contract, the contractor was not permitted to engage in dialogue with power plant site personnel; near-site railroad personnel; or any local, state, or regional stakeholders about the studies. DOE provided site-specific data. The contractor was tasked to rely on past corporate experience, data available from DOE, and publicly available data when making site-specific recommendations for removing all SNF from each site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Transition From Operation to Decommissioning with a Sudden Decision to Permanent Shutdown and with Parallel Operation On-site - 18559

In a short time the Swedish nuclear operator OKG have to shift from a operator of three nuclear power plants at one site to a company where both decommissioning and operation will take place in parallel for numerous of years as a result of a decision to permanently shut down two of the three reactors. The paper will be focusing on the challenges to change organization and mind-set from operations to decommissioning within parts of a company while still maintaining and developing the highest standard for the plant still in operation, and this in the combination with a decision of permanent shutdown earlier than expected. The paper will also address how OKG managed to create and maintain momentum in essential processes like permits, defueling and project planning before, during and short after the organization was changed and some of the employees had to leave the company. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Proven Safe and Effective Process for the Treatment of Bulk Sodium at Shutdown Reactors and Research Facilities - 18601

Elemental sodium is an alkaline metal that reacts when exposed to air or water. This reaction is exothermic in nature and presents the risk of attendant fire as well as hydrogen generation. Within the United States sodium was used in liquid metal fast breeder reactors (LMFBR) during the 1960's and 70's as a heat transfer medium for commercial, experimental and prototype reactors as well as for nuclear energy related research and development experiments due to the element's good thermal conductivity. Shutdown of these reactors and the end of these experiments present a challenge to decommissioning teams as the wastes associated with each project have different characteristics affecting treatment based on the sodium's origin as well as different volumes requiring treatment. French CEA (Commissariat a l'Energie Atomique) and FRAMATOME worked together in the 1980's to develop a treatment process called the NOAH (NaOH - sodium hydroxide) process for the disposal of large volumes of bulk sodium. This paper provides an overview of the treatment process, details past uses of the technology and discusses the process' applications within the United States Department of Energy complex. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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↗
DOE OSTI2014

Analysis of Shutdown Dose Rate in Fusion Energy Systems Using Hybrid Monte Carlo/Deterministic Techniques - Paper 138

Accurate assessments of shutdown dose rate (SDDR) are critical to support operation, maintenance, and waste disposal planning and to guide possible design changes of critical components in fusion energy systems. An SDDR calculation involves three steps: 1. a neutron transport calculation to determine the space and energy neutron flux distributions, 2. an activation calculation for computing the photon source distribution, and 3. a photon transport calculation for the estimation of the final SDDR. A companion paper describes the challenges of SDDR computations in fusion energy systems and the status of the techniques that are currently being used in SDDR analysis. Of these techniques, the rigorous 2-step (R2S) computational system entails Monte Carlo (MC) neutron and photon transport calculations coupled with a comprehensive activation step using a dedicated inventory code and library. The use of global MC variance reduction techniques was suggested for accelerating the SDDR MC neutron transport calculation. These techniques, which attempt to calculate MC tallies with nearly uniform relative uncertainties in both the low-flux space-energy regions and the high-flux space-energy regions, do not preferentially focus the MC computational efforts toward space-energy regions of high importance to the final decay dose. The ability of these approaches to accurately predict SDDR is inhibited by their prohibitive computational costs, which will be on the order of thousands of processor-years for full-scale modeling of an entire fusion plant. The companion paper describes the theoretical background of the Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method, which has been proposed to speed up SDDR MC neutron calculations. The MS-CADIS method uses the CADIS method, which has been successfully used for more than a decade in shielding calculations, but focuses on multistep shielding calculations such as SDDR analyses. The companion paper also describes a new method for calculating the uncertainties in SDDR due to uncertainties in the MC neutron calculation. This new method of uncertainty propagation uses the MS-CADIS neutron adjoint source to propagate the uncertainties in the neutron fluxes to the SDDR. This paper describes the application of the MS-CADIS method for speeding up the SDDR MC calculations and for calculating the SDDR uncertainty due to the neutron flux uncertainties. A new metric for assessing the reliability of SDDR calculations was also suggested and used in this analysis. The ITER benchmark problem was used in this analysis. The problem resembles the configuration and geometrical arrangement of an upper port plug in ITER. The analysis compared the efficiency of the MS-CADIS method to the traditional approach of using global MC variance reduction techniques for speeding up SDDR MC neutron calculation. The application of the MS-CADIS method to SDDR calculations in fusion energy systems was tested using the ITER benchmark problem. Compared to the standard FW-CADIS method, the increase in the efficiency of the SDDR neutron MC calculation due to the use of the MS-CADIS method was between 18% and 69%. The MS-CADIS method also increases the fraction of non-zero-scoring mesh tally elements in the space-energy regions of high importance to the final SDDR. Implementation of the MS-CADIS method in the SCALE and the ADVANTG code systems is currently under way. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗