Rapid Discontinuation of Prednisone in Kidney Transplant Recipients: 15-Year Outcomes From the University of Minnesota.

PubMed

Serrano, Oscar Kenneth; Kandaswamy, Raja; Gillingham, Kristen; Chinnakotla, Srinath; Dunn, Ty B; Finger, Erik; Payne, William; Ibrahim, Hassan; Kukla, Aleksandra; Spong, Richard; Issa, Naim; Pruett, Timothy L; Matas, Arthur

2017-10-01

Short- and intermediate-term results have been reported after rapid discontinuation of prednisone (RDP) in kidney transplant recipients. Yet there has been residual concern about late graft failure in the absence of maintenance prednisone. From October 1, 1999, through June 1, 2015, we performed a total of 1553 adult first and second kidney transplants-1021 with a living donor, 532 with a deceased donor-under our RDP protocol. We analyzed the 15-year actuarial overall patient survival (PS), graft survival (GS), death-censored GS (DCGS), and acute rejection-free survival (ARFS) rates for RDP compared with historical controls on maintenance prednisone. For living donor recipients, the actuarial 15-year PS rates were similar between groups. But RDP was associated with increased GS (P = 0.02) and DCGS (P = 0.01). For deceased donor recipients, RDP was associated with significantly better PS (P < 0.01), GS (P < 0.01) and DCGS (P < 0.01). There was no difference between groups in the rate of acute or chronic rejection, or in the mean estimated glomerular filtration rate at 15 years. However, RDP-treated recipients had significantly lower rates of avascular necrosis, cytomegalovirus, cataracts, new-onset diabetes after transplant, and cardiac complications. Importantly, for recipients with GS longer than 5 years, there was no difference between groups in subsequent actuarial PS, GS, and DCGS. In summary, at 15 years postkidney transplant, RDP did not lead to decreased in PS or GS, or an increase in graft dysfunction but as associated with reduced complication rates.

Effect of supplementation of mustard oil cake on intake, digestibility and microbial protein synthesis of cattle in a straw-based diet in Bangladesh.

PubMed

Khandaker, Zahirul Haque; Uddin, Mohammad Mohi; Sultana, Nadira; Peters, Kurt J

2012-04-01

The objective of this study was to analyse the effects of different levels of rumen-degradable protein (RDP) on intake, digestibility and microbial protein synthesis by supplementing mustard oil cake (MOC) on rice straw-based diet of cattle (Bos indicus) in Bangladesh. A 4 × 4 Latin square design was applied. Four diets having constant energy (7.0 MJ/kg of dry matter (DM)) with varying levels of RDP (M(0) = 4.1 g/MJ (control), M(1) = 6.3 g/MJ, M(2) = 8.3 g/MJ and M(3) = 12.4 g/MJ of metabolizable energy (ME)) were received by each animal for a period of 28 days. A metabolism trial was conducted for 7 days. Results indicate that with increasing levels of RDP, crude protein (CP) and RDP intake increased significantly (P < 0.01). The significant (P < 0.01) increase in digestibility values are obtained for DM, organic matter, CP and digestible organic matter in the rumen. The digestibility of neutral detergent fibre and acid detergent fibre was also increased significantly (P < 0.05). The total nitrogen (N), ammonia-N and total volatile fatty acids increase significantly (P < 0.01) while the rumen pH increased from M(0) to M(2) and decreased thereafter. The efficiency microbial N intake increased significantly (P < 0.01) but showed a curvilinear response with higher RDP level (12.40 g/RDP/MJ ME). This study concludes that supplementation of RDP from MOC enhances the intake, digestibility and microbial protein synthesis which ultimately increases utilization of low-quality feed resources that can be used for developing cost-effective feeding systems on a straw-based diet in tropical regions.

Impact of Implementing a Primary Enforcement Seat Belt Law in Florida : A Case Study

DOT National Transportation Integrated Search

2012-08-01

On June 30, 2009, Florida implemented a primary seat belt law. The State was already engaged in a Rural : Demonstration Program (RDP) to increase belt usage in rural areas in the northern part of the State and participated : regularly in annual Click...

Conceptual design of a divertor Thomson scattering diagnostic for NSTX-U

DOE Office of Scientific and Technical Information (OSTI.GOV)

McLean, A. G., E-mail: mclean@fusion.gat.com; Soukhanovskii, V. A.; Allen, S. L.

2014-11-15

A conceptual design for a divertor Thomson scattering (DTS) diagnostic has been developed for the NSTX-U device to operate in parallel with the existing multipoint Thomson scattering system. Higher projected peak heat flux in NSTX-U will necessitate application of advanced magnetics geometries and divertor detachment. Interpretation and modeling of these divertor scenarios will depend heavily on local measurement of electron temperature, T{sub e}, and density, n{sub e}, which DTS provides in a passive manner. The DTS design for NSTX-U adopts major elements from the successful DIII-D DTS system including 7-channel polychromators measuring T{sub e} to 0.5 eV. If implemented onmore » NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.« less

Effect of protein degradability on milk production of dairy ewes.

PubMed

Mikolayunas-Sandrock, C; Armentano, L E; Thomas, D L; Berger, Y M

2009-09-01

The objective of this experiment was to determine the effect of protein degradability of dairy sheep diets on milk yield and protein utilization across 2 levels of milk production. Three diets were formulated to provide similar energy concentrations and varying concentrations of rumen-degradable protein (RDP) and rumen-undegradable protein (RUP): 12% RDP and 4% RUP (12-4) included basal levels of RDP and RUP, 12% RDP and 6% RUP (12-6) included additional RUP, and 14% RDP and 4% RUP (14-4) included additional RDP. Diets were composed of alfalfa-timothy cubes, whole and ground corn, whole oats, dehulled soybean meal, and expeller soybean meal (SoyPlus, West Central, Ralston, IA). Estimates of RDP and RUP were based on the Small Ruminant Nutrition System model (2008) and feed and orts were analyzed for Cornell N fractions. Eighteen multiparous dairy ewes in midlactation were divided by milk yield (low and high) into 2 blocks of 9 ewes each and were randomly assigned within block (low and high) to 3 pens of 3 ewes each. Dietary treatments were arranged in a 3 x 3 Latin square within each block and applied to pens for 14-d periods. We hypothesized that pens consuming high-RUP diets (12-6) would produce more milk and milk protein than the basal diet (12-4) and pens consuming high-RDP diets (14-4) would not produce more milk than the basal diet (12-4). Ewes in the high-milk-yield square consumed more dry matter and produced more milk, milk fat, and milk protein than ewes in the low-milk-yield square. There was no effect of dietary treatment on dry matter intake. Across both levels of milk production, the 12-6 diet increased milk yield by 14%, increased milk fat yield by 14%, and increased milk protein yield by 13% compared with the 14-4 and 12-4 diets. Gross N efficiency (milk protein N/intake protein N) was 11 and 15% greater in the 12-6 and 12-4 diets, respectively, compared with the 14-4 diet. Milk urea N concentration was greater in the 12-6 diet and tended to be greater in the 14-4 diet compared with the 12-4 diet, indicating that the excretion of urea N in this study was more closely related to dietary crude protein concentration than to protein degradability.

Effects of parity and supply of rumen-degraded and undegraded protein on production and nitrogen balance in Holsteins.

PubMed

Flis, S A; Wattiaux, M A

2005-06-01

Eight Holstein cows (4 primiparous and 4 multiparous) were used in a replicated 4 x 4 Latin square design to determine milk production response and N balance when diets had no NRC-predicted excess of rumen-undegradable protein (RUP) or rumen-degradable protein (RDP), 10% RUP excess, 10% RDP excess, or 10% excess of both RUP and RDP. Diets were fed as a total mixed ration with (dry matter basis) 25% alfalfa silage, 25% corn silage, 19 to 21% corn grain, and varying proportions of solvent soybean meal and expeller soybean meal as primary sources of supplemental RDP and RUP, respectively. Milk yield and dry matter intake (DMI) were recorded daily, and total collection of feces and urine was completed in the last 3 d of each 21-d period. Dietary crude protein averaged 17.5 and 18.5% for the recommended and excess RDP diets, respectively, and 17.3 and 18.4% for the recommended and excess RUP diets, respectively. When cows were fed excess RUP diets in the form of expeller soybean meal, DMI and milk production increased, but the opposite was true when the diets contained excess RDP in the form of solvent soybean meal. Milk composition was not affected by RDP, RUP, or by parity, and there were no parity x RDP interactions for any of the measurements. However, apparent digestibility of neutral detergent fiber, dry matter, and N increased in multiparous cows but not in primiparous cows because of excess RUP. The increase in the yield of milk N with excess RUP was not influenced by parity, but multiparous cows retained more of the additional N apparently absorbed, whereas primiparous cows excreted the additional apparently absorbed N in the urine. Overall, the difference in urinary N due to parity (70 g/d) was about 4 times greater than the impact of dietary treatments (17 g/d). Our results suggest that multiparous cows have either a much larger urea pool or a greater demand to restore body protein mobilized earlier in lactation compared with primiparous cows. Reduction in urinary N excretion in commercial dairy herds could be obtained by separately balancing rations for first and later lactations.

'Responsible drinking' programs and the alcohol industry in Brazil: killing two birds with one stone?

PubMed

Pantani, Daniela; Sparks, Robert; Sanchez, Zila M; Pinsky, Ilana

2012-10-01

Over the last decade, the Brazilian alcohol industry - which for years has ignored alcohol problems - inaugurated responsible drinking programs (RDPs). This paper reports findings from an exploratory study that investigated the RDP-related activities of six leading alcohol companies in Brazil (three national, three transnational) focusing on program goals and components, target populations and evaluation methods. Interviews were conducted from October 2007 to February 2008 with nine key-informants, and 71 corporate documents were collected along with additional web information about the programs. Content analysis of interviews and institutional documents was used to identify the companies' RDP activities. Three types of RDPs were found that focused respectively on institutional action, drinking and driving, and underage drinking. All three transnational firms were involved in RDPs, whereas national firms demonstrated limited involvement. The majority of RDPs were implemented using television. No targeted research appears to have been undertaken by the companies to assess the efficacy of the strategies in terms of changes in drinking behavior. The evidence for both national and transnational firms means that is difficult to confirm that the responsible drinking programs produced so far in Brazil have been undertaken to systematically reduce alcohol problems, or mainly as part of a public relations strategy to reduce criticism and potentially forestall government regulations (Babor, 2006, 2009; Jernigan, 2009). Copyright © 2012 Elsevier Ltd. All rights reserved.

Methods and compositions for diagnosing and preventing a group B streptococcal infection

DOEpatents

Brady, Linda Jeannine [Gainesville, FL; Seifert, Kyle N [Harrisonburg, VA; Adderson, Elisabeth E [Memphis, TN; Bohnsack, John F [Salt Lake City, UT

2009-09-15

The present invention provides a group B streptococcal (GBS) surface antigen, designated epsilon antigen, that is co-expressed with the delta antigen on a subset of serotype III GBS. Epsilon is expressed on more pathogenic Restriction Digest Pattern (RDP) III-3 GBS, but not on RDP types 1, 2, or 4. Accordingly, the present invention provides compositions and methods for detecting a group B streptococcus serotype III, RDP III-3 strain. Vaccines and methods of identifying agents which inhibit adhesion of a group B streptococcal cell to a host cell are also provided.

The Actions of Lyophilized Apple Peel on the Electrical Activity and Organization of the Ventricular Syncytium of the Hearts of Diabetic Rats

PubMed Central

Martínez-Ladrón de Guevara, Elideth; Pérez-Hernández, Nury; Villalobos-López, Miguel Ángel; Pérez-Ishiwara, David Guillermo; Salas-Benito, Juan Santiago; Martínez Martínez, Alejandro; Hernández-García, Vicente

2016-01-01

This study was designed to examine the effects of lyophilized red delicious apple peel (RDP) on the action potentials (APs) and the input resistance-threshold current relationship. The experiments were performed on isolated papillary heart muscles from healthy male rats, healthy male rats treated with RDP, diabetic male rats, and diabetic male rats treated with RDP. The preparation was superfused with oxygenated Tyrode's solution at 37°C. The stimulation and the recording of the APs, the input resistance, and the threshold current were made using conventional electrophysiological methods. The RDP presented no significant effect in normal rats. Equivalent doses in diabetic rats reduced the APD and ARP. The relationship between input resistance and threshold current established an inverse correlation. The results indicate the following: (1) The functional structure of the cardiac ventricular syncytium in healthy rats is heterogeneous, in terms of input resistance and threshold current. Diabetes further accentuates the heterogeneity. (2) As a consequence, conduction block occurs and increases the possibility of reentrant arrhythmias. (3) These modifications in the ventricular syncytium, coupled with the increase in the ARP, are the adequate substrate so that, with diabetes, the heart becomes more arrhythmogenic. (4) RDP decreases the APD, the ARP, and most syncytium irregularity caused by diabetes. PMID:26839897

Survival outcomes and prognostic factors in mycosis fungoides/Sézary syndrome: validation of the revised International Society for Cutaneous Lymphomas/European Organisation for Research and Treatment of Cancer staging proposal.

PubMed

Agar, Nita Sally; Wedgeworth, Emma; Crichton, Siobhan; Mitchell, Tracey J; Cox, Michael; Ferreira, Silvia; Robson, Alistair; Calonje, Eduardo; Stefanato, Catherine M; Wain, Elizabeth Mary; Wilkins, Bridget; Fields, Paul A; Dean, Alan; Webb, Katherine; Scarisbrick, Julia; Morris, Stephen; Whittaker, Sean J

2010-11-01

We have analyzed the outcome of mycosis fungoides (MF) and Sézary syndrome (SS) patients using the recent International Society for Cutaneous Lymphomas (ISCL)/European Organisation for Research and Treatment of Cancer (EORTC) revised staging proposal. Overall survival (OS), disease-specific survival (DSS), and risk of disease progression (RDP) were calculated for a cohort of 1,502 patients using univariate and multivariate models. The mean age at diagnosis was 54 years, and 71% of patients presented with early-stage disease. Disease progression occurred in 34%, and 26% of patients died due to MF/SS. A significant difference in survival and progression was noted for patients with early-stage disease having patches alone (T1a/T2a) compared with those having patches and plaques (T1b/T2b). Univariate analysis established that (1) advanced skin and overall clinical stage, increased age, male sex, increased lactate dehydrogenase (LDH), and large-cell transformation were associated with reduced survival and increased RDP; (2) hypopigmented MF, MF with lymphomatoid papulosis, and poikilodermatous MF were associated with improved survival and reduced RDP; and (3) folliculotropic MF was associated with an increased RDP. Multivariate analysis established that (1) advanced skin (T) stage, the presence in peripheral blood of the tumor clone without Sézary cells (B0b), increased LDH, and folliculotropic MF were independent predictors of poor survival and increased RDP; (2) large-cell transformation and tumor distribution were independent predictors of increased RDP only; and (3) N, M, and B stages; age; male sex; and poikilodermatous MF were only significant for survival. This study has validated the recently proposed ISCL/EORTC staging system and identified new prognostic factors.

Substitution of Asp-309 by Asn in the Arg-Asp-Pro (RDP) motif of Acetobacter diazotrophicus levansucrase affects sucrose hydrolysis, but not enzyme specificity.

PubMed Central

Batista, F R; Hernández, L; Fernández, J R; Arrieta, J; Menéndez, C; Gómez, R; Támbara, Y; Pons, T

1999-01-01

beta-Fructofuranosidases share a conserved aspartic acid-containing motif (Arg-Asp-Pro; RDP) which is absent from alpha-glucopyranosidases. The role of Asp-309 located in the RDP motif of levansucrase (EC 2.4.1.10) from Acetobacter diazotrophicus SRT4 was studied by site-directed mutagenesis. Substitution of Asp-309 by Asn did not affect enzyme secretion. The kcat of the mutant levansucrase was reduced 75-fold, but its Km was similar to that of the wild-type enzyme, indicating that Asp-309 plays a major role in catalysis. The two levansucrases showed optimal activity at pH 5.0 and yielded similar product profiles. Thus the mutation D309N affected the efficiency of sucrose hydrolysis, but not the enzyme specificity. Since the RDP motif is present in a conserved position in fructosyltransferases, invertases, levanases, inulinases and sucrose-6-phosphate hydrolases, it is likely to have a common functional role in beta-fructofuranosidases. PMID:9895294

Improved feed protein fractionation schemes for formulating rations with the cornell net carbohydrate and protein system.

PubMed

Lanzas, C; Broderick, G A; Fox, D G

2008-12-01

Adequate predictions of rumen-degradable protein (RDP) and rumen-undegradable protein (RUP) supplies are necessary to optimize performance while minimizing losses of excess nitrogen (N). The objectives of this study were to evaluate the original Cornell Net Carbohydrate Protein System (CNCPS) protein fractionation scheme and to develop and evaluate alternatives designed to improve its adequacy in predicting RDP and RUP. The CNCPS version 5 fractionates CP into 5 fractions based on solubility in protein precipitant agents, buffers, and detergent solutions: A represents the soluble nonprotein N, B1 is the soluble true protein, B2 represents protein with intermediate rates of degradation, B3 is the CP insoluble in neutral detergent solution but soluble in acid detergent solution, and C is the unavailable N. Model predictions were evaluated with studies that measured N flow data at the omasum. The N fractionation scheme in version 5 of the CNCPS explained 78% of the variation in RDP with a root mean square prediction error (RMSPE) of 275 g/d, and 51% of the RUP variation with RMSPE of 248 g/d. Neutral detergent insoluble CP flows were overpredicted with a mean bias of 128 g/d (40% of the observed mean). The greatest improvements in the accuracy of RDP and RUP predictions were obtained with the following 2 alternative schemes. Alternative 1 used the inhibitory in vitro system to measure the fractional rate of degradation for the insoluble protein fraction in which A = nonprotein N, B1 = true soluble protein, B2 = insoluble protein, C = unavailable protein (RDP: R(2) = 0.84 and RMSPE = 167 g/d; RUP: R(2) = 0.61 and RMSPE = 209 g/d), whereas alternative 2 redefined A and B1 fractions as the non-amino-N and amino-N in the soluble fraction respectively (RDP: R(2) = 0.79 with RMSPE = 195 g/d and RUP: R(2) = 0.54 with RMSPE = 225 g/d). We concluded that implementing alternative 1 or 2 will improve the accuracy of predicting RDP and RUP within the CNCPS framework.

Leukemia, lymphoma and multiple myeloma mortality (1950–1999) and incidence (1969–1999) in the Eldorado uranium workers cohort

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zablotska, Lydia B., E-mail: Lydia.Zablotska@ucsf.edu; Lane, Rachel S.D.; Frost, Stanley E.

Uranium workers are chronically exposed to low levels of radon decay products (RDP) and gamma (γ) radiation. Risks of leukemia from acute and high doses of γ-radiation are well-characterized, but risks from lower doses and dose-rates and from RDP exposures are controversial. Few studies have evaluated risks of other hematologic cancers in uranium workers. The purpose of this study was to analyze radiation-related risks of hematologic cancers in the cohort of Eldorado uranium miners and processors first employed in 1932–1980 in relation to cumulative RDP exposures and γ-ray doses. The average cumulative RDP exposure was 100.2 working level months andmore » the average cumulative whole-body γ-radiation dose was 52.2 millisievert. We identified 101 deaths and 160 cases of hematologic cancers in the cohort. Overall, male workers had lower mortality and cancer incidence rates for all outcomes compared with the general Canadian male population, a likely healthy worker effect. No statistically significant association between RDP exposure or γ-ray doses, or a combination of both, and mortality or incidence of any hematologic cancer was found. We observed consistent but non-statistically significant increases in risks of chronic lymphocytic leukemia (CLL) and Hodgkin lymphoma (HL) incidence and non-Hodgkin lymphoma (NHL) mortality with increasing γ-ray doses. These findings are consistent with recent studies of increased risks of CLL and NHL incidence after γ-radiation exposure. Further research is necessary to understand risks of other hematologic cancers from low-dose exposures to γ-radiation. - Highlights: • We analyzed long-term follow-up for hematologic cancers of the Eldorado uranium workers. • Workers were exposed to a unique combination of radon decay products (RDP) and gamma (γ) ray doses. • Exposures to RDP and γ-ray doses were not associated with significantly increased risks of cancers. • Radiation risks of chronic lymphocytic leukemia (CLL) and Hodgkin lymphoma were increased. • Study findings provide additional support for radiation-related risks of CLL.« less

A meta-analysis of bacterial diversity in the feces of cattle

USDA-ARS?s Scientific Manuscript database

In this study, we conducted a meta-analysis on 16S rRNA gene sequences of bovine fecal origin that are publicly available in the RDP database. A total of 13663 sequences including 603 isolate sequences were identified in the RDP database (Release 11, Update 1), where 13447 sequences were assigned t...

Rumen-degradable protein in roll conditioned or macerated legume hays and silages estimated by in situ kinetics vs. alternative methods

USDA-ARS?s Scientific Manuscript database

Alternatives to the in situ method for estimating rumen-degradable protein (RDP) in diverse forage legumes should be validated. In this study, RDP in roll conditioned or macerated silages and hays of Medicago, Lotus, and Trifolium species with differing polyphenol compositions were estimated from in...

Validation of yield component traits identified by GWA mapping in a rice tropical japonica x tropical japonica RIL mapping population

USDA-ARS?s Scientific Manuscript database

The Rice Diversity Panel 1 (RDP1) was developed for genome-wide association (GWA) mapping to explore the five diverse rice (Oryza sativa) subpopulations (indica, aus, aromatic, temperate japonica and tropical japonica). RDP1 was evaluated for over 30 agronomic and morphological traits, most of whic...

Heuristic reusable dynamic programming: efficient updates of local sequence alignment.

PubMed

Hong, Changjin; Tewfik, Ahmed H

2009-01-01

Recomputation of the previously evaluated similarity results between biological sequences becomes inevitable when researchers realize errors in their sequenced data or when the researchers have to compare nearly similar sequences, e.g., in a family of proteins. We present an efficient scheme for updating local sequence alignments with an affine gap model. In principle, using the previous matching result between two amino acid sequences, we perform a forward-backward alignment to generate heuristic searching bands which are bounded by a set of suboptimal paths. Given a correctly updated sequence, we initially predict a new score of the alignment path for each contour to select the best candidates among them. Then, we run the Smith-Waterman algorithm in this confined space. Furthermore, our heuristic alignment for an updated sequence shows that it can be further accelerated by using reusable dynamic programming (rDP), our prior work. In this study, we successfully validate "relative node tolerance bound" (RNTB) in the pruned searching space. Furthermore, we improve the computational performance by quantifying the successful RNTB tolerance probability and switch to rDP on perturbation-resilient columns only. In our searching space derived by a threshold value of 90 percent of the optimal alignment score, we find that 98.3 percent of contours contain correctly updated paths. We also find that our method consumes only 25.36 percent of the runtime cost of sparse dynamic programming (sDP) method, and to only 2.55 percent of that of a normal dynamic programming with the Smith-Waterman algorithm.

Initial experience with robotic pancreatic surgery in Singapore: single institution experience with 30 consecutive cases.

PubMed

Goh, Brian K P; Low, Tze-Yi; Lee, Ser-Yee; Chan, Chung-Yip; Chung, Alexander Y F; Ooi, London L P J

2018-05-24

Presently, the worldwide experience with robotic pancreatic surgery (RPS) is increasing although widespread adoption remains limited. In this study, we report our initial experience with RPS. This is a retrospective review of a single institution prospective database of 72 consecutive robotic hepatopancreatobiliary surgeries performed between 2013 and 2017. Of these, 30 patients who underwent RPS were included in this study of which 25 were performed by a single surgeon. The most common procedure was robotic distal pancreatectomy (RDP) which was performed in 20 patients. This included eight subtotal pancreatectomies, two extended pancreatecto-splenectomies (en bloc gastric resection) and 10 spleen-saving-RDP. Splenic preservation was successful in 10/11 attempted spleen-saving-RDP. Eight patients underwent pancreaticoduodenectomies (five hybrid with open reconstruction), one patient underwent a modified Puestow procedure and one enucleation of uncinate tumour. Four patients had extended resections including two RDP with gastric resection and two pancreaticoduodenectomies with vascular resection. There was one (3.3%) open conversion and seven (23.3%) major (>Grade II) morbidities. Overall, there were four (13.3%) clinically significant (Grade B) pancreatic fistulas of which three required percutaneous drainage. These occurred after three RDP and one robotic enucleation. There was one reoperation for port-site hernia and no 30-day/in-hospital mortalities. The median post-operative stay was 6.5 (range: 3-36) days and there were six (20%) 30-day readmissions. Our initial experience showed that RPS can be adopted safely with a low open conversion rate for a wide variety of procedures including pancreaticoduodenectomy. © 2018 Royal Australasian College of Surgeons.

LEUKEMIA, LYMPHOMA AND MULTIPLE MYELOMA MORTALITY (1950–1999) AND INCIDENCE (1969–1999) IN THE ELDORADO URANIUM WORKERS COHORT

PubMed Central

Zablotska, Lydia B.; Lane, Rachel S.D.; Frost, Stanley E.; Thompson, Patsy A.

2014-01-01

Uranium workers are chronically exposed to low levels of radon decay products (RDP) and gamma (γ) radiation. Risks of leukemia from acute and high doses of γ-radiation are well-characterized, but risks from lower doses and dose-rates and from RDP exposures are controversial. Few studies have evaluated risks of other hematologic cancers in uranium workers. The purpose of this study was to analyze radiation-related risks of hematologic cancers in the cohort of Eldorado uranium miners and processors first employed in 1932–1980 in relation to cumulative RDP exposures and γ-ray doses. The average cumulative RDP exposure was 100.2 working level months and the average cumulative whole-body γ-radiation dose was 52.2 millisievert. We identified 101 deaths and 160 cases of hematologic cancers in the cohort. Overall, male workers had lower mortality and cancer incidence rates for all outcomes compared with the general Canadian male population, a likely healthy worker effect. No statistically significant association between RDP exposure or γ-ray doses, or a combination of both, and mortality or incidence of any hematologic cancer was found. We observed consistent but non-statistically significant increases in risks of chronic lymphocytic leukemia (CLL) and Hodgkin lymphoma (HL) incidence and non-Hodgkin lymphoma (NHL) mortality with increasing γ-ray doses. These findings are consistent with recent studies of increased risks of CLL and NHL incidence after γ-radiation exposure. Further research is necessary to understand risks of other hematologic cancers from low-dose exposures to γ-radiation. PMID:24583244

Studying long 16S rDNA sequences with ultrafast-metagenomic sequence classification using exact alignments (Kraken).

PubMed

Valenzuela-González, Fabiola; Martínez-Porchas, Marcel; Villalpando-Canchola, Enrique; Vargas-Albores, Francisco

2016-03-01

Ultrafast-metagenomic sequence classification using exact alignments (Kraken) is a novel approach to classify 16S rDNA sequences. The classifier is based on mapping short sequences to the lowest ancestor and performing alignments to form subtrees with specific weights in each taxon node. This study aimed to evaluate the classification performance of Kraken with long 16S rDNA random environmental sequences produced by cloning and then Sanger sequenced. A total of 480 clones were isolated and expanded, and 264 of these clones formed contigs (1352 ± 153 bp). The same sequences were analyzed using the Ribosomal Database Project (RDP) classifier. Deeper classification performance was achieved by Kraken than by the RDP: 73% of the contigs were classified up to the species or variety levels, whereas 67% of these contigs were classified no further than the genus level by the RDP. The results also demonstrated that unassembled sequences analyzed by Kraken provide similar or inclusively deeper information. Moreover, sequences that did not form contigs, which are usually discarded by other programs, provided meaningful information when analyzed by Kraken. Finally, it appears that the assembly step for Sanger sequences can be eliminated when using Kraken. Kraken cumulates the information of both sequence senses, providing additional elements for the classification. In conclusion, the results demonstrate that Kraken is an excellent choice for use in the taxonomic assignment of sequences obtained by Sanger sequencing or based on third generation sequencing, of which the main goal is to generate larger sequences. Copyright © 2016 Elsevier B.V. All rights reserved.

HARVESTING ROOFTOP RUNOFF FROM RDP HOUSING IN SOUTH AFRICA: A PRAGMATIC APPROACH TO SUPPLEMENT WATER SUPPLY WHILE REDUCING FLOODING, SOIL EROSION, AND DISEASE

EPA Science Inventory

The team was able to design a shuttle valve and full scale rainwater collection system that, in times of enough rainfall, provided an extra free source of water to flush toilets. Based of the dimensions of the RDP house and the tank size, the system should be able to collec...

Effects of molasses and corn grain at 2 levels of ruminally degradable protein on lactating cow ruminal fermentation and rumen content mass

USDA-ARS?s Scientific Manuscript database

The objective of this study was to evaluate lactating dairy cow ruminal fermentation and rumen content mass with diets containing molasses (M) or finely ground dry corn grain at 3 levels of M (0, 5.25, 10.5% DM) and with differing levels of ruminally degradable protein (+RDP or –RDP). Twelve ruminal...

Metabolic and productive response to ruminal protein degradability in early lactation cows fed untreated or xylose-treated soybean meal-based diets.

PubMed

Jahani-Moghadam, M; Amanlou, H; Nikkhah, A

2009-12-01

Effects of different dietary rumen undegradable (RUP) to degradable (RDP) protein ratios on ruminal nutrient degradation, feed intake, blood metabolites and milk production were determined in early lactation cows. Four multiparous (43 ± 5 days in milk) and four primiparous (40 ± 6 days in milk) tie-stall-housed Holstein cows were used in a duplicated 4 × 4 Latin square design with four 21-day periods. Each period had 14-day of adaptation and 7-day of sampling. Diets contained on a dry matter (DM) basis, 23.3% alfalfa hay, 20% corn silage and 56.7% concentrate. Cows were first offered alfalfa hay at 7:00, 15:00 and 23:00 hours, and 30 min after each alfalfa hay delivery were offered a mixture of corn silage and concentrate. Treatments were diets with RUP:RDP ratios of (i) 5.2:11.6 (control), (ii) 6.1:10.6, (iii) 7.1:9.5 and (iv) 8.1:8.5, on a dietary DM% basis. Different RUP:RDP ratios were obtained by partial and total replacement of untreated soybean meal (SBM) with xylose-treated SBM (XSBM). In situ study using three rumen-cannulated non-lactating cows showed that DM and crude protein (CP) of SBM had greater rapidly degradable fractions. The potentially degradable fractions were degraded more slowly in XSBM. Treatment cows produced greater milk, protein, lactose, solids-non-fat and total solids than control cows. Increasing RUP:RDP reduced blood urea linearly. Feed costs dropped at RUP:RDP ratios of 6.1:10.6 and 7.1:9.5, but not at 8.1:8.5, compared with the 5.2:11.6 ratio. Intake of DM and CP, rumen pH, blood glucose, albumin and total protein, faecal and urine pH, changes in body weight and body condition score, and milk lactose and solids-non-fat percentages did not differ among treatments. Results provide evidence that increasing dietary RUP:RDP ratio from 5.2:11.6 to 7.1:9.5 optimizes nitrogen metabolism and milk production and reduces feed costs in early lactation cows. Reduced blood urea suggests reprodutive benefits.

Effects of ruminal protein degradability and frequency of supplementation on nitrogen retention, apparent digestibility, and nutrient flux across visceral tissues in lambs fed low-quality forage.

PubMed

Atkinson, R L; Toone, C D; Robinson, T J; Harmon, D L; Ludden, P A

2010-02-01

Two experiments were conducted to determine the effect of ruminal protein degradability and supplementation frequency on intake, apparent digestibility, N retention, and nutrient flux across visceral tissues of lambs fed a low-quality forage diet. In both experiments, wethers were fed a basal diet of mature crested wheatgrass hay (4.2% CP) for ad libitum consumption plus 1 of 4 supplements: 1) a high RDP supplement provided daily (RDP-D), 2) the high RDP supplement provided on alternate days (RDP-A), 3) a high RUP provided on alternate days (RUP-A), or 4) a 50:50 mixture of the RDP and RUP supplements provided on alternate days. In Exp. 1, 12 lambs (29.9 +/- 2.7 kg initial BW) were used. Forage OM, NDF, and ADF intake were not affected by treatment. Total tract digestibilities (OM, NDF, ADF, and N) were unaffected (P >or= 0.15) by treatment. Neither protein degradability nor supplementation frequency had an effect (P >or= 0.52) on N retention. In Exp. 2, 15 lambs (34 +/- 4 kg initial BW) fitted with indwelling catheters in a hepatic vein, the hepatic portal vein, a mesenteric vein, and a mesenteric artery were used. Release of ammonia N by the portal-drained viscera (PDV) was reduced (P = 0.004) in alternate-day-supplemented lambs compared with RDP-D. Consequently, hepatic uptake of ammonia N was least (P = 0.003) in all alternate-day lambs. Alpha-amino nitrogen (AAN) release by the PDV and hepatic uptake of AAN were not affected by treatment or supplementation frequency. Additionally, hepatic output and PDV uptake of urea N were not affected by treatment. Hepatic N uptake (ammonia N + AAN) accounted for urea synthesized by the liver in all treatments; however, hepatic urea synthesis was approximately 4.5-fold less for RUP-A lambs. This suggests that the provision of AA as RUP may provide a delay in ureagenesis, thus altering the timing of N recycling.

Leukemia, lymphoma and multiple myeloma mortality (1950-1999) and incidence (1969-1999) in the Eldorado uranium workers cohort.

PubMed

Zablotska, Lydia B; Lane, Rachel S D; Frost, Stanley E; Thompson, Patsy A

2014-04-01

Uranium workers are chronically exposed to low levels of radon decay products (RDP) and gamma (γ) radiation. Risks of leukemia from acute and high doses of γ-radiation are well-characterized, but risks from lower doses and dose-rates and from RDP exposures are controversial. Few studies have evaluated risks of other hematologic cancers in uranium workers. The purpose of this study was to analyze radiation-related risks of hematologic cancers in the cohort of Eldorado uranium miners and processors first employed in 1932-1980 in relation to cumulative RDP exposures and γ-ray doses. The average cumulative RDP exposure was 100.2 working level months and the average cumulative whole-body γ-radiation dose was 52.2 millisievert. We identified 101 deaths and 160 cases of hematologic cancers in the cohort. Overall, male workers had lower mortality and cancer incidence rates for all outcomes compared with the general Canadian male population, a likely healthy worker effect. No statistically significant association between RDP exposure or γ-ray doses, or a combination of both, and mortality or incidence of any hematologic cancer was found. We observed consistent but non-statistically significant increases in risks of chronic lymphocytic leukemia (CLL) and Hodgkin lymphoma (HL) incidence and non-Hodgkin lymphoma (NHL) mortality with increasing γ-ray doses. These findings are consistent with recent studies of increased risks of CLL and NHL incidence after γ-radiation exposure. Further research is necessary to understand risks of other hematologic cancers from low-dose exposures to γ-radiation. Copyright © 2014 The Authors. Published by Elsevier Inc. All rights reserved.

Occurrence and distribution of oligomeric organophosphorus flame retardants in different treatment stages of a sewage treatment plant.

PubMed

Liang, Kang; Shi, Fengqiong; Liu, Jingfu

2018-01-01

Oligomeric organophosphate esters (OOPEs) like 2,2-bis(chloromethyl)-propane-1,3-diyltetrakis (2-chloroethyl) bisphosphate (V6), resorcinol bis(diphenylphosphate) (RDP) and bisphenol A bis(diphenylphosphate) (BDP), are widely used as alternatives of Deca-BDE in plastic and electronic consumer products. However, studies on the environmental occurrence and fate of OOPEs are very scarce. This work studied the occurrence, distribution and fate of V6, BDP and RDP during the different treatment stages of a sewage treatment plant (STP) in Beijing, China. To accomplish this, a method to analyze trace V6, BDP and RDP in suspended solids samples and aqueous samples of sewage and sludge was developed by using liquid chromatography tandem mass spectrometry (LC-MS/MS). Using this method, BDP and RDP were detected for the first time in suspended solids of sewage and sludge with a concentration of 2.06-5.82 ng/g dry weight and 0.44-3.45 ng/g dry weight, respectively, whereas their concentration level in the aqueous phase of these samples were below the detection limits of the method. However, V6 was detected in all treatment stages of the STP, with concentrations in the range of 10.2-27.1 ng/L in aqueous phase and 0.40-1.73 ng/g dw in solid phase. Mass balance results indicated that 75.6% of the original V6 mass flow was discharged along with effluent, while 83.3% and 72.2% of the initial RDP and BDP mass flow were lost due to biodegradation, respectively. Nevertheless, compared to the 14 widely used monomeric organophosphate esters (MOPEs), the concentration levels of OOPEs in this studied STP were relatively low. Copyright © 2017 Elsevier Ltd. All rights reserved.

ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

NASA Astrophysics Data System (ADS)

Whyte, Dennis; ADX Team

2015-11-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

NASA Astrophysics Data System (ADS)

Park, J.; Im, K.; Kwon, S.; Kim, J.; Kim, D.; Woo, M.; Shin, C.

2017-12-01

This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

Power exhaust scenarios and control for projected high-power NSTX-U operation

NASA Astrophysics Data System (ADS)

Menard, Jonathan; Gerhardt, S. P.; Myers, C. E.; Reinke, M. L.; Brooks, A.; Mardenfeld, M.; NSTX Upgrade Team

2017-10-01

An important goal of the NSTX Upgrade (NSTX-U) research program is to characterize energy confinement in the low-aspect-ratio spherical tokamak configuration over a significantly expanded range of plasma current, toroidal field, and heating power, while increasing flattop durations up to 5 seconds. However, the narrowing of the scrape-off layer at higher current combined with an improved understanding of expected halo-current loads has motivated a significant re-design of NSTX-U plasma facing components in the high-heat-flux regions of the divertor. In order to reduce the expected divertor heat flux to acceptable levels, a combination of mitigation techniques will be used: increased divertor poloidal flux expansion, increased divertor radiation, and controlled strike-point sweeping. The machine requirements for these various mitigation techniques are studied here using a newly implemented reduced heat-flux model. Systematic equilibrium scans are used to quantify the required divertor coil currents and to verify vertical stability for a range of plasma shapes. Free-boundary control schemes to constrain the strike-point location and field-line angle-of-incidence will also be discussed. Work supported by DOE contract DE-AC02- 09CH11466.

Global Emerging Infection Surveillance and Response (GEIS)- Avian Influenza Pandemic Influenza (AI/PI) Program

DTIC Science & Technology

2014-10-01

amplicon of Corona Virus RdP gene. Finally, one PCR amplicon of a Chikungunya virus gene from the VHF group was sequenced. These sequence data are...suggestive of STIs ( discharge or genital ulcer) often go undiagnosed, and are treated empirically with broad spectrum antibiotics. The drug resistance... discharge are offered anonymous screening for gonorrhea and chlamydia (GC) and specimen taken for detection and isolation of Neisseria gonorrhoeae

Experiences with tungsten coatings in high heat flux tests and under plasma load in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Herrmann, A.; Greuner, H.; Fuchs, J. C.; de Marné, P.; Neu, R.; ASDEX Upgrade Team

2009-12-01

ASDEX Upgrade was operated with about 6400 s plasma discharge during the scientific program in 2007/2008 exploring tungsten as a first wall material in tokamaks. In the first phase, the heating power was restricted to 10 MW. It was increased to 15 MW in the second phase. During this operational period, a delamination of the 200 μm W-VPS coating happened at 2 out of 128 tiles of the outer divertor and an unscheduled opening was required. In the third phase, ASDEX Upgrade was operated with partly predamaged tiles and up to 15 MW heating power. The target load was actively controlled by N2-seeding. This paper presents the screening test of target tiles in the high heat flux test facility GLADIS, experiences with operation and detected damages of the outer divertor as well as the heat load to the outer divertor and the reasons for the toroidal asymmetry of the divertor load.

Divertor extreme ultraviolet (EUV) survey spectroscopy in DIII-D

NASA Astrophysics Data System (ADS)

McLean, Adam; Allen, Steve; Ellis, Ron; Jarvinen, Aaro; Soukhanovskii, Vlad; Boivin, Rejean; Gonzales, Eduardo; Holmes, Ian; Kulchar, James; Leonard, Anthony; Williams, Bob; Taussig, Doug; Thomas, Dan; Marcy, Grant

2017-10-01

An extreme ultraviolet spectrograph measuring resonant emissions of D and C in the lower divertor has been added to DIII-D to help resolve an 2X discrepancy between bolometrically measured radiated power and that predicted by boundary codes for DIII-D, JET and ASDEX-U. With 290 and 450 gr/mm gratings, the DivSPRED spectrometer, an 0.3 m flat-field McPherson model 251, measures ground state transitions for D (the Lyman series) and C (e.g., C IV, 155 nm) which account for >75% of radiated power in the divertor. Combined with Thomson scattering and imaging in the DIII-D divertor, measurements of position, temperature and fractional power emission from plasma components are made and compared to UEDGE/SOLPS-ITER. Mechanical, optical, electrical, vacuum, and shielding aspects of DivSPRED are presented. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698 and DE-AC52-07NA27344, and by the LLNL Laboratory Directed R&D Program, project #17-ERD-020.

Pomalidomide desensitization in a patient hypersensitive to immunomodulating agents

PubMed Central

Seki, J.T.; Sakurai, N.; Lam, W.; Reece, D.E.

2017-01-01

Despite progressive treatments with tandem stem-cell transplantation, patients with incurable myeloma eventually succumb to relapsed or refractory disease if left untreated. Promising agents such as proteasome inhibitors and immunomodulating imide drugs (imids), including the newer-generation agent pomalidomide, in combination with lower-dose dexamethasone, have been shown to be effective and to significantly improve and prolong survival in pretreated patients. Although the incidence of pomalidomide hypersensitivity reaction (hsr) in this class of drugs is not as well known, we have documented cutaneous toxicity (grade 3 by the Common Terminology Criteria for Adverse Events, version 4) in 2 separate cases (not yet published). Because the imids are chemically, structurally, and pharmacologically similar, it is not unreasonable to consider possible cross-reactivity in pomalidomide recipients who developed hsr when receiving previous lines of imids. As a patient’s advocate, it is only prudent to provide a responsible, and yet practical, means to better address cross-sensitivity for patients. Intervention with the use of a rapid desensitization program (rdp) as a preventive measure should be introduced before initiating pomalidomide. Such a proactive measure for the patient’s safety will ensure a smooth transition into pomalidomide treatment. A hsr can be either related or non-related to immunoglobulin E. As imids become an essential treatment backbone for myeloma and other plasma-cell diseases, an increasing number of patients could experience skin and other life-threatening toxicities, resulting in unnecessary discontinuation of these life-prolonging agents. An extemporaneously prepared pomalidomide suspension developed at our centre enables patients to undergo rdp safely. Patients enjoy a good quality of life and clinical response after the rdp procedure. PMID:28874903

Putative recombination events and evolutionary history of five economically important viruses of fruit trees based on coat protein-encoding gene sequence analysis.

PubMed

Boulila, Moncef

2010-06-01

To enhance the knowledge of recombination as an evolutionary process, 267 accessions retrieved from GenBank were investigated, all belonging to five economically important viruses infecting fruit crops (Plum pox, Apple chlorotic leaf spot, Apple mosaic, Prune dwarf, and Prunus necrotic ringspot viruses). Putative recombinational events were detected in the coat protein (CP)-encoding gene using RECCO and RDP version 3.31beta algorithms. Based on RECCO results, all five viruses were shown to contain potential recombination signals in the CP gene. Reconstructed trees with modified topologies were proposed. Furthermore, RECCO performed better than the RDP package in detecting recombination events and exhibiting their evolution rate along the sequences of the five viruses. RDP, however, provided the possible major and minor parents of the recombinants. Thus, the two methods should be considered complementary.

Plasma-wall interactions in ITER

NASA Astrophysics Data System (ADS)

Parker, R.; Janeschitz, G.; Pacher, H. D.; Post, D.; Chiocchio, S.; Federici, G.; Ladd, P.; Iter Joint Central Team; Home Teams

1997-02-01

This paper reviews the status of the design of the divertor and first-wall/shield, the main in-vessel components for ITER. Under nominal ignited conditions, 300 MW of alpha power will be produced and must be removed from the divertor and first-wall. Additional power from auxiliary sources up to the level of 100 MW must also be removed in the case of driven burns. In the ignited case, about 100 MW will be radiated to the first wall as bremsstrahlung. Allowing the remaining power to be conducted to the divertor target plates would result in excessive heat fluxes. The power handling strategy is to radiate an additional 100-150 MW in the SOL and the divertor channel via a combination of radiation from hydrogen, and intrinsic and seeded impurities. Vertical targets have been adopted for the baseline divertor configuration. This geometry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densities on the target plates can be ≤ 5 MW/ m2. Such regimes promote relatively high pressure (> 1 Pa) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping speed to pump helium is ≤ 50 m3/ s. An important physics question is the quality of core confinement in these attractive divertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performance of in-vessel components. Both centered disruptions and VDE's produce stresses in the first-wall/shield modules, backplate and the divertor wings and cassettes that are near or even somewhat in excess of allowables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties are major factors determining component lifetime. Considering the potential for impurity contamination and minimizing tritium inventory as well as thermomechanical performance, the present material selection calls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux could be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the final selection for the first assembly will be made at the end of the EDA.

Effects of rumen-degradable protein:rumen-undegradable protein ratio and corn processing on production performance, nitrogen efficiency, and feeding behavior of Holstein dairy cows.

PubMed

Savari, M; Khorvash, M; Amanlou, H; Ghorbani, G R; Ghasemi, E; Mirzaei, M

2018-02-01

This study was conducted to investigate the effects of the ratio of rumen-degradable protein (RDP) to rumen-undegradable protein (RUP) and corn processing method on production performance, nitrogen (N) efficiency, and feeding behavior of high-producing Holstein dairy cows. Twelve multiparous Holstein cows (second parity; milk yield = 48 ± 3 kg/d) were assigned to a replicated 4 × 4 Latin square design with a 2 × 2 factorial arrangement of treatments. Factor 1 was corn processing method [ground corn (GC) or steam flaked corn (SFC) with a flake density of about 390 g/L], and factor 2 was RDP:RUP ratio [low ratio (LR) = 60:40; high ratio (HR) = 65:35] based on crude protein (%). The crude protein concentrations were kept constant across the treatments (16.7% of DM). No significant interactions of main treatment effects occurred for lactation performance data. Cows fed 2 different RDP:RUP ratios exhibited similar dry matter intake (DMI), but those fed SFC showed decreased feed intake compared with those receiving GC (25.1 ± 0.48 vs. 26.2 ± 0.47 kg/d, respectively). Cows fed HR diets produced more milk than did those fed LR diets (44.4 ± 1.05 vs. 43.2 ± 1.05 kg/d, respectively). Milk fat content decreased but milk protein content increased in cows fed SFC compared with those fed GC. Feed efficiency (i.e., milk yield/DMI) was enhanced with increasing ratio of RDP:RUP (1.68 ± 0.04 vs. 1.74 ± 0.04 for LR and HR, respectively). Apparent N efficiency was higher in cows fed HR than in those fed LR (30.4 ± 0.61 vs. 29.2 ± 0.62, respectively). Compared with cows fed the GC-based diet, those receiving SFC exhibited lower values of N intake, N-NH 3 concentration, and fecal N excretion. Cows receiving SFC-based diets spent more time ruminating (min/kg of DMI) than did those fed GC. Although these results showed no interaction effects of RDP:RUP ratio and corn processing method on performance, higher RDP:RUP ratios and ground corn can be effective feeding strategies for feed to lactating cows receiving high-concentrate diets. Copyright © 2018 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

NASA Astrophysics Data System (ADS)

Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

2015-09-01

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

Apparent Explosion Moments from Rg Waves Recorded on SPE: Implications for the Late-Time Damage Source Model

NASA Astrophysics Data System (ADS)

Patton, H. J.; Larmat, C. S.; Rougier, E.

2016-12-01

Seismic moments for chemical shots making up Phase I of the Source Physics Experiments (SPE) are estimated from 6 Hz Rg waves under the assumption that the shots are pure explosions. These apparent explosion moments are compared to moments determined using the Reduced Displacement Potential (RDP) method applied to free field data. LIDAR/photogrammetry observations, strong ground motions on the free surface near ground zero, and moment tensor inversion results are evidence in support of the fourth shot SPE-4P being essentially a pure explosion. The apparent moment for SPE-4P is 9 × 1010 Nm in good agreement with the RDP moment 8 × 1010 Nm. In stark contrast, apparent moments for the first three shots are three to four times smaller than RDP moments. Data show that spallation occurred on these shots, as well as permanent deformations detected with ground-based LIDAR. As such, the source medium suffered late-time damage. The late-time damage source model predicts destructive interference between Rg waves radiated by explosion and damage sources, which reduces amplitudes and explains why apparent moments are smaller than RDP moments based on compressional energy emitted directly from the source. SPE-5 was conducted at roughly the same yield-scaled burial depth as SPE-2 and -3, but with five times the yield. As such, the damage source model predicts less reduction of apparent moment. At this writing, preliminary results from Rg interferometry and RDP moments confirm this prediction. SPE-6 is scheduled for the fall of 2016, and it should have the strongest damage source of all SPE shots. The damage model predicts that the polarity of Rg waves could be reversed. Realization of this prediction will be strong confirmation of the late-time damage source model. This abstract has a Los Alamos National Laboratory Unlimited Release Number LA-UR-16-25709.

Using Divertor Strike Point Splitting to Study Plasma Response and Its Sensitivity to Equilibrium Uncertainties

NASA Astrophysics Data System (ADS)

Teklu, Abraham; Orlov, D. M.; Moyer, R. A.; Bykov, I.; Evans, T. E.; Wu, W.; Trevisan, G. L.; Lyons, B. C.; Abrams, T.; Makowski, M. A.; Lasnier, C. S.; Fenstermacher, M. E.

2017-10-01

Resonant magnetic perturbations (RMPs) from 3D coils have been varied to modify the splitting of the divertor strike points in DIII-D. This splitting is imaged in filtered visible and infrared emission from the divertor to determine the particle and heat flux patterns on the target plates. The observed splitting is compared to vacuum and plasma response modeling in discharges where a subset of the RMP coils were ramped to shift the divertor footprints from dominantly n = 3 to n = 2 pattern. These results will be used to determine if the plasma response model can be validated with the measured splitting. We will also study the sensitivity of the modeled splitting to details of the 2D equilibrium. This RMP ramp technique could be used in ITER to spread out the heat flux while avoiding excessive forces on the RMP coils. Work supported by U.S. DOE under the Science Undergraduate Laboratory Internship (SULI) program and DE-FC02-04ER54698, DE-FG02-07ER54917, DE-FG02-05ER54809 and DE-AC52-07NA27344.

Distinguishing Internet-facing ICS Devices Using PLC Programming Information

DTIC Science & Technology

2014-06-19

the course of this research. I would also like to thank the faculty and students at AFIT who helped me think through some of these problems. Finally, I...21 FTP 143 IMAP 1900 UPnP 6379 Redis 22 SSH 161 SNMP 2323 Telnet 7777 Oracle 23 Telnet 443 HTTPS 3306 MySQL 8000 Qconn 25 SMTP 445 SMB 3389 RDP 8080...250ms over the course of 10,000 samples provided sufficient data to test for statistically significant changes to ladder logic execution times with a

Engineering Biodegradable Flame Retardant Wood-Plastic Composites

NASA Astrophysics Data System (ADS)

Zhang, Linxi

Wood-plastic composites (WPCs), which are produced by blending wood and polymer materials, have attracted increasing attentions in market and industry due to the low cost and excellent performance. In this research, we have successfully engineered WPC by melt blending Polylactic Acid (PLA) and Poly(butylene adipate-co-terphthalate) (PBAT) with recycled wood flour. The thermal property and flammability of the composite are significantly improved by introducing flame retardant agent resorcinol bis(biphenyl phosphate) (RDP). The mechanical and morphological properties are also investigated via multiple techniques. The results show that wood material has increased toughness and impact resistance of the PLA/PBAT polymer matrix. SEM images have confirmed that PLA and PBAT are immiscible, but the incompatibility is reduced by the addition of wood. RDP is initially dispersed in the blends evenly. It migrates to the surface of the sample after flame application, and serves as a barrier between the fire and underlying polymers and wood mixture. It is well proved in the research that RDP is an efficient flame retardant agent in the WPC system.

Advanced divertor configurations with large flux expansion

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; McLean, A.; Menard, J. E.; Paul, S. F.; Podesta, M.; Raman, R.; Ryutov, D. D.; Scotti, F.; Kaita, R.; Maingi, R.; Mueller, D. M.; Roquemore, A. L.; Reimerdes, H.; Canal, G. P.; Labit, B.; Vijvers, W.; Coda, S.; Duval, B. P.; Morgan, T.; Zielinski, J.; De Temmerman, G.; Tal, B.

2013-07-01

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3-7 MW/m2 to 0.5-1 MW/m2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L-H power threshold, enhanced stability of the peeling-ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2-3) Type I ELM frequency and slightly increased (20-30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX and TCV experiments are providing support for the snowflake divertor as a viable solution for the outstanding tokamak plasma-material interface issues.

Snowflake divertor configuration studies for NSTX-Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V A

2011-11-12

Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand controlmore » of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.« less

Effects of supplemental ruminally degradable protein versus increasing amounts of supplemental ruminally undegradable protein on site and extent of digestion and ruminal characteristics in lambs fed low-quality forage.

PubMed

Atkinson, R L; Toone, C D; Ludden, P A

2007-12-01

Four ruminally and duodenally cannulated Suffolk wether lambs (34.5 +/- 2 kg initial BW) were used in a 4 x 4 Latin square designed experiment to compare effects of supplemental ruminally degradable protein (RDP) vs. increasing amounts of supplemental ruminally undegradable protein (RUP) on ruminal characteristics and site and extent of digestion in lambs. Lambs were fed a basal diet of crested wheatgrass hay (4.2% CP) for ad libitum consumption, plus 1 of 4 protein supplements: isolated soy protein (RDP source) fed to meet estimated RDP requirements assuming a microbial efficiency of 11% of TDN (CON) or corn gluten meal (RUP source) fed at 50, 100, or 150% of the supplemental N provided by CON (C50, C100, and C150, respectively). Neither NDF nor ADF intake was affected (P >/= 0.18) by protein degradability, but they increased or tended to increase (P /= 0.26) for CON and C100, but increased (P /= 0.33) by protein degradability. However, true ruminal N digestibility was greater (P = 0.03) for CON compared with C100. Ruminal ammonia concentrations were greater (P = 0.002) for CON compared with C100 lambs, and increased (P = 0.001) with increasing RUP. Microbial N flows were not affected (P >/= 0.12) by protein degradability or increasing RUP. Likewise, neither ruminal urease activity (P >/= 0.11) nor microbial efficiency (P >/= 0.50) were affected by protein degradability or level of RUP. Total tract OM, NDF, and ADF digestibility was greater (P

Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-06-02

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike pointmore » was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower n e, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at P SOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).« less

Developing physics basis for the snowflake divertor in the DIII-D tokamak

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2018-02-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies in the NSTX and DIII-D tokamaks and contribute to the physics basis of the SF divertor as a power exhaust concept for future tokamaks.« less

Developing physics basis for the snowflake divertor in the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies in the NSTX and DIII-D tokamaks and contribute to the physics basis of the SF divertor as a power exhaust concept for future tokamaks.« less

Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-11-16

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment ofmore » the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (see standard divertor).« less

Developing physics basis for the snowflake divertor in the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies conducted in high-power H-mode discharges in the NSTX and DIII-D tokamaks, and, along with snowflake divertor results from TCV and other tokamaks, contribute to the physics basis of the SF divertor as a power exhaust concept for future high power density tokamaks.

Studies in Intelligence. Volume 58, Number 4 (Unclassified Articles from December 2014)

DTIC Science & Technology

2014-12-01

Hastert addressing special joint ses- sion of the Senate and House in New York City’s Federal Hall on 6 September 2012. (Photo © Martin H. Simon/Corbis...RDP80M01133A000800110003-6. 11. Ibid. 12. Memorandum for Dr. Martin , “Displaying Uncertainty to Decisionmakers,” 28 June 1973, CIA-RDP84B00506R000100080023-2...from Truman to Obama, by Peter Kross. The Zhivago Affair: The Kremlin, the CIA, and the Battle Over a Forbidden Book, by Peter Finn and Petra Couvée

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Si, Hang; Guo, Houyang Y.; Covele, Brent

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment frommore » $$1.18\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$ to $$0.88\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$. Furthermore, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of $$0.67\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$, thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.« less

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

DOE PAGES

Si, Hang; Guo, Houyang Y.; Covele, Brent; ...

2018-04-04

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment frommore » $$1.18\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$ to $$0.88\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$. Furthermore, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of $$0.67\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$, thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.« less

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

NASA Astrophysics Data System (ADS)

Si, H.; Guo, H. Y.; Covele, B.; Leonard, A. W.; Watkins, J. G.; Thomas, D.; Ding, R.

2018-05-01

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment from 1.18× {{10}19} {{m}-3} to 0.88× {{10}19} {{m}-3} . Moreover, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of 0.67× {{10}19} {{m}-3} , thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.

A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

DOE PAGES

Soukhanovskii, V. A.

2017-04-28

The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertormore » configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.« less

A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.

The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertormore » configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

We conducted physics experiments: record normalized {Beta} = 4.9 achieved in VH-mode, {Beta} limits of ITER-like configurations evaluated, FWCD commissioning. The tokamak vessel was opened to atmosphere for six weeks and a number of key diagnostics for understanding the divertor were installed. The DIII-D Advisory Committee met in January to review the DIII-D program and plan. They commended us for recent progress and supported the vanadium divertor design. The U.S./Japan DIII-D steering committee met and recommended extending the agreement to the year 2000. The field work proposal for FY 96/97 was presented in Washington on March 29, 1995. A reviewmore » of the DIII-D plan to install vanadium structural components as part of the new radiative divertor modification was held in Washington 31, 1995 and the panel endorsed the plans. Preliminary plans were developed with PPPL for collaborations in FY96,« less

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

NASA Astrophysics Data System (ADS)

Frerichs, H.; Schmitz, O.; Covele, B.; Feng, Y.; Guo, H. Y.; Hill, D.

2018-05-01

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

Pre-irradiation testing of actively cooled Be-Cu divertor modules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Duwe, R.; Kuehnlein, W.

1995-09-01

A set of neutron irradiation tests is prepared on different plasma facing materials (PFM) candidates and miniaturized components for ITER. Beside beryllium the irradiation program which will be performed in the High Flux Reactor (HFR) in Petten, includes different carbon fiber composites (CFQ) and tungsten alloys. The target values for the neutron irradiation will be 0.5 dpa at temperatures of 350{degrees}C and 700{degrees}C, resp.. The post irradiation examination (PIE) will cover a wide range of mechanical tests; in addition the degradation of thermal conductivity will be investigated. To determine the high heat flux (HHF) performance of actively cooled divertor modules,more » electron beam tests which simulate the expected heat loads during the operation of ITER, are scheduled in the hot cell electron beam facility JUDITH. These tests on a selection of different actively cooled beryllium-copper and CFC-copper divertor modules are performed before and after neutron irradiation; the pre-irradiation testing is an essential part of the program to quantify the zero-fluence high heat flux performance and to detect defects in the modules, in particular in the brazed joints.« less

Advanced Design Program (ARIES) Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tillack, Mark

2016-02-16

Progress is reported for the ARIES 3-year research program at UC San Diego, including three main tasks: 1. Completion of ARIES research on PMI/PFC issues. 2. Detailed engineering design and analysis of divertors and first wall/blankets. 3. Mission & requirements of FNSF.

Influence of impurity seeding on the plasma radiation in the EAST tokamak

NASA Astrophysics Data System (ADS)

Liping, DONG; Yanmin, DUAN; Kaiyun, CHEN; Xiuda, YANG; Ling, ZHANG; Feng, XU; Jingbo, CHEN; Songtao, MAO; Zhenwei, WU; Liqun, HU

2018-04-01

Plasma radiation characteristics in EAST argon (Ar) gas and neon (Ne) gas seeding experiments are studied. The radiation profiles reconstructed from the fast bolometer measurement data by tomography method are compared with the ones got from the simulation program based on corona model. And the simulation results coincide roughly with the experimental data. For Ar seeding discharges, the substantial enhanced radiations can be generally observed in the edge areas at normalized radius ρ pol∼0.7–0.9, while the enhanced regions are more outer for Ne seeding discharges. The influence of seeded Ar gas on the core radiation is related to the injected position. In discharges with LSN divertor configuration, the Ar ions can permeate into the core region more easily when being injected from the opposite upper divertor ports. In USN divertor configuration, the W impurity sputtered from the upper divertor target plates are observed to be an important contributor to the increase of the core radiation no matter impurity seeding from any ports. The maximum radiated power fractions f rad (P rad/P heat) about 60%–70% have been achieved in the recent EAST experimental campaign in 2015–2016.

Two-point modeling of SOL losses of HHFW power in NSTX

NASA Astrophysics Data System (ADS)

Kish, Ayden; Perkins, Rory; Ahn, Joon-Wook; Diallo, Ahmed; Gray, Travis; Hosea, Joel; Jaworski, Michael; Kramer, Gerrit; Leblanc, Benoit; Sabbagh, Steve

2017-10-01

High-harmonic fast-wave (HHFW) heating is a heating and current-drive scheme on the National Spherical Torus eXperiment (NSTX) complimentary to neutral beam injection. Previous experiments suggest that a significant fraction, up to 50%, of the HHFW power is promptly lost to the scrape-off layer (SOL). Research indicates that the lost power reaches the divertor via wave propagation and is converted to a heat flux at the divertor through RF rectification rather than heating the SOL plasma at the midplane. This counter-intuitive hypothesis is investigated using a simplified two-point model, relating plasma parameters at the divertor to those at the midplane. Taking measurements at the divertor region of NSTX as input, this two-point model is used to predict midplane parameters, using the predicted heat flux as an indicator of power input to the SOL. These predictions are compared to measurements at the midplane to evaluate the extent to which they are consistent with experiment. This work was made possible by funding from the Department of Energy for the Summer Undergraduate Laboratory Internship (SULI) program. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

Effects of varying ruminally undegradable protein supplementation on forage digestion, nitrogen metabolism, and urea kinetics in Nellore cattle fed low-quality tropical forage.

PubMed

Batista, E D; Detmann, E; Titgemeyer, E C; Valadares Filho, S C; Valadares, R F D; Prates, L L; Rennó, L N; Paulino, M F

2016-01-01

Effects of supplemental RDP and RUP on nutrient digestion, N metabolism, urea kinetics, and muscle protein degradation were evaluated in Nellore heifers () consuming low-quality signal grass hay (5% CP and 80% NDF, DM basis). Five ruminally and abomasally cannulated Nellore heifers (248 ± 9 kg) were used in a 5 × 5 Latin square. Treatments were the control (no supplement) and RDP supplementation to meet 100% of the RDP requirement plus RUP provision to supply 0, 50, 100, or 150% of the RUP requirement. Supplemental RDP (casein plus NPN) was ruminally dosed twice daily, and RUP supply (casein) was continuously infused abomasally. Jugular infusion of [NN]-urea with measurement of enrichment in urine was used to evaluate urea kinetics. The ratio of urinary 3-methylhistidine to creatinine was used to estimate skeletal muscle protein degradation. Forage NDF intake (2.48 kg/d) was not affected ( ≥ 0.37) by supplementation, but supplementation did increase ruminal NDF digestion ( < 0.01). Total N intake (by design) and N retention increased ( < 0.001) with supplementation and also linearly increased with RUP provision. Urea entry rate and gastrointestinal entry rate of urea were increased by supplementation ( < 0.001). Supplementation with RUP linearly increased ( = 0.02) urea entry rate and tended ( = 0.07) to linearly increase gastrointestinal entry rate of urea. Urea use for anabolic purposes tended ( = 0.07) to be increased by supplementation, and RUP provision also tended ( = 0.08) to linearly increase the amount of urea used for anabolism. The fraction of recycled urea N incorporated into microbial N was greater ( < 0.001) for control (22%) than for supplemented (9%) heifers. Urinary 3-methylhistidine:creatinine of control heifers was more than double that of supplemented heifers ( < 0.001). Control heifers reabsorbed a greater ( < 0.001) fraction of urea from the renal tubule than did supplemented heifers. Overall, unsupplemented heifers had greater mobilization of AA from myofibrillar protein, which provided N for urea synthesis and subsequent recycling. Supplemental RUP, when RDP was supplied, not only increased N retention but also supported increased urea N recycling and increased ruminal microbial protein synthesis.

Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

2013-10-01

Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical "metric," the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes.

Plasma detachment in divertor tokamaks

NASA Astrophysics Data System (ADS)

Leonard, A. W.

2018-04-01

Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E× B drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

SOLPS modeling of the effect on plasma detachment of closing the lower divertor in DIII-D

DOE PAGES

Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; ...

2016-12-15

SOLPS modeling has been carried out to assess the effect of tightly closing the lower divertor in DIII-D, which at present is almost fully open, on the achievement of cold dissipative/detached divertor conditions. To isolate the impact of other factors on the divertor plasma solution and to make direct comparisons, most of the parameters including the meshes were kept as similar as possible. Only the neutral baffling was modified to compare a fully open divertor with a tightly closed one. The modeling shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchangemore » losses in the divertor and reducing the electron temperature T et and deposited power density q dep at the target plate. Furthermore, the closed structure enables the divertor plasma to enter into highly dissipative and detached divertor conditions at a significantly lower upstream density. The effects of divertor closure on the neutral density and pressure, and their correlation with the divertor plasma conditions are also demonstrated. The effect of molecular D 2- ion D + elastic collisions and neutral-neutral collisions on the divertor plasma solution are assessed.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lumsdaine, A.; Bjorholm, T.; Harris, J.

The Wendelstein 7-X stellarator is in final stages of commissioning, and will begin operation in late 2015. In the first phase, the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current will evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A highmore » heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a test divertor unit scraper element (TDU-SE) in 2017 during an earlier operational phase. Several U.S. fusion energy science laboratories have been involved in the design, analysis (structural and thermal finite element, as well as computational fluid dynamics), plasma simulation, planning, prototyping, and diagnostic development around the scraper element program (both TDU-SE and HHF-SE). As a result, this paper presents an overview of all of these activities and their current status.« less

Overview of design and analysis activities for the W7-X scraper element

DOE PAGES

Lumsdaine, A.; Bjorholm, T.; Harris, J.; ...

2016-08-18

The Wendelstein 7-X stellarator is in final stages of commissioning, and will begin operation in late 2015. In the first phase, the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current will evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A highmore » heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a test divertor unit scraper element (TDU-SE) in 2017 during an earlier operational phase. Several U.S. fusion energy science laboratories have been involved in the design, analysis (structural and thermal finite element, as well as computational fluid dynamics), plasma simulation, planning, prototyping, and diagnostic development around the scraper element program (both TDU-SE and HHF-SE). As a result, this paper presents an overview of all of these activities and their current status.« less

Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

NASA Astrophysics Data System (ADS)

Bathke, C. G.; Krakowski, R. A.; Miller, R. L.

Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line trackings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented.

Plasma detachment in divertor tokamaks

DOE PAGES

Leonard, A. W.

2018-02-07

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Plasma detachment in divertor tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leonard, A. W.

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Divertor scenario development for NSTX Upgrade

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; McLean, A. G.; Meier, E. T.; Rognlien, T. D.; Ryutov, D. D.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; Kaita, R.; Kolemen, E.; Leblanc, B. P.; Menard, J. E.; Podesta, M.; Scotti, F.

2012-10-01

In the NSTX-U tokamak, initial plans for divertor plasma-facing components (PFCs) include lithium and boron coated graphite, with a staged transition to molybdenum. Steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m^2 in 2 MA, 12 MW NBI-heated discharges of up to 5 s duration, thus challenging PFC thermal limits. Based on the recent NSTX divertor experiments and modeling with edge transport code UEDGE, a favorable basis for divertor power handling in NSTX-U is developed. The snowflake divertor geometry and feedback-controlled divertor impurity seeding applied to the lower and upper divertors are presently envisioned. In the NSTX snowflake experiments with lithium-coated graphite PFCs, the peak divertor heat fluxes from Type I ELMs and between ELMs were significantly reduced due to geometry effects, increased volumetric losses and null-point convective redistribution between strike points. H-mode core confinement was maintained at H98(y,2)<=1 albeit the radiative detachment. Additional CD4 seeding demonstrated potential for a further increase of divertor radiation.

Small angle slot divertor concept for long pulse advanced tokamaks

NASA Astrophysics Data System (ADS)

Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

2017-04-01

SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

Detachment experiments in new DIII-D upper divertor

NASA Astrophysics Data System (ADS)

Moser, A. L.; Leonard, A. W.; Groebner, R. J.; Guo, H.; Wang, H.; Watkins, J. G.; McLean, A. G.; Fenstermacher, M. E.; Shafer, M. W.; Briesemeister, A. R.; Hinson, E. T.

2017-10-01

Installation of the Small Angle Slot (SAS) in the upper divertor of DIII-D enables new studies of the effect of target and baffle geometry on divertor detachment. This structure provides a more-closed upper divertor as well as the SAS divertor itself. Initial SAS experiment results indicate that divertor detachment occurs at a lower line-averaged density than in the more-open, lower single null divertor configurations on DIII-D. In contrast, the increased divertor closure of the new installation did not reduce the upstream density required for detachment beyond that achieved with the previous upper divertor structure. Particle pumping in the upper divertor structure is found to produce a 10 % reduction in the pedestal density required for detachment compared to the case with no pumping. Comparisons focus on both the onset of detachment (measured by in-target Langmuir probes) as a function of upstream density, as well as the effect of the new divertor configurations on pedestal density profiles. Work supported by US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-SC00013911.

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

DOE PAGES

Frerichs, H.; Schmitz, O.; Covele, B.; ...

2018-02-28

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the diverter slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frerichs, H.; Schmitz, O.; Covele, B.

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the diverter slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

Magnetic configuration flexibility of snowflake divertor for HL-2M [Analysis of snowflake divertor configurations for HL-2M

DOE PAGES

Zheng, G. Y.; Xu, X. Q.; Ryutov, D. D.; ...

2014-07-09

HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reducemore » the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. Furthermore, a new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.« less

Probabilistic topic modeling for the analysis and classification of genomic sequences

PubMed Central

2015-01-01

Background Studies on genomic sequences for classification and taxonomic identification have a leading role in the biomedical field and in the analysis of biodiversity. These studies are focusing on the so-called barcode genes, representing a well defined region of the whole genome. Recently, alignment-free techniques are gaining more importance because they are able to overcome the drawbacks of sequence alignment techniques. In this paper a new alignment-free method for DNA sequences clustering and classification is proposed. The method is based on k-mers representation and text mining techniques. Methods The presented method is based on Probabilistic Topic Modeling, a statistical technique originally proposed for text documents. Probabilistic topic models are able to find in a document corpus the topics (recurrent themes) characterizing classes of documents. This technique, applied on DNA sequences representing the documents, exploits the frequency of fixed-length k-mers and builds a generative model for a training group of sequences. This generative model, obtained through the Latent Dirichlet Allocation (LDA) algorithm, is then used to classify a large set of genomic sequences. Results and conclusions We performed classification of over 7000 16S DNA barcode sequences taken from Ribosomal Database Project (RDP) repository, training probabilistic topic models. The proposed method is compared to the RDP tool and Support Vector Machine (SVM) classification algorithm in a extensive set of trials using both complete sequences and short sequence snippets (from 400 bp to 25 bp). Our method reaches very similar results to RDP classifier and SVM for complete sequences. The most interesting results are obtained when short sequence snippets are considered. In these conditions the proposed method outperforms RDP and SVM with ultra short sequences and it exhibits a smooth decrease of performance, at every taxonomic level, when the sequence length is decreased. PMID:25916734

The expanding spectrum of neurological phenotypes in children with ATP1A3 mutations, Alternating Hemiplegia of Childhood, Rapid-onset Dystonia-Parkinsonism, CAPOS and beyond.

PubMed

Sweney, Matthew T; Newcomb, Tara M; Swoboda, Kathryn J

2015-01-01

ATP1A3 mutations have now been recognized in infants and children presenting with a diverse group of neurological phenotypes, including Rapid-onset Dystonia-Parkinsonism (RDP), Alternating Hemiplegia of Childhood (AHC), and most recently, Cerebellar ataxia, Areflexia, Pes cavus, Optic atrophy, and Sensorineural hearing loss (CAPOS) syndrome. Existing literature on ATP1A3-related disorders in the pediatric population were reviewed, with attention to clinical features and associated genotypes among those with RDP, AHC, or CAPOS syndrome phenotypes. While classically defined phenotypes associated with AHC, RDP, and CAPOS syndromes are distinct, common elements among ATP1A3-related neurological disorders include characteristic episodic neurological symptoms and signs that vary in severity, duration, and frequency of occurrence. Affected children typically present in the context of an acute onset of paroxysmal, episodic neurological symptoms ranging from oculomotor abnormalities, hypotonia, paralysis, dystonia, ataxia, seizure-like episodes, or encephalopathy. Neurodevelopmental delays or persistence of dystonia, chorea, or ataxia after resolution of an initial episode are common, providing important clues for diagnosis. The phenotypic spectrum of ATP1A3-related neurological disorders continues to expand beyond the distinct yet overlapping phenotypes in patients with AHC, RDP, and CAPOS syndromes. ATP1A3 mutation analysis is appropriate to consider in the diagnostic algorithm for any child presenting with episodic or fluctuating ataxia, weakness or dystonia whether they manifest persistence of neurological symptoms between episodes. Additional work is needed to better identify and classify affected patients and develop targeted treatment approaches. Copyright © 2015 Elsevier Inc. All rights reserved.

Pseudotyped adeno-associated virus 2/9-delivered CCL11 shRNA alleviates lung inflammation in an allergen-sensitized mouse model.

PubMed

Wu, Chia-Jen; Huang, Wen-Chung; Chen, Li-Chen; Shen, Chia-Rui; Kuo, Ming-Ling

2012-11-01

Airway infiltration by eosinophils is a major characteristic of chronic asthma. CCL11 (eotaxin-1) is secreted by lung epithelial cells and functions as the major chemokine for eosinophil recruitment. Pseudotyped adeno-associated virus (AAV) 2/9, composed by the AAV2 rep and AAV9 cap genes, can efficiently target lung epithelial cells and might carry gene sequences with therapeutic potential for asthma. This study aimed to determine whether pseudotyped AAV2/9 virus carrying the small hairpin RNA targeting CCL11 and expressed by CMV/U6 promoter could reduce eosinophilia and asthmatic responses in mite allergen-sensitized mice. Mice were sensitized by intraperitoneal and challenged by intratracheal injection with recombinant Dermatophagoides pteronyssinus group 2 allergen (rDp2). AAV2/9 viral vectors were intratracheally injected three days before the first challenge. AAV2/9 sh47 virus significantly reduced airway hyperresponsiveness, airway resistance, CCL11 levels, and eosinophilia in the lungs of sensitized mice. Th2 cytokines, including interleukins (IL)-4, IL-5, and IL-10, were also significantly reduced in the bronchoalveolar lavage fluid of AAV2/9 sh47 virus-treated mice. Th2 cytokine levels were also reduced in rDp2-stimulated mediastinal lymphocytes in treated mice. However, serum levels of rDp2-specific IgG1 and IgE, as well as Th2 cytokine levels in rDp2-stimulated splenocyte culture supernatants, were comparable to the sensitized control group. The results suggest that AAV2/9 sh47 virus relieved local instead of systemic inflammatory responses. Therefore, the CMV/U6 promoter with AAV2/9 viral vector, which is preferable to target lung epithelia cells, might be applied as a novel therapeutic approach for asthma.

Pseudotyped Adeno-Associated Virus 2/9-Delivered CCL11 shRNA Alleviates Lung Inflammation in an Allergen-Sensitized Mouse Model

PubMed Central

Wu, Chia-Jen; Huang, Wen-Chung; Chen, Li-Chen; Shen, Chia-Rui

2012-01-01

Abstract Airway infiltration by eosinophils is a major characteristic of chronic asthma. CCL11 (eotaxin-1) is secreted by lung epithelial cells and functions as the major chemokine for eosinophil recruitment. Pseudotyped adeno-associated virus (AAV) 2/9, composed by the AAV2 rep and AAV9 cap genes, can efficiently target lung epithelial cells and might carry gene sequences with therapeutic potential for asthma. This study aimed to determine whether pseudotyped AAV2/9 virus carrying the small hairpin RNA targeting CCL11 and expressed by CMV/U6 promoter could reduce eosinophilia and asthmatic responses in mite allergen-sensitized mice. Mice were sensitized by intraperitoneal and challenged by intratracheal injection with recombinant Dermatophagoides pteronyssinus group 2 allergen (rDp2). AAV2/9 viral vectors were intratracheally injected three days before the first challenge. AAV2/9 sh47 virus significantly reduced airway hyperresponsiveness, airway resistance, CCL11 levels, and eosinophilia in the lungs of sensitized mice. Th2 cytokines, including interleukins (IL)-4, IL-5, and IL-10, were also significantly reduced in the bronchoalveolar lavage fluid of AAV2/9 sh47 virus-treated mice. Th2 cytokine levels were also reduced in rDp2-stimulated mediastinal lymphocytes in treated mice. However, serum levels of rDp2-specific IgG1 and IgE, as well as Th2 cytokine levels in rDp2-stimulated splenocyte culture supernatants, were comparable to the sensitized control group. The results suggest that AAV2/9 sh47 virus relieved local instead of systemic inflammatory responses. Therefore, the CMV/U6 promoter with AAV2/9 viral vector, which is preferable to target lung epithelia cells, might be applied as a novel therapeutic approach for asthma. PMID:22913580

Overview of Compact Toroidal Hybrid research program progress and plans

NASA Astrophysics Data System (ADS)

Maurer, David; Ennis, David; Hanson, James; Hartwell, Gregory; Herfindal, Jeffrey; Knowlton, Stephen; Ma, Xingxing; Pandya, Mihir; Roberds, Nicholas; Ross, Kevin; Traverso, Peter

2016-10-01

disruptive behavior on the level of applied 3D magnetic shaping; (2) test and advance the V3FIT reconstruction code and NIMROD modeling of CTH; and (3) study the implementation of an island divertor. Progress towards these goals and other developments are summarized. The disruptive density limit exceeds the Greenwald limit as the vacuum transform is increased, but a threshold for avoidance is not observed. Low- q disruptions, with 1.1 < q (a) <2.0, cease to occur if the vacuum transform is raised above 0.07. Application of vacuum transform can reduce and eliminate the vertical drift of elongated discharges that would otherwise be vertically unstable. Reconstructions using external magnetics give accurate estimates for quantities near the plasma boundary, and internal diagnostics have been implemented to extend the range of accuracy into the plasma core. Sawtooth behavior has been reproducibly modified with external transform and NIMROD is used to model these observations and reproduces experimental trends. An island divertor design has begun with connection length studies to model energy deposition on divertor plates located in an edge 1/3 island as well as the study of a non-resonant divertor configuration. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

DOE PAGES

Jaworski, M. A.; Brooks, A.; Kaita, R.; ...

2016-08-08

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physicsmore » and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.« less

Advanced Divertor Design and Application under Modern Superconducting Tokamak Constraints

NASA Astrophysics Data System (ADS)

Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh; Valanju, Prashant

2013-10-01

With current ITER projections already predicting divertor exhaust heat loads in the 5-10 MW/m2 range, i.e. at the maximum tolerance, it is clear that the divertor heat load problem will only be exacerbated for future superconducting tokamaks, as well as perhaps some modern tokamaks today. Thus, an advanced divertor, such as the X-Divertor (XD), Super-X Divertor (SXD), or Snowflake (SF) will become a virtual necessity to reduce incident heat flux at the target plates. Using the 2D magnetic equilibrium code CORSICA, we explore the possibilities of creating an advanced divertor for a next-generation superconducting tokamak (Ip = 15 MA, BT = 5.3 T, R = 6.2 m) under nominal engineering constraints. Advanced divertors were achieved with no in-vessel PF coils, PF current densities below 30 MA/m2, and vertical maintenance access, all of which are favorable conditions for tokamaks today. Both the XD and SF divertors are readily achievable while maintaining core plasma performance, and the advantages and disadvantages of each are discussed in turn. Some thought is given as to how the divertor cassette will need to be modified to accommodate advanced divertors. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

Investigations on the heat flux and impurity for the HL-2M divertor

NASA Astrophysics Data System (ADS)

Zheng, G. Y.; Cai, L. Z.; Duan, X. R.; Xu, X. Q.; Ryutov, D. D.; Cai, L. J.; Liu, X.; Li, J. X.; Pan, Y. D.

2016-12-01

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with {{I}\\text{p}}   =  0.5 MA at the first stage of HL-2M and the high parameters with {{I}\\text{p}}   =  2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with {{d}x}=30 \\text{cm} present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, {{d}x} , to achieve a better performance in terms of the parameters of discharges.

"Snowflake" divertor configuration in NSTX

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.

2011-08-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

NASA Astrophysics Data System (ADS)

Frerichs, Heinke; Schmitz, Oliver; Covele, Brent; Guo, Houyang; Hill, David; Feng, Yuhe

2017-10-01

In the Small Angle Slot (SAS) divertor in DIII-D, the combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field causes the strike point to vary radially along the divertor slot and even leave it at some toroidal locations. This effect essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade performance of the slot divertor. This effect has been approximated by a finite gap in the divertor baffle. Simulations with EMC3-EIRENE show that a toroidally localized loss of divertor closure can result in non-axisymmetric divertor densities and temperatures. This introduces a density window of 10-15% on top of the nominal threshold separatrix density during which a non-axisymmetric onset of local detachment occurs, initially leaving the gap and up to 60 deg beyond that still attached. Conversely, the impact of such toroidally localized divertor perturbations on the toroidal symmetry of midplane separatrix conditions is small. This work has been funded by the U.S. Department of Energy under Early Career Award Grant DE-SC0013911, and Grant DE-FC02-04ER54698.

Mortality (1950–1999) and cancer incidence (1969–1999) of workers in the Port Hope cohort study exposed to a unique combination of radium, uranium and γ-ray doses

PubMed Central

Zablotska, Lydia B; Lane, Rachel S D; Frost, Stanley E

2013-01-01

Objectives Uranium processing workers are exposed to uranium and radium compounds from the ore dust and to γ-ray radiation, but less to radon decay products (RDP), typical of the uranium miners. We examined the risks of these exposures in a cohort of workers from Port Hope radium and uranium refinery and processing plant. Design A retrospective cohort study with carefully documented exposures, which allowed separation of those with primary exposures to radium and uranium. Settings Port Hope, Ontario, Canada, uranium processors with no mining experience. Participants 3000 male and female workers first employed (1932–1980) and followed for mortality (1950–1999) and cancer incidence (1969–1999). Outcome measures Cohort mortality and incidence were compared with the general Canadian population. Poisson regression was used to evaluate the association between cumulative RDP exposures and γ-ray doses and causes of death and cancers potentially related to radium and uranium processing. Results Overall, workers had lower mortality and cancer incidence compared with the general Canadian population. In analyses restricted to men (n=2645), the person-year weighted mean cumulative RDP exposure was 15.9 working level months (WLM) and the mean cumulative whole-body γ-ray dose was 134.4 millisieverts. We observed small, non-statistically significant increases in radiation risks of mortality and incidence of lung cancer due to RDP exposures (excess relative risks/100 WLM=0.21, 95% CI <−0.45 to 1.59 and 0.77, 95% CI <−0.19 to 3.39, respectively), with similar risks for those exposed to radium and uranium. All other causes of death and cancer incidence were not significantly associated with RDP exposures or γ-ray doses or a combination of both. Conclusions In one of the largest cohort studies of workers exposed to radium, uranium and γ-ray doses, no significant radiation-associated risks were observed for any cancer site or cause of death. Continued follow-up and pooling with other cohorts of workers exposed to by-products of radium and uranium processing could provide valuable insight into occupational risks and suspected differences in risk with uranium miners. PMID:23449746

Comparative divertor-transport study for helical devices

NASA Astrophysics Data System (ADS)

Feng, Y.; Kobayashi, M.; Sardei, F.; Masuzaki, S.; Kisslinger, J.; Morisaki, T.; Grigull, P.; Yamada, H.; McCormick, K.; Ohyabu, N.; König, R.; Yamada, I.; Giannone, L.; Narihara, K.; Wenzel, U.; Morita, S.; Thomsen, H.; Miyazawa, J.; Hildebrandt, D.; Watanabe, T.; Wagner, F.; Ashikawa, N.; Ida, K.; Komori, A.; Motojima, O.; Nakamura, Y.; Peterson, B. J.; Sato, K.; Shoji, M.; Tamura, N.; Tokitani, M.; LHD experimental Group

2009-09-01

Using the island divertors (IDs) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine sizes following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. The fundamental role of low-order magnetic islands in both divertor concepts is emphasized. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime, which is absent from W7-AS and LHD, is predicted to exist for W7-X. The paper focuses on identifying and understanding the role of divertors for high density plasma operations in helical devices. In this regard, special attention is paid to investigating the divertor function for controlling intrinsic impurities. Impurity transport behaviour and wall-sputtering processes of CX-neutrals are studied under different divertor plasma conditions. A divertor retention effect on intrinsic impurities at high SOL collisonalities is predicted for all the three devices. The required SOL plasma conditions and the underlying mechanisms are analysed in detail. Numerical results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. Different SOL transport regimes are numerically identified for the standard divertor configuration of W7-X and the possible consequences on high density plasmas are assessed. All the EMC3-EIRENE simulations presented in this paper are based on vacuum fields and comparisons with local diagnostics are made for low-ß plasmas.

Divertor heat flux mitigation in the National Spherical Torus Experimenta)

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Maingi, R.; Gates, D. A.; Menard, J. E.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Bell, M. G.; Bell, R. E.; Boedo, J. A.; Bush, C. E.; Kaita, R.; Kugel, H. W.; Leblanc, B. P.; Mueller, D.; NSTX Team

2009-02-01

Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

Divertor detachment

NASA Astrophysics Data System (ADS)

Krasheninnikov, Sergei

2015-11-01

The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Footprint of Field lines

NASA Astrophysics Data System (ADS)

Figgins, Montoya; Ali, Halima; Punjabi, Alkesh

2000-10-01

We are working with the Simple Map^1 to find the footprint of field lines on the diverter plate in a single-null tokamak. Footprint of a field line is the position of the line when it escapes across the divertor plate. The Simple Map represents the magnetic field in a single-null divertor tokamak. The path of a field line is given by the equations: X_n+1=X_n-kY_n(1-Y_n) and Y_n+1=Y_n+kX_n+1. In order to find the footprint, we must first find the last good surface which is Y=0.997135768 and X=0. The value of k is fixed at 0.6. The starting values X0 are fixed at X_0=0. We use 10,000 points between the last good surface and the X-point. The X-point is located at (0,1). We also use the Continuous Analog of the Simple Map given by the equations: X(φ)=X_0-kY0 (1-Y_0)φ and Y(φ)=Y_0+kX(φ)φ. This will tell us what the (φ,X) is which represents the field lines crossing the divertor plate. The divertor plate is located at Y=1. When graphed, the footprint of field lines looks like the rings of Saturn. This work is supported by US DOES OFES. Ms. Montoya Figgins is HU CFRT Summer Fusion High School Scholar from E. E. Smith High School in North Carolina. She is supported by NASA under its NASA SHARP Plus Program. 1. Punjabi A, Verma A, and Boozer A, Phys Rev Lett, 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994)

A review of direct experimental measurements of detachment

DOE PAGES

Boedo, J.; McLean, A. G.; Rudakov, D. L.; ...

2018-02-22

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. Here, we review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson Scattering (TS) in the divertor regionmore » and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.« less

A review of direct experimental measurements of detachment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boedo, J.; McLean, A. G.; Rudakov, D. L.

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. Here, we review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson Scattering (TS) in the divertor regionmore » and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.« less

A super-cusp divertor configuration for tokamaks

NASA Astrophysics Data System (ADS)

Ryutov, D. D.

2015-10-01

> This study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase's cusp divertor. It turns out that the set of remote coils can indeed produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough control that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called `a super-cusp'. General geometrical features of the three-null configurations produced by remote coils are described. Issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils.

A review of direct experimental measurements of detachment

NASA Astrophysics Data System (ADS)

Boedo, J.; McLean, A. G.; Rudakov, D. L.; Watkins, J. G.

2018-04-01

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. We review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson scattering in the divertor region and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.

Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoder Jr, Graydon L; Harvey, Karen; Ferrada, Juan J

A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

Advantages and Challenges of Radiative Liquid Lithium Divertor

NASA Astrophysics Data System (ADS)

Ono, Masayuki

2017-10-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

The effect of feedback-controlled divertor nitrogen seeding on the boundary plasma and power exhaust channel width in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Labombard, B.; Brunner, D.; Kuang, A. Q.; McCarthy, W.; Terry, J. L.

2017-10-01

The scrape-off layer (SOL) power channel width, λq, is projected to be 0.5 mm in power reactors, based on multi-machine measurements of divertor target heat fluxes in H-mode at low levels of divertor dissipation. An important question is: does λq change with the level of divertor dissipation? We report results in which feedback controlled nitrogen seeding in the divertor was used to systematically vary divertor dissipation in a series of otherwise identical L-mode plasmas at three plasma currents: 0.55, 0.8 and 1.1 MA. Outer midplane profiles were recorded with a scanning Mirror Langmuir Probe; divertor plasma conditions were monitored with `rail' Langmuir probe and surface thermocouple arrays. Despite an order of magnitude reduction in divertor target heat fluxes (q// 400 MW m-2 to 40 MW m-2) and corresponding change in divertor regime from sheath-limited through high-recycling to near-detached, the upstream electron temperature profile is found to remain unchanged or to become slightly steeper in the near SOL and to drop significantly in the far SOL. Thus heat in the SOL appears to take advantage of this impurity radiation `heat sink' in the divertor by preferentially draining via the narrow (and perhaps an increasingly narrow) λq of the near SOL. Supported by USDoE award DE-FC02-99ER54512.

Description of a Pressure Measurement Technique for Obtaining Surface Static Pressures of a Radial Turbine

NASA Technical Reports Server (NTRS)

Dicicco, L. Danielle; Nowlin, Brent C.; Tirres, Lizet

1992-01-01

The aerodynamic performance of a solid uncooled version of a cooled radial turbine was evaluated in the Small Engine Components Test Facility Turbine rig at the NASA Lewis Research Center. Specifically, an experiment was conducted to rotor surface static pressures. This was the first time surface static pressures had been measured on a radial turbine at NASA Lewis. These pressures were measured by a modified Rotating Data Package (RDP), a standard product manufactured by Scanivalve, Inc. Described here are the RDP, and the modifications that were made, as well as the checkout, installation, and testing procedures. The data presented are compared to analytical results obtained from NASA's MERIDL TSONIC BLAYER (MTSB) code.

Description of a pressure measurement technique for obtaining surface static pressures of a radial turbine

NASA Technical Reports Server (NTRS)

Dicicco, L. D.; Nowlin, Brent C.; Tirres, Lizet

1992-01-01

The aerodynamic performance of a solid uncooled version of a cooled radial turbine was evaluated in the Small Engine Components Test Facility Turbine rig at the NASA Lewis Research Center. Specifically, an experiment was conducted to rotor surface static pressures. This was the first time surface static pressures had been measured on a radial turbine at NASA Lewis. These pressures were measured by a modified Rotating Data Package (RDP), a standard product manufactured by Scanivalve, Inc. Described here are the RDP, and the modifications that were made, as well as the checkout, installation, and testing procedures. The data presented are compared to analytical results obtained from NASA's MERIDL TSONIC BLAYER (MTSB) code.

A Technical Assessment of Seismic Yield Estimation. Appendix. Part 2

DTIC Science & Technology

1981-01-01

VALurs or -log b ...... 4 4% 5 534 6 64 7 734 8 log b ..... 4.60 4.74 4.87 5.00 5.12 5.23 5.33 6 ..43 5.52 834 9 934 10 1034 11 1134 12 12,1. log b......5...305.A0. 4S. 5 0. S5. j j . - 7 0.-7S. 70. 7S. p 0. ;5,00 Tr*E (SEC! Figure 6 . Comparison of exact and analytically continued half space Rayleigh...34equivalent RDP," Te’ for the two-dimen- sional PILEDRIVER calculation is compared to two other estimates for the RDP. 23 - * a .K~ W.S - - r!L7-- 7 W

Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, W.R.; Smith, J.P.; Trester, P.W.

1997-04-01

V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding,more » is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy.« less

OEDGE modeling of plasma contamination efficiency of Ar puffing from different divertor locations in EAST

NASA Astrophysics Data System (ADS)

Pengfei, ZHANG; Ling, ZHANG; Zhenwei, WU; Zong, XU; Wei, GAO; Liang, WANG; Qingquan, YANG; Jichan, XU; Jianbin, LIU; Hao, QU; Yong, LIU; Juan, HUANG; Chengrui, WU; Yumei, HOU; Zhao, JIN; J, D. ELDER; Houyang, GUO

2018-04-01

Modeling with OEDGE was carried out to assess the initial and long-term plasma contamination efficiency of Ar puffing from different divertor locations, i.e. the inner divertor, the outer divertor and the dome, in the EAST superconducting tokamak for typical ohmic plasma conditions. It was found that the initial Ar contamination efficiency is dependent on the local plasma conditions at the different gas puff locations. However, it quickly approaches a similar steady state value for Ar recycling efficiency >0.9. OEDGE modeling shows that the final equilibrium Ar contamination efficiency is significantly lower for the more closed lower divertor than that for the upper divertor.

Phase Transition Control for High Performance Ruddlesden-Popper Perovskite Solar Cells.

PubMed

Zhang, Xu; Munir, Rahim; Xu, Zhuo; Liu, Yucheng; Tsai, Hsinhan; Nie, Wanyi; Li, Jianbo; Niu, Tianqi; Smilgies, Detlef-M; Kanatzidis, Mercouri G; Mohite, Aditya D; Zhao, Kui; Amassian, Aram; Liu, Shengzhong Frank

2018-05-01

Ruddlesden-Popper reduced-dimensional hybrid perovskite (RDP) semiconductors have attracted significant attention recently due to their promising stability and excellent optoelectronic properties. Here, the RDP crystallization mechanism in real time from liquid precursors to the solid film is investigated, and how the phase transition kinetics influences phase purity, quantum well orientation, and photovoltaic performance is revealed. An important template-induced nucleation and growth of the desired (BA) 2 (MA) 3 Pb 4 I 13 phase, which is achieved only via direct crystallization without formation of intermediate phases, is observed. As such, the thermodynamically preferred perpendicular crystal orientation and high phase purity are obtained. At low temperature, the formation of intermediate phases, including PbI 2 crystals and solvate complexes, slows down intercalation of ions and increases nucleation barrier, leading to formation of multiple RDP phases and orientation randomness. These insights enable to obtain high quality (BA) 2 (MA) 3 Pb 4 I 13 films with preferentially perpendicular quantum well orientation, high phase purity, smooth film surface, and improved optoelectronic properties. The resulting devices exhibit high power conversion efficiency of 12.17%. This work should help guide the perovskite community to better control Ruddlesden-Popper perovskite structure and further improve optoelectronic and solar cell devices. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Characteristics of the Secondary Divertor on DIII-D

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Lasnier, C. J.; Leonard, A. W.; Evans, T. E.; Pitts, R.; Stangeby, P. C.; Boedo, J. A.; Moyer, R. A.; Rudakov, D. L.

2009-11-01

In order to address a concern that the ITER secondary divertor strike plates may be insufficiently robust to handle the incident pulses of particles and energy from ELMs, we performed dedicated studies of the secondary divertor plasma and scrape-off layer (SOL). Detailed measurements of the ELM energy and particle deposition footprint on the secondary divertor target plates were made with a fast IR camera and Langmuir probes and SOL profile and transport measurements were made with reciprocating probes. The secondary divertor and SOL conditions depended on changes in the magnetic balance and the core plasma density. Larger density resulted in smaller ELMs and the magnetic balance affected how many ELM particles coupled to the secondary SOL and divertor. Particularly striking are the images from a new fast IR camera that resolve ELM heat pulses and show spiral patterns with multiple peaks during ELMs in the secondary divertor.

Divertor-leg instability for finite beta and radially-tilted divertor plate

NASA Astrophysics Data System (ADS)

Cohen, R. H.; Ryutov, D. D.

2004-11-01

Plasma in the divertor leg may experience a fast instability caused by sheath boundary conditions (BC). Perturbations cannot penetrate beyond the X point because of very strong shearing in its vicinity. Accordingly, this instability could increase cross-field transport in the divertor leg, and thereby reduce the heat load on the divertor plate, without having any appreciable negative effect on core plasma confinement. A way of describing the role of shearing in terms of the surface resistivity attributed to a ``control plane'' below the X point has recently been suggested (Contr. Plasma Phys., v. 44, p. 168, 2004). We use this BC, plus sheath BC at the divertor plate. We include effects of finite beta and of the radial tilt of the divertor plate. We optimize the radial tilt in order to maximize radial transport in divertor legs. We discuss experimental signatures of the instability: i) phase velocity and wave-numbers of the most unstable modes; ii) correlations between fluctuations of various parameters; and iii) the differences between fluctuations in the common and private flux regions.

A super-cusp divertor configuration for tokamaks

DOE PAGES

Ryutov, D. D.

2015-08-26

Our study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase’s cusp divertor. It turns out that the set of remote coils can produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough controlmore » that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called ‘a super-cusp’. General geometrical features of the three-null configurations produced by remote coils are described. Furthermore, issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils.« less

Implementation of a long leg X-point target divertor in the ARC fusion pilot plant

NASA Astrophysics Data System (ADS)

Kuang, A. Q.; Cao, N. M.; Creely, A. J.; Dennett, C. A.; Hecla, J.; Hoffman, H.; Major, M.; Ruiz Ruiz, J.; Tinguely, R. A.; Tolman, E. A.; Brunner, D.; Labombard, B.; Sorbom, B. N.; Whyte, D. G.; Grover, P.; Laughman, C.

2017-10-01

A long leg X-point target divertor geometry in a double null geometry has been implemented in the ARC pilot plant design, exploiting ARC's demountable toroidal field (TF) coils and FLiBe immersion blanket, which allow superconducting poloidal field coils to be located inside the TF coils, adequately shielded from neutrons. This new design maintains the original TF coil size, core plasma shape, and attains a tritium breedin ratio 1.08. The long leg divertor geometry provides significant advantages. Neutron transport computations indicate a factor of 10 reduction in divertor material neutron damage rate compared to the first wall, easing requirements for high heat flux components. Simulations have shown that long legged divertors are able to maintain a passively stable detachment front that stays in the divertor leg over a wide power window, in principle, responding immediately to fast changes in power exhaust. The ARC design exploits this new paradigm for divertor heat flux control: fewer concerns about coping with fast transients and a focus on neutron-tolerant diagnostics to measure and adjust detachment front locations in the outer divertor legs over long timescales.

On heat loading, novel divertors, and fusion reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, M.; Valanju, P. M.; Mahajan, S. M.; Wiley, J. C.

2007-07-01

The limited thermal power handling capacity of the standard divertors (used in current as well as projected tokamaks) is likely to force extremely high (˜90%) radiation fractions frad in tokamak fusion reactors that have heating powers considerably larger than ITER [D. J. Campbell, Phys. Plasmas 8, 2041 (2001)]. Such enormous values of necessary frad could have serious and debilitating consequences on the core confinement, stability, and dependability for a fusion power reactor, especially in reactors with Internal Transport Barriers. A new class of divertors, called X-divertors (XD), which considerably enhance the divertor thermal capacity through a flaring of the field lines only near the divertor plates, may be necessary and sufficient to overcome these problems and lead to a dependable fusion power reactor with acceptable economics. X-divertors will lower the bar on the necessary confinement to bring it in the range of the present experimental results. Its ability to reduce the radiative burden imparts the X-divertor with a key advantage. Lower radiation demands allow sharply peaked density profiles that enhance the bootstrap fraction creating the possibility for a highly increased beta for the same beta normal discharges. The X-divertor emerges as a beta-enhancer capable of raising it by up to roughly a factor of 2.

Fish diversity in the Río de la Plata and adjacent waters: an overview of environmental influences on its spatial and temporal structure.

PubMed

Jaureguizar, A J; Solari, A; Cortés, F; Milessi, A C; Militelli, M I; Camiolo, M D; Luz Clara, M; García, M

2016-07-01

The fish diversity and the main environmental factors affecting the spatial distribution of species, life history stages and community structure in the Río de la Plata (RdP) and adjacent waters are reviewed and analysed, with emphasis on the functional guild classification. The functional guild classification indicated that most species in the RdP were marine stragglers, zoobenthivores and oviparous species, although the biomass was dominated by estuarine species. Salinity had a stronger influence than temperature on the spatial pattern for all life stages, shallower and fresher waters are the preferred habitats of neonates and juveniles. During the breeding season (spring-summer), adults showed an intrusion into the inner part of RdP or to its adjacent nearshore waters from the offshore waters for spawning or mating, respectively. Variations in river discharge and wind patterns greatly affected the spatial extent of estuarine water, which ultimately influenced the domain of the main life-history stages (juveniles or adults) for both marine and estuarine fishes, as well as species and fish assemblage composition. The strong environmental gradient restricts some species and life-history stages to a particular section and defines three main fish assemblage areas. The composition of the fish assemblage is indicative of the recruitment of freshwater and marine species to the estuary in opposite ways, determined by the vertical stratification. Seasonal changes in the species composition were related to migration as a result of salinity and temperature variations and reproductive migrations to spawning and mating areas. This overview reveals that the RdP is under environmental variations that are likely to produce modifications to fish distribution and abundance that affect its fisheries. This context plus fish stock declines and changes in exploitation patterns could amplify the magnitude of the variations in the fisheries resources availability and affect the sustainability of fishing communities. © 2016 The Fisheries Society of the British Isles.

Soil information in the Strategic Environmental Assessment of Rural Development Plans

NASA Astrophysics Data System (ADS)

Costantini, Edoardo

2016-04-01

Soil information is essential for land planning at different administrative levels. Currently, in Europe there is a hierarchy of at least seven main territorial administrative entities which consider soil in their land planning policies. In this study, European, national and regional regulations that affect soil are discussed, considering themes, priorities, and focus areas. A better attention on the Strategic Environmental Assessment (SEA) of the Rural Development Plans (RDP) 2014-2020 is given, by analysing the Environmental Reports produced in the framework of the SEA of the RDP 2014-2020 of some European Regions. Both old and new soil indicators are introduced, as well as the consequences of their adoption for soil science and soil scientists. It is evident from this study that soil information is treated very variably, not only because of the different kinds of available information, but mainly as a consequence of the expertise and sensitivity to the soil issues of the authors, since they can be different kinds of public or private bodies hired by regional authorities. Therefore, despite having the same reference, the SEAs are very different. In most cases, the amount of soil information is little or negligible, even when available. The ex-ante impacts are often only qualitative, or quantified only in terms of the areal extent of applied measures. On the other hand, following the European recommendations, the ongoing and ex-post reports of the monitoring activities of RDP are expected to provide a wealth of soil information, both qualitative and quantitative. Soil scientists, soil and water conservationists, and management experts will be requested to provide more sophisticated and dynamic types of information. Therefore, the SEA of RDP and similar land planning activities will provide greater scientific and technical opportunities for soil science, provided that the public bodies in charge of evaluations, namely Member States, Regions and their auditing counterparts within the European Commission, are able to implement their own recommendations.

Overview of the EUROfusion Medium Size Tokamak scientific program

NASA Astrophysics Data System (ADS)

Bernert, Matthias; Bolzonella, Tommaso; Coda, Stefano; Hakola, Antti; Meyer, Hendrik; Eurofusion Mst1 Team; Tcv Team; Mast-U Team; ASDEX Upgrade Team

2017-10-01

Under the EUROfusion MST1 program, coordinated experiments are conducted at three European medium sized tokamaks (ASDEX Upgrade, TCV and MAST-U). It complements the JET program for preparing a safe and efficient operation for ITER and DEMO. Work under MST1 benefits from cross-machine comparisons but also makes use of the unique capabilities of each device. For the 2017/2018 campaign 25 topic areas were defined targeting three main objectives: 1) Development towards an edge and wall compatible H-mode scenario with small or no ELMs. 2) Investigation of disruptions in order to achieve better predictions and improve avoidance or mitigation schemes. 3) Exploring conventional and alternative divertor configurations for future high P/R scenarios. This contribution will give an overview of the work done under MST1 exemplified by the highlight results for each top objective from the last campaigns, such as evaluation of natural small ELM scenarios, runaway mitigation and control, assessment of detachment in alternative divertor configurations and highly radiative scenarios. See author list of ``H. Meyer et al. 2017 Nucl. Fusion 57, 102014''.

Divertor impurity monitor for the International Thermonuclear Experimental Reactor

NASA Astrophysics Data System (ADS)

Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

1999-01-01

The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λ<450 nm) to prevent the transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

Response to "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

2014-05-01

Relying on coil positions relative to the plasma, the "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the "proximity condition," used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

Theory of Advanced Magnetic Divertors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Michael; Valanju, Prashant; Mahajan, Swadesh; Covele, Brent

2013-10-01

The magnetic field structure in the SOL is the most important determinant of divertor physics. A comprehensive analytical and numerical methodology is developed to investigate SOL magnetic fields in the backdrop of two advanced divertor geometries- the X-divertor (XD) proposed and discussed in 2004, and the snowflake divertor (SFD) of 2007-2010. The analysis shows that XD and SFD represent very distinct and readily distinguishable magnetic geometries, epitomized through a differentiating metric, the Divertor Index (DI). In terms of this simple metric, the XD (DI > 1) and the SFD (DI < 1) fall on opposite sides of the standard divertor SD (DI = 1). Amongst other things, DI signifies the rate of convergence (divergence) of the flux surfaces near the divertor plate; the flux surfaces of SFD are more convergent contracting) than the SD while the XD flux surfaces are less convergent, in fact, divergent (flaring). These different SOL magnetics imply different physics, particularly with respect to detachment dynamics. It is also shown that some experiments on NSTX and DIII-D match both the prescription and the predictions of the 2004 XD paper. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

DOE PAGES

Rognlien, Thomas D.; McLean, Adam G.; Fenstermacher, Max E.; ...

2017-01-27

A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult H-mode regime. The data set, which spans a range of plasmas densities for both forward and reverse toroidal magnetic field (B t) over a range of plasma densities, is provided by divertor Thomson scattering (DTS). Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (T e) and density (n e) across both divertor legs for individual discharges. The calculations show the same features of in/out plasma asymmetries as measured inmore » the experiment, with the normal B t direction (ion ∇B drift toward the X-point) having higher n e and lower T e in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. Furthermore, these 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.« less

Alternative divertor target concepts for next step fusion devices

NASA Astrophysics Data System (ADS)

Mazul, I. V.

2016-12-01

The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.

2017-09-13

A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.

NASA Astrophysics Data System (ADS)

Soukhanovskii, Vsevolod

2007-11-01

Steady-state handling of the heat flux is a critical divertor issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) devices. Because of an inherently compact divertor, it was thought that ST-based devices might not be able to fully utilize radiative and dissipative divertor techniques based on induced power and momentum loss. However, initial experiments conducted in the National Spherical Torus Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA 2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of elongations κ=1.8-2.4 and triangularities δ=0.45-0.75 demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m^2, could be reduced by 50-75% using a high-recycling radiative divertor regime with D2 injection. Furthermore, similar reduction was obtained with a partially detached divertor (PDD) at high D2 injection rates, however, it was accompanied by an X-point MARFE that quickly led to confinement degradation. Another approach takes advantage of the ST relation between strong shaping and high performance, and utilizes the poloidal magnetic flux expansion in the divertor region. Up to 60 % reduction in divertor peak heat flux was achieved at similar levels of scrape-off layer power by varying plasma shaping and thereby increasing the outer strike point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with divertor D2 injection at rates up to 10^22 s-1, a PDD regime with OSP peak heat flux 0.5-1.5 MW/m^2 was obtained without noticeable confinement degradation. Calculations based on a two point scrape-off layer model with parameterized power and momentum losses show that the short parallel connection length at the OSP sets the upper limit on the radiative exhaust channel, and both the impurity radiation and large momentum sink achievable only at high divertor neutral pressures are required for detachment.

Structural evaluation of a DTHR bundle divertor particle collector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prevenslik, T.V.

1980-09-01

The purpose of this report is to present a structural evaluation of the current bundle divertor particle collector BDPC design under a peak heat flux in relation to criteria that protect against coolant leakage into the plasma over replacement schedules planned during DTHR operation. In addition, an assessment of the BDPC structural integrity at higher heat fluxes is presented. Further, recommendations for modifications in the current BDPC design that would improve design reliability to be considered in future design studies are described. Finally, experimental test programs directed to establishing materials data necessary in providing greater confidence in subsequent structural evaluationsmore » of BDPC designs in relation to coolant leakage over planned replacement schedules are identified.« less

Divertor target shape optimization in realistic edge plasma geometry

NASA Astrophysics Data System (ADS)

Dekeyser, W.; Reiter, D.; Baelmans, M.

2014-07-01

Tokamak divertor design for next-step fusion reactors heavily relies on numerical simulations of the plasma edge. Currently, the design process is mainly done in a forward approach, where the designer is strongly guided by his experience and physical intuition in proposing divertor shapes, which are then thoroughly assessed by numerical computations. On the other hand, automated design methods based on optimization have proven very successful in the related field of aerodynamic design. By recasting design objectives and constraints into the framework of a mathematical optimization problem, efficient forward-adjoint based algorithms can be used to automatically compute the divertor shape which performs the best with respect to the selected edge plasma model and design criteria. In the past years, we have extended these methods to automated divertor target shape design, using somewhat simplified edge plasma models and geometries. In this paper, we build on and extend previous work to apply these shape optimization methods for the first time in more realistic, single null edge plasma and divertor geometry, as commonly used in current divertor design studies. In a case study with JET-like parameters, we show that the so-called one-shot method is very effective is solving divertor target design problems. Furthermore, by detailed shape sensitivity analysis we demonstrate that the development of the method already at the present state provides physically plausible trends, allowing to achieve a divertor design with an almost perfectly uniform power load for our particular choice of edge plasma model and design criteria.

Plasma-Facing Component and Materials Testing for the NSTX-U

NASA Astrophysics Data System (ADS)

Jaworski, Michael; Brooks, A.; Gerhardt, S.; Loesser, D.; Mardenfeld, M.; Menard, J.; Gray, T.; Reinke, M.

2017-10-01

The NSTX-U Recovery Project is developing plasma-facing components for use in the divertor of NSTX-U. The extreme conditions of the NSTX-U divertor make it possible to stress even graphite surfaces to the material limits leading to the possibility of component failures. In addition, the complex, mixed-material environment of the NSTX-U due to the use of boron and lithium wall conditioning techniques creates significant uncertainties in the monitoring of the PFCs. A testing program has been developed to inform on the material and design limitations of the NSTX-U high-heat flux components. These tests include high-heat flux testing in electron beam facilities as well as plasma-based testing. The NSTX-U components could experience perpendicular heat fluxes as high as 45 MW/m2. Parallel heat fluxes onto leading edges could reach 475 MW/m2. The testing program and material survey plan will be presented. Work supported by DOE contract DE-AC02-09CH11466 and DE-AC05-00OR22725.

Study of radial die-wall pressure changes during pharmaceutical powder compaction.

PubMed

Abdel-Hamid, Sameh; Betz, Gabriele

2011-04-01

In tablet manufacturing, less attention is paid to the measurement of die-wall pressure than to force-displacement diagrams. Therefore, the aim of this study was to investigate radial stress change during pharmaceutical compaction. The Presster(TM), a tablet-press replicator, was used to characterize compaction behavior of microcrystalline cellulose (viscoelastic), calcium hydrogen phosphate dihydrate (brittle), direct compressible mannitol (plastic), pre-gelatinized starch (plastic/elastic), and spray dried lactose monohydrate (plastic/brittle) by measuring radial die-wall pressure; therefore powders were compacted at different (pre) compaction pressures as well as different speeds. Residual die-wall pressure (RDP) and maximum die-wall pressure (MDP) were measured. Various tablet physical properties were correlated to radial die-wall pressure. With increasing compaction pressure, RDP and MDP (P < 0.0001) increased for all materials, with increasing precompaction RDP decreased for plastic materials (P < 0.05), whereas with increasing speed MDP decreased for all materials (P < 0.05). During decompression, microcrystalline cellulose and pre-gelatinized starch showed higher axial relaxation, whereas mannitol and lactose showed higher radial relaxation, calcium hydrogen phosphate showed high axial and radial relaxations. Plastic and brittle materials showed increased tendencies for friction because of high radial relaxation. Die-wall monitoring is suggested as a valuable tool for characterizing compaction behavior of materials and detecting friction phenomena in the early stage of development.

Tokamak power exhaust with the snowflake divertor: Present results and outstanding issues

DOE PAGES

Soukhanovskii, V. A.; Xu, X.

2015-09-15

Here, a snowflake divertor magnetic configuration (Ryutov in Phys Plasmas 14(6):064502, 2007) with the second-order poloidal field null offers a number of possible advantages for tokamak plasma heat and particle exhaust in comparison with the standard poloidal divertor with the first-order null. Results from snowflake divertor experiments are briefly reviewed and future directions for research in this area are outlined.

Interpretations of the impact of cross-field drifts on divertor flows in DIII-D with UEDGE

DOE PAGES

Jaervinen, Aaro E.; Allen, Steve L.; Groth, Mathias; ...

2017-01-27

Simulations using the multi-fluid code UEDGE indicates that, in low confinement (Lmode) plasmas in DIII-D, recycling driven flows dominate poloidal particle flows in the divertor, whereas E×B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E×B drift flows dominate both poloidal and radial particle flows in the divertor. UEDGE indicates that the toroidal C 2+ flow velocities in the divertor plasma are entrained within 30% to the background deuterium flow in both Land H-mode plasmas in the plasma region where the CIII 465 nm emission is measured. Therefore, UEDGE indicates that the Carbon Dopplermore » Coherence Imaging System (CIS), measuring the toroidal velocity of the C 2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations. Due to strong poloidal E×B flows in H-mode, net poloidal particle transport can be in opposite direction than the poloidal component of the parallel-B plasma flow.« less

Molecular characterization of Plum pox virus Rec isolates from Russia suggests a new insight into evolution of the strain.

PubMed

Chirkov, Sergei; Ivanov, Peter; Sheveleva, Anna; Kudryavtseva, Anna; Mitrofanova, Irina

2018-04-01

Field isolates of Plum pox virus (PPV), belonging to the strain Rec, have been found for the first time in Russia. Full-size genomes of the isolates K28 and Kisl-1pl from myrobalan and plum, respectively, were sequenced on the 454 platform. Analysis of all known PPV-Rec complete genomes using the Recombination Detection Program (RDP4) revealed yet another recombination event in the 5'-terminal region. This event was detected by seven algorithms, implemented in the RDP4, with statistically significant P values and supported by a phylogenetic analysis with the bootstrap value of 87%. A putative PPV-M-derived segment, encompassing the C-terminus of the P1 gene and approximately two-thirds of the HcPro gene, is bordered by breakpoints at positions 760-940 and 1838-1964, depending on the recombinant isolate. The predicted 5'-distal breakpoint for the isolate Valjevka is located at position 2804. The Dideron (strain D) and SK68 (strain M) isolates were inferred as major and minor parents, respectively. Finding of another recombination event suggests more complex evolutionary history of PPV-Rec than previously assumed. Perhaps the first recombination event led to the formation of a PPV-D variant harboring the PPV-M-derived fragment within the 5'-proximal part of the genome. Subsequent recombination of its descendant with PPV-M in the 3'-proximal genomic region resulted in the emergence of the evolutionary successful strain Rec.

Testing the role of molecular physics in dissipative divertor operations through helium plasmas at DIII-D

DOE PAGES

Canik, John M.; Briesemeister, Alexis R.; McLean, Adam G.; ...

2017-05-10

Recent experiments in DIII-D helium plasmas are examined to resolve the role of atomic and molecular physics in major discrepancies between experiment and modeling of dissipative divertor operation. Helium operation removes the complicated molecular processes of deuterium plasmas that are a prime candidate for the inability of standard fluid models to reproduce dissipative divertor operation, primarily the consistent under-prediction of radiated power. Modeling of these experiments shows that the full divertor radiation can be accounted for, but only if measures are taken to ensure that the model reproduces the measured divertor density. Relying on upstream measurements instead results in amore » lower divertor density and radiation than is measured, indicating a need for improved modeling of the connection between the diverter and the upstream scrape-off layer. Furthermore, these results show that fluid models are able to quantitatively describe the divertor-region plasma, including radiative losses, and indicate that efforts to improve the fidelity of the molecular deuterium models are likely to help resolve the discrepancy in radiation for deuterium plasmas.« less

Exploration of magnetic perturbation effects on advanced divertor configurations in NSTX-U

DOE Data Explorer

Frerichs, H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Waters, I. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Schmitz, O. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Canal, G. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Evans, T. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Feng, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Soukhanovskii, V. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2016-06-01

The control of divertor heat loads - both steady state and transient - remains a key challenge for the successful operation of ITER and FNSF. Magnetic perturbations provide a promising technique to control ELMs (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads are so called 'advanced divertors' which aim at optimizing the magnetic field configuration: the snowflake and the (super-)X-divertor. It is likely that both concepts - magnetic perturbations and advanced divertors - will have to work together, and we explore their interaction based on the NSTX-U setup. An overview of different divertor configurations under the impact of magnetic perturbations is presented, and the resulting impact on plasma edge transport is investigated with the EMC3-EIRENE code. Variations in size of the magnetic footprint of the perturbed separatrix are found, which is related to the level of flux expansion on the divertor target. Non-axisymmetric peaking of the heat flux related to the perturbed separatrix is found at the outer strike point, but only in locations where flux expansion is not too large.

Evaluation of heat and particle controllability on the JT-60SA divertor

NASA Astrophysics Data System (ADS)

Kawashima, H.; Hoshino, K.; Shimizu, K.; Takizuka, T.; Ide, S.; Sakurai, S.; Asakura, N.

2011-08-01

The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41 MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15 MW/m2. Dependence of the heat flux mitigation on a D2 gas-puff is evaluated by SONIC simulations for high density (ne_ave ˜ 1 × 1020 m-3) high current plasmas. It is found that the peak heat load 10 MW/m2 with dense (ned > 4 × 1020 m-3) and cold (Ted, Tid ⩽ 1 eV) divertor plasmas are obtained at a moderate gas-puff of Γpuff = 15 × 1021 s-1. Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100 m3/s. In full non-inductive current drive plasmas with low density (ne_ave ˜ 5 × 1019 m-3), the reduction of divertor heat load is achieved with the Ar injection.

Optimized tokamak power exhaust with double radiative feedback in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Eich, T.; Fuchs, J. C.; Giannone, L.; Herrmann, A.; Schweinzer, J.; Treutterer, W.; the ASDEX Upgrade Team

2012-12-01

A double radiative feedback technique has been developed on the ASDEX Upgrade tokamak for optimization of power exhaust with a standard vertical target divertor. The main chamber radiation is measured in real time by a subset of three foil bolometer channels and controlled by argon injection in the outer midplane. The target heat flux is in addition controlled by nitrogen injection in the divertor private flux region using either a thermoelectric sensor or the scaled divertor radiation obtained by a bolometer channel in the outer divertor. No negative interference of the two radiation controllers has been observed so far. The combination of main chamber and divertor radiative cooling extends the operational space of a standard divertor configuration towards high values of P/R. Pheat/R = 14 MW m-1 has been achieved so far with nitrogen seeding alone as well as with combined N + Ar injection, with the time-averaged divertor peak heat flux below 5 MW m-2. Good plasma performance can be maintained under these conditions, namely H98(y,2) = 1 and βN = 3.

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

NASA Astrophysics Data System (ADS)

Covele, B.; Kotschenreuther, M.; Mahajan, S.; Valanju, P.; Leonard, A.; Watkins, J.; Makowski, M.; Fenstermacher, M.; Si, H.

2017-08-01

The X-divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at 10-20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. However, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. The model also points to carbon radiation as the primary driver of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency for core operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.

Reactor application of an improved bundle divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, T.F.; Ruck, G.W.; Lee, A.Y.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

Evaluating Stellarator Divertor Designs with EMC3

NASA Astrophysics Data System (ADS)

Bader, Aaron; Anderson, D. T.; Feng, Y.; Hegna, C. C.; Talmadge, J. N.

2013-10-01

In this paper various improvements of stellarator divertor design are explored. Next step stellarator devices require innovative divertor solutions to handle heat flux loads and impurity control. One avenue is to enhance magnetic flux expansion near strike points, somewhat akin to the X-Divertor concept in Tokamaks. The effect of judiciously placed external coils on flux deposition is calculated for configurations based on the HSX stellarator. In addition, we attempt to optimize divertor plate location to facilitate the external coil placement. Alternate areas of focus involve altering edge island size to elucidate the driving physics in the edge. The 3-D nature of stellarators complicates design and necessitates analysis of new divertor structures with appropriate simulation tools. We evaluate the various configurations with the coupled codes EMC3-EIRENE, allowing us to benchmark configurations based on target heat flux, impurity behavior, radiated power, and transitions to high recycling and detached regimes. Work supported by DOE-SC0006103.

Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak.

PubMed

Xu, J C; Wang, L; Xu, G S; Luo, G N; Yao, D M; Li, Q; Cao, L; Chen, L; Zhang, W; Liu, S C; Wang, H Q; Jia, M N; Feng, W; Deng, G Z; Hu, L Q; Wan, B N; Li, J; Sun, Y W; Guo, H Y

2016-08-01

In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

Experimental study of heating scheme effect on the inner divertor power footprint widths in EAST lower single null discharges

NASA Astrophysics Data System (ADS)

Deng, G. Z.; Xu, J. C.; Liu, X.; Liu, X. J.; Liu, J. B.; Zhang, H.; Liu, S. C.; Chen, L.; Yan, N.; Feng, W.; Liu, H.; Xia, T. Y.; Zhang, B.; Shao, L. M.; Ming, T. F.; Xu, G. S.; Guo, H. Y.; Xu, X. Q.; Gao, X.; Wang, L.

2018-04-01

A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λ q and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λ q and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor.

Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, J. C.; Jia, M. N.; Feng, W.

2016-08-15

In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triplemore » probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.« less

Real-time control of divertor detachment in H-mode with impurity seeding using Langmuir probe feedback in JET-ITER-like wall

NASA Astrophysics Data System (ADS)

Guillemaut, C.; Lennholm, M.; Harrison, J.; Carvalho, I.; Valcarcel, D.; Felton, R.; Griph, S.; Hogben, C.; Lucock, R.; Matthews, G. F.; Perez Von Thun, C.; Pitts, R. A.; Wiesen, S.; contributors, JET

2017-04-01

Burning plasmas with 500 MW of fusion power on ITER will rely on partially detached divertor operation to keep target heat loads at manageable levels. Such divertor regimes will be maintained by a real-time control system using the seeding of radiative impurities like nitrogen (N), neon or argon as actuator and one or more diagnostic signals as sensors. Recently, real-time control of divertor detachment has been successfully achieved in Type I ELMy H-mode JET-ITER-like wall discharges by using saturation current (I sat) measurements from divertor Langmuir probes as feedback signals to control the level of N seeding. The degree of divertor detachment is calculated in real-time by comparing the outer target peak I sat measurements to the peak I sat value at the roll-over in order to control the opening of the N injection valve. Real-time control of detachment has been achieved in both fixed and swept strike point experiments. The system has been progressively improved and can now automatically drive the divertor conditions from attached through high recycling and roll-over down to a user-defined level of detachment. Such a demonstration is a successful proof of principle in the context of future operation on ITER which will be extensively equipped with divertor target probes.

Impurity seeding for tokamak power exhaust: from present devices via ITER to DEMO

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Dux, R.; Casali, L.; Eich, T.; Giannone, L.; Herrmann, A.; McDermott, R.; Mlynek, A.; Müller, H. W.; Reimold, F.; Schweinzer, J.; Sertoli, M.; Tardini, G.; Treutterer, W.; Viezzer, E.; Wenninger, R.; Wischmeier, M.; the ASDEX Upgrade Team

2013-12-01

A future fusion reactor is expected to have all-metal plasma facing materials (PFMs) to ensure low erosion rates, low tritium retention and stability against high neutron fluences. As a consequence, intrinsic radiation losses in the plasma edge and divertor are low in comparison to devices with carbon PFMs. To avoid localized overheating in the divertor, intrinsic low-Z and medium-Z impurities have to be inserted into the plasma to convert a major part of the power flux into radiation and to facilitate partial divertor detachment. For burning plasma conditions in ITER, which operates not far above the L-H threshold power, a high divertor radiation level will be mandatory to avoid thermal overload of divertor components. Moreover, in a prototype reactor, DEMO, a high main plasma radiation level will be required in addition for dissipation of the much higher alpha heating power. For divertor plasma conditions in present day tokamaks and in ITER, nitrogen appears most suitable regarding its radiative characteristics. If elevated main chamber radiation is desired as well, argon is the best candidate for the simultaneous enhancement of core and divertor radiation, provided sufficient divertor compression can be obtained. The parameter Psep/R, the power flux through the separatrix normalized by the major radius, is suggested as a suitable scaling (for a given electron density) for the extrapolation of present day divertor conditions to larger devices. The scaling for main chamber radiation from small to large devices has a higher, more favourable dependence of about Prad,main/R2. Krypton provides the smallest fuel dilution for DEMO conditions, but has a more centrally peaked radiation profile compared to argon. For investigation of the different effects of main chamber and divertor radiation and for optimization of their distribution, a double radiative feedback system has been implemented in ASDEX Upgrade (AUG). About half the ITER/DEMO values of Psep/R have been achieved so far, and close to DEMO values of Prad,main/R2, albeit at lower Psep/R. Further increase of this parameter may be achieved by increasing the neutral pressure or improving the divertor geometry.

The influence of Filaments in the Private Flux Region on Divertor Power and Particle Deposition

NASA Astrophysics Data System (ADS)

Harrison, James

2014-10-01

Recent advances in imaging of the MAST divertor have revealed, for the first time, evidence for filaments in the private flux region (PFR). Detailed analysis of the image data shows 3 distinct types of fluctuations occurring within the divertor volume: highly sheared filaments in the SOL originating from the outer midplane, high frequency (>50 kHz) filaments near the separatrix of the outer divertor leg and filaments in the private flux region originating from inner divertor leg. With the need to extrapolate divertor performance from existing machines to future devices, these observations can contribute to our quantitative understanding of transport in the PFR. In particular, they suggest that transport in the PFR is, at least in part, driven by turbulence, which may not be well captured by the Eich/Wagner description of the divertor footprint, expressed in terms of exponential decay in space above the X-point and Gaussian spreading below the X-point. The PFR filaments are observed to move largely parallel with the flux surfaces in a way equivalent to a toroidal angular velocity of order 2 ×104 rad/s in H-mode, and slower by a factor of order 2 in L-mode. During their transit parallel to the flux surfaces across the PFR, the filaments eject plasma in bursts, away from the separatrix, deeper into the private flux region. Correlation analysis suggests that they are generated by processes local to the inner divertor leg, as there is a weak correlation between fluctuations in the SOL and PFR above what is expected from line integration effects. Scaling of filament properties with machine operating parameters, such as plasma current, density and auxiliary heating power will be presented, together with a comparison with data from divertor Langmuir probes and IR thermography to estimate the role PFR filaments play in determining the width of the divertor footprint.

An innovative small angle slot divertor concept for long pulse advanced tokamaks

NASA Astrophysics Data System (ADS)

Guo, Houyang

2017-10-01

A new Small Angle Slot (SAS) divertor is being developed in DIII-D to address the challenge of efficient divertor heat dispersal at the relatively low plasma density required for non-inductive current drive in future advanced tokamaks. SAS features a small incident angle near the plasma strike point on the divertor target plate with a progressively opening slot. SOLPS (B2-Eirene) edge code analysis finds that SAS can achieve strong plasma cooling when the strike point is placed near the small angle target plate in the slot, leading to low electron temperature Te across the entire divertor target. This is enabled by strong coupling between a gas tight slot and directed neutral recycling by the small angle target to enhance neutral buildup near the target. SOLPS analysis reveals a strong correlation between Te and D2 density at the target for various divertor configurations including the flat target, slanted target, and lower single null divertor. The strong correlation suggests that achievement of low Te may reduce essentially to identifying the divertor baffle geometry that achieves the highest target gas density at a given upstream condition. The SAS divertor concept has recently been tested in DIII-D for a range of plasma configurations and conditions with precise control of slot strike point location. In confirmation of SOLPS predictions, a sharp transition is observed when the strike point is moved to the critical outer corner of SAS. A set of Langmuir probes imbedded in SAS show that the Te radial profile, which is peaked at the strike point when it is located away from the SAS corner, becomes low across the target when the strike point is located near the corner. With further increase in density, deep-slot detachment occurs with Te 1 eV, measured by the unique DIII-D divertor Thomson Scattering diagnostic. Work supported by US DOE under DE-FC02-04ER54698.

Exploration of magnetic perturbation effects on advanced divertor configurations in NSTX-U

DOE PAGES

Frerichs, H.; Schmitz, O.; Waters, I.; ...

2016-06-01

The control of divertor heat loads - both steady state and transient - remains a key challenge for the successful operation of ITER and FNSF. Magnetic perturbations provide a promising technique to control ELMs (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads are so called 'advanced divertors' which aim at optimizing the magnetic field configuration: the snowflake and the (super-)X-divertor. It is likely that both concepts - magnetic perturbations and advanced divertors - will have to work together, and we explore their inter- action based onmore » the NSTX-U setup. An overview of different divertor con gurations under the impact of magnetic perturbations is presented, and the resulting impact on plasma edge transport is investigated with the EMC3-EIRENE code. Variations in size of the magnetic footprint of the perturbed separatrix are found, which is related to the level of flux expansion on the divertor target. Non-axisymmetric peaking of the heat flux related to the perturbed separatrix is found at the outer strike point, but only in locations where flux expansion is not too large.« less

Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

NASA Astrophysics Data System (ADS)

Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

2016-10-01

The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

NASA Astrophysics Data System (ADS)

Chen, Bin; Xu, Xueqiao; Xia, Tianyang; Ye, Minyou

2015-11-01

China Fusion Engineering Testing Reactor is under conceptual design, acting as a bridge between ITER and DEMO. The detached divertor operation offers great promise for a reduction of heat flux onto divertor target plates for acceptable erosion. Therefore, a density scan is performed via an increase of D2 gas puffing rates in the range of 0 . 0 ~ 5 . 0 ×1023s-1 by using the B2-Eirene/SOLPS 5.0 code package to study the heat flux control and impurity screening property. As the density increases, it shows a gradually change of the divertor operation status, from low-recycling regime to high-recycling regime and finally to detachment. Significant radiation loss inside the confined plasma in the divertor region during detachment leads to strong parallel density and temperature gradients. Based on the SOLPS simulations, BOUT + + simulations will be presented to investigate the stability and turbulent transport under divertor plasma detachment, particularly the strong parallel gradient driven instabilities and enhanced plasma turbulence to spread heat flux over larger surface areas. The correlation between outer mid-plane and divertor turbulence and the related transport will be analyzed. Prepared by LLNL under Contract DE-AC52-07NA27344. LLNL-ABS-675075.

Favorable effects of turbulent plasma mixing on the performance of innovative tokamak divertors

NASA Astrophysics Data System (ADS)

Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Umansky, M. V.

2013-10-01

The problem of reducing the heat load on plasma-facing components is one of the most demanding issues for MFE devices. The general approach to the solution of this problem is the use of a specially configured poloidal magnetic field, so called magnetic divertors. In recent years, novel divertors possessing the 2-nd and 3-rd order nulls of the poloidal field (PF) have been proposed. They are called a ``snowflake'' (SF) and a ``cloverleaf'' (CL) divertor, respectively, due to characteristic shape of the magnetic separatrix. Among several beneficial features of such divertors is an effect of strong turbulent plasma mixing that is intrinsic to the zone of weak PF near the null-point. The turbulence spreads the heat flux between multiple divertor exhaust channels and increases the heat flux width within each channel. Among physical processes affecting the onset of convection the curvature-driven mode of axisymmetric rolls is most prominent. The effect is quite significant for the SF and is even stronger for the CL divertor. Projections to future ITER-scale facilities are discussed. Work performed for U.S. DoE by LLNL under Contract DE-AC52-07NA27344.

Design of ITER divertor VUV spectrometer and prototype test at KSTAR tokamak

NASA Astrophysics Data System (ADS)

Seon, Changrae; Hong, Joohwan; Song, Inwoo; Jang, Juhyeok; Lee, Hyeonyong; An, Younghwa; Kim, Bosung; Jeon, Taemin; Park, Jaesun; Choe, Wonho; Lee, Hyeongon; Pak, Sunil; Cheon, MunSeong; Choi, Jihyeon; Kim, Hyeonseok; Biel, Wolfgang; Bernascolle, Philippe; Barnsley, Robin; O'Mullane, Martin

2017-12-01

Design and development of the ITER divertor VUV spectrometer have been performed from the year 1998, and it is planned to be installed in the year 2027. Currently, the design of the ITER divertor VUV spectrometer is in the phase of detail design. It is optimized for monitoring of chord-integrated VUV signals from divertor plasmas, chosen to contain representative lines emission from the tungsten as the divertor material, and other impurities. Impurity emission from overall divertor plasmas is collimated through the relay optics onto the entrance slit of a VUV spectrometer with working wavelength range of 14.6-32 nm. To validate the design of the ITER divertor VUV spectrometer, two sets of VUV spectrometers have been developed and tested at KSTAR tokamak. One set of spectrometer without the field mirror employs a survey spectrometer with the wavelength ranging from 14.6 nm to 32 nm, and it provides the same optical specification as the spectrometer part of the ITER divertor VUV spectrometer system. The other spectrometer with the wavelength range of 5-25 nm consists of a commercial spectrometer with a concave grating, and the relay mirrors with the same geometry as the relay mirrors of the ITER divertor VUV spectrometer. From test of these prototypes, alignment method using backward laser illumination could be verified. To validate the feasibility of tungsten emission measurement, furthermore, the tungsten powder was injected in KSTAR plasmas, and the preliminary result could be obtained successfully with regard to the evaluation of photon throughput. Contribution to the Topical Issue "Atomic and Molecular Data and their Applications", edited by Gordon W.F. Drake, Jung-Sik Yoon, Daiji Kato, Grzegorz Karwasz.

Development of a Method for Local Electron Temperature and Density Measurements in the Divertor of the JET Tokamak

NASA Technical Reports Server (NTRS)

Jupen, C.; Meigs, A.; Bhatia, A. K.; Brezinsek, S.; OMullane, M.

2004-01-01

Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)<1 eV) and at high electron density (n(sub e)>10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

Divertor electron temperature and impurity diffusion measurements with a spectrally resolved imaging radiometer.

PubMed

Clayton, D J; Jaworski, M A; Kumar, D; Stutman, D; Finkenthal, M; Tritz, K

2012-10-01

A divertor imaging radiometer (DIR) diagnostic is being studied to measure spatially and spectrally resolved radiated power P(rad)(λ) in the tokamak divertor. A dual transmission grating design, with extreme ultraviolet (~20-200 Å) and vacuum ultraviolet (~200-2000 Å) gratings placed side-by-side, can produce coarse spectral resolution over a broad wavelength range covering emission from impurities over a wide temperature range. The DIR can thus be used to evaluate the separate P(rad) contributions from different ion species and charge states. Additionally, synthetic spectra from divertor simulations can be fit to P(rad)(λ) measurements, providing a powerful code validation tool that can also be used to estimate electron divertor temperature and impurity transport.

Electron pressure balance in the SOL through the transition to detachment

DOE PAGES

McLean, A. G.; Leonard, A. W.; Makowski, M. A.; ...

2015-02-07

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of P inj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure T e down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurementsmore » of n e, T e and p e in the divertor versus Lparallel demonstrate a rapid transition from Te ≥ 15 eV to ≤3 eV occurring both at the outer strike point and upstream of the X-point. Furthermore, these observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.« less

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

DOE PAGES

Covele, Brent; Kotschenreuther, M.; Mahajan, S.; ...

2017-06-23

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Electron pressure balance in the SOL through the transition to detachment

NASA Astrophysics Data System (ADS)

McLean, A. G.; Leonard, A. W.; Makowski, M. A.; Groth, M.; Allen, S. L.; Boedo, J. A.; Bray, B. D.; Briesemeister, A. R.; Carlstrom, T. N.; Eldon, D.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Liu, C.; Osborne, T. H.; Petrie, T. W.; Soukhanovskii, V. A.; Stangeby, P. C.; Tsui, C.; Unterberg, E. A.; Watkins, J. G.

2015-08-01

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Covele, Brent; Kotschenreuther, M.; Mahajan, S.

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Impurity ion flow and temperature measured in a detached divertor with externally applied non-axisymmetric fields on DIII-D

DOE PAGES

Briesemeister, A. R.; Isler, R. C.; Allen, S. L.; ...

2014-11-15

In this study, externally applied non-axisymmetric magnetic fields are shown to have little effect on the impurity ion flow velocity and temperature as measured by the multichord divertor spectrometer in the DIII-D divertor for both attached and detached conditions. These experiments were performed in H-mode plasmas with the grad-B drift toward the target plates, with and without n = 3 resonant magnetic perturbations (RMPs). The flow velocity in the divertor is shown to change by as much as 30% when deuterium gas puffing is used to create detachment of the divertor plasma. No measurable changes in the C III flowmore » were observed in response to the RMP fields for the conditions used in this work. Images of the C III emission are used along with divertor Thomson scattering to show that the local electron and C III temperatures are equilibrated for the conditions shown.« less

Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

NASA Astrophysics Data System (ADS)

Masuzaki, S.; Tokitani, M.; Otsuka, T.; Oya, Y.; Hatano, Y.; Miyamoto, M.; Sakamoto, R.; Ashikawa, N.; Sakurada, S.; Uemura, Y.; Azuma, K.; Yumizuru, K.; Oyaizu, M.; Suzuki, T.; Kurotaki, H.; Hamaguchi, D.; Isobe, K.; Asakura, N.; Widdowson, A.; Heinola, K.; Jachmich, S.; Rubel, M.; contributors, JET

2017-12-01

Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011-2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

Oral health-related quality of life in socially endangered persons in Copenhagen, Denmark.

PubMed

Øzhayat, Esben Boeskov; Østergaard, Peter; Gotfredsen, Klaus

2016-11-01

The aims of this study were to investigate and describe the Oral Health Related Quality of Life (OHRQoL) in a socially endangered group of people and to compare the OHRQoL to other patient groups. About 294 socially endangered persons attending a volunteer clinic in Copenhagen Denmark filled in the OHIP-14 questionnaire. The group was compared in mean score and reported problems to a group of patients with tooth loss and about to have a removable dental prosthesis (RDP), a group with tooth loss about to have a fixed dental prosthesis (FDP) and a control group without tooth loss. Significantly higher OHIP-14 score was seen in the socially endangered group (15.5 (SD 12.6)) compared with the control (1.9 (SD 2.7)) and the FDP group (9.4 (SD 8.2)) but not the RDP group (13.1 (SD 10.5)). This difference was not changed after stratifying in age groups. Problems related to psychological disability, social disability, and handicap were more frequent in the social endangered group than for the other groups. The items pain, tense, diet, relax, life, and function stand out as problems in the socially endangered group compared to the other groups. The OHRQoL is highly impaired in the socially endangered persons and at least to the level of persons with great tooth loss about to have an RDP. The problems seem to be more handicapping in the socially endangered compared with other patient groups known to have high impairment.

Environmental Protection Tools in Agricultural Management Works

NASA Astrophysics Data System (ADS)

Glowacka, Agnieszka; Taszakowski, Jaroslaw; Janus, Jaroslaw; Bozek, Piotr

2016-10-01

Land consolidation is a fundamental instrument for agricultural management. It facilitates comprehensive changes in the agricultural, social, and ecological domains. Consolidation and post-consolidation development-related investments are an opportunity to improve living conditions in rural areas, and simultaneously ensure its positive impact on the environment. One of the primary goals of consolidation, directly specified in the Act on land consolidation, is to improve farming conditions. In Poland, consolidation is possible due to EU funds: RDP 2007-2013 and RDP 2014-2020. In order for individual villages to be granted EU funds for consolidation and post-consolidation development under the Rural Development Programme 2014-2020, their consolidation has to implement actions with positive impact on the environment and the landscape. The goal of this paper is to analyse documentation in the form of assumptions for a land consolidation project enclosed to an RDP 2014-2020 grant application and project information sheets as the basis for environmental impact assessment in the context of detailed presentation of environmental protection solutions that ensure a positive impact of the project on the environment and landscape. The detailed study involved 9 villages in the Malopolskie Voivodeship, which applied for EU grants for land consolidation in the current financial perspective. The paper specifies the existing state of the analysed villages as regards the natural environment, lists agricultural management instruments that have a positive impact on the environment, and demonstrates that planning of actions aimed at environmental protection is a necessary element of assumptions for land consolidation projects.

Morphodynamics of dome dunes under unimodal wind regimes

NASA Astrophysics Data System (ADS)

Gao, Xin; Narteau, Clement; Rozier, Olivier

2017-04-01

Dome dunes are isolated sand piles with a rounded shape and no slip face. They are not only incipient or disappearing dunes, they can also reach a giant size and form dome-dune fields. Nevertheless, unlike other types of dunes, they have not been the subject of intense research, certainly because they result from complex multidirectional wind regimes. Here we analyze the morphodynamics of dome dunes under unimodal wind regimes. From numerical modeling using a normal distribution of sand flux orientation, we show that the transition from barchan to dome dunes occur when the standard deviation is larger than 40°. As confirmed by sand flux roses of dome-dune fields in arid deserts on Earth, it corresponds to RDP/DP-value of 0.8 (RDP/DP is the ratio between the resultant drift potential and the drift potential). Both in the field and in the numerical model, the transition from barchan to dome-dunes can also be captured from the coefficient of variation of the planar dune shape. Not surprisingly, smaller dome dunes are faster than larger ones. However, the dependence of dune migration rate on the RDP-value changes according to the presence or absence of slip faces because of the speed-up effect. Transient finger dunes may develop in dome-dune fields, but they rapidly break-up into smaller bodies. This shows that, contrary to bidirectional wind regimes, a large dispersion of sand flux orientation is not efficient in building longitudinal dunes.

X-Divertor Geometries for Deeper Detachment Without Degrading the DIII-D H-Mode

NASA Astrophysics Data System (ADS)

Covele, Brent; Kotschenreuther, M. T.; Valanju, P. M.; Mahajan, S. M.; Leonard, A. W.; Hyatt, A. W.; McLean, A. G.; Thomas, D. M.; Guo, H. Y.; Watkins, J. G.; Makowski, M. A.; Hill, D. N.

2015-11-01

Recent DIII-D experiments comparing the standard divertor (SD) and X-Divertor (XD) geometries show heat and particle flux reduction at the divertor target plate. The XD features large poloidal flux expansion, increased connection length, and poloidal field line flaring, quantified by the Divertor Index. Both SD and XD were pushed deep into detachment with increased gas puffing, until core energy confinement and pedestal pressure were substantially reduced. As expected, outboard target heat fluxes are significantly reduced in the XD compared to the SD under similar upstream plasma conditions, even at low Greenwald fraction. The high-triangularity (floor) XD cases show larger reduction in temperature, heat, and particle flux relative to the SD in all cases, while low-triangularity (shelf) XD cases show more modest reductions over the SD. Consequently, heat flux reduction and divertor detachment may be achieved in the XD with less gas puffing and higher pedestal pressures. Further causative analysis, as well as detailed modeling with SOLPS, is underway. These initial experiments suggest the XD as a promising candidate to achieve divertor heat flux control compatible with robust H-mode operation. Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-04ER54754, and DE-FG02-04ER54742.

Critical need for MFE: the Alcator DX advanced divertor test facility

NASA Astrophysics Data System (ADS)

Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

2013-10-01

Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

The dependence of divertor power sharing on magnetic flux balance in near double-null configurations on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Brunner, D.; Kuang, A. Q.; LaBombard, B.; Terry, J. L.

2018-07-01

Management of power exhaust will be a crucial task for tokamak fusion reactors. Reactor concepts are often proposed with double-null divertors, i.e. having two magnetic separatrices in an up-down symmetric configuration. This arrangement is potentially advantageous since the majority of the tokamak exhaust power tends to flow to the outer pair of divertor legs at large major radius, where the geometry is favorable for spreading the heat over a large surface area and there is more room for advanced divertor configurations. Despite the importance, there have been relatively few studies of divertor power sharing in near double null configurations and no studies at the poloidal magnetic fields and scrape-off layer power widths anticipated for a reactor. Motivated by this need we have undertaken a systematic study on Alcator C-Mod, examining the effect of magnetic flux balance on the power sharing among the four divertor legs in near double-null plasmas. Ohmic L-modes at three values of plasma current and ICRF-heated enhanced D-alpha (EDA) H-modes and I-modes at a single value of plasma current are explored, producing poloidal magnetic fields of 0.42, 0.62 and 0.85 Tesla. For Ohmic L-modes and ICRF-heated EDA H-modes, we find that the point of equal power sharing between upper and lower divertors occurs remarkably close to a balanced double null. Power sharing amongst the outer (upper versus lower) and inner (upper versus lower) pairs of divertors can be described in terms of a logistic function of magnetic flux balance, consistent with heat flux mapping along magnetic field lines to the outer midplane. Power sharing between inner and outer legs is found to follow a Gaussian-like function of magnetic flux balance with non-zero power to the inner divertors at double null. The overall behavior of H-modes operated near double null and for I-modes operating to within one heat flux e-folding of double null are found similar to Ohmic L-modes, with a significant reduction of power on the inner divertor legs. The results are encapsulated in terms of empirically-informed analytic functions of magnetic flux balance. When combined with magnetic equilibrium control system specifications, these relationships can be used to specify the power flux handling requirements for each of the four divertor target plates.

Basic physical processes and reduced models for plasma detachment

NASA Astrophysics Data System (ADS)

Stangeby, P. C.

2018-04-01

The divertor of a tokamak reactor will have to satisfy a number of critical constraints, the first of which is that the divertor targets not fail due to excessive heating or sputter-erosion. This paramount constraint of target survival defines the operating window for the principal plasma properties at the divertor target, the density n t and temperature, T t. In particular T et < 10 eV is shown to be required. Code and experimental studies show that the pressure–momentum loss by the plasma that occurs along flux tubes in the edge, between the divertor entrance and target, (i) correlates strongly with T et, and (ii) begins to increase as T et falls below 10 eV, becoming very strong by 1 eV. The transition between the high-recycling regime and the detached divertor regime has therefore been defined here to occur when T et < 10 eV. Simple analytic models are developed (i) to relate (T t, n t) to the controlling conditions ‘upstream’ e.g. at the divertor entrance, and (ii) in turn to relate (T t, n t) to other important divertor quantities including (a) the required level of radiative cooling in the divertor, and (b) the ion flux to the target in the presence of volumetric loss of particles, momentum and power in the divertor. The 2 Point Model, 2PM, is a widely used analytic model for relating (T t, n t) to the controlling upstream conditions. The 2PM is derived here for various levels of complexity regarding the effects included. Analytic models of divertor detachment provide valuable insight and useful approximations, but more complete modeling requires the use of edge codes such as EDGE2D, SOLPS, SONIC, UEDGE, etc. Edge codes have grown to become quite sophisticated and now constitute, in effect, ‘code-experiments’ that—just as for actual experiments—can benefit from interpretation in terms of simple conceptual frameworks. 2 Point Model Formatting, 2PMF, of edge code output can provide such a conceptual framework. Methods of applying 2PMF are illustrated here with some examples.

Overview of long pulse H-mode operation on EAST

NASA Astrophysics Data System (ADS)

Gong, X.; Garofalo, A. M.; Wan, B.; Li, J.; Qian, J.; Li, E.; Liu, F.; Zhao, Y.; Wang, M.; Xu, H.; EAST Team

2017-10-01

The EAST research program aims to demonstrate steady-state long-pulse high-performance H-mode operations with ITER-like poloidal configuration and RF-dominated heating schemes. In the recent experimental campaign, a long pulse fully non-inductive H-mode discharge lasting over 100 seconds using the upper ITER-like tungsten divertor has been achieved in EAST. This scenario used only RF heating and current drive, but also benefitted from an integrated control of the wall conditioning, plasma configuration, divertor heat flux, particle exhaust, impurity management and superconducting coils safety. Maintaining effective coupling of multiple RF heating and current drive sources on EAST is a critical ingredient. This long pulse discharge had good energy confinement, H98,y2 1.1-1.2, and all of the plasma parameters reach a true steady-state. Power balance indicates that the confinement improvement is due partly to a significantly reduced core electron transport inside minor radius rho<0.4. This work was supported by the National Magnetic Confinement Fusion Program of China Contract No. 2015GB10200 and the US Department of Energy Contract No. DE-SC0010685.

Design, R&D and commissioning of EAST tungsten divertor

NASA Astrophysics Data System (ADS)

Yao, D. M.; Luo, G. N.; Zhou, Z. B.; Cao, L.; Li, Q.; Wang, W. J.; Li, L.; Qin, S. G.; Shi, Y. L.; Liu, G. H.; Li, J. G.

2016-02-01

After commissioning in 2005, the EAST superconducting tokamak had been operated with its water cooled divertors for eight campaigns up to 2012, employing graphite as plasma facing material. With increase in heating power over 20 MW in recent years, the heat flux going to the divertors rises rapidly over 10 MW m-2 for steady state operation. To accommodate the rapid increasing heat load in EAST, the bolting graphite tile divertor must be upgraded. An ITER-like tungsten (W) divertor has been designed and developed; and firstly used for the upper divertor of EAST. The EAST upper W divertor is modular structure with 80 modules in total. Eighty sets of W/Cu plasma-facing components (PFC) with each set consisting of an outer vertical target (OVT), an inner vertical target (IVT) and a DOME, are attached to 80 stainless steel cassette bodies (CB) by pins. The monoblock W/Cu-PFCs have been developed for the strike points of both OVT and IVT, and the flat type W/Cu-PFCs for the DOME and the baffle parts of both OVT and IVT, employing so-called hot isostatic pressing (HIP) technology for tungsten to CuCrZr heat sink bonding, and electron beam welding for CuCrZr to CuCrZr and CuCrZr to other material bonding. Both monoblock and flat type PFC mockups passed high heat flux (HHF) testing by means of electron beam facilities. The 80 divertor modules were installed in EAST in 2014 and results of the first commissioning are presented in this paper.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

NASA Astrophysics Data System (ADS)

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

2016-12-01

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

DOE PAGES

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; ...

2016-09-14

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, whichmore » we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

Impurity-induced divertor plasma oscillations

DOE PAGES

Smirnov, R. D.; Kukushkin, A. S.; Krasheninnikov, S. I.; ...

2016-01-07

Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ionmore » transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. As a result, the implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.« less

The effects of particle recycling on the divertor plasma: A particle-in-cell with Monte Carlo collision simulation

NASA Astrophysics Data System (ADS)

Chang, Mingyu; Sang, Chaofeng; Sun, Zhenyue; Hu, Wanpeng; Wang, Dezhen

2018-05-01

A Particle-In-Cell (PIC) with Monte Carlo Collision (MCC) model is applied to study the effects of particle recycling on divertor plasma in the present work. The simulation domain is the scrape-off layer of the tokamak in one-dimension along the magnetic field line. At the divertor plate, the reflected deuterium atoms (D) and thermally released deuterium molecules (D2) are considered. The collisions between the plasma particles (e and D+) and recycled neutral particles (D and D2) are described by the MCC method. It is found that the recycled neutral particles have a great impact on divertor plasma. The effects of different collisions on the plasma are simulated and discussed. Moreover, the impacts of target materials on the plasma are simulated by comparing the divertor with Carbon (C) and Tungsten (W) targets. The simulation results show that the energy and momentum losses of the C target are larger than those of the W target in the divertor region even without considering the impurity particles, whereas the W target has a more remarkable influence on the core plasma.

Modelling of Divertor Detachment in MAST Upgrade

NASA Astrophysics Data System (ADS)

Moulton, David; Carr, Matthew; Harrison, James; Meakins, Alex

2017-10-01

MAST Upgrade will have extensive capabilities to explore the benefits of alternative divertor configurations such as the conventional, Super-X, x divertor, snowflake and variants in a single device with closed divertors. Initial experiments will concentrate on exploring the Super-X and conventional configurations, in terms of power and particle loads to divertor surfaces, access to detachment and its control. Simulations have been carried out with the SOLPS5.0 code validated against MAST experiments. The simulations predict that the Super-X configuration has significant advantages over the conventional, such as lower detachment threshold (2-3x lower in terms of upstream density and 4x higher in terms of PSOL). Synthetic spectroscopy diagnostics from these simulations have been created using the Raysect ray tracing code to produce synthetic filtered camera images, spectra and foil bolometer data. Forward modelling of the current set of divertor diagnostics will be presented, together with a discussion of future diagnostics and analysis to improve estimates of the plasma conditions. Work supported by the RCUK Energy Programme [Grant Number EP/P012450/1] and EURATOM.

ELM mitigation techniques

NASA Astrophysics Data System (ADS)

Evans, T. E.

2013-07-01

Large edge-localized mode (ELM) control techniques must be developed to help ensure the success of burning and ignited fusion plasma devices such as tokamaks and stellarators. In full performance ITER tokamak discharges, with QDT = 10, the energy released by a single ELM could reach ˜30 MJ which is expected to result in an energy density of 10-15 MJ/m2on the divertor targets. This will exceed the estimated divertor ablation limit by a factor of 20-30. A worldwide research program is underway to develop various types of ELM control techniques in preparation for ITER H-mode plasma operations. An overview of the ELM control techniques currently being developed is discussed along with the requirements for applying these techniques to plasmas in ITER. Particular emphasis is given to the primary approaches, pellet pacing and resonant magnetic perturbation fields, currently being considered for ITER.

Area of Stochastic Scrape-Off Layer for a Single-Null Divertor Tokamak Using Simple Map

NASA Astrophysics Data System (ADS)

Fisher, Tiffany; Verma, Arun; Punjabi, Alkesh

1996-11-01

The magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The Simple map is characterized by a single parameter k representing the toroidal asymmetry. The width of the stochastic scrape-off layer and its area varies with the map parameter k. We calculate the area of the stochastic scrape-off layer for different k's and obtain a parametric expression for the area in terms of k and y _LastGoodSurface(k). This work is supported by US DOE OFES. Tiffany Fisher is a HU CFRT Summer Fusion High school Workshop Scholar from New Bern High School in North Carolina. She is supported by NASA SHARP Plus Program.

High Adjustable Shear Map for a Single-null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Whitten, Michaelangelo; Lam, Maria; Punjabi, Alkesh

1996-11-01

An explicit map that has an adjustable shear s is x1 = x0 - k y0 [ ( 1 - y0 ) ( 1 + s y0 ) + s x ^21 ] y1 = y0 + k x1 [ 1 + s ( x^21 + y^20 ) ] Tokamak shear corresponds to negative s. Thus we can construct maps for variable shear for a single-null divertor tokamak (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 3322, 69 (1992) ^, (Punjabi A, Verma A and Boozer A, J Plasma Phys), 52, 91 (1994). Here we present the results from an initial study of this map. This work is supported by US DOE OFES. Michelangelo Whitten is a HU CFRT Summer Fusion High School Workshop scholar from Bowie High School in El Paso, Texas. He is supported by NASA SHARP Plus Program.

ADX: a high field, high power density, Advanced Divertor test eXperiment

NASA Astrophysics Data System (ADS)

Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

2014-10-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

Thermal Fatigue Study on the Divertor Plate Materials

NASA Astrophysics Data System (ADS)

Wu, Ji-hong; Zhang, Fu; Xu, Zeng-yu; Yan, Jian-cheng

2002-10-01

Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.

Modelling of 13CH4 injection and local carbon deposition at the outer divertor of ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Aho-Mantila, L.; Airila, M. I.; Wischmeier, M.; Krieger, K.; Pugno, R.; Coster, D. P.; Chankin, A. V.; Neu, R.; Rohde, V.

2009-12-01

Numerical modelling of 13CH4 injection into the outer divertor plasma of the full tungsten, vertical target of ASDEX Upgrade is presented. The SOLPS5.0 code package is used to calculate a realistic scrape-off layer plasma background corresponding to L-mode discharges in the attached divertor plasma regime. The ERO code is then used for detailed modelling of the hydrocarbon break-up, re-deposition and re-erosion processes. The deposition patterns observed at two different poloidal locations are shown to strongly reflect the cross-field gradients in divertor plasma density and temperature, as well as the local plasma collisionality. Experimental results with forward and reversed BT, accompanied by numerical modelling, also point towards a significant poloidal hydrocarbon E×B drift in the divertor region.

Tritium analysis of divertor tiles used in JET ITER-like wall campaigns by means of β-ray induced x-ray spectrometry

NASA Astrophysics Data System (ADS)

Hatano, Y.; Yumizuru, K.; Koivuranta, S.; Likonen, J.; Hara, M.; Matsuyama, M.; Masuzaki, S.; Tokitani, M.; Asakura, N.; Isobe, K.; Hayashi, T.; Baron-Wiechec, A.; Widdowson, A.; contributors, JET

2017-12-01

Energy spectra of β-ray induced x-rays from divertor tiles used in ITER-like wall campaigns of the Joint European Torus were measured to examine tritium (T) penetration into tungsten (W) layers. The penetration depth of T evaluated from the intensity ratio of W(Lα) x-rays to W(Mα) x-rays showed clear correlation with poloidal position; the penetration depth at the upper divertor region reached several micrometers, while that at the lower divertor region was less than 500 nm. The deep penetration at the upper part was ascribed to the implantation of high energy T produced by DD fusion reactions. The poloidal distribution of total x-ray intensity indicated higher T retention in the inboard side than the outboard side of the divertor region.

Correlation of 3D Shift and 3D Tilt of the Patella in Patients With Recurrent Dislocation of the Patella and Healthy Volunteers: An In Vivo Analysis Based on 3-Dimensional Computer Models.

PubMed

Yamada, Yuzo; Toritsuka, Yukiyoshi; Nakamura, Norimasa; Horibe, Shuji; Sugamoto, Kazuomi; Yoshikawa, Hideki; Shino, Konsei

2017-11-01

The concepts of lateral deviation and lateral inclination of the patella, characterized as shift and tilt, have been applied in combination to evaluate patellar malalignment in patients with patellar dislocation. It is not reasonable, however, to describe the 3-dimensional (3D) positional relation between the patella and the femur according to measurements made on 2-dimensional (2D) images. The current study sought to clarify the relation between lateral deviation and inclination of the patella in patients with recurrent dislocation of the patella (RDP) by redefining them via 3D computer models as 3D shift and 3D tilt. Descriptive laboratory study. Altogether, 60 knees from 56 patients with RDP and 15 knees from 10 healthy volunteers were evaluated. 3D shift and tilt of the patella were analyzed with 3D computer models created by magnetic resonance imaging scans obtained at 10° intervals of knee flexion (0°-50°). 3D shift was defined as the spatial distance between the patellar reference point and the midsagittal plane of the femur; it is expressed as a percentage of the interepicondylar width. 3D tilt was defined as the spatial angle between the patellar reference plane and the transepicondylar axis. Correlations between the 2 parameters were assessed with the Pearson correlation coefficient. The patients' mean Pearson correlation coefficient was 0.895 ± 0.186 (range, -0.073 to 0.997; median, 0.965). In all, 56 knees (93%) had coefficients >0.7 (strong correlation); 1 knee (2%), >0.4 (moderate correlation); 2 knees (3%), >0.2 (weak correlation); and 1 knee (2%), <0.2 (no correlation). The mean correlation coefficient of the healthy volunteers was 0.645 ± 0.448 (range, -0.445 to 0.982; median, 0.834). A statistically significant difference was found in the distribution of the correlation coefficients between the patients and the healthy volunteers ( P = .0034). When distribution of the correlation coefficients obtained by the 3D analyses was compared with that by the 2D (conventional) analyses, based on the bisect offset index and patellar tilt angle, the 3D analyses showed statistically higher correlations between the lateral deviation and inclination of the patella ( P < .01). 3D shift and 3D tilt of the patella were moderately or strongly correlated in 95% of patients with RDP at 0° to 50° of knee flexion. It is not always necessary to use both parameters when evaluating patellar alignment, at least for knees with RDP at 0° to 50° of flexion. Such a description may enable surgeons to describe patellar alignment more simply, leading to a better, easier understanding of the characteristics of each patient with RDP.

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE PAGES

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.; ...

2017-06-22

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

DTT: a divertor tokamak test facility for the study of the power exhaust issues in view of DEMO

NASA Astrophysics Data System (ADS)

Albanese, R.; WPDTT2 Team; DTT Project Proposal Contributors, the

2017-01-01

In parallel with the programme to optimize the operation with a conventional divertor based on detached conditions to be tested on the ITER device, a project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a divertor tokamak test facility (DTT). The DTT project proposal refers to a set of parameters selected so as to have edge conditions as close as possible to DEMO, while remaining compatible with DEMO bulk plasma performance in terms of dimensionless parameters and given constraints. The paper illustrates the DTT project proposal, referring to a 6 MA plasma with a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6-1.8, and a toroidal field of 6 T. This selection will guarantee sufficient flexibility to test a wide set of divertor concepts and techniques to cope with large heat loads, including conventional tungsten divertors; liquid metal divertors; both conventional and advanced magnetic configurations (including single null, snow flake, quasi snow flake, X divertor, double null); internal coils for strike point sweeping and control of the width of the scrape-off layer in the divertor region; and radiation control. The Poloidal Field system is planned to provide a total flux swing of more than 35 Vs, compatible with a pulse length of more than 100 s. This is compatible with the mission of studying the power exhaust problem and is obtained using superconducting coils. Particular attention is dedicated to diagnostics and control issues, especially those relevant for plasma control in the divertor region, designed to be as compatible as possible with a DEMO-like environment. The construction is expected to last about seven years, and the selection of an Italian site would be compatible with a budget of 500 M€.

X-Divertors on ITER - with no hardware changes

NASA Astrophysics Data System (ADS)

Valanju, Prashant; Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh; Kessel, Charles

2014-10-01

Using CORSICA, we have discovered that X-Divertor (XD) equilibria are possible on ITER - without any extra PF coils inside the TF coils, and with no changes to ITER's poloidal field (PF) coil set, divertor cassette, strike points, or first wall. Starting from the Standard Divertor (SD), a sequence of XD configurations (with increasing flux expansions at the divertor plate) can be made by reprogramming ITER PF coil currents while keeping them all under their design limits (Lackner and Zohm have shown this to be impossible for Snowflakes). The strike point is held fixed, so no changes in the divertor or pumping hardware will be needed. The main plasma shape is kept very close to the SD case, so no hardware changes to the main chamber will be needed. Time-dependent ITER-XD operational scenarios are being checked using TSC. This opens the possibility that many XDs could be tested and used to assist in high-power operation on ITER. Because of the toroidally segmented ITER divertor plates, strongly detached operation may be critical for making use of the largest XD flux expansion possible. The flux flaring in XDs is expected to increase the stability of detachment, so that H-mode confinement is not affected. Detachment stability is being examined with SOLPS. This work supported by US DOE Grants DE-FG02-04ER54742 and DE-FG02-04ER54754 and by TACC at UT Austin.

Impact of the plasma geometry on divertor power exhaust: experimental evidence from TCV and simulations with SolEdge2D and TOKAM3X

NASA Astrophysics Data System (ADS)

Gallo, A.; Fedorczak, N.; Elmore, S.; Maurizio, R.; Reimerdes, H.; Theiler, C.; Tsui, C. K.; Boedo, J. A.; Faitsch, M.; Bufferand, H.; Ciraolo, G.; Galassi, D.; Ghendrih, P.; Valentinuzzi, M.; Tamain, P.; the EUROfusion MST1 Team; the TCV Team

2018-01-01

A deep understanding of plasma transport at the edge of magnetically confined fusion plasmas is needed for the handling and control of heat loads on the machine first wall. Experimental observations collected on a number of tokamaks over the last three decades taught us that heat flux profiles at the divertor targets of X-point configurations can be parametrized by using two scale lengths for the scrape-off layer (SOL) transport, separately characterizing the main SOL ({λ }q) and the divertor SOL (S q ). In this work we challenge the current interpretation of these two scale lengths as well as their dependence on plasma parameters by studying the effect of divertor geometry modifications on heat exhaust in the Tokamak à Configuration Variable. In particular, a significant broadening of the heat flux profiles at the outer divertor target is diagnosed while increasing the length of the outer divertor leg in lower single null, Ohmic, L-mode discharges. Efforts to reproduce this experimental finding with both diffusive (SolEdge2D-EIRENE) and turbulent (TOKAM3X) modelling tools confirm the validity of a diffusive approach for simulating heat flux profiles in more traditional, short leg, configurations while highlighting the need of a turbulent description for modified, long leg, ones in which strongly asymmetric divertor perpendicular transport develops.

Flow reversal, convection, and modeling in the DIII-D divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

1998-12-01

Measurements of the parallel Mach number of background plasma in the DIII-D tokamak divertor [M. A. Mahdavi {ital et al.} in {ital Proceedings, 16th International Conference}, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997) Vol. I, p. 397] were performed using a fast scanning Mach probe. The parallel particle flow shows evidence of complex behavior such as reverse flow, i.e., flow away from the target plate, stagnant flow, and large scale convection. For detached discharges, measurements confirm predictions of convective flow towards the divertor target plate at near sound speed over large regions in the divertor. The resulting convected heatmore » flux is a dominant heat transport mechanism in the divertor. For attached discharges with high recycling, particle flow reversal in a thin region at or near the outer separatrix, thereby confirming the existence of a mechanism by which impurities can be transported away from the divertor target plates. Modeling results from the two-dimensional fluid code UEDGE [G. D. Porter and the DIII-D Team, {open_quotes}Divertor characterization experiments and modelling in DIII-D,{close_quotes} in {ital Proceedings of the 23rd European Conference on Controlled Fusion and Plasma Physics}, 24{endash}28 June 1996, Kiev, Ukraine (European Physical Society, Petit-Lancy, Switzerland, 1996), Vol. 20C, Part II, p. 699] can reproduce the main features of the experimental observations. {copyright} {ital 1998 American Institute of Physics.}« less

Measurement of the Density of Base Fluids at Pressures 0.422 to 2.20 Gpa

NASA Technical Reports Server (NTRS)

Hamrock, B. J.; Jacobson, B. O.; Bergstroem, S. I.

1985-01-01

The influence of pressure on the density of six base fluids is experimentally studied for a range of pressures from 0.422 to 2.20 GPa. An important parameter used to describe the results is the change in relative volume with change in pressure dv sub r/dp. For pressures less than the solidification pressure (p ps) a small change in pressure results in a large change in dv sub r/ps. For pressures greater than the solidification pressure (p ps) there is no change in dv sub r/dp with changing pressure. The solidification pressures of the base fluids varies considerably, as do the slopes that the experimental data assumes for p ps. A new formula is developed that describes the effect of pressure on density in terms of four constants. These constants vary for the different base fluids tested.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, M.; Jaworski, M. A.; Kaita, R.

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition to those issues, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues while potentially improving the reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-freemore » core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 500°C than the first wall ~ 600 – 700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover in tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory issue. With an expected T fraction of ≤ 0.7 %, an acceptable level of T inventory can be achieved. In NSTX-U, preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issue on the ARLLD concept. A LL-loop is also being prepared off line for prototyping future use on NSTX-U.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Covele, Brent; Kotschenreuther, M.; Mahajan, S.

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Status of National Spherical Torus Experiment Liquid Lithium Divertor

NASA Astrophysics Data System (ADS)

Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

2009-11-01

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Numerical Calculation of the Peaking Factor of a Water-Cooled W/Cu Monoblock for a Divertor

NASA Astrophysics Data System (ADS)

Han, Le; Chang, Haiping; Zhang, Jingyang; Xu, Tiejun

2015-09-01

In order to accurately predict the incident critical heat flux (ICHF, the heat flux at the heated surface when CHF occurs) of a water-cooled W/Cu monoblock for a divertor, the exact knowledge of its peaking factors (fp) under one-sided heating conditions with different design parameters is a key issue. In this paper, the heat conduction in the solid domain of a water-cooled W/Cu monoblock is calculated numerically by assuming the local heat transfer coefficients (HTC) of the cooling wall to be functions of the local wall temperature, so as to obtain fp. The reliability of the calculation method is validated by an experimental example result, with the maximum error of 2.1% only. The effects of geometric and flow parameters on the fp of a water-cooled W/Cu monoblock are investigated. Within the scope of this study, it is shown that the fp increases with increasing dimensionless W/Cu monoblock width and armour thickness (the shortest distance between the heated surface and Cu layer), and the maximum increases are 43.8% and 22.4% respectively. The dimensionless W/Cu monoblock height and Cu thickness have little effect on fp. The increase of Reynolds number and Jakob number causes the increase of fp, and the maximum increases are 6.8% and 9.6% respectively. Based on the calculated results, an empirical correlation on peaking factor is obtained via regression. These results provide a valuable reference for the thermal-hydraulic design of water-cooled divertors. supported by National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005) and Funding of Jiangsu Innovation Program for Graduate Education, China (CXLX12_0170), the Fundamental Research Funds for the Central Universities of China

Partial pneumonectomy of telomerase null mice carrying shortened telomeres initiates cell growth arrest resulting in a limited compensatory growth response

PubMed Central

Jackson, Sha-Ron; Lee, Jooeun; Reddy, Raghava; Williams, Genevieve N.; Kikuchi, Alexander; Freiberg, Yael; Warburton, David

2011-01-01

Telomerase mutations and significantly shortened chromosomal telomeres have recently been implicated in human lung pathologies. Natural telomere shortening is an inevitable consequence of aging, which is also a risk factor for development of lung disease. However, the impact of shortened telomeres and telomerase dysfunction on the ability of lung cells to respond to significant challenge is still largely unknown. We have previously shown that lungs of late generation, telomerase null B6.Cg-Terctm1Rdp mice feature alveolar simplification and chronic stress signaling at baseline, a phenocopy of aged lung. To determine the role telomerase plays when the lung is challenged, B6.Cg-Terctm1Rdp mice carrying shortened telomeres and wild-type controls were subjected to partial pneumonectomy. We found that telomerase activity was strongly induced in alveolar epithelial type 2 cells (AEC2) of the remaining lung immediately following surgery. Eighty-six percent of wild-type animals survived the procedure and exhibited a burst of early compensatory growth marked by upregulation of proliferation, stress response, and DNA repair pathways in AEC2. In B6.Cg-Terctm1Rdp mice carrying shortened telomeres, response to pneumonectomy was characterized by decreased survival, diminished compensatory lung growth, attenuated distal lung progenitor cell response, persistent DNA damage, and cell growth arrest. Overall, survival correlated strongly with telomere length. We conclude that functional telomerase and properly maintained telomeres play key roles in both long-term survival and the early phase of compensatory lung growth following partial pneumonectomy. PMID:21460122

Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER

DOE PAGES

Chang, Choong Seock; Ku, Seung -Hoe; Loarte, Alberto; ...

2017-07-11

Here, the XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates.

Innovative divertor concept development on DIII-D and EAST

DOE PAGES

Guo, H. Y.; Allen, S.; Canik, J.; ...

2016-06-02

A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lomanowski, B. A., E-mail: b.a.lomanowski@durham.ac.uk; Sharples, R. M.; Meigs, A. G.

2014-11-15

The mirror-linked divertor spectroscopy diagnostic on JET has been upgraded with a new visible and near-infrared grating and filtered spectroscopy system. New capabilities include extended near-infrared coverage up to 1875 nm, capturing the hydrogen Paschen series, as well as a 2 kHz frame rate filtered imaging camera system for fast measurements of impurity (Be II) and deuterium Dα, Dβ, Dγ line emission in the outer divertor. The expanded system provides unique capabilities for studying spatially resolved divertor plasma dynamics at near-ELM resolved timescales as well as a test bed for feasibility assessment of near-infrared spectroscopy.

Real-time radiative divertor feedback control development for the NSTX-U tokamak using a vacuum ultraviolet spectrometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A., E-mail: vlad@llnl.gov; Kaita, R.; Stratton, B.

2016-11-15

A radiative divertor technique is planned for the NSTX-U tokamak to prevent excessive erosion and thermal damage of divertor plasma-facing components in H-mode plasma discharges with auxiliary heating up to 12 MW. In the radiative (partially detached) divertor, extrinsically seeded deuterium or impurity gases are used to increase plasma volumetric power and momentum losses. A real-time feedback control of the gas seeding rate is planned for discharges of up to 5 s duration. The outer divertor leg plasma electron temperature T{sub e} estimated spectroscopically in real time will be used as a control parameter. A vacuum ultraviolet spectrometer McPherson Modelmore » 251 with a fast charged-coupled device detector is developed for temperature monitoring between 5 and 30 eV, based on the Δn = 0, 1 line intensity ratios of carbon, nitrogen, or neon ion lines in the spectral range 300–1600 Å. A collisional-radiative model-based line intensity ratio will be used for relative calibration. A real-time T{sub e}-dependent signal within a characteristic divertor detachment equilibration time of ∼10–15 ms is expected.« less

Real-time radiative divertor feedback control development for the NSTX-U tokamak using a vacuum ultraviolet spectrometer

DOE PAGES

Soukhanovskii, V. A.; Kaita, R.; Stratton, B.

2016-08-04

Here, a radiative divertor technique is planned for the NSTX-U tokamak to prevent excessive erosion and thermal damage of divertor plasma-facing components in H-mode plasma discharges with auxiliary heating up to 12 MW. In the radiative (partially detached) divertor, extrinsically seeded deuterium or impurity gases are used to increase plasma volumetric power and momentum losses. A real-time feedback control of the gas seeding rate is planned for discharges of up to 5 s duration. The outer divertor leg plasma electron temperature T e estimated spectroscopically in real time will be used as a control parameter. A vacuum ultraviolet spectrometer McPhersonmore » Model 251 with a fast charged-coupled device detector is developed for temperature monitoring between 5 and 30 eV, based on the Δn = 0, 1 line intensity ratios of carbon, nitrogen, or neon ion lines in the spectral range 300–1600 Å. A collisional-radiative model-based line intensity ratio will be used for relative calibration. A real-time T e-dependent signal within a characteristic divertor detachment equilibration time of ~10–15 ms is expected.« less

Tungsten migration in Alcator C-Mod: sputtering and melting

NASA Astrophysics Data System (ADS)

Wright, G. M.; Barnard, H.; Lipschultz, B.; Whyte, D. G.

2010-11-01

A row of bulk tungsten (W) tiles were installed near the typical outer strike-point location in the Alcator C-Mod divertor in 2007. In the 2009/2010 campaign, one of the W tiles mechanically failed resulting in significant W melting at that location. Post-campaign PIXE surface analysis has been used to observe tungsten (W) deposition and migration patterns in the divertor for the typical operations (sputtering only) and operation with melted components. For sputtering conditions, W deposition of up to 20 nm equivalent thickness is observed at various divertor surfaces indicating prompt re-deposition at the outer divertor, neutral and ion transport through the private-flux region and ion transport in the scrape off layer. For melting conditions, W deposition of up to 400 nm equivalent thickness is observed at some locations at the outer divertor. However, the toroidal distribution of W on the outer divertor is strongly non-uniform. There is no W deposition measured on the inner wall limiter. These results indicate that impurity migration is affected by the erosion mechanism and source, with the migration from melting being less predictable and uniform than from the sputtering case. Supported by USDoE award DE-SC00-02060.

Scrape off layer modelling studies for SST-I

NASA Astrophysics Data System (ADS)

Warrier, M.; Jaishankar, S.; Deshpande, S.; Coster, D.; Schneider, R.; Chaturvedi, S.; Srinivasan, R.; Braams, B. J.; SST Team

SOL modelling results for SST-1 (SST Team, Proceedings of the 16th IEEE/NPSS Symposium on Fusion Engineering, Champaign, IL, vol. II, 1995, p. 481) show a sheath limited flow regime. This is due to the low edge densities required by lower hybrid current drive (LHCD), coupled with high power input per unit volume. Coupled plasma-neutral transport studies using B2-Eirene [R. Schneider et al., J. Nucl. Mater. 196-198 (1992) 810] show significantly high charge exchange losses and radiated power from the core. It also shows that the heat flux to the inner divertor is higher than that to the outer divertor due to thinner inner SOL widths. The Monte-Carlo neutral transport code DEGAS [D. Heifitz et al., J. Comput. Phys. 46 (1982) 309] was used to optimise the baffle plate geometry and it was seen that a configuration where the baffle plate shields the main plasma from the divertor strike point results in reduced backflow of neutrals. The divertor erosion code DIVER (M. Warrier et al., SST Divertor Modelling Report, 1996-1997) was used to predict a steady state operating temperature for the SST divertor plate lying in the range 750-1000°C for which the erosion will be minimum.

Results from core-edge experiments in high Power, high performance plasmas on DIII-D

DOE PAGES

Petrie, T. W.; Fenstermacher, M. E.; Holcomb, C. T.; ...

2016-12-24

Here, significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q ⊥ P) ∝ [P SOL x I P] 0.92 for P SOL = 8-19 MW and I P = 1.0–1.4 MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-Dmore » plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q ⊥ P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core.« less

Electric field divertor plasma pump

DOEpatents

Schaffer, Michael J.

1994-01-01

An electric field plasma pump includes a toroidal ring bias electrode (56) positioned near the divertor strike point of a poloidal divertor of a tokamak (20), or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix (40) of the poloidal divertor contacts the ring electrode (56), which then also acts as a divertor plate. A plenum (54) or other duct near the electrode (56) includes an entrance aperture open to receive electrically-driven plasma. The electrode (56) is insulated laterally with insulators (63,64), one of which (64) is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode (56) and a vacuum vessel wall (22), with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E.times.B/B.sup.2 drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable.

Electric field divertor plasma pump

DOEpatents

Schaffer, M.J.

1994-10-04

An electric field plasma pump includes a toroidal ring bias electrode positioned near the divertor strike point of a poloidal divertor of a tokamak, or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix of the poloidal divertor contacts the ring electrode, which then also acts as a divertor plate. A plenum or other duct near the electrode includes an entrance aperture open to receive electrically-driven plasma. The electrode is insulated laterally with insulators, one of which is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode and a vacuum vessel wall, with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E [times] B/B[sup 2] drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable. 11 figs.

ELM elimination with lithium aerosol injection in upper-single null discharges using the tungsten divertor in EAST

NASA Astrophysics Data System (ADS)

Sun, Z.; Maingi, R.; Hu, J.; Lunsford, R.; Diallo, A.; Tritz, K.; Osborne, T.; Canik, J.; Zuo, G.; Wang, L.; Xu, G.; Gong, X.; EAST Team Team

2017-10-01

A reproducible, fully non-inductive H-mode regime devoid of large ELMs has been achieved by continuous Li injection in EAST into the upper `ITER-like' tungsten divertor, extending previous results on the graphite divertor. These discharges did not suffer from density or impurity accumulation, and maintained constant core radiated power. The new results extend the energy confinement multiplier H98(y,2) 1.2, as compared to H98(y,2) 0.75 previously on the graphite divertor. The observed ELM elimination is correlated with a decrease in particle recycling, as expected from the strong Li coating before the experiment, and real-time Li aerosol injection. In addition, core W concentration was reduced during the Li injection. ELM elimination is likely related to the reduced recycling and density /temperature profile changes. A low-n electromagnetic coherent mode (MCM) at 40kHz became stronger in amplitude and also more coherent. The MCM shows strong magnetic fluctuations as measured by fast Mirnov coils, but weak density fluctuations. As compared to the graphite divertor, Li injection into the tungsten divertor eliminated ELMs at twice the previous auxiliary heating power, and reduced pedestal collisionality.

Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

NASA Astrophysics Data System (ADS)

Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

2017-12-01

Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part of the coolant pipe. These results were acceptable for the plasma facing and structural materials.

A Fast Visible Camera Divertor-Imaging Diagnostic on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roquemore, A; Maingi, R; Lasnier, C

2007-06-19

In recent campaigns, the Photron Ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the National Spherical Torus Experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE Localized Modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX smallmore » type V ELM regime [2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the Photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera [4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.« less

Using the tritium plasma experiment to evaluate ITER PFC safety

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Anderl, Robert A.; Bartlit, John R.; Causey, Rion A.; Haines, John R.

1993-06-01

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 1023 ions/m2.s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.

Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

DOE PAGES

Ono, M.; Jaworski, M. A.; Kaita, R.; ...

2016-08-05

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition to those issues, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues while potentially improving the reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-freemore » core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 500°C than the first wall ~ 600 – 700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover in tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory issue. With an expected T fraction of ≤ 0.7 %, an acceptable level of T inventory can be achieved. In NSTX-U, preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issue on the ARLLD concept. A LL-loop is also being prepared off line for prototyping future use on NSTX-U.« less

The snowflake divertor

DOE PAGES

Ryutov, D. D.; Soukhanovskii, V. A.

2015-11-17

The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. One of the most interesting effects of the snowflake geometry is the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation tomore » the existing theoretical models is described. Divertor concepts utilizing the properties of a snowflake configuration are briefly discussed.« less

Divertor target for magnetic containment device

DOEpatents

Luzzi, Jr., Theodore E.

1982-01-01

In a plasma containment device of a type having superconducting field coils for magnetically shaping the plasma into approximately the form of a torus, an improved divertor target for removing impurities from a "scrape off" region of the plasma comprises an array of water cooled swirl tubes onto which the scrape off flux is impinged. Impurities reflected from the divertor target are removed from the target region by a conventional vacuum getter system. The swirl tubes are oriented and spaced apart within the divertor region relative to the incident angle of the scrape off flux to cause only one side of each tube to be exposed to the flux to increase the burnout rating of the target. The divertor target plane is oriented relative to the plane of the path of the scrape off flux such that the maximum heat flux onto a swirl tube is less than the tube design flux. The containment device is used to contain the plasma of a tokamak fusion reactor and is applicable to other long pulse plasma containment systems.

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

DOE PAGES

Maingi, R.; Hu, J. S.; Sun, Z.; ...

2018-01-05

Here, we report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previousmore » ELM elimination results via Li injection into the lower carbon divertor in EAST. These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs, highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.« less

Minimum magnetic curvature for resilient divertors using Compact Toroidal Hybrid geometry

NASA Astrophysics Data System (ADS)

Bader, A.; Hegna, C. C.; Cianciosa, M.; Hartwell, G. J.

2018-05-01

The properties of resilient divertors are explored using equilibria derived from Compact Toroidal Hybrid (CTH) geometries. Resilience is defined here as the robustness of the strike point patterns as the plasma geometry and/or plasma profiles are changed. The addition of plasma current in the CTH configurations significantly alters the shape of the last closed flux surface and the rotational transform profile, however, it does not alter the strike point pattern on the target plates, and hence has resilient divertor features. The limits of when a configuration transforms to a resilient configuration is then explored. New CTH-like configurations are generated that vary from a perfectly circular cross section to configurations with increasing amounts of toroidal shaping. It is found that even small amounts of toroidal shaping lead to strike point localization that is similar to the standard CTH configuration. These results show that only a small degree of three-dimensional shaping is necessary to produce a resilient divertor, implying that any highly shaped optimized stellarator will possess the resilient divertor property.

Development of heat sink concept for near-term fusion power plant divertor

NASA Astrophysics Data System (ADS)

Rimza, Sandeep; Khirwadkar, Samir; Velusamy, Karupanna

2017-04-01

Development of an efficient divertor concept is an important task to meet in the scenario of the future fusion power plant. The divertor, which is a vital part of the reactor has to discharge the considerable fraction of the total fusion thermal power (∼15%). Therefore, it has to survive very high thermal fluxes (∼10 MW/m2). In the present paper, an efficient divertor heat exchanger cooled by helium is proposed for the fusion tokamak. The Plasma facing surface of divertor made-up of several modules to overcome the stresses caused by high heat flux. The thermal hydraulic performance of one such module is numerically investigated in the present work. The result shows that the proposed design is capable of handling target heat flux values of 10 MW/m2. The computational model has been validated against high-heat flux experiments and a satisfactory agreement is noticed between the present simulation and the reported results.

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

NASA Astrophysics Data System (ADS)

Maingi, R.; Hu, J. S.; Sun, Z.; Tritz, K.; Zuo, G. Z.; Xu, W.; Huang, M.; Meng, X. C.; Canik, J. M.; Diallo, A.; Lunsford, R.; Mansfield, D. K.; Osborne, T. H.; Gong, X. Z.; Wang, Y. F.; Li, Y. Y.; EAST Team

2018-02-01

We report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3-5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previous ELM elimination results via Li injection into the lower carbon divertor in EAST (Hu et al 2015 Phys. Rev. Lett. 114 055001). These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs (Lang et al 2017 Nucl. Fusion 57 016030), highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maingi, R.; Hu, J. S.; Sun, Z.

Here, we report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previousmore » ELM elimination results via Li injection into the lower carbon divertor in EAST. These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs, highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.« less

Double-null divertor configuration discharge and disruptive heat flux simulation using TSC on EAST

NASA Astrophysics Data System (ADS)

Bo, SHI; Jinhong, YANG; Cheng, YANG; Desheng, CHENG; Hui, WANG; Hui, ZHANG; Haifei, DENG; Junli, QI; Xianzu, GONG; Weihua, WANG

2018-07-01

The tokamak simulation code (TSC) is employed to simulate the complete evolution of a disruptive discharge in the experimental advanced superconducting tokamak. The multiplication factor of the anomalous transport coefficient was adjusted to model the major disruptive discharge with double-null divertor configuration based on shot 61 916. The real-time feed-back control system for the plasma displacement was employed. Modeling results of the evolution of the poloidal field coil currents, the plasma current, the major radius, the plasma configuration all show agreement with experimental measurements. Results from the simulation show that during disruption, heat flux about 8 MW m‑2 flows to the upper divertor target plate and about 6 MW m‑2 flows to the lower divertor target plate. Computations predict that different amounts of heat fluxes on the divertor target plate could result by adjusting the multiplication factor of the anomalous transport coefficient. This shows that TSC has high flexibility and predictability.

Experience on divertor fuel retention after two ITER-Like Wall campaigns

NASA Astrophysics Data System (ADS)

Heinola, K.; Widdowson, A.; Likonen, J.; Ahlgren, T.; Alves, E.; Ayres, C. F.; Baron-Wiechec, A.; Barradas, N.; Brezinsek, S.; Catarino, N.; Coad, P.; Guillemaut, C.; Jepu, I.; Krat, S.; Lahtinen, A.; Matthews, G. F.; Mayer, M.; Contributors, JET

2017-12-01

The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during the operation of a future fusion reactor. Present work summarizes and reports the plasma fuel retention in the divertor resulting from the two first experimental campaigns with the ITER-Like Wall. The deposition pattern in the divertor after the second campaign shows same trend as was observed after the first campaign: highest deposition of 10-15 μm was found on the top part of the inner divertor. Due to the change in plasma magnetic configurations from the first to the second campaign, and the resulted strike point locations, an increase of deposition was observed on the base of the divertor. The deuterium retention was found to be affected by the hydrogen plasma experiments done at the end of second experimental campaign.

Simple Map with Low MN Perturbation for a Single-Null Divertor Tokamak with Constant Width of Stochastic Layer

NASA Astrophysics Data System (ADS)

Verma, Arun; Smith, Terry; Punjabi, Alkesh; Boozer, Allen

1996-11-01

In this work, we investigate the effects of low MN perturbations in a single-null divertor tokamak with stochastic scrape-off layer. The unperturbed magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). We choose the combinations of the map parameter k, and the strength of the low MN perturbation such that the width of stochastic layer remains unchanged. We give detailed results on the effects of low MN perturbation on the magnetic topology of the stochastic layer and on the footprint of field lines on the divertor plate given the constraint of constant width of the stochastic layer. The low MN perturbations occur naturally and therefore their effects are of considerable importance in tokamak divertor physics. This work is supported by US DOE OFES. Use of CRAY at HU and at NERSC is gratefully acknowledged.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Last Good Surface

NASA Astrophysics Data System (ADS)

Phan, Huong; Ali, Halima; Punjabi, Alkesh

2000-10-01

The Simple Map^1 is a representation of the magnetic field inside a single-null divertor tokamak. It is given by the equations: X_n+1=X_n kYn (1-Y_n), Y_n+1= Y_n+kX_n+1. These equations mimic the motion of the magnetic field lines in a single-null divertor tokamak. The fixed stable point is (0,0) and the unstable fixed oint is (0,1). k is fixed at 0.60. In our work, the starting values of Y in the map is kept in the interval of 0 to 1, and the starting value of X is 0. Using the successive bifurcation method, we first run these equations for 10^6 iterations to find the approximate value of Y when chaos occurs. We examine the neighborhood of this Y value to find the exact value of Y for the last good surface. We call this value Y_lgs. We find Y_lgs to be 0.997135768 for k=0.60 and X=0. This work is supported by US DOE OFES. Ms. Huong Phan is a HU CFRT Summer Fusion High School Workshop Scholar from Andrew P. Hill High School in California. She is supported by NASA SHARP Plus Program. 1. Punjabi A, Verma A and Boozer A, Phys Rev Lett 69 3322 (1992) and J Plasma Phys 52 91 (1994)

On Heat Loading, Novel Divertors, and Fusion Reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike

2006-10-01

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction (fRad) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of fRad will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core fRad within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.

Upgraded divertor Thomson scattering system on DIII-D

NASA Astrophysics Data System (ADS)

Glass, F.; Carlstrom, T. N.; Du, D.; McLean, A. G.; Taussig, D. A.; Boivin, R. L.

2016-11-01

A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (Te in the range of 0.5 eV-2 keV, ne in the range of 5 × 1018-1 × 1021 m3) for both low Te in detachment and high Te measurement up beyond the separatrix.

Plasma power recycling at the divertor surface

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, Xian -Zhu; Guo, Zehua

With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

Plasma power recycling at the divertor surface

DOE PAGES

Tang, Xian -Zhu; Guo, Zehua

2016-12-03

With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

Conceptual design of divertor and first wall for DEMO-FNS

NASA Astrophysics Data System (ADS)

Sergeev, V. Yu.; Kuteev, B. V.; Bykov, A. S.; Gervash, A. A.; Glazunov, D. A.; Goncharov, P. R.; Dnestrovskij, A. Yu.; Khayrutdinov, R. R.; Klishchenko, A. V.; Lukash, V. E.; Mazul, I. V.; Molchanov, P. A.; Petrov, V. S.; Rozhansky, V. A.; Shpanskiy, Yu. S.; Sivak, A. B.; Skokov, V. G.; Spitsyn, A. V.

2015-11-01

Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m-2. Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280-300 °С at 5 MW m-2 and 600-650 °С at 10.5 MW m-2. The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

Upgraded divertor Thomson scattering system on DIII-D.

PubMed

Glass, F; Carlstrom, T N; Du, D; McLean, A G; Taussig, D A; Boivin, R L

2016-11-01

A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard - beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror - and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (T e in the range of 0.5 eV-2 keV, n e in the range of 5 × 10 18 -1 × 10 21 m 3 ) for both low T e in detachment and high T e measurement up beyond the separatrix.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono, M. Jaworski, R. Kaita, C. N. Skinner, J.P. Allain, R. Maingi, F. Scotti, V.A. Soukhanovskii, and the NSTX-U Team

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTXU, the PMI research has received a strong emphasis. With ~ 15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m2 . To support the PMI research, a comprehensive set of PMI diagnostic tools are being implemented. The snow-flake configuration can produce exceptionally high divertor flux expansion of up to ~ 50.more » Combined with the radiative divertor concept, the snow-flake configuration has reduced the divertor heat flux by an order of magnitude in NSTX. Another area of active PMI investigation is the effect of divertor lithium coating (both in solid and liquid phases). The overall NSTX lithium PFC coating results suggest exciting opportunities for future magnetic confinement research including significant electron energy confinement improvements, Hmode power threshold reduction, the control of Edge Localized Modes (ELMs), and high heat flux handling. To support the NSTX-U/PPPL PMI research, there are also a number of associated PMI facilities implemented at PPPL/Princeton University including the Liquid Lithium R&D facility, Lithium Tokamak Experiment, and Laboratories for Materials Characterization and Surface Chemistry.« less

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

NASA Astrophysics Data System (ADS)

Dennett, C. A.; Cao, N. M.; Creely, A. J.; Hecla, J.; Hoffman, H.; Kuang, A. Q.; Major, M.; Ruiz Ruiz, J.; Tinguely, R. A.; Tolman, E. A.; Brunner, D.; Labombard, B.; Sorbom, B. N.; Whyte, D. G.; Grover, P.; Laughman, C.

2017-10-01

The ARC pilot plant conceptual design study has been extended to explore solutions for managing heat exhaust resulting from 525 MW of fusion power in a compact (R 3.3 m) tokamak. Superconducting poloidal field coils are configured to produce double-null equilibria that support X-point target divertors while maintaining the original core plasma shape and toroidal field coil size. Long outer divertor legs are appended to the original vacuum vessel, providing both large surface areas for surface dissipation of radiative heat and significantly reduced neutron damage for divertor components. A molten salt FLiBe blanket adequately shields all superconductors and functions as a tritium breeder, with advanced neutronics calculations indicating a tritium breeding ratio of 1.08. In addition, FLiBe is used as the active coolant for the entire vessel. A tungsten swirl-tube cooling channel is implemented in the divertor, capable of exhausting 12 MW/m2, heat flux while keeping total FliBe pumping power below 1% of fusion power. Finally, three novel diagnostics are explored: Cherenkov radiation emitted in FLiBe to measure fusion reaction rate, microwave interferometry to measure divertor detachment front location, and IR imaging through the FLiBe blanket to monitor selected divertor ``hotspots.''

Actively convected liquid metal divertor

NASA Astrophysics Data System (ADS)

Shimada, Michiya; Hirooka, Yoshi

2014-12-01

The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated with magnetic fusion experiments such as ITER. This issue will be aggravated even more for DEMO and power reactors because the divertor heat load will be significantly higher and yet the use of copper would not be allowed as the heat sink material. Instead, reduced activation ferritic/martensitic steel alloys with heat conductivities substantially lower than that of copper, will be used as the structural materials. The present proposal is to fill the lower part of the vacuum vessel with liquid metals with relatively low melting points and low chemical activities including Ga and Sn. The divertor modules, equipped with electrodes and cooling tubes, are immersed in the liquid metal. The electrode, placed in the middle of the liquid metal, can be biased positively or negatively with respect to the module. The j × B force due to the current between the electrode and the module provides a rotating motion for the liquid metal around the electrodes. The rise in liquid temperature at the separatrix hit point can be maintained at acceptable levels from the operation point of view. As the rotation speed increases, the current in the liquid metal is expected to decrease due to the v × B electromotive force. This rotating motion in the poloidal plane will reduce the divertor heat load significantly. Another important benefit of the convected liquid metal divertor is the fast recovery from unmitigated disruptions. Also, the liquid metal divertor concept eliminates the erosion problem.

Ratio of dietary rumen degradable protein to rumen undegradable protein affects nitrogen partitioning but does not affect the bovine milk proteome produced by mid-lactation Holstein dairy cows.

PubMed

Tacoma, R; Fields, J; Ebenstein, D B; Lam, Y-W; Greenwood, S L

2017-09-01

Little is known about the bovine milk proteome or whether it can be affected by diet. The objective of this study was to determine if the dietary rumen degradable protein (RDP):rumen undegradable protein (RUP) ratio could alter the bovine milk proteome. Six Holstein cows (parity: 2.5 ± 0.8) in mid lactation were blocked by days in milk (80 ± 43 d in milk) and milk yield (57.5 ± 6.0 kg) and randomly assigned to treatment groups. The experiment was conducted as a double-crossover design consisting of three 21-d periods. Within each period, treatment groups received diets with either (1) a high RDP:RUP ratio (RDP treatment: 62.4:37.6% of crude protein) or (2) a low RDP:RUP ratio (RUP treatment: 51.3:48.7% of crude protein). Both diets were isonitrogenous and isoenergetic (crude protein: 18.5%, net energy for lactation: 1.8 Mcal/kg of dry matter). To confirm N and energy status of cows, dry matter intake was determined daily, rumen fluid samples were collected for volatile fatty acid analysis, blood samples were collected for plasma glucose, β-hydroxybutyrate, urea nitrogen, and fatty acid analysis, and total 24-h urine and fecal samples were collected for N analysis. Milk samples were collected to determine the general milk composition and the protein profile. Milk samples collected for high-abundance protein analysis were subjected to HPLC analysis to determine the content of α-casein, β-casein, and κ-casein, as well as α-lactalbumin and β-lactoglobulin. Samples collected for low-abundance protein analysis were fractionated, enriched using ProteoMiner treatment, and separated using sodium dodecyl sulfate-PAGE. After excision and digestion, the peptides were analyzed using liquid chromatography (LC) tandem mass spectrometry (MS/MS). The LC-MS/MS data were analyzed using PROC GLIMMIX of SAS (version 9.4, SAS Institute Inc., Cary, NC) and adjusted using the MULTTEST procedure. All other parameters were analyzed using PROC MIXED of SAS. No treatment differences were observed in dry matter intake, milk yield, general milk composition, plasma parameters, or rumen volatile fatty acid concentrations, indicating no shift in total energy or protein available. Milk urea N and plasma urea N concentrations were higher in the RDP group, indicating some shift in N partitioning due to diet. A total of 595 milk proteins were identified, with 83% of these proteins known to be involved in cellular processes. Although none of the low-abundance proteins identified by LC-MS/MS were affected by diet, feeding a diet high in RUP decreased β-casein, κ-casein, and total milk casein concentration. Further investigations of the interactions between diet and the milk protein profile are needed to manipulate the milk proteome using diet. The Authors. Published by the Federation of Animal Science Societies and Elsevier Inc. on behalf of the American Dairy Science Association®. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/3.0/).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

El efecto de la panfotocoagulación con láser en edema macular diabético con el fotocoagulador Pascal® versus el láser de argón convencional.

PubMed

Mahgoub, Mohamed M; Macky, Tamer A

2017-07-11

Objetivo: El objetivo de este estudio fue comparar el efecto de la panfotocoagulación (PFC) en el edema macular diabético (EMD) en pacientes con retinopatía diabética proliferativa (RDP) con el fotocoagulador Pascal® (FP) vs. un fotocoagulador con láser de argón convencional (FLAC). Métodos: Se aleatorizó el uso de FP o FLAC en ochenta ojos con RDP y EMD con afectación central de la mácula. Ambos grupos tuvieron una evaluación de base de mejor agudeza visual corregida y fueron examinados con tomografía de coherencia óptica y angiografía con fluoresceína. Resultados: El número medio de disparos de láser en los grupos de FP y FLAC fue 1.726,10 y 752,00 en la sesión 1 y 1.589,00 y 830,00 (p < 0,001) en la sesión 2, respectivamente. El grosor foveal central (GFC) medio antes de comenzar el estudio fue 306 ± 100 y 314 ± 98 en los grupos de FP y FLAC, respectivamente. A las 8 semanas, el GFC medio fue 332 ± 116 y 347 ± 111 en los grupos de FP y FLAC, respectivamente (p > 0,05). La MAVC media fue similar durante el periodo de estudio y no hubo ninguna diferencia significativa entre los grupos (p > 0,05). Conclusiones: El FP y el FLAC mostraron efectos similares en el EMD en ojos con RDP y fueron igualmente seguros sin un aumento significativo del GFC. © 2017 S. Karger AG, Basel.

Activation and Environmental Aspects of In-Vacuum Vessel Components of CFETR

NASA Astrophysics Data System (ADS)

Zhang, Xiaokang; Liu, Songlin; Zhu, Qingjun; Gao, Fangfang; Li, Jia

2016-11-01

The water-cooled ceramic breeder (WCCB) blanket is one of the three candidates of China's Fusion Engineering Test Reactor (CFETR). The evaluation of the radioactivity and decay heat produced by neutrons for the in-vacuum vessel components is essential for the assessment of radioactive wastes and the safety of CFETR. The activation calculation of CFETR in-vacuum vessel components was carried out by using the Monte Carlo N-Particle Transport Code MCNP, IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, and the nuclear inventory code FISPACT-2007 and corresponding EAF-2007 libraries. In these analyses, the three-dimensional (3-D) neutronics model was employed and the WCCB blanket, the divertor, and the shield were modeled in detail to provide the detailed spatial distribution of the neutron flux and energy spectra. Then the neutron flux, energy spectra and the materials specification were transferred to FISPACT for the activation calculation with an assumed irradiation scenario of CFETR. This paper presents the main results of the activation analysis to evaluate the radioactivity, the decay heat, the contact dose, and the waste classification of the radioactive materials. At the time of shutdown, the activity of the WCCB blanket is 1.88×1019 Bq and the specific activity, the decay heat and the contact dose rate are 1.7 × 1013 Bq/kg, 3.05 MW, and 2.0 × 103 Sv/h respectively. After cooling for 100 years, 79% (4166.4 tons) radioactive wastes produced from the blanket, divertor, high temperature shield (HTS) and low temperature shield (LTS) need near surface disposal, while 21% (1112.3 tons) need geological disposal. According to results of the contact dose rate, all the components of the blanket, divertor, HTS and LTS could potentially be recycled after shutdown by using advanced remote handling equipment. In addition, the selection of Eurofer97 or RAFM for the divertor is better than that of SS316 because SS316 makes the activity of the divertor-body keep at a relatively high level. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015BG108002, 2014GB122000, 2014GB119000), National Natural Science Foundation of China (No. 11175207)

Divertor power and particle fluxes between and during type-I ELMs in the ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Dux, R.; Eich, T.; Fischer, R.; Giannone, L.; Harhausen, J.; Herrmann, A.; Müller, H. W.; Pautasso, G.; Wischmeier, M.; ASDEX Upgrade Team

2008-08-01

Particle, electric charge and power fluxes for type-I ELMy H-modes are measured in the divertor of the ASDEX Upgrade tokamak by triple Langmuir probes, shunts, infrared (IR) thermography and spectroscopy. The discharges are in the medium to high density range, resulting in predominantly convective edge localized modes (ELMs) with moderate fractional stored energy losses of 2% or below. Time resolved data over ELM cycles are obtained by coherent averaging of typically one hundred similar ELMs, spatial profiles from the flush-mounted Langmuir probes are obtained by strike point sweeps. The application of simple physics models is used to compare different diagnostics and to make consistency checks, e.g. the standard sheath model applied to the Langmuir probes yields power fluxes which are compared with the thermographic measurements. In between ELMs, Langmuir probe and thermography power loads appear consistent in the outer divertor, taking into account additional load due to radiation and charge exchange neutrals measured by thermography. The inner divertor is completely detached and no significant power flow by charged particles is measured. During ELMs, quite similar power flux profiles are found in the outer divertor by thermography and probes, albeit larger uncertainties in Langmuir probe evaluation during ELMs have to be taken into account. In the inner divertor, ELM power fluxes from thermography are a factor 10 larger than those derived from probes using the standard sheath model. This deviation is too large to be caused by deficiencies of probe analysis. The total ELM energy deposition from IR is about a factor 2 higher in the inner divertor compared with the outer divertor. Spectroscopic measurements suggest a quite moderate contribution of radiation to the target power load. Shunt measurements reveal a significant positive charge flow into the inner target during ELMs. The net number of elementary charges correlates well with the total core particle loss obtained from highly resolved density profiles. As a consequence, the discrepancy between probe and IR measurements is attributed to the ion power channel via a high mean impact energy of the ions at the inner target. The dominant contributing mechanism is proposed to be the directed loss of ions from the pedestal region into the inner divertor.

A multichannel visible spectroscopy system for the ITER-like W divertor on EAST.

PubMed

Mao, Hongmin; Ding, Fang; Luo, Guang-Nan; Hu, Zhenhua; Chen, Xiahua; Xu, Feng; Yang, Zhongshi; Chen, Jingbo; Wang, Liang; Ding, Rui; Zhang, Ling; Gao, Wei; Xu, Jichan; Wu, Chengrui

2017-04-01

To facilitate long-pulse high power operation, an ITER-like actively cooled tungsten (W) divertor was installed in Experimental Advanced Superconducting Tokamak (EAST) to replace the original upper graphite divertor in 2014. A dedicated multichannel visible spectroscopic diagnostic system has been accordingly developed for the characterization of the plasma and impurities in the W divertor. An array of 22 lines-of-sight (LOSs) provides a profile measurement of the light emitted from the plasma along upper outer divertor, and the other 17 vertical LOSs view the upper inner divertor, achieving a 13 mm poloidal resolution in both regions. The light emitted from the plasma is collected by a specially designed optical lens assembly and then transferred to a Czerny-Turner spectrometer via 40 m quartz fibers. At the end, the spectra dispersed by the spectrometer are recorded with an Electron-Multiplying Charge Coupled Device (EMCCD). The optical throughput and quantum efficiency of the system are optimized in the wavelength range 350-700 nm. The spectral resolution/coverage can be adjusted from 0.01 nm/3 nm to 0.41 nm/140 nm by switching the grating with suitable groove density. The frame rate depends on the setting of LOS number in EMCCD and can reach nearly 2 kHz for single LOS detection. The light collected by the front optical lens can also be divided and partly transferred to a photomultiplier tube array with specified bandpass filter, which can provide faster sampling rates by up to 200 kHz. The spectroscopic diagnostic is routinely operated in EAST discharges with absolute optical calibrations applied before and after each campaign, monitoring photon fluxes from impurities and H recycling in the upper divertor. This paper presents the technical details of the diagnostic and typical measurements during EAST discharges.

The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, J. R.; Fishpool, G. M.; Thornton, A. J.

2015-09-15

The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFRmore » of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.« less

A multichannel visible spectroscopy system for the ITER-like W divertor on EAST

NASA Astrophysics Data System (ADS)

Mao, Hongmin; Ding, Fang; Luo, Guang-Nan; Hu, Zhenhua; Chen, Xiahua; Xu, Feng; Yang, Zhongshi; Chen, Jingbo; Wang, Liang; Ding, Rui; Zhang, Ling; Gao, Wei; Xu, Jichan; Wu, Chengrui

2017-04-01

To facilitate long-pulse high power operation, an ITER-like actively cooled tungsten (W) divertor was installed in Experimental Advanced Superconducting Tokamak (EAST) to replace the original upper graphite divertor in 2014. A dedicated multichannel visible spectroscopic diagnostic system has been accordingly developed for the characterization of the plasma and impurities in the W divertor. An array of 22 lines-of-sight (LOSs) provides a profile measurement of the light emitted from the plasma along upper outer divertor, and the other 17 vertical LOSs view the upper inner divertor, achieving a 13 mm poloidal resolution in both regions. The light emitted from the plasma is collected by a specially designed optical lens assembly and then transferred to a Czerny-Turner spectrometer via 40 m quartz fibers. At the end, the spectra dispersed by the spectrometer are recorded with an Electron-Multiplying Charge Coupled Device (EMCCD). The optical throughput and quantum efficiency of the system are optimized in the wavelength range 350-700 nm. The spectral resolution/coverage can be adjusted from 0.01 nm/3 nm to 0.41 nm/140 nm by switching the grating with suitable groove density. The frame rate depends on the setting of LOS number in EMCCD and can reach nearly 2 kHz for single LOS detection. The light collected by the front optical lens can also be divided and partly transferred to a photomultiplier tube array with specified bandpass filter, which can provide faster sampling rates by up to 200 kHz. The spectroscopic diagnostic is routinely operated in EAST discharges with absolute optical calibrations applied before and after each campaign, monitoring photon fluxes from impurities and H recycling in the upper divertor. This paper presents the technical details of the diagnostic and typical measurements during EAST discharges.

Various divertor biasing configurations and improved divertor performance with biasing on Tokamak de Varennes (TdeV)*

DOE Office of Scientific and Technical Information (OSTI.GOV)

Decoste, R.; Lachambre, J.; Abel, G.

1994-05-01

Electrically insulated divertor plates are used on TdeV (Tokamak de Varennes) [18[ital th] [ital EPS] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Physics] Berlin (European Physical Society, Petit-Lancy, 1991), Vol. 15C, Part I, pp. 1--141] to produce various biasing configurations, which can be decomposed into two basic modes. Plasma biasing, with a radial electric field [ital E][sub [ital r

Upgraded divertor Thomson scattering system on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.

2016-11-15

A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, beforemore » being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (T{sub e} in the range of 0.5 eV–2 keV, n{sub e} in the range of 5 × 10{sup 18}–1 × 10{sup 21} m{sup 3}) for both low T{sub e} in detachment and high T{sub e} measurement up beyond the separatrix.« less

Active Control of Power Exhaust in Strongly Heated ASDEX Upgrade Plasmas

NASA Astrophysics Data System (ADS)

Dux, Ralph; Kallenbach, Arne; Bernert, Matthias; Eich, Thomas; Fuchs, Christoph; Giannone, Louis; Herrmann, Albrecht; Schweinzer, Josef; Treutterer, Wolfgang

2012-10-01

Due to the absence of carbon as an intrinsic low-Z radiator, and tight limits for the acceptable power load on the divertor target, ITER will rely on impurity seeding for radiative power dissipation and for generation of partial detachment. The injection of more than one radiating species is required to optimise the power removal in the main plasma and in the divertor region, i.e. a low-Z species for radiation in the divertor and a medium-Z species for radiation in the outer core plasma. In ASDEX Upgrade, a set of robust sensors, which is suitable to feedback control the radiated power in the main chamber and the divertor as well as the electron temperature at the target, has been developed. Different feedback schemes were applied in H-mode discharges with a maximum heating power of up to 23,W, i.e. at ITER values of P/R (power per major radius) to control all combinations of power flux into the divertor region, power flux onto the target or electron temperature at the target through injection of nitrogen as the divertor radiator and argon as the main chamber radiator. Even at the highest heating powers the peak heat flux density at the target is kept at benign values. The control schemes and the plasma behaviour in these discharges will be discussed.

Erosion and deposition in the JET divertor during the second ITER-like wall campaign

NASA Astrophysics Data System (ADS)

Mayer, M.; Krat, S.; Baron-Wiechec, A.; Gasparyan, Yu; Heinola, K.; Koivuranta, S.; Likonen, J.; Ruset, C.; de Saint-Aubin, G.; Widdowson, A.; Contributors, JET

2017-12-01

Erosion of plasma-facing materials and successive transport and redeposition of eroded material are crucial processes determining the lifetime of plasma-facing components and the trapped tritium inventory in redeposited material layers. Erosion and deposition in the JET divertor were studied during the second JET ITER-like wall campaign ILW-2 in 2013-2014 by using a poloidal row of specially prepared divertor marker tiles including the tungsten bulk tile 5. The marker tiles were analyzed using elastic backscattering with 3-4.5 MeV incident protons and nuclear reaction analysis using 0.8-4.5 MeV 3He ions before and after the campaign. The erosion/deposition pattern observed during ILW-2 is qualitatively comparable to the first campaign ILW-1 in 2011-2012: deposits consist mainly of beryllium with 5-20 at.% of carbon and oxygen and small amounts of Ni and W. The highest deposition with deposited layer thicknesses up to 30 μm per campaign is still observed on the upper and horizontal parts of the inner divertor. Outer divertor tiles 5, 6, 7 and 8 are net W erosion areas. The observed D inventory is roughly comparable to the inventory observed during ILW-1. The results obtained during ILW-2 therefore confirm the positive results observed in ILW-1 with respect to reduced material deposition and hydrogen isotopes retention in the divertor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dr. Ricardo Maqueda; Dr. Fred M. Levinton

Nova Photonics, Inc. has a collaborative effort at the National Spherical Torus Experiment (NSTX). This collaboration, based on fast imaging of visible phenomena, has provided key insights on edge turbulence, intermittency, and edge phenomena such as edge localized modes (ELMs) and multi-faceted axisymmetric radiation from the edge (MARFE). Studies have been performed in all these areas. The edge turbulence/intermittency studies make use of the Gas Puff Imaging diagnostic developed by the Principal Investigator (Ricardo Maqueda) together with colleagues from PPPL. This effort is part of the International Tokamak Physics Activity (ITPA) edge, scrape-off layer and divertor group joint activity (DSOL-15:more » Inter-machine comparison of blob characteristics). The edge turbulence/blob study has been extended from the current location near the midplane of the device to the lower divertor region of NSTX. The goal of this effort was to study turbulence born blobs in the vicinity of the X-point region and their circuit closure on divertor sheaths or high density regions in the divertor. In the area of ELMs and MARFEs we have studied and characterized the mode structure and evolution of the ELM types observed in NSTX, as well as the study of the observed interaction between MARFEs and ELMs. This interaction could have substantial implications for future devices where radiative divertor regions are required to maintain detachment from the divertor plasma facing components.« less

Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

NASA Astrophysics Data System (ADS)

Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

2018-05-01

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

Divertor Coil Design and Implementation on Pegasus

NASA Astrophysics Data System (ADS)

Shriwise, P. C.; Bongard, M. W.; Cole, J. A.; Fonck, R. J.; Kujak-Ford, B. A.; Lewicki, B. T.; Winz, G. R.

2012-10-01

An upgraded divertor coil system is being commissioned on the Pegasus Toroidal Experiment in conjunction with power system upgrades in order to achieve higher β plasmas, reduce impurities, and possibly achieve H-mode operation. Design points for the divertor coil locations and estimates of their necessary current ratings were found using predictive equilibrium modeling based upon a 300 kA target plasma. This modeling represented existing Pegasus coil locations and current drive limits. The resultant design calls for 125 kA-turns from the divertor system to support the creation of a double null magnetic topology in plasmas with Ip<=300 kA. Initial experiments using this system will employ 900 V IGBT power supply modules to provide IDIV<=4 kA. The resulting 20 kA-turn capability of the existing divertor coil will be augmented by a new coil providing additional A-turns in series. Induced vessel wall current modeling indicates the time response of a 28 turn augmentation coil remains fast compared to the poloidal field penetration rate through the vessel. First results operating the augmented system are shown.

Parallel Energy Transport in Detached DIII-D Divertor Plasmas

NASA Astrophysics Data System (ADS)

Leonard, A. W.; Lore, J. D.; Canik, J. M.; McLean, A. G.; Makowski, M. A.

2017-10-01

A comparison of experiment and modeling of detached divertor plasmas is examined in the context of parallel energy transport. Experimental estimates of power carried by electron thermal conduction versus plasma convection are experimentally inferred from power balance measurements of radiated power and target plate heat flux combined with Thomson scattering measurements of the Te profile along the divertor leg. Experimental profiles of Te exhibit relatively low gradients with Te < 15 eV from the X-point to the target implying transport dominated by convection. In contrast, fluid modeling with SOLPS produces sharp Te gradients for Te > 3 eV, characteristic of transport dominated by electron conduction through the bulk of the divertor. This discrepancy with experimental transport dominated by convection and modeling by conduction has significant implications for the radiative capacity of divertor plasmas and may explain at least part of the difficulty for fluid modeling to obtain the experimentally observed radiative losses. Comparisons are also made for helium plasmas where the match between experiment and modeling is much better. Work supported by the US DOE under DE-FC02-04ER54698.

Minimum magnetic curvature for resilient divertors using Compact Toroidal Hybrid geometry

DOE PAGES

Bader, Aaron; Hegna, C. C.; Cianciosa, Mark R.; ...

2018-03-16

The properties of resilient divertors are explored using equilibria derived from Compact Toroidal Hybrid (CTH) geometries. Resilience is defined here as the robustness of the strike point patterns as the plasma geometry and/or plasma profiles are changed. The addition of plasma current in the CTH configurations significantly alters the shape of the last closed flux surface and the rotational transform profile, however, it does not alter the strike point pattern on the target plates, and hence has resilient divertor features. The limits of when a configuration transforms to a resilient configuration is then explored. New CTH-like configurations are generated thatmore » vary from a perfectly circular cross section to configurations with increasing amounts of toroidal shaping. It is found that even small amounts of toroidal shaping lead to strike point localization that is similar to the standard CTH configuration. Lastly, these results show that only a small degree of three-dimensional shaping is necessary to produce a resilient divertor, implying that any highly shaped optimized stellarator will possess the resilient divertor property.« less

Designing divertor targets for uniform power load

NASA Astrophysics Data System (ADS)

Dekeyser, W.; Reiter, D.; Baelmans, M.

2015-08-01

Divertor design for next step fusion reactors heavily relies on 2D edge plasma modeling with codes as e.g. B2-EIRENE. While these codes are typically used in a design-by-analysis approach, in previous work we have shown that divertor design can alternatively be posed as a mathematical optimization problem, and solved very efficiently using adjoint methods adapted from computational aerodynamics. This approach has been applied successfully to divertor target shape design for more uniform power load. In this paper, the concept is further extended to include all contributions to the target power load, with particular focus on radiation. In a simplified test problem, we show the potential benefits of fully including the radiation load in the design cycle as compared to only assessing this load in a post-processing step.

Fiber digestion kinetics and protein degradability characteristics of stockpiled Tifton 85 bermudagrass.

PubMed

Sechler, S R; Mullenix, M K; Holland, C M; Muntifering, R B

2017-09-01

A 2-yr study was conducted to determine effects of N fertilization level on fiber digestion kinetics and protein degradability characteristics of stockpiled Tifton 85 bermudagrass (T85). Six 0.76-ha pastures of stockpiled T85 were cut to a 10-cm stubble height on August 1 of each yr and fertilized with 56 (56N), 112 (112N), or 168 (168N) kg N/ha (2 pastures/treatment). Fiber digestion kinetics included the 72-hr potential extent of NDF digestion (PED), rate of NDF digestion, and lag time. In yr 1 and 2, PED decreased over the stockpile season. Rates of NDF digestion did not differ ( > 0.05) among N fertilization treatments in either yr. In yr 1, rate of NDF digestion was greatest ( < 0.05) in October/November, and decreased beginning in December. In yr 2, rate of NDF digestion decreased ( < 0.05) in January compared with November, but digestion rates were similar for November and January 21 sampling dates. Lag time was greater ( < 0.05) for the 112N than 56N and 168N treatments, and increased ( < 0.05) across sampling dates in yr 1. In yr 2, lag time increased ( < 0.05) from 9.0 to 17.7 h across the season. In yr 1 and 2, a negative correlation ( < 0.05) between forage lignin concentration and both PED ( = -0.91 and -0.87 in yr 1 and 2, respectively) and rate of NDF digestion ( = -0.60 and -0.25 in yr 1 and 2, respectively) was observed. There was a trend ( = 0.06) for lignin concentration to be positively correlated with lag time ( = 0.39) in yr 1, and a strong relationship was observed in yr 2 ( = 0.91; < 0.05). The RDP fraction as a % of CP was ≥ 90% throughout both years. Concentration of RDP (% of total DM) decreased across the stockpiling season through January in yr 1 and 2. Results suggest that kinetic parameters of NDF digestion in stockpiled T85 were influenced more by temporal changes over the stockpile season than by N fertilization level. Supplement formulations based on kinetic parameters of fiber digestion may require periodic adjustment to insure that energy-yielding components of NDF are sufficient to meet animal requirements throughout the stockpile season. The CP fraction in stockpiled T85 contains sufficient RDP to support fibrolytic activity and growth of ruminal microorganisms throughout the stockpile season. Toward the latter end of the season, supplementation with sources of digestible fiber and RDP could be expected to increase MP supply to the host animal.

Processes forcing the suspended sediments distribution in a wide, shallow and microtidal estuary: a numerical case study for the Río de la Plata

NASA Astrophysics Data System (ADS)

Simionato, Claudia; Moreira, Diego

2017-04-01

The impact of the diverse mechanisms driving the suspended sediments distribution in the wide, shallow and microtidal Río de la Plata (RdP) estuary and the adjacent shelf is studied by means of a set of process-oriented numerical simulations. With that aim, a regional application of the hydro-sedimentological Model for Applications at Regional Scale (MARS) is implemented, tested and run under diverse conditions. Even the simulations are idealized, they reproduce both qualitatively and quantitatively well the main features of the suspended sediments observed distribution, particularly the mean values of concentration and its gradients: perpendicular to the estuary axis at the upper and intermediate RdP and parallel to the estuary axis at its outer part. Even though naturally the diameter of the sediments that deposit decays with the distance to the sources (with sands and silts dominating in the upper estuary and fine silts and clays over the Barra del Indio), model results show that the large width and the geometry of the estuary play an important role in the sedimentation process. The widening and deepening, and the associated significant reduction of the currents speed that occurs after (i) the confluence of the tributaries and (ii) downstream the Barra del Indio Shoal, favors sediments deposition downstream those areas. Even though tides are of small amplitude in the study area, they have a significant impact on the lateral mixing and the re-suspension of bottom sediments; this last augments the concentration of fine sediments in the layers close to the bottom but their energy is not enough to rise them up to the surface. The model reproduces the increment in the concentration of fine sediments observed in the areas where tidal dissipation energy by bottom friction maximizes (over the southern coast of the RdP and around Punta Piedras and Punta Rasa), but shows that tides alone cannot account for the observed maxima. Winds (which can be quite large over this area) enhance horizontal mixing, smoothing the pattern produced by the tides. Wind waves are the most important forcing for the vertical mixing of the sediments. Their effect is most evident along the southern coast of the RdP and the Barra del Indio Shoal, where wind waves rise to the surface the sediments resuspended by tides. The bottom salinity front acts retaining the sediments upstream the Barra del Indio shoal; there, estuarine currents and flocculation play an important role in sediments deposition.

Integrated simulations of H-mode operation in ITER including core fuelling, divertor detachment and ELM control

NASA Astrophysics Data System (ADS)

Polevoi, A. R.; Loarte, A.; Dux, R.; Eich, T.; Fable, E.; Coster, D.; Maruyama, S.; Medvedev, S. Yu.; Köchl, F.; Zhogolev, V. E.

2018-05-01

ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on W transport assumptions during the ELM, a maximum ELM frequency is also identified above which core tungsten accumulation takes place.

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Anderl, Robert A.; Bartlit, John R.; Causey, Rion A.; Haines, John R.

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10(exp 19) ions/((sq cm)(s)) and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment.

Nitrogen source and concentration affect utilization of glucose by mixed ruminal microbes in vitro

USDA-ARS?s Scientific Manuscript database

Availability of ruminally degradable protein (RDP) changes the utilization of carbohydrates by ruminal microbes. However, the effects are not well described, though such information is needed to understand the potential impact on nutrient supplies for ruminants. The objective of this study was to co...

Improved Feed Protein Fractional Schemes for Formulating Rations With the Cornell Net Carbohydrate and Protein System

USDA-ARS?s Scientific Manuscript database

Adequate predictions of rumen-degradable protein (RDP) and rumen-undegradable protein (RUP) supplies are necessary to optimize performance while minimizing losses of excess nitrogen (N). The objectives of this study were to evaluate the original Cornell Net Carbohydrate Protein System (CNCPS) protei...

Simulation of turbulence in the divertor region of tokamak edge plasma

NASA Astrophysics Data System (ADS)

Umansky, M. V.; Rognlien, T. D.; Xu, X. Q.

2005-03-01

Results are presented for turbulence simulations with the fluid edge turbulence code BOUT [X.Q. Xu, R.H. Cohen, Contr. Plas. Phys. 36 (1998) 158]. The present study is focussed on turbulence in the divertor leg region and on the role of the X-point in the structure of turbulence. Results of the present calculations indicate that the ballooning effects are important for the divertor fluctuations. The X-point shear leads to weak correlation of turbulence across the X-point regions, in particular for large toroidal wavenumber. For the saturated amplitudes of the divertor region turbulence it is found that amplitudes of density fluctuations are roughly proportional to the local density of the background plasma. The amplitudes of electron temperature and electric potential fluctuations are roughly proportional to the local electron temperature of the background plasma.

Divertor with a third-order null of the poloidal field

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryutov, D. D.; Umansky, M. V.

2013-09-15

A concept and preliminary feasibility analysis of a divertor with the third-order poloidal field null is presented. The third-order null is the point where not only the field itself but also its first and second spatial derivatives are zero. In this case, the separatrix near the null-point has eight branches, and the number of strike-points increases from 2 (as in the standard divertor) to six. It is shown that this magnetic configuration can be created by a proper adjustment of the currents in a set of three divertor coils. If the currents are somewhat different from the required values, themore » configuration becomes that of three closely spaced first-order nulls. Analytic approach, suitable for a quick orientation in the problem, is used. Potential advantages and disadvantages of this configuration are briefly discussed.« less

Initial development of the DIII–D snowflake divertor control

NASA Astrophysics Data System (ADS)

Kolemen, E.; Vail, P. J.; Makowski, M. A.; Allen, S. L.; Bray, B. D.; Fenstermacher, M. E.; Humphreys, D. A.; Hyatt, A. W.; Lasnier, C. J.; Leonard, A. W.; McLean, A. G.; Maingi, R.; Nazikian, R.; Petrie, T. W.; Soukhanovskii, V. A.; Unterberg, E. A.

2018-06-01

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5×  reduction in the peak heat flux for many energy confinement times (2–3 s) without any adverse effects on core plasma performance.

Detecting and Analyzing Genetic Recombination Using RDP4.

PubMed

Martin, Darren P; Murrell, Ben; Khoosal, Arjun; Muhire, Brejnev

2017-01-01

Recombination between nucleotide sequences is a major process influencing the evolution of most species on Earth. The evolutionary value of recombination has been widely debated and so too has its influence on evolutionary analysis methods that assume nucleotide sequences replicate without recombining. When nucleic acids recombine, the evolution of the daughter or recombinant molecule cannot be accurately described by a single phylogeny. This simple fact can seriously undermine the accuracy of any phylogenetics-based analytical approach which assumes that the evolutionary history of a set of recombining sequences can be adequately described by a single phylogenetic tree. There are presently a large number of available methods and associated computer programs for analyzing and characterizing recombination in various classes of nucleotide sequence datasets. Here we examine the use of some of these methods to derive and test recombination hypotheses using multiple sequence alignments.

A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

2012-10-01

Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

The Simple Map for a Single-null Divertor Tokamak: How to Look for Self-Similarity in Chaos

NASA Astrophysics Data System (ADS)

Nguyen, Christina; Ali, Halima; Punjabi, Alkesh

2000-10-01

The movement of magnetic field lines inside a single-null divertor tokamak can be described by the Simple Map^1. The Simple Map in the Poincaré Surface of Section is given by the equations: X_1=X_0-KY_0(1-Y_0) and Y_1=Y_0+KX_1. In these equations, K remains constant at 0.60. However, the values for X0 and Y0 are changed. These values are changed so that we can zoom into chaos. Chaos lies between the region (0,0.997) and (0,1). In chaos, there lies order. As we zoom into chaos, we again find chaos and order that looks like the original good surfaces and chaos. This phenomenon is called self-similarity. Self-similarity can occur for an infinite number of times if one magnifies into the chaotic region. For this work, we write a program in a computer language called Fortran 77 and Gnuplot. This work is supported by US DOE OFES. Ms. Christina Nguyen is a HU CFRT Summer Fusion High School Workshop Scholar from Andrew Hill High School in California. She is supported by NASA SHARP Plus Program. 1. Punjabi A, Verma A and Boozer A, Phys Rev Lett 69 3322 (1992) and J Plasma Phys 52 91 (1994)

Exposures of tungsten nanostructures to divertor plasmas in DIII-D

DOE PAGES

Rudakov, D. L.; Wong, C. P. C.; Doerner, R. P.; ...

2016-01-22

Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with littlemore » obvious damage except in the areas where unipolar arcing occurred. In conclusion, arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.« less

Plasma transport in a simulated magnetic-divertor configuration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strawitch, C. M.

1981-03-01

The transport properties of plasma on magnetic field lines that intersect a conducting plate are studied experimentally in the Wisconsin internal ring D.C. machine. The magnetic geometry is intended to simulate certain aspects of plasma phenomena that may take place in a tokamak divertor. It is found by a variety of measurements that the cross field transport is non-ambipolar; this may have important implications in heat loading considerations in tokamak divertors. The undesirable effects of nonambipolar flow make it preferable to be able to eliminate it. However, we find that though the non-ambipolarity may be reduced, it is difficult tomore » eliminate entirely. The plasma flow velocity parallel to the magnetic field is found to be near the ion acoustic velocity in all cases. The experimental density and electron temperature profiles are compared to the solutions to a one dimensional transport model that is commonly used in divertor theory.« less

Structural investigation of re-deposited layers in JET

NASA Astrophysics Data System (ADS)

Likonen, J.; Vainonen-Ahlgren, E.; Khriachtchev, L.; Coad, J. P.; Rubel, M.; Renvall, T.; Arstila, K.; Hole, D. E.; Contributors to the EFDA-JET Work-programme

2008-07-01

JET Mk-II Gas Box divertor tiles exposed in 1998-2001 have been analysed with various ion beam techniques, secondary ion mass spectrometry (SIMS) and Raman spectroscopy. Inner divertor wall tiles removed in 2001 were covered with a duplex film. The inner layer was very rich in metallic impurities, with Be/C ˜ 1 and H-isotopes only present at low concentrations. The outer layer contained higher concentrations of D than normal for plasma-facing surfaces in JET (D/C ˜ 0.4), and Be/C ˜ 0.14. Raman and SIMS analyses show that the deposited films on inner divertor tiles are hydrogenated amorphous carbon with low sp 3 fractions. The deposits have polymeric structure and low density. Both Raman scattering and SIMS indicate that films on inner divertor wall Tiles 1 and 3, and on floor Tile 4 have some differences in the chemical structure of the deposited films

Initial development of the DIII–D snowflake divertor control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kolemen, Egemen; Vail, P. J.; Makowski, M. A.

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasmamore » and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. In conclusion, the SFD resulted in a 2.5×reduction in the peak heat flux for many energy confinement times (2–3s) without any adverse effects on core plasma performance.« less

Initial development of the DIII–D snowflake divertor control

DOE PAGES

Kolemen, Egemen; Vail, P. J.; Makowski, M. A.; ...

2018-04-11

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasmamore » and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. In conclusion, the SFD resulted in a 2.5×reduction in the peak heat flux for many energy confinement times (2–3s) without any adverse effects on core plasma performance.« less

The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, Christopher; Nygren, R. E.; Chrobak, C P.

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels ofmore » isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.« less

Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

NASA Astrophysics Data System (ADS)

Schmitz, O.; Becoulet, M.; Cahyna, P.; Evans, T. E.; Feng, Y.; Frerichs, H.; Loarte, A.; Pitts, R. A.; Reiser, D.; Fenstermacher, M. E.; Harting, D.; Kirschner, A.; Kukushkin, A.; Lunt, T.; Saibene, G.; Reiter, D.; Samm, U.; Wiesen, S.

2016-06-01

Results from three-dimensional modeling of plasma edge transport and plasma-wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q  =  10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q 95  =  4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP coil current yield a reduction of the width of the divertor flux spreading to about 20-25 cm and cause increased peak heat fluxes back to values similar to those in the axisymmetric case. The dependencies of these features on the divertor recycling regime and the perpendicular transport assumptions, as well as toroidal averaged effects mimicking rotation of the RMP field, are discussed in the paper.

Farmers' Knowledge Attributes Contribute to Attaining Higher Farm Technical Efficiency: A Transition Economy Case

ERIC Educational Resources Information Center

Manevska-Tasevska, Gordana

2013-01-01

Purpose: This study sought to explore how farmers' knowledge attributes influence the technical efficiency of their farms. In addition, farm efficiency was compared to the actual Macedonian Rural Development Programme (RDP) (2007-2013) and instruments considered to improve Macedonian education potential were evaluated. Design/methodology/approach:…

Use of the Rice Diversity Panel 1 to map traits important for rice improvement

USDA-ARS?s Scientific Manuscript database

The ‘Rice Diversity Panel 1’ (RDP1) is composed of 421 diverse Oryza sativa accessions from 79 countries, including indica and aus which belong to the Indica varietal group, and tropical japonica, temperate japonica, and aromatic (Group V) which comprise the Japonica varietal group. This panel was ...

Using rice genome-wide association studies to identify DNA markers for marker-assisted selection

USDA-ARS?s Scientific Manuscript database

Rice association mapping panels are collections of rice (Oryza sativa L.) accessions developed for genome-wide association (GWA) studies. One of these panels, the Rice Diversity Panel 1 (RDP1) was phenotyped by various research groups for several traits of interest, and more recently, genotyped with...

Heat Stress Alters Ruminal Fermentation and Digesta Characteristics, and Behavior in Lactating Dairy Cattle

USDA-ARS?s Scientific Manuscript database

In a study designed to assess the impact and interaction of nonfiber carbohydrates (NFC) and ruminally degradable protein (RDP) on ruminal characteristics and animal behavior, animals experienced heat stress in the first period (HS), and no/greatly reduced heat stress (NHS) in the second period, all...

Assessing foliar ascorbate content in the rice diversity panel 1

USDA-ARS?s Scientific Manuscript database

Early spring plantings of rice can have poor stands due to cold temperatures. Our previous studies have shown that high vitamin C (ascorbate AsA) Arabidopsis lines are tolerant to cold stress. The rice diversity panel 1 (RDP1) represents the genetic diversity of Oryza sativa and has been extensively...

α-synuclein assemblies sequester neuronal α3-Na+/K+-ATPase and impair Na+ gradient

PubMed Central

Shrivastava, Amulya Nidhi; Redeker, Virginie; Fritz, Nicolas; Pieri, Laura; Almeida, Leandro G; Spolidoro, Maria; Liebmann, Thomas; Bousset, Luc; Renner, Marianne; Léna, Clément; Aperia, Anita; Melki, Ronald; Triller, Antoine

2015-01-01

Extracellular α-synuclein (α-syn) assemblies can be up-taken by neurons; however, their interaction with the plasma membrane and proteins has not been studied specifically. Here we demonstrate that α-syn assemblies form clusters within the plasma membrane of neurons. Using a proteomic-based approach, we identify the α3-subunit of Na+/K+-ATPase (NKA) as a cell surface partner of α-syn assemblies. The interaction strength depended on the state of α-syn, fibrils being the strongest, oligomers weak, and monomers none. Mutations within the neuron-specific α3-subunit are linked to rapid-onset dystonia Parkinsonism (RDP) and alternating hemiplegia of childhood (AHC). We show that freely diffusing α3-NKA are trapped within α-syn clusters resulting in α3-NKA redistribution and formation of larger nanoclusters. This creates regions within the plasma membrane with reduced local densities of α3-NKA, thereby decreasing the efficiency of Na+ extrusion following stimulus. Thus, interactions of α3-NKA with extracellular α-syn assemblies reduce its pumping activity as its mutations in RDP/AHC. PMID:26323479

Safety characteristics of the monolithic CFC divertor

NASA Astrophysics Data System (ADS)

Zucchetti, M.; Merola, M.; Matera, R.

1994-09-01

The main distinguishing feature of the monolithic CFC divertor is the use of a single material, a carbon fibre reinforced carbon, for the protective armour, the heat sink and the cooling channels. This removes joint interface problems which are one of the most important concerns related to the reference solutions of the ITER CDA divertor. An activation analysis of the different coolant options for this concept is presented. It turns out that neither short-term nor long-term activation are a concern for any coolants investigated. Therefore the proposed concept proves to be attractive from a safety stand-point also.

Upstream Density for Plasma Detachment with Conventional and Lithium Vapor-Box Divertors

NASA Astrophysics Data System (ADS)

Goldston, Rj; Schwartz, Ja

2016-10-01

Fusion power plants are likely to require detachment of the divertor plasma from material targets. The lithium vapor box divertor is designed to achieve this, while limiting the flux of lithium vapor to the main plasma. We develop a simple model of near-detachment to evaluate the required upstream plasma density, for both conventional and lithium vapor-box divertors, based on particle and dynamic pressure balance between up- and down-stream, at near-detachment conditions. A remarkable general result is found, not just for lithium-induced detachment, that the upstream density divided by the Greenwald-limit density scales as (P 5 / 8 /B 3 / 8) Tdet1 / 2 / (ɛcool + γTdet) , with no explicit size scaling. Tdet is the temperature just before strong pressure loss, 1/2 of the ionization potential of the dominant recycling species, ɛcool is the average plasma energy lost per injected hydrogenic and impurity atom, and γ is the sheath heat transmission factor. A recent 1-D calculation agrees well with this scaling. The implication is that the plasma exhaust problem cannot be solved by increasing R. Instead significant innovation, such as the lithium vapor box divertor, will be required. This work supported by DOE Contract No. DE-AC02-09CH11466.

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

NASA Astrophysics Data System (ADS)

Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

2016-02-01

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak.

PubMed

Brunner, D; Burke, W; Kuang, A Q; LaBombard, B; Lipschultz, B; Wolfe, S

2016-02-01

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

Fast imaging of filaments in the X-point region of Alcator C-Mod

DOE PAGES

Terry, J. L.; Ballinger, S.; Brunner, D.; ...

2017-01-27

A rich variety of field-aligned fluctuations has been revealed using fast imaging of D α emission from Alcator C-Mod's lower X-point region. Field-aligned filamentary fluctuations are observed along the inner divertor leg, within the Private-Flux-Zone (PFZ), in the Scrape-Off Layer (SOL) outside the outer divertor leg, and, under some conditions, at or above the X-point. The locations and dynamics of the filaments in these regions are strikingly complex in C-Mod. Changes in the filaments’ generation appear to be ordered by plasma density and magnetic configuration. Filaments are not observed for plasmas with n/nGreenwald ≲ 0.12 nor are they observed inmore » Upper Single Null configurations. In a Lower Single Null with 0.12 ≲ n/nGreenwald ≲ 0.45 and Bx∇B directed down, filaments typically move up the inner divertor leg toward the X-point. Reversing the field direction results in the appearance of filaments outside of the outer divertor leg. With the divertor targets “detached”, filaments inside the LCFS are seen. Lastly, these studies were motivated by observations of filaments in the X-point and PFZ regions in MAST, and comparisons with those observations are made.« less

Fabrication of titanium removable dental prosthesis frameworks with a 2-step investment coating method.

PubMed

Koike, Mari; Hummel, Susan K; Ball, John D; Okabe, Toru

2012-06-01

Although pure titanium is known to have good biocompatibility, a titanium alloy with better strength is needed for fabricating clinically acceptable, partial removable dental prosthesis (RDP) frameworks. The mechanical properties of an experimental Ti-5Al-5Cu alloy cast with a 2-step investment technique were examined for RDP framework applications. Patterns for tests for various properties and denture frameworks for a preliminary trial casting were invested with a 2-step coating method using 2 types of mold materials: a less reactive spinel compound (Al(2)O(3)·MgO) and a less expensive SiO(2)-based material. The yield and tensile strength (n=5), modulus of elasticity (n=5), elongation (n=5), and hardness (n=8) of the cast Ti-5Al-5Cu alloy were determined. The external appearance and internal porosities of the preliminary trial castings of denture frameworks (n=2) were examined with a conventional dental radiographic unit. Cast Ti-6Al-4V alloy and commercially pure titanium (CP Ti) were used as controls. The data for the mechanical properties were statistically analyzed with 1-way ANOVA (α=.05). The yield strength of the cast Ti-5Al-5Cu alloy was 851 MPa and the hardness was 356 HV. These properties were comparable to those of the cast Ti-6Al-4V and were higher than those of CP Ti (P<.05). One of the acrylic resin-retention areas of the Ti-5Al-5Cu frameworks was found to have been incompletely cast. The cast biocompatible experimental Ti-5Al-5Cu alloy exhibited high strength when cast with a 2-step coating method. With a dedicated study to determine the effect of sprue design on the quality of castings, biocompatible Ti-5Al-5Cu RDP frameworks for a clinical trial can be produced. Copyright © 2012 The Editorial Council of the Journal of Prosthetic Dentistry. Published by Mosby, Inc. All rights reserved.

Establishment of Bacterial Herbicide Degraders in a Rapid Sand Filter for Bioremediation of Phenoxypropionate-Polluted Groundwater

PubMed Central

Feld, Louise; Nielsen, Tue Kjærgaard; Hansen, Lars Hestbjerg; Aamand, Jens

2015-01-01

In this study, we investigated the establishment of natural bacterial degraders in a sand filter treating groundwater contaminated with the phenoxypropionate herbicides (RS)-2-(4-chloro-2-methylphenoxy)propanoic acid (MCPP) and (RS)-2-(2,4-dichlorophenoxy)propanoic acid (DCPP) and the associated impurity/catabolite 4-chlorophenoxypropanoic acid (4-CPP). A pilot facility was set up in a contaminated landfill site. Anaerobic groundwater was pumped up and passed through an aeration basin and subsequently through a rapid sand filter, which is characterized by a short residence time of the water in the filter. For 3 months, the degradation of DCPP, MCPP, and 4-CPP in the sand filter increased to 15 to 30% of the inlet concentration. A significant selection for natural bacterial herbicide degraders also occurred in the sand filter. Using a most-probable-number (MPN) method, we found a steady increase in the number of culturable phenoxypropionate degraders, reaching approximately 5 × 105 degraders per g sand by the end of the study. Using a quantitative PCR targeting the two phenoxypropionate degradation genes, rdpA and sdpA, encoding stereospecific dioxygenases, a parallel increase was observed, but with the gene copy numbers being about 2 to 3 log units higher than the MPN. In general, the sdpA gene was more abundant than the rdpA gene, and the establishment of a significant population of bacteria harboring sdpA occurred faster than the establishment of an rdpA gene-carrying population. The identities of the specific herbicide degraders in the sand filter were assessed by Illumina MiSeq sequencing of 16S rRNA genes from sand filter samples and from selected MPN plate wells. We propose a list of potential degrader bacteria involved in herbicide degradation, including representatives belonging to the Comamonadaceae and Sphingomonadales. PMID:26590282

Effects of dietary urea concentration and zilpaterol hydrochloride on performance and carcass characteristics of finishing steers.

PubMed

Samuelson, K L; Hubbert, M E; Löest, C A

2016-12-01

Cattle receiving zilpaterol hydrochloride () may recycle less N and require a greater supply of RDP. This study evaluated effects of ZH on performance and carcass characteristics of steers fed diets with increasing dietary RDP concentrations supplied as urea. Steers (429 animals; BW = 423 ± 4.5 kg) were sorted into 3 blocks according to BW and assigned to 1 of 6 treatments (6 pens per treatment) in a randomized complete block design. Treatments were a 2 × 3 factorial arrangement of either no ZH or ZH (75 mg ZH per steer daily) supplemented to finishing diets containing 0, 0.5, or 1.0% urea of dietary DM. Pen weights were recorded before treatment initiation; urea was fed for 27 d, and ZH treatments were fed for 24 d with a 3-d withdrawal period. Pen weights were recorded before transporting steers to a commercial abattoir. Continuous response variables were analyzed using the MIXED procedure and categorical data were analyzed using the GLIMMIX procedure of SAS. No ZH × dietary urea interactions ( ≥ 0.14) occurred for all performance and carcass response variables. Feeding ZH for the last 27 d (included a 3-d withdrawal period) of the finishing period increased ( < 0.01) ADG, decreased ( < 0.01) DMI, and increased ( < 0.01) G:F compared with no ZH. In addition, ZH increased HCW ( < 0.01), dressing percentage ( < 0.01), LM area ( < 0.01), and decreased ( = 0.01) yield grade. Increasing dietary urea linearly decreased ( = 0.01) ADG and DMI. A tendency for a linear decrease ( = 0.10) in HCW, and a tendency for a quadratic increase ( = 0.07) in marbling score were observed as urea increased in the diet. Results indicate that cattle supplemented with ZH do not require additional RDP in the diet, and that performance and carcass characteristics were negatively affected when urea was increased in the diet.

Effect of protein supplementation on expression and distribution of urea transporter-B in lambs fed low-quality forage.

PubMed

Ludden, P A; Stohrer, R M; Austin, K J; Atkinson, R L; Belden, E L; Harlow, H J

2009-04-01

Two experiments were conducted to determine the effects of ruminal protein degradability, supplementation frequency, and increasing dietary protein on the expression and distribution of urea transporter-B (UT-B) in lambs fed low-quality forage (mature crested wheatgrass hay; 4.2 to 4.7% CP). In Exp. 1, 15 Dorset wether lambs (initial BW=45.8+/-1.3 kg) were blocked by initial BW and assigned to 1 of 3 treatments within a randomized complete block design for 28 d, with supplements fed to achieve 7, 10, or 13% total dietary CP. In Exp. 2, 13 Dorset wether lambs (initial BW=34+/-4 kg) were used in a completely randomized design and given 1 of 4 isonitrogenous supplements: 1) ruminally degradable protein (RDP) fed daily (n=3), 2) RDP fed on alternate days (n=3), 3) ruminally undegradable protein (RUP) fed on alternate days (n=3), or 4) a 50:50 mixture of RDP and RUP fed on alternate days (n=4) for 18 d. Alternate-day treatments were fed at twice that of daily supplementation. On the last day of both experiments, lambs were killed and samples taken for Western blot analyses for UT-B. Immunoblotting using a rabbit polyclonal antibody to UT-B confirmed the presence of distinct 32-kDa (consistent with a nonglycosylated UT-B protein) and 47-kDa (probable N-glycosylated form of UT-B) protein bands in all 9 tissues analyzed. In both experiments, the liver, dorsal rumen, reticulum, and ventral rumen displayed strong bands at 32 kDa and lighter bands at 47 kDa, whereas the cecum, large colon, spiral colon, and parotid salivary gland displayed slight 32-kDa bands and stronger, more visible bands at 47 kDa. Both protein bands were apparent in the kidney at similar visual intensities in Exp. 1, whereas the relative intensities of the 2 UT-B bands in the kidney were variable, and appeared somewhat reciprocal among animals in Exp. 2. Although the abundance of the 47-kDa UT-B band in the ventral rumen was greater (P=0.03) in lambs fed RDP daily in Exp. 2, no other treatment differences (P >or= 0.15 to 0.99) in the abundance of the 32- or 47-kDa UT-B proteins within tissues were observed in either experiment. Although protein supplementation strategy had little effect on UT-B expression in tissues other than the ventral rumen, differences in the degree of glycosylation of UT-B across tissues may provide insight into its regulation.

Advanced simulation of mixed-material erosion/evolution and application to low and high-Z containing plasma facing components

NASA Astrophysics Data System (ADS)

Brooks, J. N.; Hassanein, A.; Sizyuk, T.

2013-07-01

Plasma interactions with mixed-material surfaces are being analyzed using advanced modeling of time-dependent surface evolution/erosion. Simulations use the REDEP/WBC erosion/redeposition code package coupled to the HEIGHTS package ITMC-DYN mixed-material formation/response code, with plasma parameter input from codes and data. We report here on analysis for a DIII-D Mo/C containing tokamak divertor. A DIII-D/DiMES probe experiment simulation predicts that sputtered molybdenum from a 1 cm diameter central spot quickly saturates (˜4 s) in the 5 cm diameter surrounding carbon probe surface, with subsequent re-sputtering and transport to off-probe divertor regions, and with high (˜50%) redeposition on the Mo spot. Predicted Mo content in the carbon agrees well with post-exposure probe data. We discuss implications and mixed-material analysis issues for Be/W mixing at the ITER outer divertor, and Li, C, Mo mixing at an NSTX divertor.

Constrained ripple optimization of Tokamak bundle divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hively, L.M.; Rome, J.A.; Lynch, V.E.

1983-02-01

Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have lowmore » on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.« less

Initial results from divertor heat-flux instrumentation on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Labombard, B.; Brunner, D.; Payne, J.; Reinke, M.; Terry, J. L.; Hughes, J. W.; Lipschultz, B.; Whyte, D.

2009-11-01

Physics-based plasma transport models that can accurately simulate the heat-flux power widths observed in the tokamak boundary are lacking at the present time. Yet this quantity is of fundamental importance for ITER and most critically important for DEMO, a reactor similar to ITER but with ˜4 times the power exhaust. In order to improve our understanding, C-Mod, DIII-D and NSTX will aim experiments in FY10 towards characterizing the divertor ``footprint'' and its connection to conditions ``upstream'' in the boundary and core plasmas [2]. Standard IR-based heat-flux measurements are particularly difficult in C-Mod, due to its vertical-oriented divertor targets. To overcome this, a suite of embedded heat-flux sensor probes (tile thermocouples, calorimeters, surface thermocouples) combined with IR thermography was installed during the FY09 opening, along with a new divertor bolometer system. This paper will report on initial experiments aimed at unfolding the heat-flux dependencies on plasma operating conditions. [2] a proposed US DoE Joint Facilities Milestone.

HSX as an example of a resilient non-resonant divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bader, A.; Boozer, A. H.; Hegna, C. C.

This study describes an initial description of the resilient divertor properties of quasi-symmetric (QS) stellarators using the HSX (Helically Symmetric eXperiment) configuration as a test-case. Divertors in high-performance QS stellarators will need to be resilient to changes in plasma configuration that arise due to evolution of plasma pressure profiles and bootstrap currents for divertor design. Resiliency is tested by examining the changes in strike point patterns from the field line following, which arise due to configurational changes. A low strike point variation with high configuration changes corresponds to high resiliency. The HSX edge displays resilient properties with configuration changes arisingmore » from the (1) wall position, (2) plasma current, and (3) external coils. The resilient behavior is lost if large edge islands intersect the wall structure. The resilient edge properties are corroborated by heat flux calculations from the fully 3-D plasma simulations using EMC3-EIRENE. Additionally, the strike point patterns are found to correspond to high curvature regions of magnetic flux surfaces.« less

Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

NASA Astrophysics Data System (ADS)

Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

2009-12-01

Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

Design of snowflake-diverted equilibria of CFETR

NASA Astrophysics Data System (ADS)

Hang, LI; Xiang, GAO; Guoqiang, LI; Zhengping, LUO; Damao, YAO; Yong, GUO

2018-03-01

The Chinese Fusion Engineering Test Reactor (CFETR) represents the next generation of full superconducting fusion reactors in China. Recently, CFETR was redesigned with a larger size and will be operated in two phases. To reduce the heat flux on the target plate, a snowflake (SF) divertor configuration is proposed. In this paper we show that by adding two dedicated poloidal field (PF) coils, the SF configuration can be achieved in both phases. The equilibria were calculated by TEQ code for a range of self-inductances l i3. The coil currents were calculated at some fiducial points in the flattop phase. The results indicate that the PF coil system has the ability to maintain a long flattop phase in 7.5 and 10 MA inductive scenarios for the single null divertor (SND) and SF divertor configurations. The properties of the SF configuration were also analyzed. The connection length and flux expansion of the SF divertor were both increased significantly over the SND.

HSX as an example of a resilient non-resonant divertor

DOE PAGES

Bader, A.; Boozer, A. H.; Hegna, C. C.; ...

2017-03-16

This study describes an initial description of the resilient divertor properties of quasi-symmetric (QS) stellarators using the HSX (Helically Symmetric eXperiment) configuration as a test-case. Divertors in high-performance QS stellarators will need to be resilient to changes in plasma configuration that arise due to evolution of plasma pressure profiles and bootstrap currents for divertor design. Resiliency is tested by examining the changes in strike point patterns from the field line following, which arise due to configurational changes. A low strike point variation with high configuration changes corresponds to high resiliency. The HSX edge displays resilient properties with configuration changes arisingmore » from the (1) wall position, (2) plasma current, and (3) external coils. The resilient behavior is lost if large edge islands intersect the wall structure. The resilient edge properties are corroborated by heat flux calculations from the fully 3-D plasma simulations using EMC3-EIRENE. Additionally, the strike point patterns are found to correspond to high curvature regions of magnetic flux surfaces.« less

Characterization of novel human papillomavirus types 157, 158 and 205 from healthy skin and recombination analysis in genus γ-Papillomavirus.

PubMed

Bolatti, Elisa M; Chouhy, Diego; Casal, Pablo E; Pérez, Germán R; Stella, Emma J; Sanchez, Adriana; Gorosito, Mario; Bussy, Ramón Fernandez; Giri, Adriana A

2016-08-01

Gammapapillomavirus (γ-PV) is a diverse and rapidly expanding genus, currently consisting of 79 fully characterized human PV (HPV) types. In this study, three novel types, HPV157, HPV158 and HPV205, obtained from healthy sun-exposed skin of two immunocompetent individuals, were amplified by the "Hanging droplet" long PCR technique, cloned, sequenced and characterized. HPV157, HPV158 and HPV205 genomes comprise 7154-bp, 7192-bp and 7298-bp, respectively, and contain four early (E1, E2, E6 and E7) and two late genes (L1 and L2). Phylogenetic analysis of the L1 ORF placed all novel types within the γ-PV genus: HPV157 was classified as a new member of species γ-12 while HPV158 and HPV205 belong to species γ-1. We then explored potential recombination events in genus γ-PV with the RDP4 program in a dataset of 74 viruses (71 HPV types with available full-length genomes and the 3 novel types). Two events, both located in the E1 ORF, met the inclusion criterion (p-values <0.05 with at least four methods) and persisted in different ORF combinations: an inter-species recombination in species γ-8 (major and minor parents: species γ-24 and γ-11, respectively), and an intra-species recombination in species γ-7 (recombinant strain: HPV170; major and minor parents: HPV-109 and HPV-149, respectively). These findings were confirmed by phylogenetic tree incongruence analysis. An additional incongruence was found in members of species γ-9 but it was not detected by the RDP4. This report expands our knowledge of the family Papillomaviridae and provides for the first time in silico evidence of recombination in genus γ-PV. Copyright © 2016 Elsevier B.V. All rights reserved.

Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, T.F.; Lee, A.Y.; Ruck, G.W.

1979-01-25

The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm/sup 2/) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm/sup 2/) and for ISX-B/sup 2/ (11 kA/cm/sup 2/). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure.

Scaling Relationships for ELM Diverter Heat Flux on DIII D

NASA Astrophysics Data System (ADS)

Peters, E. A.; Makowski, M. A.; Leonard, A. W.

2015-11-01

Edge Localized Modes (ELMs) are periodic plasma instabilities that occur during H-mode operation in tokamaks. Left unmitigated, these instabilities result in concentrated particle and heat fluxes at the divertor and stand to cause serious damage to the plasma facing components of tokamaks. The purpose of this research is to find scaling relationships that predict divertor heat flux due to ELMs based on plasma parameters at the time of instability. This will be accomplished by correlating characteristic ELM parameters with corresponding plasma measurements and analyzing the data for trends. One early assessment is the effect of the heat transmission coefficient ? on the in/out asymmetry of the calculated ELM heat fluxes. Using IR camera data, further assessments in this study will continue to emphasize in/out asymmetry in ELMs, as this has important implications for ITER operation. Work supported in part by the US DOE, DE-AC52-07NA27344, DE-FC02-04ER54698, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

Final Report on ITER Task Agreement 81-08

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard L. Moore

As part of an ITER Implementing Task Agreement (ITA) between the ITER US Participant Team (PT) and the ITER International Team (IT), the INL Fusion Safety Program was tasked to provide the ITER IT with upgrades to the fusion version of the MELCOR 1.8.5 code including a beryllium dust oxidation model. The purpose of this model is to allow the ITER IT to investigate hydrogen production from beryllium dust layers on hot surfaces inside the ITER vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). Also included in the ITER ITA was a task to construct a RELAP5/ATHENA model of themore » ITER divertor cooling loop to model the draining of the loop during a large ex-vessel pipe break followed by an in-vessel divertor break and compare the results to a simular MELCOR model developed by the ITER IT. This report, which is the final report for this agreement, documents the completion of the work scope under this ITER TA, designated as TA 81-08.« less

OEDGE Modeling of Collector Probe measurements in L-mode from the DIII-D tungsten ring campaign

NASA Astrophysics Data System (ADS)

Elder, J. D.; Stangeby, P. C.; Unterberg, Z.; Donovan, D.; Wampler, W. R.; Watkins, J.; Abrams, T.; McLean, A. G.

2017-10-01

During the tungsten ring campaign on DIII-D, a collector probe system with multiple diameter, dual-facing collector rods was inserted into the far scrape off layer (SOL) near the outer midplane to measure the plasma tungsten content. For most probes more tungsten was observed on the side connected along field lines to the inner divertor, with the larger probes showing largest divertor-facing asymmetries The OEDGE code is used to model the tungsten erosion, transport and deposition. It has been enhanced with (i) a peripheral particle transport and deposition model to record the impurity content in the peripheral region outside the regular mesh, and (ii) a collector probe model. The OEDGE results approximately reproduce both the divertor-facing asymmetries and the radial decay of each collector probe profile. The effect of changing impurity transport assumptions and wall location are examined. The measured divertor-facing asymmetries imply a higher tungsten density in the plasma upstream of the probe; this was expected theoretically from the effect of the parallel ion temperature gradient force driving upstream transport of tungsten from the outer divertor and was also found in the code analysis. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-NA0003525, DE-AC05-00OR22725, and DE-AC52-07NA27344.

Characterization of chemical sputtering using the Mark II DiMES porous plug injector in attached and semi-detached divertor plasmas of DIII-D

NASA Astrophysics Data System (ADS)

McLean, A. G.; Davis, J. W.; Stangeby, P. C.; Allen, S. L.; Boedo, J. A.; Bray, B. D.; Brezinsek, S.; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Haasz, A. A.; Hollmann, E. M.; Isler, R. C.; Lasnier, C. J.; Mu, Y.; Petrie, T. W.; Rudakov, D. L.; Watkins, J. G.; West, W. P.; Whyte, D. G.; Wong, C. P. C.

2009-06-01

An improved, self-contained gas injection system for the divertor material evaluation system (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH 4 are found to be 58 ± 12 in an attached divertor ( ne ˜ 1.5 × 10 13/cm 3, Te ˜ 25 eV, Tsurf ˜ 450 K), and 94 ± 20 in a semi-detached cold divertor ( ne ˜ 6.0 × 10 13/cm 3, Te ˜ 2-3 eV, Tsurf ˜ 350 K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C + at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH 4 from the PPI, then compared with that seen in intrinsic sputtering.

Characterizing Tungsten Sourcing and SOL Transport during the Metal Rings Campaign

NASA Astrophysics Data System (ADS)

Thomas, D. M.; Abrams, T.; Unterberg, E. A.; Donovan, D.; Elder, J. D.; Wampler, W. R.; DIII-D Team

2017-10-01

The Metal Rings Campaign on DIII-D utilized two isotopically and poloidally distinct toroidal arrays of tungsten coated inserts in the lower divertor to study W divertor erosion near the outer strike point (OSP) and divertor entrance and subsequent migration in a mixed-material (C-W) environment. In AT hybrid discharges (PAUX = 14 MW, H98 = 1.6, βN = 3.7) with rapid ELMs (fELM 200 Hz, δW/W 0.7%) W impurities are seen to reach the midplane predominantly from the OSP region rather than the divertor entrance (far-SOL). Conversely, in scenarios with less frequent larger ELMs (fELM 60 Hz, δW/W 3.6%), the W impurities are found to transport equally from the OSP and entrance region. ELM-resolved spectroscopic measurements of W sourcing indicate that large ELMs can source W at many times the inter ELM rate. The peak W erosion rate can shift radially outwards consistent with the ELM energy flux, thereby shifting the balance between strikepoint and far-SOL sources. Changes in the peak erosion locations between forward and reversed Bt discharges are consistent with ExB ion drift effects. Evidence for a near-SOL impurity buildup between the divertors driven by the parallel grad-Ti force is also seen. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698.

Controlling marginally detached divertor plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eldon, David; Kolemen, Egemen; Barton, Joseph L.

A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e = 5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportionalintegral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate. However, the observed bifurcation in plasma conditions at the outer strike point with the ion B ×more » $$\

Progress in extrapolating divertor heat fluxes towards large fusion devices

NASA Astrophysics Data System (ADS)

Sieglin, B.; Faitsch, M.; Eich, T.; Herrmann, A.; Suttrop, W.; Collaborators, JET; the MST1 Team; the ASDEX Upgrade Team

2017-12-01

Heat load to the plasma facing components is one of the major challenges for the development and design of large fusion devices such as ITER. Nowadays fusion experiments can operate with heat load mitigation techniques, e.g. sweeping, impurity seeding, but do not generally require it. For large fusion devices however, heat load mitigation will be essential. This paper presents the current progress of the extrapolation of steady state and transient heat loads towards large fusion devices. For transient heat loads, so-called edge localized modes are considered a serious issue for the lifetime of divertor components. In this paper, the ITER operation at half field (2.65 T) and half current (7.5 MA) will be discussed considering the current material limit for the divertor peak energy fluence of 0.5 {MJ}/{{{m}}}2. Recent studies were successful in describing the observed energy fluence in the JET, MAST and ASDEX Upgrade using the pedestal pressure prior to the ELM crash. Extrapolating this towards ITER results in a more benign heat load compared to previous scalings. In the presence of magnetic perturbation, the axisymmetry is broken and a 2D heat flux pattern is induced on the divertor target, leading to local increase of the heat flux which is a concern for ITER. It is shown that for a moderate divertor broadening S/{λ }{{q}}> 0.5 the toroidal peaking of the heat flux disappears.

The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

DOE PAGES

Perkins, R. J.; Hosea, J. C.; Jaworski, M. A.; ...

2015-04-13

The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. We demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heatmore » flux transmission coefficient in the presence of the RF field. Though the precise comparison between computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. Our work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.« less

Numerical analyses of baseline JT-60SA design concepts with the COREDIV code

NASA Astrophysics Data System (ADS)

Zagórski, R.; Gałązka, K.; Ivanova-Stanik, I.; Stępniewski, W.; Garzotti, L.; Giruzzi, G.; Neu, R.; Romanelli, M.

2017-06-01

JT-60SA reference design scenarios at high (#3) and low (#2) density have been analyzed with the help of the self-consistent COREDIV code. Simulations results for a standard C wall and full W wall have been compared in terms of the influence of impurities, both intrinsic (C, W) and seeded (N, Ar, Ne, Kr), on the radiation losses and plasma parameters. For scenario #3 in a C environment, the regime of detachment on divertor plates can be achieved with N or Ne seeding, whereas for the low density and high power scenario (#2), the C and seeding impurity radiation does not effectively reduce power to the targets. In this case, only an increase of either average density or edge density together with Kr seeding might help to develop conditions with strong radiation losses and semi-detached conditions in the divertor. The calculations show that, in the case of a W divertor, the power load to the plate is mitigated by seeding and the central plasma dilution is smaller compared to the C divertor. For the high density case (#3) with Ne seeding, operation in full detachment mode is predicted. Ar seems to be an optimal choice for the low-density high-power scenario #2, showing a wide operating window, whereas Ne leads to high plasma dilution at high seeding levels albeit not achieving semi-detached conditions in the divertor.

Quiescence near the X-point of MAST measured by high speed visible imaging

NASA Astrophysics Data System (ADS)

Walkden, N. R.; Harrison, J.; Silburn, S. A.; Farley, T.; Henderson, S. S.; Kirk, A.; Militello, F.; Thornton, A.; The MAST Team

2017-12-01

Using high speed imaging of the divertor volume, the region close to the X-point in MAST is shown to be quiescent. This is confirmed by three different analysis techniques and the quiescent X-point region (QXR) spans from the separatrix to the \\psiN = 1.02 flux surface. Local reductions to the atomic density and effects associated with the camera viewing geometry are ruled out as causes of the QXR, leaving quiescence in the local plasma conditions as being the most likely cause. The QXR is found to be ubiquitous across a significant operational space in MAST including L-mode and H-mode discharges across maximal ranges of 9.8×1019~m-2 in line integrated density, 0.36 MA in plasma current, 0.11 T in toroidal magnetic field and 3.2 MW in NBI power. When mapped to the divertor target the QXR occupies approximately an e-folding length of the heat-flux profile, containing  ∼60% of the total heat flux to the target, and also shows a tendency towards higher frequency shorter lived fluctuations in the ion-saturation current. This is consistent with short-lived divertor localised filamentary structures observed further down the outer divertor leg in the camera images, and suggests a complex multi-region picture of filamentary transport in the divertor.

Controlling marginally detached divertor plasmas

DOE PAGES

Eldon, David; Kolemen, Egemen; Barton, Joseph L.; ...

2017-05-04

A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e = 5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportionalintegral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate. However, the observed bifurcation in plasma conditions at the outer strike point with the ion B ×more » $$\

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

The snowflake divertor

NASA Astrophysics Data System (ADS)

Ryutov, D. D.; Soukhanovskii, V. A.

2015-11-01

The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. Among potential beneficial effects of this geometry are: increased volume of a low poloidal field around the null, increased connection length, and the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described.

Suppression of tritium retention in remote areas of ITER by nonperturbative reactive gas injection.

PubMed

Tabarés, F L; Ferreira, J A; Ramos, A; van Rooij, G; Westerhout, J; Al, R; Rapp, J; Drenik, A; Mozetic, M

2010-10-22

A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4  nm/min deposition can be suppressed by addition of 1  Pa·m³ s⁻¹ ammonia flow at 10 cm from the plasma. These results bolster the concept of nonperturbative scavenger injection for tritium inventory control in carbon-based fusion plasma devices, thus paving the way for ITER operation in the active phase under a carbon-dominated, plasma facing component background.

Toroidally symmetric plasma vortex at tokamak divertor null point

DOE PAGES

Umansky, M. V.; Ryutov, D. D.

2016-03-09

Reduced MHD equations are used for studying toroidally symmetric plasma dynamics near the divertor null point. Numerical solution of these equations exhibits a plasma vortex localized at the null point with the time-evolution defined by interplay of the curvature drive, magnetic restoring force, and dissipation. Convective motion is easier to achieve for a second-order null (snowflake) divertor than for a regular x-point configuration, and the size of the convection zone in a snowflake configuration grows with plasma pressure at the null point. In conclusion, the trends in simulations are consistent with tokamak experiments which indicate the presence of enhanced transportmore » at the null point.« less

Do Farm Advisory Services Improve Adoption of Rural Development Policies? An Empirical Analysis in GI Areas

ERIC Educational Resources Information Center

De Rosa, Marcello; Bartoli, Luca

2017-01-01

Purpose: The aim of the paper is to evaluate how advisory services stimulate the adoption of rural development policies (RDP) aiming at value creation. Design/methodology/approach: By linking the use of agricultural extension services (AES) to policies for value creation, we will put forward an empirical analysis in Italy, with the aim of…

Molecular Characterization of Wetland Soil Bacterial Communities in Constructed Mesocosms

DTIC Science & Technology

2008-03-01

Acido., Acidobacteria; Actino., Actinobacteria ; Bacter., Bacteroidetes; Chloro., Chloroflexi; Firm., Firmicutes; Gemma., Gemmatimonadetes; Lenti...Planted sequences (B) using RDP Classifier. Abbreviations: Acidobacteria; Actino., Actinobacteria ; Bacter., Bacteroidetes; Chloro., Chloroflexi...4.2 4.79 Lentisphaerae 0.13 0 Actinobacteria 2.5 3.04 Nitrospira 1.18 1.07 Chloroflexi 3.55 3.04 Acidobacteria 16.16 12.87 Proteobacteria 34.95

Divertor sheath power studies in DIII-D using fixed Langmuir probes and three-dimensional modeling of tile heat flows

NASA Astrophysics Data System (ADS)

Donovan, D.; Nygren, R.; Buchenauer, D.; Watkins, J.; Rudakov, D.; Leonard, A.; Wong, C. P. C.; Makowski, M.

2014-04-01

Experimental results are presented from the three-Langmuir probe (LP) diagnostic head of the divertor material evaluation system (DiMES) on DIII-D that confirm the size of the projected current collection area of the LPs, which is essential for properly measuring ion saturation current density (Jsat) and the sheath power transmission factor (SPTF). Also using the 3-LP DiMES head, the hypothesis that collisional effects on plasma density occurring in the magnetic sheath of the tile are responsible for a lower than expected SPTF is tested and deemed not to have a significant impact on the SPTF. Three-dimensional thermal modeling of wall tiles is presented that accounts for lateral heat conduction, temperature dependence of tile material properties and radiative heat loss from the tile surface. This modeling was developed to be used in the analysis of temperature profiles of the divertor embedded thermocouple (TC) array to obtain more accurate interpretations of TC temperature profiles to infer divertor surface heat flux than have previously been accomplished using more basic one-dimensional methods.

Fast plasma shutdown by killer pellet injection in JT-60U with reduced heat flux on the divertor plate and avoiding runaway electron generation

NASA Astrophysics Data System (ADS)

Yoshino, R.; Kondoh, T.; Neyatani, Y.; Itami, K.; Kawano, Y.; Isei, N.

1997-02-01

A killer pellet is an impurity pellet that is injected into a tokamak plasma in order to terminate a discharge without causing serious damage to the tokamak machine. In JT-60U neon ice pellets have been injected into OH and NB heated plasmas and fast plasma shutdowns have been demonstrated without large vertical displacement. The heat pulse on the divertor plate has been greatly reduced by killer pellet injection (KPI), but a low-power heat flux tail with a long time duration is observed. The total energy on the divertor plate increases with longer heat flux tail, so it has been reduced by shortening the tail. Runaway electron (RE) generation has been observed just after KPI and/or in the later phase of the plasma current quench. However, RE generation has been avoided when large magnetic perturbations are excited. These experimental results clearly show that KPI is a credible fast shutdown method avoiding large vertical displacement, reducing heat flux on the divertor plate, and avoiding (or minimizing) RE generation.

ADX - Advanced Divertor and RF Tokamak Experiment

NASA Astrophysics Data System (ADS)

Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

2015-11-01

The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

FLIT: Flowing LIquid metal Torus

NASA Astrophysics Data System (ADS)

Kolemen, Egemen; Majeski, Richard; Maingi, Rajesh; Hvasta, Michael

2017-10-01

The design and construction of FLIT, Flowing LIquid Torus, at PPPL is presented. FLIT focuses on a liquid metal divertor system suitable for implementation and testing in present-day fusion systems, such as NSTX-U. It is designed as a proof-of-concept fast-flowing liquid metal divertor that can handle heat flux of 10 MW/m2 without an additional cooling system. The 72 cm wide by 107 cm tall torus system consisting of 12 rectangular coils that give 1 Tesla magnetic field in the center and it can operate for greater than 10 seconds at this field. Initially, 30 gallons Galinstan (Ga-In-Sn) will be recirculated using 6 jxB pumps and flow velocities of up to 10 m/s will be achieved on the fully annular divertor plate. FLIT is designed as a flexible machine that will allow experimental testing of various liquid metal injection techniques, study of flow instabilities, and their control in order to prove the feasibility of liquid metal divertor concept for fusion reactors. FLIT: Flowing LIquid metal Torus. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

Coupled Kinetic-MHD Simulations of Divertor Heat Load with ELM Perturbations

NASA Astrophysics Data System (ADS)

Cummings, Julian; Chang, C. S.; Park, Gunyoung; Sugiyama, Linda; Pankin, Alexei; Klasky, Scott; Podhorszki, Norbert; Docan, Ciprian; Parashar, Manish

2010-11-01

The effect of Type-I ELM activity on divertor plate heat load is a key component of the DOE OFES Joint Research Target milestones for this year. In this talk, we present simulations of kinetic edge physics, ELM activity, and the associated divertor heat loads in which we couple the discrete guiding-center neoclassical transport code XGC0 with the nonlinear extended MHD code M3D using the End-to-end Framework for Fusion Integrated Simulations, or EFFIS. In these coupled simulations, the kinetic code and the MHD code run concurrently on the same massively parallel platform and periodic data exchanges are performed using a memory-to-memory coupling technology provided by EFFIS. The M3D code models the fast ELM event and sends frequent updates of the magnetic field perturbations and electrostatic potential to XGC0, which in turn tracks particle dynamics under the influence of these perturbations and collects divertor particle and energy flux statistics. We describe here how EFFIS technologies facilitate these coupled simulations and discuss results for DIII-D, NSTX and Alcator C-Mod tokamak discharges.

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

NASA Astrophysics Data System (ADS)

Deng, Guozhong; Liu, Xiaoju; Wang, Liang; Liu, Shaocheng; Xu, Jichan; Feng, Wei; Liu, Jianbin; Liu, Huan; Gao, Xiang

2017-04-01

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak (EAST) L-mode and ELM-free H-mode plasmas. The divertor power footprint widths, which consist of the scrape-off layer (SOL) width λ q and heat spreading S, are important physical parameters for edge plasmas. In this work, a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I p. Strong inverse scaling of the SOL width with I p has been achieved for both L-mode and H-mode plasmas in the forms of {λ }q,{{L}\\text-\\text{mode}}=4.98× {I}{{p}}-0.68 and {λ }q,{{H}\\text-\\text{mode}}=1.86× {I}{{p}}-1.08. Similar trends have also been demonstrated in the study of heat spreading with {S}{{L}\\text-\\text{mode}}=1.95× {I}{{p}}-0.542 and {S}{{H}\\text-\\text{mode}}=0.756× {I}{{p}}-0.872. In addition, studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current. The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor (CFETR).

Overview of decade-long development of plasma-facing components at ASIPP

NASA Astrophysics Data System (ADS)

Luo, G.-N.; Liu, G. H.; Li, Q.; Qin, S. G.; Wang, W. J.; Shi, Y. L.; Xie, C. Y.; Chen, Z. M.; Missirlian, M.; Guilhem, D.; Richou, M.; Hirai, T.; Escourbiac, F.; Yao, D. M.; Chen, J. L.; Wang, T. J.; Bucalossi, J.; Merola, M.; Li, J. G.; EAST Team

2017-06-01

The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m-2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m-2 and 1000 cycles at 20 MW m-2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m-2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.

Divertor-localized fluctuations in NSTX-U L-mode discharges

NASA Astrophysics Data System (ADS)

Scotti, Filippo; Soukhanovskii, V. A.; Zweben, S.; Myra, J.; Baver, D.; Sabbagh, S. A.

2017-10-01

The 3-D structure of divertor turbulence is characterized in NSTX-U by means of fast camera imaging. Edge and divertor turbulence can be important in determining the heat flux width in fusion devices. Field-aligned filaments are found on the divertor legs via imaging of C III and D- α emission in NBI-heated diverted L-mode discharges, similar to observations in Alcator C-Mod and MAST. These flute-like fluctuations of up to 10-20% in RMS/mean are radially localized around the separatrix and limited to the region below the X-point. Poloidal and parallel correlation lengths are a few cm (10-50ρi) and several meters, respectively. For the outer leg filaments, poloidal correlation lengths decrease along the leg away from the strike point and typical effective toroidal mode numbers are in the range of 10-20. Opposite toroidal rotation is observed for inner (co-current rotation) and outer leg (counter-current rotation) filaments with apparent poloidal propagation of 1 km/s. The poloidal motion of outer leg filaments is opposite to the one typically observed for NSTX upstream blobs in the scrape-off layer. The shape, dynamics and absence of correlation with upstream turbulence suggest that these fluctuations are generated and localized in the divertor region. Supported by US DOE DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FG02- 02ER54678, DE-FG02-99ER54524.

Controlling marginally detached divertor plasmas

NASA Astrophysics Data System (ADS)

Eldon, D.; Kolemen, E.; Barton, J. L.; Briesemeister, A. R.; Humphreys, D. A.; Leonard, A. W.; Maingi, R.; Makowski, M. A.; McLean, A. G.; Moser, A. L.; Stangeby, P. C.

2017-06-01

A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e  =  5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B   ×  \

Lesson from Tungsten Leading Edge Heat Load Analysis in KSTAR Divertor

NASA Astrophysics Data System (ADS)

Hong, Suk-Ho; Pitts, Richard Anthony; Lee, Hyeong-Ho; Bang, Eunnam; Kang, Chan-Soo; Kim, Kyung-Min; Kim, Hong-Tack; ITER Organization Collaboration; Kstar Team Team

2016-10-01

An important design issue for the ITER tungsten (W) divertor and in fact for all such components using metallic plasma-facing elements and which are exposed to high parallel power fluxes, is the question of surface shaping to avoid melting of leading edges. We have fabricated a series of tungsten blocks with a variety of leading edge heights (0.3, 0.6, 1.0, and 2.0 mm), from the ITER worst case to heights even beyond the extreme value tested on JET. They are mounted into adjacent, inertially cooled graphite tile installed in the central divertor region of KSTAR, within the field of view of an infra-red (IR) thermography system with a spatial resolution to 0.4 mm/pixel. Adjustment of the outer divertor strike point position is used to deposit power on the different blocks in different discharges. The measured power flux density on flat regions of the surrounding graphite tiles is used to obtain the parallel power flux, q|| impinging on the various W blocks. Experiments have been performed in Type I ELMing H-mode with Ip = 600 kA, BT = 2 T, PNBI = 3.5 MW, leading to a hot attached divertor with typical pulse lengths of 10 s. Three dimensional ANSYS simulations using q|| and assuming geometric projection of the heat flux are found to be consistent with the observed edge loading. This research was partially supported by Ministry of Science, ICT, and Future Planning under KSTAR project.

Changes in divertor conditions in response to changing core density with RMPs

DOE PAGES

Briesemeister, Alexis R.; Ahn, Joon -Wook; Canik, John M.; ...

2017-06-07

The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicated non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have components which has at least one but typically many resonances with the rotational transform of the plasma. RMPs are found to alter inter-ELM heat flux to the divertormore » by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that, local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. Furthermore, these trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity, is seen after RMPs are applied.« less

Effects of low-Z and high-Z impurities on divertor detachment and plasma confinement

DOE PAGES

Wang, H. Q.; Guo, Houyang Y.; Petrie, Thomas W.; ...

2017-03-18

The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N 2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N 2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss powermore » to the L-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection leads to an increase in core radiation, up to 50% of the injected power, and a reduction in pedestal temperature over 60%, thus significantly degrading the confinement, i.e., with H 98 reducing from 1.1 to below 0.7. As for Ne seeding, H 98 near 0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N 2 seeded plasmas, radiation is predominately confined in the boundary plasma, with up to 50% of heating power being radiated in the divertor region and less than 25% in the core at the onset of detachment. In the case of strong N 2 gas puffing, the confinement recovers during the detachment, from ~20% reduction at the onset of the detachment to greater than that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.« less

Fabrication of divertor mock-up with ODS-Cu and W by the improved brazing technique

NASA Astrophysics Data System (ADS)

Tokitani, M.; Hamaji, Y.; Hiraoka, Y.; Masuzaki, S.; Tamura, H.; Noto, H.; Tanaka, T.; Muroga, T.; Sagara, A.; FFHR Design Group

2017-07-01

Copper alloy has been considered as a divertor cooling tube or heat sink not only in the helical reactor FFHR-d1 but also in the tokamak DEMO reactor, because it has a high thermal conductivity. This work focused on applying an oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu-0.3 wt%Al2O3) as the divertor heat sink material of FFHR-d1. This alloy has superior high temperature yield strength exceeding 300 MPa at room temperature even after annealing up to ~1000 °C. The change in material properties of Pure-Cu, GlidCop® and CuCrZr by neutron irradiation are summarized in this paper. A primary dose limit is the radiation-induced hardening/softening (~0.2 dpa/1-2 dpa) which has a temperature dependence. According to such an evaluation, the GlidCop® can be selected as the current best candidate material in the commercial base of the divertor heat sink, and its temperature should be maintained as close as possible to 300 °C during operation. Bonding between the W armour and the GlidCop® heat sink was successfully performed by using an improved brazing technique with BNi-6 (Ni-11%P) filler material. The bonding strength was measured by a three-point bending test and reached up to approximately 200 MPa. Surprisingly, several specimens showed an obvious yield point. This means that the BNi-6 brazing (bonding) layer caused relaxation of the applied stress. The small-scale divertor mock-up of the W/BNi-6/GlidCop® was successfully fabricated by using the improved brazing technique. The heat loading test was carried out by the electron beam device ACT2 in NIFS. The mock-up showed an excellent heat removal capability for use in the FFHR-d1 divertor.

Changes in divertor conditions in response to changing core density with RMPs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Briesemeister, Alexis R.; Ahn, Joon -Wook; Canik, John M.

The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicated non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have components which has at least one but typically many resonances with the rotational transform of the plasma. RMPs are found to alter inter-ELM heat flux to the divertormore » by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that, local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. Furthermore, these trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity, is seen after RMPs are applied.« less

Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

NASA Astrophysics Data System (ADS)

Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

2018-06-01

The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

Liquid lithium loop system to solve challenging technology issues for fusion power plant

DOE PAGES

Ono, Masayuki; Majeski, Richard P.; Jaworski, Michael A.; ...

2017-07-12

Here, steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peakmore » heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned. We examined two key technology issues: 1) dust or solid particle removal and 2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust / impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~ 1 l/sec LL flow, even a small 0.1% dust content by weight (or 0.5 g per sec) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW fusion power) fusion power plant, about 0.5 g / sec of tritium is needed to maintain the fusion fuel cycle assuming ~ 1 % fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap (SCT) concept.« less

Liquid lithium loop system to solve challenging technology issues for fusion power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, Masayuki; Majeski, Richard P.; Jaworski, Michael A.

Here, steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peakmore » heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned. We examined two key technology issues: 1) dust or solid particle removal and 2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust / impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~ 1 l/sec LL flow, even a small 0.1% dust content by weight (or 0.5 g per sec) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW fusion power) fusion power plant, about 0.5 g / sec of tritium is needed to maintain the fusion fuel cycle assuming ~ 1 % fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap (SCT) concept.« less

Liquid lithium loop system to solve challenging technology issues for fusion power plant

NASA Astrophysics Data System (ADS)

Ono, M.; Majeski, R.; Jaworski, M. A.; Hirooka, Y.; Kaita, R.; Gray, T. K.; Maingi, R.; Skinner, C. H.; Christenson, M.; Ruzic, D. N.

2017-11-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s-1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s-1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s-1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s-1 of tritium is needed to maintain the fusion fuel cycle assuming ~1% fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap concept.

Nitrogen efficiency of eastern Canadian dairy herds: Effect on production performance and farm profitability.

PubMed

Fadul-Pacheco, L; Pellerin, D; Chouinard, P Y; Wattiaux, M A; Duplessis, M; Charbonneau, É

2017-08-01

Nitrogen efficiency (milk N/dietary N; NE) can be used as a tool for the nutritional, economic, and environmental management of dairy farms. The aim of this study was to identify the characteristics of herds with varying NE and assess the effect on farm profitability. One hundred dairy herds located in Québec, Canada, comprising on average 42 ± 18 cows in lactation were visited from October 2014 to June 2015. Feed intake was measured over 24 h. Samples of each feedstuff were taken and sent to a commercial laboratory for analysis of chemical composition. Feeding management and feed prices were recorded. Milk yield was recorded and milk samples were collected over 2 consecutive milkings. Fat, protein, and milk urea N were analyzed. Balances of metabolizable protein (MP; MP supply - MP requirements) and rumen degradable protein (RDP; RDP supply - RDP requirement) were calculated. A hierarchical cluster analysis was conducted and allowed grouping the farms by their NE. Four clusters were identified with an average NE of 22.1 (NE22), 26.9 (NE27), 30.0 (NE30), and 35.8% (NE36). Herds in clusters NE30 and NE36 were fed diets with greater concentrations of starch, net energy for lactation, and nonfiber carbohydrates than those in the other 2 clusters. Moreover, the average proportion of corn silage was lower for herds in cluster NE22 compared with NE30 and NE36 (8.23 vs. 31.8 and 31.3% of total forages, respectively). In addition, crude protein of the diets declined from an average of 16.0 to 14.9% with increasing NE among clusters. Average dry matter intake declined from 26.1 to 22.5 kg/d as NE of clusters increased. Herds in cluster NE22 had lower yields of milk (28.7 vs. 31.8 kg/d), fat (1.15 vs. 1.29 kg/d), and protein (0.94 vs. 1.05 kg/d) than the other clusters. Also, milk urea N was greater for farms in cluster NE22 (13.2 mg/dL) than for farms in the other clusters (11.4 mg/dL). Furthermore, MP and RDP balances decreased from 263.2 to -153.7 g/d and from 594.7 to 486.9 g/d, respectively, with increasing NE among clusters. Income over feed cost increased from $14.3 to $17.3/cow per day (Can$) as NE among clusters augmented. Results from this study showed that some farms were able to achieve high NE by using lower levels of dietary N and having cows with lower DMI while maintaining milk performance. These farms had a potentially lower environmental impact, and they were more profitable. Copyright © 2017 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Lithium As Plasma Facing Component for Magnetic Fusion Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masayuki Ono

The use of lithium in magnetic fusion confinement experiments started in the 1990's in order to improve tokamak plasma performance as a low-recycling plasma-facing component (PFC). Lithium is the lightest alkali metal and it is highly chemically reactive with relevant ion species in fusion plasmas including hydrogen, deuterium, tritium, carbon, and oxygen. Because of the reactive properties, lithium can provide strong pumping for those ions. It was indeed a spectacular success in TFTR where a very small amount (~ 0.02 gram) of lithium coating of the PFCs resulted in the fusion power output to improve by nearly a factor ofmore » two. The plasma confinement also improved by a factor of two. This success was attributed to the reduced recycling of cold gas surrounding the fusion plasma due to highly reactive lithium on the wall. The plasma confinement and performance improvements have since been confirmed in a large number of fusion devices with various magnetic configurations including CDX-U/LTX (US), CPD (Japan), HT-7 (China), EAST (China), FTU (Italy), NSTX (US), T-10, T-11M (Russia), TJ-II (Spain), and RFX (Italy). Additionally, lithium was shown to broaden the plasma pressure profile in NSTX, which is advantageous in achieving high performance H-mode operation for tokamak reactors. It is also noted that even with significant applications (up to 1,000 grams in NSTX) of lithium on PFCs, very little contamination (< 0.1%) of lithium fraction in main fusion plasma core was observed even during high confinement modes. The lithium therefore appears to be a highly desirable material to be used as a plasma PFC material from the magnetic fusion plasma performance and operational point of view. An exciting development in recent years is the growing realization of lithium as a potential solution to solve the exceptionally challenging need to handle the fusion reactor divertor heat flux, which could reach 60 MW/m2 . By placing the liquid lithium (LL) surface in the path of the main divertor heat flux (divertor strike point), the lithium is evaporated from the surface. The evaporated lithium is quickly ionized by the plasma and the ionized lithium ions can provide a strongly radiative layer of plasma ("radiative mantle"), thus could significantly reduce the heat flux to the divertor strike point surfaces, thus protecting the divertor surface. The protective effects of LL have been observed in many experiments and test stands. As a possible reactor divertor candidate, a closed LL divertor system is described. Finally, it is noted that the lithium applications as a PFC can be quite flexible and broad. The lithium application should be quite compatible with various divertor configurations, and it can be also applied to protecting the presently envisioned tungsten based solid PFC surfaces such as the ones for ITER. Lithium based PFCs therefore have the exciting prospect of providing a cost effective flexible means to improve the fusion reactor performance, while providing a practical solution to the highly challenging divertor heat handling issue confronting the steadystate magnetic fusion reactors.« less

Numerical study of the current-convective instability driven by asymmetry of detachment in inner and outer divertors

NASA Astrophysics Data System (ADS)

Stepanenko, A. A.; Krasheninnikov, S. I.

2018-01-01

One of the possible mechanisms responsible for strong radiation fluctuations observed in recent experiments with detached plasmas at ASDEX Upgrade [Potzel et al., Nucl. Fusion 54, 013001 (2014)] can be related to the onset of the current-convective instability (CCI) driven by strong asymmetry of detachment in the inner and outer divertors of the tokamak [S. Krasheninnikov and A. Smolyakov, Phys. Plasmas 23, 092505 (2016)]. In this study, we present the physical model, used to simulate the CCI, and the first numerical results of modeling of the CCI dynamics in ASDEX Upgrade-like conditions. The simulation results provide frequency spectra of turbulent divertor plasma oscillations showing reasonably good agreement with the available experimental data.

Interactions of corn meal or molasses with a soybean-sunflower meal mix or flaxseed meal on production, milk fatty acids composition, and nutrient utilization in dairy cows fed grass hay-based diets

USDA-ARS?s Scientific Manuscript database

We investigated the interactions of molasses or corn meal [nonstructural carbohydrate (NSC) sources] with flaxseed meal or a soybean-sunflower meal protein mix [rumen-degradable protein (RDP) sources] on animal production, milk fatty acids profile, and nutrient utilization in organic Jersey cows fed...

Quantifying ruminal nitrogen metabolism using the omasal sampling technique in cattle--a meta-analysis.

PubMed

Broderick, G A; Huhtanen, P; Ahvenjärvi, S; Reynal, S M; Shingfield, K J

2010-07-01

Mixed model analysis of data from 32 studies (122 diets) was used to evaluate the precision and accuracy of the omasal sampling technique for quantifying ruminal-N metabolism and to assess the relationships between nonammonia-N flow at the omasal canal and milk protein yield. Data were derived from experiments in cattle fed North American diets (n=36) based on alfalfa silage, corn silage, and corn grain and Northern European diets (n=86) composed of grass silage and barley-based concentrates. In all studies, digesta flow was quantified using a triple-marker approach. Linear regressions were used to predict microbial-N flow to the omasum from intake of dry matter (DM), organic matter (OM), or total digestible nutrients. Efficiency of microbial-N synthesis increased with DM intake and there were trends for increased efficiency with elevated dietary concentrations of crude protein (CP) and rumen-degraded protein (RDP) but these effects were small. Regression of omasal rumen-undegraded protein (RUP) flow on CP intake indicated that an average 32% of dietary CP escaped and 68% was degraded in the rumen. The slope from regression of observed omasal flows of RUP on flows predicted by the National Research Council (2001) model indicated that NRC predicted greater RUP supply. Measured microbial-N flow was, on average, 26% greater than that predicted by the NRC model. Zero ruminal N-balance (omasal CP flow=CP intake) was obtained at dietary CP and RDP concentrations of 147 and 106 g/kg of DM, corresponding to ruminal ammonia-N and milk urea N concentrations of 7.1 and 8.3mg/100mL, respectively. Milk protein yield was positively related to the efficiency of microbial-N synthesis and measured RUP concentration. Improved efficiency of microbial-N synthesis and reduced ruminal CP degradability were positively associated with efficiency of capture of dietary N as milk N. In conclusion, the results of this study indicate that the omasal sampling technique yields valuable estimates of RDP, RUP, and ruminal microbial protein supply in cattle. Copyright (c) 2010 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

NASA Astrophysics Data System (ADS)

Brunner, D.; Wolfe, S. M.; LaBombard, B.; Kuang, A. Q.; Lipschultz, B.; Reinke, M. L.; Hubbard, A.; Hughes, J.; Mumgaard, R. T.; Terry, J. L.; Umansky, M. V.; The Alcator C-Mod Team

2017-08-01

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m-2 down to  <10 MW m-2 while maintaining excellent core confinement, H 98  >  1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

Achievement of radiative feedback control for long-pulse operation on EAST

NASA Astrophysics Data System (ADS)

Wu, K.; Yuan, Q. P.; Xiao, B. J.; Wang, L.; Duan, Y. M.; Chen, J. B.; Zheng, X. W.; Liu, X. J.; Zhang, B.; Xu, J. C.; Luo, Z. P.; Zang, Q.; Li, Y. Y.; Feng, W.; Wu, J. H.; Yang, Z. S.; Zhang, L.; Luo, G.-N.; Gong, X. Z.; Hu, L. Q.; Hu, J. S.; Li, J.

2018-05-01

The active feedback control of radiated power to prevent divertor target plates overheating during long-pulse operation has been developed and implemented on EAST. The radiation control algorithm, with impurity seeding via a supersonic molecular beam injection (SMBI) system, has shown great success in both reliability and stability. By seeding a sequence of short neon (Ne) impurity pulses with the SMBI from the outer mid-plane, the radiated power of the bulk plasma can be well controlled, and the duration of radiative control (feedforward and feedback) is 4.5 s during a discharge of 10 s. Reliable control of the total radiated power of bulk plasma has been successfully achieved in long-pulse upper single null (USN) discharges with a tungsten divertor. The achieved control range of {{f}rad} is 20%–30% in L-mode regimes and 18%–36% in H-mode regimes. The temperature of the divertor target plates was maintained at a low level during the radiative control phase. The peak particle flux on the divertor target was decreased by feedforward Ne injection in the L-mode discharges, while the Ne pulses from the SMBI had no influence on the peak particle flux because of the very small injecting volume. It is shown that although the radiated power increased, no serious reduction of plasma-stored energy or confinement was observed during the control phase. The success of the radiation control algorithm and current experiments in radiated power control represents a significant advance for steady-state divertor radiation and heat flux control on EAST for near-future long-pulse operation.

Design and Calibration of a Dispersive Imaging Spectrometer Adaptor for a Fast IR Camera on NSTX-U

NASA Astrophysics Data System (ADS)

Reksoatmodjo, Richard; Gray, Travis; Princeton Plasma Physics Laboratory Team

2017-10-01

A dispersive spectrometer adaptor was designed, constructed and calibrated for use on a fast infrared camera employed to measure temperatures on the lower divertor tiles of the NSTX-U tokamak. This adaptor efficiently and evenly filters and distributes long-wavelength infrared photons between 8.0 and 12.0 microns across the 128x128 pixel detector of the fast IR camera. By determining the width of these separated wavelength bands across the camera detector, and then determining the corresponding average photon count for each photon wavelength, a very accurate measurement of the temperature, and thus heat flux, of the divertor tiles can be calculated using Plank's law. This approach of designing an exterior dispersive adaptor for the fast IR camera allows accurate temperature measurements to be made of materials with unknown emissivity. Further, the relative simplicity and affordability of this adaptor design provides an attractive option over more expensive, slower, dispersive IR camera systems. This work was made possible by funding from the Department of Energy for the Summer Undergraduate Laboratory Internship (SULI) program. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

The Effects of Aperiodic Perturbation on the Last Good Surface of a Single-Null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Jordan, Joseph; Punjabi, Alkesh; Ali, Halima

2003-10-01

An area-preserving discrete mapping, called the Simple Map is used to analyze the effects of aperiodic perturbation on the last good magnetic surface of a single-null divertor tokamak. The SM was developed by Punjabi and Boozer /1/. The SM equations are modified to reflect the effect of aperiodic perturbation: Xn + 1 = Xn - (k+δ_aperiodics(n))Y_n(1-Y_n), Y_n+1 = Yn + kX_n+1 where δ_aperiodic is the amplitude of the aperiodic perturbation. The equations are coded into FORTRAN and used in conjunction with a graphing program to show the trajectories for different amplitudes of aperiodic perturbation. As the aperiodic perturbation is increased, the trajectories break from a clean curve and develop strange behavior as they approach chaos. The progression of the trajectories from a clean curve to a chaotic one is observed, and the results are presented in this paper. This project is supported by NASA SHARP and US DOE Grant Number DE-F02-02ER54673. The research was done under the mentorship of Profs. Punjabi and Ali. 1. Punjabi A., Verma A., and Boozer A., Phys. Rev. Lett, 69, 3322 (1992)

High-resolution disruption halo current measurements using Langmuir probes in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Tinguely, R. A.; Granetz, R. S.; Berg, A.; Kuang, A. Q.; Brunner, D.; LaBombard, B.

2018-01-01

Halo currents generated during disruptions on Alcator C-Mod have been measured with Langmuir ‘rail’ probes. These rail probes are embedded in a lower outboard divertor module in a closely-spaced vertical (poloidal) array. The dense array provides detailed resolution of the spatial dependence (~1 cm spacing) of the halo current distribution in the plasma scrape-off region with high time resolution (400 kHz digitization rate). As the plasma limits on the outboard divertor plate, the contact point is clearly discernible in the halo current data (as an inversion of current) and moves vertically down the divertor plate on many disruptions. These data are consistent with filament reconstructions of the plasma boundary, from which the edge safety factor of the disrupting plasma can be calculated. Additionally, the halo current ‘footprint’ on the divertor plate is obtained and related to the halo flux width. The voltage driving halo current and the effective resistance of the plasma region through which the halo current flows to reach the probes are also investigated. Estimations of the sheath resistance and halo region resistivity and temperature are given. This information could prove useful for modeling halo current dynamics.

Low temperature tungsten spectroscopy on a Penning Ionization Discharge

NASA Astrophysics Data System (ADS)

Kumar, Deepak; Englesbe, Alexander; Stutman, Dan; Finkenthal, Michael

2011-10-01

Complete Tungsten divertor operation is being planned on many tokamaks including Tore Supra and ITER. Thus, low temperature tungsten spectroscopy is important for aiding the divertor diagnostics on larger machines. A Penning Ionization Discharge (PID) at the Johns Hopkins University produces steady state plasmas with Te ~ 2 eV, ne ~1013 cm-3 and a fast electron fraction at ~ 10 s eV. Similar bi-Maxwellian distributions, but with slightly higher electron temperatures, are found in the divertor plasmas of tokamaks. The two significant populating mechanisms for higher charge states in the PID are: (a) collisional excitation from bulk electrons, and (b) inner shell ionization from the fast electrons. The PID is diagnosed in a wide wavelength range - XUV, VUV and visible, to differentiate the two populating mechanisms. W is introduced in the PID by the sputtering of cathodes made of CuW alloy. Spectral emission from significantly higher charge states of W (up to W IV) has been observed in the experiment. This poster will describe results indicating the populating mechanism of W ions and also describe plans on upgrading the experiment to achieve higher temperatures which are closer to the divertor conditions. Supported by USDOE.

Carbon Deposition in the Inner JET Divertor Measured by Means of Quartz Microbalance

NASA Astrophysics Data System (ADS)

Esser, H. G.; Philipps, V.; Freisinger, M.; Coad, P.; Matthews, G. F.; Neill, G.; JET EFDA Contributors

A Quartz Microbalance (QMB) system was implemented in the inner divertor region of JET in order to measure in situ and time resolved (minimum exposure time ≥0.1 s) material fluxes (mainly carbon) and layer deposition. The system has been developed to operate at temperatures up to 200°C. The aim is to investigate carbon transport to the remote areas, and hence the tritium retention in dependence on plasma conditions. This question is still a major concern for the ITER operation. The mass sensitivity of the system is Sm = 1.5 A~— 10-8 [g/Hz cm2]. First reliable measurements were made during the C5 campaign (March–May 2002; â‰e1000 plasma discharges). The results presented are based on 74 selected exposures (694 s) under various conditions (strike point position, input power, neutral pressure, ELM frequency). Most influencing on the carbon deposition in the remote area seems to be the geometry i.e. the strike point position on the divertor tiles. In average 1.9 A~— 10-4 C-atom are deposited per deuterium ion flowing into the inner divertor.

Characterization of edge turbulence in different states of divertor detachment using reflectometry in the ASDEX Upgrade tokamak

NASA Astrophysics Data System (ADS)

Nikolaeva, V.; Guimarais, L.; Manz, P.; Carralero, D.; Manso, M. E.; Stroth, U.; Silva, C.; Conway, G. D.; Seliunin, E.; Vicente, J.; Brida, D.; Aguiam, D.; Santos, J.; Silva, A.; ASDEX Upgrade team; MST1 team

2018-05-01

Transport in the scrape-off layer (SOL) depends on the state of divertor detachment. L-mode discharges were analyzed where the state of divertor detachment is varied through a density ramp-up. By means of reflectometry measurements at the low (LFS) and the high field side (HFS), midplane density fluctuations are studied for the first time in ASDEX Upgrade simultaneously at both sides of the tokamak. Radial density fluctuation profiles (δ {n}e/{n}e) increase with radius in both the HFS and the LFS. It is found that in the SOL density fluctuations at the LFS have about a factor of two larger amplitude than at the HFS in agreement with ballooned transport. Density fluctuations at the LFS show a modest variation with increasing background density resulting mainly from a rise of low frequency components. Experimental results are in good agreement with an enhanced convection of filaments at the LFS at the beginning of outer divertor detachment leading to a flatter SOL density profile. In this phase of the discharge, density fluctuations measured at the HFS far-SOL display a strong increase, which may be associated with the presence of faster filaments originated at the LFS.

Heat removal capability of divertor coaxial tube assembly

NASA Astrophysics Data System (ADS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications.

Overview of the JET results

NASA Astrophysics Data System (ADS)

Romanelli, F.; JET Contributors,

2015-10-01

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

Rapid change of blob structure in the outer scrape-off layer (SOL)

NASA Astrophysics Data System (ADS)

Cohen, R. H.

2005-10-01

Nonlinear structures (``blobs'') driven by the magnetic field curvature and highly elongated along the field lines may exist in the tokamak SOL.footnotetextS.I. Krasheninnikov. Phys. Lett. A 283, 368 (2001) The contact of the blob end with the divertor plate significantly affects the blob structure and velocity. However, the strong shearing of the flux-tube near the X-point makes impossible direct electrical contact of the blob in the upper SOL and the divertor, so that the sheath boundary condition (BC) has to be replaced by a BC imposed near the X point.footnotetextD. Ryutov, R.H. Cohen. Contr. Pl. Phys 44, 168 (2004) We show that, at larger distances from the separatrix, in the far SOL, the connection between the upper SOL and the divertor plate is re-established, and the sheath BC becomes again relevant. During the blob's outward radial motion, this event is reflected in a sudden change of its length, from the blob extending only to the X point to the blob extending down to the plate. Likewise, a blob initially existing only in the divertor leg becomes suddenly longer, and extends to the whole SOL.

Automated divertor target design by adjoint shape sensitivity analysis and a one-shot method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dekeyser, W., E-mail: Wouter.Dekeyser@kuleuven.be; Reiter, D.; Baelmans, M.

As magnetic confinement fusion progresses towards the development of first reactor-scale devices, computational tokamak divertor design is a topic of high priority. Presently, edge plasma codes are used in a forward approach, where magnetic field and divertor geometry are manually adjusted to meet design requirements. Due to the complex edge plasma flows and large number of design variables, this method is computationally very demanding. On the other hand, efficient optimization-based design strategies have been developed in computational aerodynamics and fluid mechanics. Such an optimization approach to divertor target shape design is elaborated in the present paper. A general formulation ofmore » the design problems is given, and conditions characterizing the optimal designs are formulated. Using a continuous adjoint framework, design sensitivities can be computed at a cost of only two edge plasma simulations, independent of the number of design variables. Furthermore, by using a one-shot method the entire optimization problem can be solved at an equivalent cost of only a few forward simulations. The methodology is applied to target shape design for uniform power load, in simplified edge plasma geometry.« less

Development of high poloidal beta, steady-state scenario with ITER-like tungsten divertor on EAST

NASA Astrophysics Data System (ADS)

Garofalo, A. M.; Gong, X. Z.; Qian, J.; Chen, J.; Li, G.; Li, K.; Li, M. H.; Zhai, X.; Bonoli, P.; Brower, D.; Cao, L.; Cui, L.; Ding, S.; Ding, W. X.; Guo, W.; Holcomb, C.; Huang, J.; Hyatt, A.; Lanctot, M.; Lao, L. L.; Liu, H.; Lyu, B.; McClenaghan, J.; Peysson, Y.; Ren, Q.; Shiraiwa, S.; Solomon, W.; Zang, Q.; Wan, B.

2017-07-01

Recent experiments on EAST have achieved the first long pulse H-mode (61 s) with zero loop voltage and an ITER-like tungsten divertor, and have demonstrated access to broad plasma current profiles by increasing the density in fully-noninductive lower hybrid current-driven discharges. These long pulse discharges reach wall thermal and particle balance, exhibit stationary good confinement (H 98y2 ~ 1.1) with low core electron transport, and are only possible with optimal active cooling of the tungsten armors. In separate experiments, the electron density was systematically varied in order to study its effect on the deposition profile of the external lower hybrid current drive (LHCD), while keeping the plasma in fully-noninductive conditions and with divertor strike points on the tungsten divertor. A broadening of the current profile is found, as indicated by lower values of the internal inductance at higher density. A broad current profile is attractive because, among other reasons, it enables internal transport barriers at large minor radius, leading to improved confinement as shown in companion DIII-D experiments. These experiments strengthen the physics basis for achieving high performance, steady state discharges in future burning plasmas.

Impact of the impurity seeding for divertor protection on the performance of fusion reactors

NASA Astrophysics Data System (ADS)

Siccinio, Mattia; Fable, Emiliano; Angioni, Clemente; Saarelma, Samuli; Scarabosio, Andrea; Zohm, Hartmut

2017-10-01

A 0D divertor and scrape-off layer (SOL) model has been coupled to the 1.5D core transport code ASTRA. The resulting numerical tool has been employed for various parameter scans in order to identify the most convenient choices for the operation of electricity producing fusion devices with seeded impurities for the divertor protection. In particular, the repercussions of such radiative species on the main plasma through the fuel dilution have been taken into account. The main result we found is that, when the limits on the maximum tolerable divertor heat flux are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, i.e. no improvement would descend from a further increase of R or BT once the maximum has been reached. This occurrence appears as an intrinsic physical limit for all devices where a radiative SOL is needed to deal with the power exhaust. Furthermore, the relative importance of the different power loss channels (e.g. hydrogen radiation, charge exchange, perpendicular transport and impurity radiation), through which the power entering the SOL is dissipated before reaching the target plate, is investigated with our model.

Analysis of a multi-machine database on divertor heat fluxesa)

NASA Astrophysics Data System (ADS)

Makowski, M. A.; Elder, D.; Gray, T. K.; LaBombard, B.; Lasnier, C. J.; Leonard, A. W.; Maingi, R.; Osborne, T. H.; Stangeby, P. C.; Terry, J. L.; Watkins, J.

2012-05-01

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while a Spitzer resistivity model agrees at higher collisionality where it is more valid. The SOL profile measurements and divertor heat flux scaling are consistent with a heuristic drift based model as well as a critical gradient model.

Development of high poloidal beta, steady-state scenario with ITER-like tungsten divertor on EAST

DOE PAGES

Garofalo, Andrea M.; Gong, X. Z.; Qian, J.; ...

2017-06-07

Recent experiments on EAST have achieved the first long pulse H-mode (61 s) with zero loop voltage and an ITER-like tungsten divertor, and have demonstrated access to broad plasma current profiles by increasing the density in fully-noninductive lower hybrid current-driven discharges. These long pulse discharges reach wall thermal and particle balance, exhibit stationary good confinement (H 98y2~1.1) with low core electron transport, and are only possible with optimal active cooling of the tungsten armors. In separate experiments, the electron density was systematically varied in order to study its effect on the deposition profile of the external lower hybrid current drivemore » (LHCD), while keeping the plasma in fully-noninductive conditions and with divertor strike points on the tungsten divertor. A broadening of the current profile is found, as indicated by lower values of the internal inductance at higher density. A broad current profile is attractive because, among other reasons, it enables internal transport barriers at large minor radius, leading to improved confinement as shown in companion DIII-D experiments. These experiments strengthen the physics basis for achieving high performance, steady state discharges in future burning plasmas.« less

Effects of low and high mode number tearing modes in divertor tokamaks

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima; Boozer, Allen; Evans, Todd

2007-08-01

The topological effects of magnetic perturbations on a divertor tokamak, such as DIII-D, are studied using field-line maps that were developed by Punjabi et al. [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)]. The studies consider both long-wavelength perturbations, such as those of m =1, n =1 tearing modes, and localized perturbations, which are represented as a magnetic dipole. The parameters of the dipole map are set using DIII-D data from shot 115467 in which the C-coils were activated [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The long-wavelength perturbations alter the structure of the interception of magnetic field lines with the divertor plates, but the interception is in sharp lines. The dipole perturbations cause a spreading of the interception of the field lines with the divertor plates, which alleviates problems associated with heat deposition. Magnetic field lines are the trajectories of a one-and-a-half degree of freedom Hamiltonian, which strongly constrains the topological features of the lines. Although the field line maps that we use do not accurately represent the trajectories through ordinary space of individual field lines, they do represent their topological structure.

Preliminary Results on the Heat Deposition on Divertor Plate using Low MN Map

NASA Astrophysics Data System (ADS)

Ali, Halima; Punjabi, Alkesh; Boozer, Allen

2003-10-01

The study of magnetic field line behavior and the closely related plasma behavior are important not only for their tokamak application but also for their application to other Hamiltonian, or near-Hamiltonian, systems. The behavior of field lines near a tokamak separatrix has been studied extensively using various approaches. Our approach is called method of maps. In this paper, we introduce an area-preserving map called Low MN map. We first derive the map from the general theory of maps /1/, and then use it to calculate the effects of m = 1, n = 1 perturbations on the stochastic layer and magnetic footprint in single-null divertor tokamaks. We show that there are self-similarities, singularities, and topological equivalences in the pattern of physical parameters that characterize the stochastic layer and the magnetic footprint. Preliminary results in the investigation on the heat distribution on the divertor plate indicate multiple peaked in heat flux profile distributed radially across the divertor target when the amplitude is 10-3. This, and other features, are in good agreement with experimental observations. This work is done under the DOE grant number DE-FG02-01ER54624. 1. A. Punjabi et al, J. Plasma Phys. 52, 91 (1994).

Measurements of tungsten migration in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Wampler, W. R.; Rudakov, D. L.; Watkins, J. G.; McLean, A. G.; Unterberg, E. A.; Stangeby, P. C.

2017-12-01

An experimental study of migration of tungsten in the DIII-D divertor is described, in which the outer strike point of L-mode plasmas was positioned on a toroidal ring of tungsten-coated metal inserts. Net deposition of tungsten on the divertor just outside the strike point was measured on graphite samples exposed to various plasma durations using the divertor materials evaluation system. Tungsten coverage, measured by Rutherford backscattering spectroscopy (RBS), was found to be low and nearly independent of both radius and exposure time closer to the strike point, whereas farther from the strike point the W coverage was much larger and increased with exposure time. Depth profiles from RBS show this was due to accumulation of thicker mixed-material deposits farther from the strike point where the plasma temperature is lower. These results are consistent with a low near-surface steady-state coverage on graphite undergoing net erosion, and continuing accumulation in regions of net deposition. This experiment provides data needed to validate, and further improve computational simulations of erosion and deposition of material on plasma-facing components and transport of impurities in magnetic fusion devices. Such simulations are underway and will be reported later.

Predictive modelling of JT-60SA high-beta steady-state plasma with impurity accumulation

NASA Astrophysics Data System (ADS)

Hayashi, N.; Hoshino, K.; Honda, M.; Ide, S.

2018-06-01

The integrated modelling code TOPICS has been extended to include core impurity transport, and applied to predictive modelling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce the divertor heat load. In the modelling, models and conditions are selected for a conservative prediction, which considers a lower bound of plasma performance with the maximum accumulation of impurity. The conservative prediction shows the compatibility of impurity seeding with core plasma with high-beta (β N  >  3.5) and full current drive conditions, i.e. when Ar seeding reduces the divertor heat load below 10 MW m‑2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate for Ar radiation. Due to the strong dependence of accumulation on the pedestal density gradient, high separatrix density is important for the low accumulation as well as the low divertor heat load. The conservative prediction also shows that JT-60SA has enough capability to explore the divertor heat load control by impurity seeding in high-beta steady-state plasmas.

Extreme Ultraviolet Spectra of Few-Times Ionized Tungsten for Divertor Plasma Diagnostics

DOE PAGES

Clementson, Joel; Lennartsson, Thomas; Beiersdorfer, Peter

2015-09-09

The extreme ultraviolet (EUV) emission from few-times ionized tungsten atoms has been experimentally studied at the Livermore electron beam ion trap facility. The ions were produced and confined during low-energy operations of the EBIT-I electron beam ion trap. By varying the electron-beam energy from around 30–300 eV, tungsten ions in charge states expected to be abundant in tokamak divertor plasmas were excited, and the resulting EUV emission was studied using a survey spectrometer covering 120–320 Å. It is found that the emission strongly depends on the excitation energy; below 150 eV, it is relatively simple, consisting of strong isolated linesmore » from a few charge states, whereas at higher energies, it becomes very complex. For divertor plasmas with tungsten impurity ions, this emission should prove useful for diagnostics of tungsten flux rates and charge balance, as well as for radiative cooling of the divertor volume. Several lines in the 194–223 Å interval belonging to the spectra of five- and seven-times ionized tungsten (Tm-like W VI and Ho-like W VIII) were also measured using a high-resolution spectrometer.« less

The near infrared imaging system for the real-time protection of the JET ITER-like wall

NASA Astrophysics Data System (ADS)

Huber, A.; Kinna, D.; Huber, V.; Arnoux, G.; Balboa, I.; Balorin, C.; Carman, P.; Carvalho, P.; Collins, S.; Conway, N.; McCullen, P.; Jachmich, S.; Jouve, M.; Linsmeier, Ch; Lomanowski, B.; Lomas, P. J.; Lowry, C. G.; Maggi, C. F.; Matthews, G. F.; May-Smith, T.; Meigs, A.; Mertens, Ph; Nunes, I.; Price, M.; Puglia, P.; Riccardo, V.; Rimini, F. G.; Sergienko, G.; Tsalas, M.; Zastrow, K.-D.; contributors, JET

2017-12-01

This paper describes the design, implementation and operation of the near infrared (NIR) imaging diagnostic system of the JET ITER-like wall (JET-ILW) plasma experiment and its integration into the existing JET protection architecture. The imaging system comprises four wide-angle views, four tangential divertor views, and two top views of the divertor covering 66% of the first wall and up to 43% of the divertor. The operation temperature ranges which must be observed by the NIR protection cameras are, for the materials used on JET: Be 700 °C-1400 °C W coating 700 °C-1370 °C W bulk 700 °C-1400 °C. The Real-Time Protection system operates routinely since 2011 and successfully demonstrated its capability to avoid the overheating of the main chamber beryllium wall as well as of the divertor W and W-coated carbon fibre composite (CFC) tiles. During this period, less than 0.5% of the terminated discharges were aborted by a malfunction of the system. About 2%-3% of the discharges were terminated due to the detection of actual hot spots.

Predictions of VRF on a Langmuir Probe under the RF Heating Spiral on the Divertor Floor on NSTX-U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hosea, J C; Perkins, R J; Jaworski, M A

RF heating deposition spirals are observed on the divertor plates on NSTX as shown in for a NB plus RF heating case. It has been shown that the RF spiral is tracked quite well by the spiral mapping of the strike points on the divertor plate of magnetic field lines passing in front of the high harmonic fast wave (HHFW) antenna on NSTX. Indeed, both current instrumented tiles and Langmuir probes respond to the spiral when it is positioned over them. In particular, a positive increment in tile current (collection of electrons) is obtained when the spiral is over themore » tile. This current can be due to RF rectification and/or RF heating of the scrape off layer (SOL) plasma along the magnetic field lines passing in front of the the HHFW antenna. It is important to determine quantitatively the relative contributions of these processes. Here we explore the properties of the characteristics of probes on the lower divertor plate to determine the likelyhood that the primary cause of the RF heat deposition is RF rectification.« less

A study of X-divertor in NSTX-U with SOLPS simulations

NASA Astrophysics Data System (ADS)

Chen, Zhong-Ping; Kotschenreuther, Mike; Mahajan, Swadesh; Gerhardt, Stefan

2018-03-01

The X-divertor (XD) geometry in NSTX-U is demonstrated, via SOLPS simulations, to perform better than the standard divertor (SD); in particular, it allows detachment at a lower upstream density and stabilizes the detachment front near the target, away from the main X-point. Consequently a stable detached operation becomes possible—the localization near the plate allows a vast reduction of heat fluxes without degrading the core plasma. Indeed, it is confirmed by our simulation that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures, resulting in scrape-off layers that are more favorable for advanced tokamak operation. These advantages are attributed to the unique geometric characteristics of XD—poloidal flaring near the target.

Parameter dependences of the separatrix density in nitrogen seeded ASDEX Upgrade H-mode discharges

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Sun, H. J.; Eich, T.; Carralero, D.; Hobirk, J.; Scarabosio, A.; Siccinio, M.; ASDEX Upgrade Team; EUROfusion MST1 Team

2018-04-01

The upstream separatrix electron density is an important interface parameter for core performance and divertor power exhaust. It has been measured in ASDEX Upgrade H-mode discharges by means of Thomson scattering using a self-consistent estimate of the upstream electron temperature under the assumption of Spitzer-Härm electron conduction. Its dependence on various plasma parameters has been tested for different plasma conditions in H-mode. The leading parameter determining n e,sep was found to be the neutral divertor pressure, which can be considered as an engineering parameter since it is determined mainly by the gas puff rate and the pumping speed. The experimentally found parameter dependence of n e,sep, which is dominated by the divertor neutral pressure, could be approximately reconciled by 2-point modelling.

Thermal strain measurement of EAST tungsten divertor component with bare fiber Bragg grating sensors

NASA Astrophysics Data System (ADS)

Wang, Xingli; Wang, Wanjing; Wang, Jichao; Wei, Ran; Sun, Zhaoxuan; Li, Qiang; Xie, Chunyi; Luo, Guang-Nan

2017-12-01

Fiber Bragg Gratings (FBGs) have been widely used in the sensor field to monitor temperature and strain. However, the weak mechanical property of optical fibers and insufficient heat-resistant property of general optic-fiber sensors have prevented it from being widely used, such as in some extreme engineering situations. In this work, a bare FBG sensor system had been introduced to measure thermal strain of an Experimental Advanced Superconducting Tokamak tungsten divertor component under baking condition. This strain measurement system had withstood as high temperature as 210 °C and finished the measurement experiment successfully. Meaningful measurement results had been obtained and analyzed, which showed the applicability of such a bare fiber grating sensor system and as well contributed to studying on tungsten divertor's thermal strain conditions.

Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

NASA Astrophysics Data System (ADS)

Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

2009-12-01

Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

Overview of ASDEX Upgrade results

NASA Astrophysics Data System (ADS)

Stroth, U.; Adamek, J.; Aho-Mantila, L.; Äkäslompolo, S.; Amdor, C.; Angioni, C.; Balden, M.; Bardin, S.; Barrera Orte, L.; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Birkenmeier, G.; Bobkov, V.; Boom, J.; Bottereau, C.; Bottino, A.; Braun, F.; Brezinsek, S.; Brochard, T.; Brüdgam, M.; Buhler, A.; Burckhart, A.; Casson, F. J.; Chankin, A.; Chapman, I.; Clairet, F.; Classen, I. G. J.; Coenen, J. W.; Conway, G. D.; Coster, D. P.; Curran, D.; da Silva, F.; de Marné, P.; D'Inca, R.; Douai, D.; Drube, R.; Dunne, M.; Dux, R.; Eich, T.; Eixenberger, H.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fischer, R.; Fünfgelder, H.; Fuchs, J. C.; Gál, K.; García Muñoz, M.; Geiger, B.; Giannone, L.; Görler, T.; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Guimarais, L.; Günter, S.; Haas, G.; Hakola, A. H.; Hangan, D.; Happel, T.; Härtl, T.; Hauff, T.; Heinemann, B.; Herrmann, A.; Hobirk, J.; Höhnle, H.; Hölzl, M.; Hopf, C.; Houben, A.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Kantor, M.; Käsemann, C.-P.; Kallenbach, A.; Kálvin, S.; Kantor, M.; Kappatou, A.; Kardaun, O.; Kasparek, W.; Kaufmann, M.; Kirk, A.; Klingshirn, H.-J.; Kocan, M.; Kocsis, G.; Konz, C.; Koslowski, R.; Krieger, K.; Kubic, M.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lauber, P.; Laux, M.; Lazaros, A.; Leipold, F.; Leuterer, F.; Lindig, S.; Lisgo, S.; Lohs, A.; Lunt, T.; Maier, H.; Makkonen, T.; Mank, K.; Manso, M.-E.; Maraschek, M.; Mayer, M.; McCarthy, P. J.; McDermott, R.; Mehlmann, F.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Müller, S.; Müller, H. W.; Münich, M.; Neu, G.; Neu, R.; Neuwirth, D.; Nocente, M.; Nold, B.; Noterdaeme, J.-M.; Pautasso, G.; Pereverzev, G.; Plöckl, B.; Podoba, Y.; Pompon, F.; Poli, E.; Polozhiy, K.; Potzel, S.; Püschel, M. J.; Pütterich, T.; Rathgeber, S. K.; Raupp, G.; Reich, M.; Reimold, F.; Ribeiro, T.; Riedl, R.; Rohde, V.; Rooij, G. v.; Roth, J.; Rott, M.; Ryter, F.; Salewski, M.; Santos, J.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P. A.; Schneider, W.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Scott, B.; Sempf, M.; Sertoli, M.; Siccinio, M.; Sieglin, B.; Sigalov, A.; Silva, A.; Sommer, F.; Stäbler, A.; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Tala, T.; Tardini, G.; Teschke, M.; Tichmann, C.; Told, D.; Treutterer, W.; Tsalas, M.; Van Zeeland, M. A.; Varela, P.; Veres, G.; Vicente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Viola, B.; Vorpahl, C.; Wachowski, M.; Wagner, D.; Wauters, T.; Weller, A.; Wenninger, R.; Wieland, B.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; Würsching, E.; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

2013-10-01

The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron resonance heating (ECRH) power, by installing 2 × 8 internal magnetic perturbation coils, and by improving the ion cyclotron range of frequency compatibility with the tungsten wall. With the perturbation coils, reliable suppression of large type-I edge localized modes (ELMs) could be demonstrated in a wide operational window, which opens up above a critical plasma pedestal density. The pellet fuelling efficiency was observed to increase which gives access to H-mode discharges with peaked density profiles at line densities clearly exceeding the empirical Greenwald limit. Owing to the increased ECRH power of 4 MW, H-mode discharges could be studied in regimes with dominant electron heating and low plasma rotation velocities, i.e. under conditions particularly relevant for ITER. The ion-pressure gradient and the neoclassical radial electric field emerge as key parameters for the transition. Using the total simultaneously available heating power of 23 MW, high performance discharges have been carried out where feed-back controlled radiative cooling in the core and the divertor allowed the divertor peak power loads to be maintained below 5 MW m-2. Under attached divertor conditions, a multi-device scaling expression for the power-decay length was obtained which is independent of major radius and decreases with magnetic field resulting in a decay length of 1 mm for ITER. At higher densities and under partially detached conditions, however, a broadening of the decay length is observed. In discharges with density ramps up to the density limit, the divertor plasma shows a complex behaviour with a localized high-density region in the inner divertor before the outer divertor detaches. Turbulent transport is studied in the core and the scrape-off layer (SOL). Discharges over a wide parameter range exhibit a close link between core momentum and density transport. Consistent with gyro-kinetic calculations, the density gradient at half plasma radius determines the momentum transport through residual stress and thus the central toroidal rotation. In the SOL a close comparison of probe data with a gyro-fluid code showed excellent agreement and points to the dominance of drift waves. Intermittent structures from ELMs and from turbulence are shown to have high ion temperatures even at large distances outside the separatrix.

First NATO/SIBCA Exercise in Sampling of Chemical Warfare Agents (Eerste NATO/SIBCA-oefening in monstername van chemische strijdmiddelen).

DTIC Science & Technology

1998-01-01

TNO-rapport PML1997-A81 Eerste NATO/SIBCA-oefening in monster- name van chemische strijdmiddelen TNO Prins Maurits Laboratorium Lange Kleiweg...27 Aantal pagina’s : 53 (incl. bijlagen, excl. RDP & distributielijst) Aantal bijlagen : 4 TNO Prins Maurits Laboratorium is onderdeel...van de hoofdgroep TNO Defensieonderzoek waartoe verder behoren: TNO Fysisch en Elektronisch Laboratorium TNO Technische Menskunde Nederlandse

A Metagenomic Analysis of Microbial Contamination in Aviation Fuels

DTIC Science & Technology

2009-03-01

classification by the RDP Classifier, sequences similar to members of the Acidobacteria, Actinobacteria , Bacteroidetes, Chloroflexi, Cyanobacteria... Actinobacteria 85 63 4 152 Bacteroidetes 5 0 0 5 Chloroflexi 7 0 0 7 Cyanobacteria 56 0 0 56 Deinococcus-Thermus 2 0 0 2 Firmicutes 83 99 2 184...Members of the Proteobacteria, Firmicutes and Actinobacteria were represented in all three fuel types; in Jet A and Biodiesel they were the only

Evidence for basal distortion-product otoacoustic emission components.

PubMed

Martin, Glen K; Stagner, Barden B; Lonsbury-Martin, Brenda L

2010-05-01

Distortion-product otoacoustic emissions (DPOAEs) were measured with traditional DP-grams and level/phase (L/P) maps in rabbits with either normal cochlear function or unique sound-induced cochlear losses that were characterized as either low-frequency or notched configurations. To demonstrate that emission generators distributed basal to the f(2) primary-tone contribute, in general, to DPOAE levels and phases, a high-frequency interference tone (IT) was presented at 1/3 of an octave (oct) above the f(2) primary-tone, and DPOAEs were re-measured as "augmented" DP-grams (ADP-grams) and L/P maps. The vector difference between the control and augmented functions was then computed to derive residual DP-grams (RDP-grams) and L/P maps. The resulting RDP-grams and L/P maps, which described the DPOAEs removed by the IT, supported the notion that basal DPOAE components routinely contribute to the generation of standard measures of DPOAEs. Separate experiments demonstrated that these components could not be attributed to the effects of the 1/3-oct IT on f(2), or DPOAEs generated by the addition of a third interfering tone. These basal components can "fill in" the lesion estimated by the commonly employed DP-gram. Thus, ADP-grams more accurately reveal the pattern of cochlear damage and may eventually lead to an improved DP-gram procedure.

Team-based learning in a preclinical removable denture prosthesis module in a United Arab Emirates dental school.

PubMed

Haj-Ali, Reem; Al Quran, Firas

2013-03-01

The purpose of this article is to describe the implementation of a team-based learning (TBL) approach in a removable denture prosthesis (RDP) module and present the results of students' performance in individual and group TBL activities and exam scores, students' experience with TBL and end of course evaluations, and faculty feedback. Course material at the College of Dentistry, University of Sharjah, United Arab Emirates, was transformed into seven conventional lectures and seven TBL sessions. Each TBL session consisted of pre-assigned reading (self-directed learning), in-class individual and group readiness tests (accountability), team problem-solving of patient RDP cases, and faculty-led class discussion (knowledge application). The course was assessed through scores from TBL session activities and course examinations, student satisfaction survey, and faculty feedback. Course grades were found to be higher using the TBL method then the traditional lecture-based method. Student evaluation data and faculty response indicated strong support for TBL as it was implemented in the course. The faculty noted a higher level of student engagement with team learning than in conventional class lecturing. TBL is an active-learning instructional strategy for courses with high student-to-faculty ratios. This approach provides regular feedback and the opportunity for students to develop higher reasoning skills.

Lung Cancer Mortality (1950–1999) among Eldorado Uranium Workers: A Comparison of Models of Carcinogenesis and Empirical Excess Risk Models

PubMed Central

Eidemüller, Markus; Jacob, Peter; Lane, Rachel S. D.; Frost, Stanley E.; Zablotska, Lydia B.

2012-01-01

Lung cancer mortality after exposure to radon decay products (RDP) among 16,236 male Eldorado uranium workers was analyzed. Male workers from the Beaverlodge and Port Radium uranium mines and the Port Hope radium and uranium refinery and processing facility who were first employed between 1932 and 1980 were followed up from 1950 to 1999. A total of 618 lung cancer deaths were observed. The analysis compared the results of the biologically-based two-stage clonal expansion (TSCE) model to the empirical excess risk model. The spontaneous clonal expansion rate of pre-malignant cells was reduced at older ages under the assumptions of the TSCE model. Exposure to RDP was associated with increase in the clonal expansion rate during exposure but not afterwards. The increase was stronger for lower exposure rates. A radiation-induced bystander effect could be a possible explanation for such an exposure response. Results on excess risks were compared to a linear dose-response parametric excess risk model with attained age, time since exposure and dose rate as effect modifiers. In all models the excess relative risk decreased with increasing attained age, increasing time since exposure and increasing exposure rate. Large model uncertainties were found in particular for small exposure rates. PMID:22936975

Particle simulations on transport control in divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kashiwagi, Mieko; Ido, Shunji

1995-04-01

Particle orbit simulations are carried out to study the reflection of He ions recycled from a tokamak divertor by RF electric fields, which have the frequency close to ion cyclotron resonance frequency (ICRF). The performance of particle reflection and the requirement to the intensity of RF fields are studied. The control of He recycling by ICRF fields is found to be available. 4 refs., 4 figs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono; Jaworski, M.; Kaita, R.

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

ADX: a high field, high power density, advanced divertor and RF tokamak

NASA Astrophysics Data System (ADS)

LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

2015-05-01

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv:1409.3540)) that makes use of high-temperature superconductor technology—a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

Experiments and numerical modeling of fast flowing liquid metal thin films under spatially varying magnetic field conditions

NASA Astrophysics Data System (ADS)

Narula, Manmeet Singh

Innovative concepts using fast flowing thin films of liquid metals (like lithium) have been proposed for the protection of the divertor surface in magnetic fusion devices. However, concerns exist about the possibility of establishing the required flow of liquid metal thin films because of the presence of strong magnetic fields which can cause flow disrupting MHD effects. A plan is underway to design liquid lithium based divertor protection concepts for NSTX, a small spherical torus experiment at Princeton. Of these, a promising concept is the use of modularized fast flowing liquid lithium film zones, as the divertor (called the NSTX liquid surface module concept or NSTX LSM). The dynamic response of the liquid metal film flow in a spatially varying magnetic field configuration is still unknown and it is suspected that some unpredicted effects might be lurking. The primary goal of the research work being reported in this dissertation is to provide qualitative and quantitative information on the liquid metal film flow dynamics under spatially varying magnetic field conditions, typical of the divertor region of a magnetic fusion device. The liquid metal film flow dynamics have been studied through a synergic experimental and numerical modeling effort. The Magneto Thermofluid Omnibus Research (MTOR) facility at UCLA has been used to design several experiments to study the MHD interaction of liquid gallium films under a scaled NSTX outboard divertor magnetic field environment. A 3D multi-material, free surface MHD modeling capability is under development in collaboration with HyPerComp Inc., an SBIR vendor. This numerical code called HIMAG provides a unique capability to model the equations of incompressible MHD with a free surface. Some parts of this modeling capability have been developed in this research work, in the form of subroutines for HIMAG. Extensive code debugging and benchmarking exercise has also been carried out. Finally, HIMAG has been used to study the MHD interaction of fast flowing liquid metal films under various divertor relevant magnetic field configurations through numerical modeling exercises.

Experimental and analytical studies of high heat flux components for fusion experimental reactor

NASA Astrophysics Data System (ADS)

Araki, Masanori

1993-03-01

In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 (+/-) 1 MW/sq m was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate was analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads.

Development and application of W/Cu flat-type plasma facing components at ASIPP

NASA Astrophysics Data System (ADS)

Li, Q.; Zhao, S. X.; Sun, Z. X.; Xu, Y.; Li, B.; Wei, R.; Wang, W. J.; Qin, S. G.; Shi, Y. L.; Xie, C. Y.; Wang, J. C.; Wang, X. L.; Missirlian, M.; Guilhem, D.; Liu, G. H.; Yang, Z. S.; Luo, G.-N.

2017-12-01

W/Cu flat-type plasma facing components (PFCs) were widely used in divertor of fusion device because of its advantages, such as low cost, light in weight and good machinability. However, it is very difficult to manufacture them due to the large mismatch between the thermo-mechanical properties of W and Cu. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) has successfully developed W/Cu flat-type PFCs for EAST W/Cu divertor project by hot isostatic pressing (HIP) technology. This paper presents the development and application of W/Cu flat-type PFCs at ASIPP. The optimized manufacturing process is to cast pure copper onto the rear side of W tiles at temperature of 1200 °C firstly, and then to HIP the W/Cu tiles onto CuCrZr heat sink at temperature of 600 °C, pressure of 150 MPa and duration of 3 h. W/Cu flat-type testing mock-up for EAST survived 1000 cycles at heat load of 5 MW m-2 in high heat flux tests. And then ASIPP prepared two mock-ups for CEA’s tungsten environment in steady-state tokamak (WEST) project. One mock-up withstood successfully 302 cycles of 20 MW m-2, which are far beyond the design requirement. Since 2014, W/Cu flat-type PFCs were wildly used in EAST upper divertor as baffle and dome components which showed excellent performance in 2015 and 2016 campaigns. Given the success in EAST upper divertor, W/Cu flat-type concept is as well applied in the design of actively cooled Langmuir probes which will be mounted onto EAST divertor targets soon.

Integration of uncooled scraper elements and its diagnostics into Wendelstein 7-X

DOE PAGES

Fellinger, Joris; Loesser, Doug; Neilson, Hutch; ...

2017-08-08

The modular stellarator Wendelstein 7-X in Greifswald (Germany) successfully started operation in 2015 with short pulse limiter plasmas. In 2017, the next operation phase (OP) OP1.2 will start once 10 uncooled test divertor units (TDU) with graphite armor will be installed. The TDUs allow for plasma pulses of 10 s with 8 MW heating. OP2, allowing for steady state operation, is planned for 2020 after the TDUs will be replaced by 10 water cooled CFC armored divertors. Due to the development of plasma currents like bootstrap currents in long pulse plasmas in OP2, the plasma could hit the edge ofmore » the divertor targets which has a reduced cooling capacity compared to the central part of the target tiles. To prevent overloading of these edges, a so-called scraper element can be positioned in front of the divertor, intersecting those strike lines that would otherwise hit the divertor edges. As a result, these edges are protected but as a drawback the pumping efficiency of neutrals is also reduced. As a test an uncooled scraper element with graphite tiles will be placed in two out of ten half modules in OP1.2. A decision to install ten water cooled scraper elements for OP2 is pending on the results of this test in OP1.2. To monitor the impact of the scraper element on the plasma, Langmuir probes are integrated in the plasma facing surface, and a neutral gas manometer measures the neutral density directly behind the plasma facing surface. Moreover, IR and VIS cameras observe the plasma facing surface and thermocouples monitor the temperatures of the graphite tiles and underlying support structure. This paper describes the integration of the scraper element and its diagnostics in Wendelstein 7-X.« less

Quantification of Chemical Erosion in the DIII-D Divertor

NASA Astrophysics Data System (ADS)

McLean, Adam

2009-11-01

Chemical erosion (CE) yield at the graphite divertor target in DIII-D was measured to be substantially lower in cold near-detached plasma conditions compared to well-attached ones, with major implications for ITER. Current estimates of tritium retention by co-deposition with hydrocarbons (HCs) in ITER place potentially severe restrictions on operation. However, calculations done to date have been based on excessively conservative assumptions, due to limited understanding of cold divertor plasmas (1-5eV) which bridge energy thresholds for complex atomic and molecular processes not present in attached conditions. Hydrocarbon injection through a unique porous graphite plate which realistically simulates secondary reactions of HCs with a graphite surface has been used to measure CE in-situ. For the first time in a divertor, measurements were made at extrinsic CH4 injection rates comparable to the expected intrinsic CE rate of C, with the resulting spectroscopic emissions separated from those of the intrinsic sources. Under cold plasma conditions the contribution of CE-produced C relative to total C sources in the divertor declined dramatically from ˜50% to <15%. Photon efficiencies for products from the breakup of injected CH4 were greater than previous measurements at higher puff rates, indicating the importance of minimizing perturbation to the local plasma. At 350K, the measured CE yield near the outer strike point was ˜2.6% in attachment dropping to only ˜0.5% in cold plasma; results are consistent with some theoretical predications and lab studies. Under full detachment, near total extinction of the CD band occurred, consistent with suppression of net C erosion. These findings have potentially major impact on projected target lifetime and tritium retention in future reactors, and for the PFC choice in ITER.

SOLPS simulations of X-divertor in NSTX-U

NASA Astrophysics Data System (ADS)

Chen, Zhongping; Kotschenreuther, Mike; Mahajan, Swadesh

2017-10-01

The X-divertor (XD) geometry in NSTX-U has demonstrated, in SOLPS simulations, a better performance than the standard divertor (SD) regarding detachment: achieving detachment with a lower upstream density and stabilizing the detachment front near the target. The benefits of such a localized front is that the power exhaust requirement can be satisfied without the radiation front encroaching on the core plasma. It is also found by our simulations that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures. These advantages are attributed to the unique geometric characteristics of XD - poloidal flaring near the target. The detailed physical mechanisms behind the better XD performance that is found in the simulations will be examined. Work supported by US DOE under DE-FG02-04ER54742 and SC 0012956.

Recent sheath physics studies on DIII-D

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Labombard, B.; Stangeby, P. C.; Lasnier, C. J.; McLean, A. G.; Nygren, R. E.; Boedo, J. A.; Leonard, A. W.; Rudakov, D. L.

2015-08-01

A study to examine some current issues in the physics of the plasma sheath has been recently carried out in DIII-D low power Ohmic plasmas using both flush and domed Langmuir probes, divertor Thomson scattering (DTS), an infrared camera (IRTV), and a new calorimeter triple probe assembly mounted on the Divertor Materials Evaluation System (DIMES). The sheath power transmission factor was found to be consistent with the theoretically predicted value of 7 (±2) for low power plasmas. Using this factor, the three heat flux profiles derived from the LP, DTS, and calorimeter diagnostic measurements agree. Comparison of flush and domed Langmuir probes and divertor Thomson scattering indicates that proper interpretation of flush probe data to get target plate density and temperature is feasible and could potentially yield accurate measurements of target plate conditions where the probes are located.

Upgrade of the infrared camera diagnostics for the JET ITER-like wall divertor.

PubMed

Balboa, I; Arnoux, G; Eich, T; Sieglin, B; Devaux, S; Zeidner, W; Morlock, C; Kruezi, U; Sergienko, G; Kinna, D; Thomas, P D; Rack, M

2012-10-01

For the new ITER-like wall at JET, two new infrared diagnostics (KL9B, KL3B) have been installed. These diagnostics can operate between 3.5 and 5 μm and up to sampling frequencies of ∼20 kHz. KL9B and KL3B image the horizontal and vertical tiles of the divertor. The divertor tiles are tungsten coated carbon fiber composite except the central tile which is bulk tungsten and consists of lamella segments. The thermal emission between lamellae affects the surface temperature measurement and therefore KL9A has been upgraded to achieve a higher spatial resolution (by a factor of 2). A technical description of KL9A, KL9B, and KL3B and cross correlation with a near infrared camera and a two-color pyrometer is presented.

Analysis of the plasma-wall interaction in the Heliotron E device

NASA Astrophysics Data System (ADS)

Motojima, O.; Mizuuchi, T.; Besshou, S.; Iiyoshi, A.; Uo, K.; Yamashina, T.; Mohri, M.; Satake, T.; Hashiba, M.; Amemiya, S.; Miwa, H.

1984-12-01

The plasma-wall interaction (PWI) of the currentless plasmas with temperature To, Tio ≤ 1.1 keV, density N¯e = (2-10)× 1013/cm3, and volume-averaged beta value of β$¯≤ 2% was investigated. We have observed that PWI took place mainly where the divertor field line intersected the chamber wall (called divertor traces). Boundary plasmas were measured with electrostatic probes, which showed the presence of the divertor region with the parameters in the range of Ned = 1010-1011/cm3 and Ted = 10-50 eV. Surface analysis techniques (ESCA, AES, and RBS) were applied to analyze the surface probes (Si, graphite and stainless steel) and the test pieces (SiC, TiC, and stainless steel), which were irradiated by plasmas for short and long times respectively.

Comparison of initial perturbation methods for the mesoscale ensemble prediction system of the Meteorological Research Institute for the WWRP Beijing 2008 Olympics Research and Development Project (B08RDP)

NASA Astrophysics Data System (ADS)

Saito, Kazuo; Hara, Masahiro; Kunii, Masaru; Seko, Hiromu; Yamaguchi, Munehiko

2011-05-01

Different initial perturbation methods for the mesoscale ensemble prediction were compared by the Meteorological Research Institute (MRI) as a part of the intercomparison of mesoscale ensemble prediction systems (EPSs) of the World Weather Research Programme (WWRP) Beijing 2008 Olympics Research and Development Project (B08RDP). Five initial perturbation methods for mesoscale ensemble prediction were developed for B08RDP and compared at MRI: (1) a downscaling method of the Japan Meteorological Agency (JMA)'s operational one-week EPS (WEP), (2) a targeted global model singular vector (GSV) method, (3) a mesoscale model singular vector (MSV) method based on the adjoint model of the JMA non-hydrostatic model (NHM), (4) a mesoscale breeding growing mode (MBD) method based on the NHM forecast and (5) a local ensemble transform (LET) method based on the local ensemble transform Kalman filter (LETKF) using NHM. These perturbation methods were applied to the preliminary experiments of the B08RDP Tier-1 mesoscale ensemble prediction with a horizontal resolution of 15 km. To make the comparison easier, the same horizontal resolution (40 km) was employed for the three mesoscale model-based initial perturbation methods (MSV, MBD and LET). The GSV method completely outperformed the WEP method, confirming the advantage of targeting in mesoscale EPS. The GSV method generally performed well with regard to root mean square errors of the ensemble mean, large growth rates of ensemble spreads throughout the 36-h forecast period, and high detection rates and high Brier skill scores (BSSs) for weak rains. On the other hand, the mesoscale model-based initial perturbation methods showed good detection rates and BSSs for intense rains. The MSV method showed a rapid growth in the ensemble spread of precipitation up to a forecast time of 6 h, which suggests suitability of the mesoscale SV for short-range EPSs, but the initial large growth of the perturbation did not last long. The performance of the MBD method was good for ensemble prediction of intense rain with a relatively small computing cost. The LET method showed similar characteristics to the MBD method, but the spread and growth rate were slightly smaller and the relative operating characteristic area skill score and BSS did not surpass those of MBD. These characteristic features of the five methods were confirmed by checking the evolution of the total energy norms and their growth rates. Characteristics of the initial perturbations obtained by four methods (GSV, MSV, MBD and LET) were examined for the case of a synoptic low-pressure system passing over eastern China. With GSV and MSV, the regions of large spread were near the low-pressure system, but with MSV, the distribution was more concentrated on the mesoscale disturbance. On the other hand, large-spread areas were observed southwest of the disturbance in MBD and LET. The horizontal pattern of LET perturbation was similar to that of MBD, but the amplitude of the LET perturbation reflected the observation density.

Studies of Be migration in the JET tokamak using AMS with 10Be marker

NASA Astrophysics Data System (ADS)

Bykov, I.; Bergsåker, H.; Possnert, G.; Zhou, Y.; Heinola, K.; Pettersson, J.; Conroy, S.; Likonen, J.; Petersson, P.; Widdowson, A.

2016-03-01

The JET tokamak is operated with beryllium limiter tiles in the main chamber and tungsten coated carbon fiber composite tiles and solid W tiles in the divertor. One important issue is how wall materials are migrating during plasma operation. To study beryllium redistribution in the main chamber and in the divertor, a 10Be enriched limiter tile was installed prior to plasma operations in 2011-2012. Methods to take surface samples have been developed, an abrasive method for bulk Be tiles in the main chamber, which permits reuse of the tiles, and leaching with hot HCl to remove all Be deposited at W coated surfaces in the divertor. Quantitative analysis of the total amount of Be in cm2 sized samples was made with inductively coupled plasma atomic emission spectroscopy (ICP-AES). The 10Be/9Be ratio in the samples was measured with accelerator mass spectrometry (AMS). The experimental setup and methods are described in detail, including sample preparation, measures to eliminate contributions in AMS from the 10B isobar, possible activation due to plasma generated neutrons and effects of diffusive isotope mixing. For the first time marker concentrations are measured in the divertor deposits. They are in the range 0.4-1.2% of the source concentration, with moderate poloidal variation.

In-Vessel Tritium Retention and Removal in ITER-FEAT

NASA Astrophysics Data System (ADS)

Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

Particle and Power Exhaust in EAST

NASA Astrophysics Data System (ADS)

Wang, Liang; Ding, Fang; Yu, Yaowei; Gan, Kaifu; Liang, Yunfeng; Xu, Guosheng; Xiao, Bingjia; Sun, Youwen; Luo, Guangnan; Gong, Xianzu; Hu, Jiansheng; Li, Jiangang; Wan, Baonian; Maingi, Rajesh; Guo, Houyang; Garofalo, Andrea; EAST Team

2017-10-01

A total power injection up to 0.3GJ has been achieved in EAST long pulse USN operation with ITER-like water-cooling W-monoblock divertor, which has steady-state power exhaust capability of 10 MWm-2. The peak temperature of W target saturated at t = 12 s to the value T 500oC and a heat flux 3MWm-2was maintained. Great efforts to reduce heat flux and accommodate particle exhaust simultaneously have been made towards long pulse of 102s time scale. By exploiting the observation of Pfirsch-Schlüter flow direction in the SOL, the Bt direction with Bx ∇B away from the W divertor (more particles favor outer target in USN) was adopted along with optimizing the strike point location near the pumping slot, to facilitate particle and impurity exhaust with the top cryo-pump. By tailoring the 3D divertor footprint through edge magnetic topology change, the heat load was dispersed widely and thus peak heat flux and W sputtering was well controlled. Active feedback control of total radiative power with neon seeding was achieved within frad = 17-35%, exhibiting further potential for heat flux reduction with divertor and edge radiation. Other heat flux handling techniques, including quasi snowflake configuration, will also be presented.

High heat flux Langmuir probe array for the DIII-D divertor platesa)

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Taussig, D.; Boivin, R. L.; Mahdavi, M. A.; Nygren, R. E.

2008-10-01

Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m2 for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5° surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of Jsat, Te, and Vf with 4 mm spatial resolution are shown.

2D imaging of helium ion velocity in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Samuell, C. M.; Porter, G. D.; Meyer, W. H.; Rognlien, T. D.; Allen, S. L.; Briesemeister, A.; Mclean, A. G.; Zeng, L.; Jaervinen, A. E.; Howard, J.

2018-05-01

Two-dimensional imaging of parallel ion velocities is compared to fluid modeling simulations to understand the role of ions in determining divertor conditions and benchmark the UEDGE fluid modeling code. Pure helium discharges are used so that spectroscopic He+ measurements represent the main-ion population at small electron temperatures. Electron temperatures and densities in the divertor match simulated values to within about 20%-30%, establishing the experiment/model match as being at least as good as those normally obtained in the more regularly simulated deuterium plasmas. He+ brightness (HeII) comparison indicates that the degree of detachment is captured well by UEDGE, principally due to the inclusion of E ×B drifts. Tomographically inverted Coherence Imaging Spectroscopy measurements are used to determine the He+ parallel velocities which display excellent agreement between the model and the experiment near the divertor target where He+ is predicted to be the main-ion species and where electron-dominated physics dictates the parallel momentum balance. Upstream near the X-point where He+ is a minority species and ion-dominated physics plays a more important role, there is an underestimation of the flow velocity magnitude by a factor of 2-3. These results indicate that more effort is required to be able to correctly predict ion momentum in these challenging regimes.

Symmetric Simple Map with Dipole Map for a Single-Null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Ali, Halima; Watson, Michael; Punjabi, Alkesh; Boozer, Allen

1996-11-01

This investigation focuses on the effects of an externally placed dipole coil on the magnetic topology of a single-null divertor tokamak with a stochastic scrape-off layer using the Method of Maps (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The unperturbed magnetic topology is represented by the Symmetric Simple Map (Ali H, Watson M, Mayer C, Punjabi A and Boozer A, Bull Am Phys Soc), 40, 1855 (1995). The effect of dipole perturbation is repesented by the Dipole Map (Ali H, Watson M, Punjabi A and Boozer A, Sherwood Mtg), paper 1C20 (1996). A single dipole coil is placed across from the X-point below the last good surface. The strength of the dipole perturbation and the distance of the coil from the last good surface are varied. We observe that the dipole perturbation causes spatially intermittent chaos. This has significant implications for radiative divertor concepts as well for impurity control. We also present the detailed results on the effects of the dipole coil on the properties of the stochastic layer and the footprint of the field lines on the divertor plate. This work is supported by the US DOE OFES.

Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Bolt. H.; Breitbach, G.

1994-12-31

To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors such as ITER, neutron irradiation and subsequent high heat flux tests will be most essential. The effect of neutron damage will be simulated in material test reactors (such as the HFR-Petten) in a fission neutron environment. To investigate the heat loads during normal and off-normal operation scenarios a 60 kW electron beam test stand (Juelich Divertor Test Facility in Hot Cells, JUDITH) has been installed in a hot cell which can be operated by remote handling techniques. In this facilitymore » inertially cooled test coupons can be handled as well as small actively cooled divertor mock-ups. A special clamping mechanism for small test coupons (25 mm x 25 mm x 35 mm) with an integrated coolant channel within a copper or TZM heat sink has been developed and tested in an electron beam test bed. This method is an attractive alternative to costly large scale tests on complete divertor modules. The temperature and stress fields in individual CFC or beryllium tiles brazed to metallic heat sink (e.g. copper or TZM) can be investigated before and after neutron irradiation with moderate efforts.« less

Mitigation of divertor heat loads by strike point sweeping in high power JET discharges

NASA Astrophysics Data System (ADS)

Silburn, S. A.; Matthews, G. F.; Challis, C. D.; Frigione, D.; Graves, J. P.; Mantsinen, M. J.; Belonohy, E.; Hobirk, J.; Iglesias, D.; Keeling, D. L.; King, D.; Kirov, K.; Lennholm, M.; Lomas, P. J.; Moradi, S.; Sips, A. C. C.; Tsalas, M.; Contributors, JET

2017-12-01

Deliberate periodic movement (sweeping) of the high heat flux divertor strike lines in tokamak plasmas can be used to manage the heat fluxes experienced by exhaust handling plasma facing components, by spreading the heat loads over a larger surface area. Sweeping has recently been adopted as a routine part of the main high performance plasma configurations used on JET, and has enabled pulses with 30 MW plasma heating power and 10 MW radiation to run for 5 s without overheating the divertor tiles. We present analysis of the effectiveness of sweeping for divertor temperature control on JET, using infrared camera data and comparison with a simple 2D heat diffusion model. Around 50% reduction in tile temperature rise is obtained with 5.4 cm sweeping compared to the un-swept case, and the temperature reduction is found to scale slower than linearly with sweeping amplitude in both experiments and modelling. Compatibility of sweeping with high fusion performance is demonstrated, and effects of sweeping on the edge-localised mode behaviour of the plasma are reported and discussed. The prospects of using sweeping in future JET experiments with up to 40 MW heating power are investigated using a model validated against existing experimental data.

Design requirements for plasma facing materials in ITER

NASA Astrophysics Data System (ADS)

Matera, R.; Federici, G.; ITER Joint Central Team

1996-10-01

After the official approval of the Interim Design Report, the ITER project enters the final phase of the EDA. With the definition of the design requirements of the high heat flux components, the structural and armor materials' working domain is better specified, allowing to focus the R & D program on the most critical issues and to orient the design of divertor and first wall components towards those concepts which potentially have a better chance to withstand normal and off-normal operating conditions. Among the latter, slow, high-power, high recycling transient are at present driving the design of high heat flux components. Examples of possible design solution under experimental validation in the R & D program are presented and discussed in this paper.

"ATP1A3" Mutations in Infants: A New Rapid-Onset Dystonia-Parkinsonism Phenotype Characterized by Motor Delay and Ataxia

ERIC Educational Resources Information Center

Brashear, Allison; Mink, Jonathan W.; Hill, Deborah F.; Boggs, Niki; McCall, W. Vaughn; Stacy, Mark A.; Snively, Beverly; Light, Laney S.; Sweadner, Kathleen J.; Ozelius, Laurie J.; Morrison, Leslie

2012-01-01

We report new clinical features of delayed motor development, hypotonia, and ataxia in two young children with mutations (R756H and D923N) in the "ATP1A3" gene. In adults, mutations in "ATP1A3" cause rapid-onset dystonia-Parkinsonism (RDP, DYT12) with abrupt onset of fixed dystonia. The parents and children were examined and videotaped, and…

Progress on the application of ELM control schemes to ITER scenarios from the non-active phase to DT operation

NASA Astrophysics Data System (ADS)

Loarte, A.; Huijsmans, G.; Futatani, S.; Baylor, L. R.; Evans, T. E.; Orlov, D. M.; Schmitz, O.; Becoulet, M.; Cahyna, P.; Gribov, Y.; Kavin, A.; Sashala Naik, A.; Campbell, D. J.; Casper, T.; Daly, E.; Frerichs, H.; Kischner, A.; Laengner, R.; Lisgo, S.; Pitts, R. A.; Saibene, G.; Wingen, A.

2014-03-01

Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ˜2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1 Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5-10 MA depending on modelling assumptions.

Effect of ELMs on deuterium-loaded-tungsten plasma facing components

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Rudakov, D. L.; Wampler, W.; Watkins, J. G.; Wong, C. P. C.

2011-08-01

Prior heat pulse testing of plasma facing components (PFCs) has been completed in vacuum environments without the presence of background plasma. Edge localized modes (ELMs) will not be this kind of isolated event and one should know the effect of a plasma background during these transients. Heat-pulse experiments have been conducted in the PISCES-A device utilizing laser heating in a divertor-like plasma background. Initial results indicate that the erosion of PFCs is enhanced as compared to heat pulse or plasma only tests. To determine if the enhanced erosion effect is a phenomena only witnessed in the laboratory PISCES device, tungsten and graphite samples were exposed to plasmas in the lower divertor of the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Mass loss analysis indicates that materials that contain significant deuterium prior to experiencing a transient heating event will erode faster than those that have no or little retained deuterium.

Neutral pressure behavior for diverted discharges in the Wendelstein 7-AS Stellarator

NASA Astrophysics Data System (ADS)

McCormick, K.; Grigull, P.; Burhenn, R.; Ehmler, H.; Feng, Y.; Giannone, L.; Haas, G.; Sardei, F.; NBI-, ECRH-; W7-AS Teams

2005-03-01

On the W7-AS stellarator, the subdivertor neutral pressure in an up-down divertor pair as well as at two points in the vicinity of a lower divertor module in the main chamber are measured. Results are presented for ι=5/9 island divertor discharges under conditions of normal confinement (NC) and the HDH-mode for: n˜0.1-4×1020 m-3, Pecrh = 0.5-1.5 MW, Pnbi = 2 MW, and H + and D + plasmas, with both normal- and reversed- Bt for H +. Subdivertor pressures are in the range 1-2 × 10 -3 mbar for HDH conditions. For plasma detachment at the target plates a strong up-down pressure asymmetry arises, with pup/ pdown ⩽ 5. The asymmetry reverses with reversed Bt. Main vessel pressures are a factor of 5-10 lower than the average subdivertor pressure for H +, with D + plasmas exhibiting still lower values.

Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Chung, H.M.; Smith, D.L.

1997-04-01

Test specimens made with the 832665 heat of V-4Cr-4Ti alloy were exposed in the DIII-D tokamak environment to support the installation of components made of a V-4Cr-4Ti alloy in the radiative divertor of the DIII-D. Some of the tests were conducted with the Divertor Materials Evaluation System (DiMES) to study the short-term effects of postvent bakeout, when concentrations of gaseous impurities in the DIII-D chamber are the highest. Other specimens were mounted next to the chamber wall behind the divertor baffle plate, to study the effects of longer-term exposures. By design, none of the specimens directly interacted with the plasma.more » Preliminary results from testing the exposed specimens indicate only minor degradation of mechanical properties. Additional testing and microstructural characterization are in progress.« less

Design feasibility study of a divertor component reinforced with fibrous metal matrix composite laminate

NASA Astrophysics Data System (ADS)

You, Jeong-Ha

2005-01-01

Fibrous metal matrix composites possess advanced mechanical properties compared to conventional alloys. It is expected that the application of these composites to a divertor component will enhance the structural reliability. A possible design concept would be a system consisting of tungsten armour, copper composite interlayer and copper heat sink where the composite interlayer is locally inserted into the highly stressed domain near the bond interface. For assessment of the design feasibility of the composite divertor concept, a non-linear multi-scale finite element analysis was performed. To this end, a micro-mechanics algorithm was implemented into a finite element code. A reactor-relevant heat flux load was assumed. Focus was placed on the evolution of stress state, plastic deformation and ductile damage on both macro- and microscopic scales. The structural response of the component and the micro-scale stress evolution of the composite laminate were investigated.

Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

NASA Astrophysics Data System (ADS)

Bazylev, B. N.; Janeschitz, G.; Landman, I. S.; Loarte, A.; Pestchanyi, S. E.

2007-06-01

Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

An automated approach to magnetic divertor configuration design

NASA Astrophysics Data System (ADS)

Blommaert, M.; Dekeyser, W.; Baelmans, M.; Gauger, N. R.; Reiter, D.

2015-01-01

Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrative purposes is to spread the divertor target heat load as much as possible over the entire target area. Constraints on the separatrix position are introduced to eliminate physically irrelevant magnetic field configurations during the optimization cycle. A gradient projection method is used to ensure stable cost function evaluations during optimization. The concept is applied to a configuration with typical Joint European Torus (JET) parameters and it automatically provides plausible configurations with reduced heat load.

Long-term fuel retention in JET ITER-like wall

NASA Astrophysics Data System (ADS)

Heinola, K.; Widdowson, A.; Likonen, J.; Alves, E.; Baron-Wiechec, A.; Barradas, N.; Brezinsek, S.; Catarino, N.; Coad, P.; Koivuranta, S.; Krat, S.; Matthews, G. F.; Mayer, M.; Petersson, P.; Contributors, JET

2016-02-01

Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITER-like wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be \\gt 0.3 {{mg}}.

High density operation for reactor-relevant power exhaust

NASA Astrophysics Data System (ADS)

Wischmeier, M.; ASDEX Upgrade Team; Jet Efda Contributors

2015-08-01

With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

Tokamak reactor for treating fertile material or waste nuclear by-products

DOEpatents

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Carbon Radiation Studies in the DIII-D Divertor with the Monte Carlo Impurity (MCI) Code

NASA Astrophysics Data System (ADS)

Evans, T. E.; Leonard, A. W.; West, W. P.; Finkenthal, D. F.; Fenstermacher, M. E.; Porter, G. D.; Chu, Y.

1998-11-01

Carbon sputtering and transport are modeled in the DIII--D divertor with the MCI code. Calculated 2-D radiation patterns are compared with measured radiation distributions. The results are particularly sensitive to Ti near the divertor target plates. For example, increasing the ion temperature from 8 eV to 20 eV in MCI raises P_rad^div from 1626 to 2862 kW. Although this presents difficulties in assessing which sputtering model best describes the plasma-surface interaction physics (because of experimental uncertainties in T_i), processes which either produce too much or too little radiated power compared to the measured value of 1718 kW can be eliminated. Based on this, the number of viable sputtering options has been reduced from 12 to 4. For the conditions studied, three of these options involve both physical and chemical sputtering, and one requires only physical sputtering.

Applications of Collisional Radiative Modeling of Helium and Deuterium for Image Tomography Diagnostic of Te, Ne, and ND in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Munoz Burgos, J. M.; Brooks, N. H.; Fenstermacher, M. E.; Meyer, W. H.; Unterberg, E. A.; Schmitz, O.; Loch, S. D.; Balance, C. P.

2011-10-01

We apply new atomic modeling techniques to helium and deuterium for diagnostics in the divertor and scrape-off layer regions. Analysis of tomographically inverted images is useful for validating detachment prediction models and power balances in the divertor. We apply tomographic image inversion from fast tangential cameras of helium and Dα emission at the divertor in order to obtain 2D profiles of Te, Ne, and ND (neutral ion density profiles). The accuracy of the atomic models for He I will be cross-checked against Thomson scattering measurements of Te and Ne. This work summarizes several current developments and applications of atomic modeling into diagnostic at the DIII-D tokamak. Supported in part by the US DOE under DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC05-00OR22725.

Design and Test of Wendelstein 7-X Water-Cooled Divertor Scraper

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boscary, J.; Greuner, Henri; Ehrke, Gunnar

Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipesmore » of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper.« less

Effect of Dipole Perturbation on a Good Confining Surface Near X Point

NASA Astrophysics Data System (ADS)

Prakash, Sushant; Witmer, Lisa; Punjabi, Alkesh; Ali, Halima

2002-11-01

We analyze effects of dipole perturbation on a good magnetic surface of a single null divertor tokamak using the method of maps developed by Punjabi and Boozer /1/. Unperturbed fields are represented by the Symmetric Simple Map /2/. Effects of high MN perturbation are represented by the Dipole Map (DM) /2/ given by: x_n+1 = x_n+2δ s^3x_n[(y_n-y_s+s) / [x_n+1^2+(y_n-y_s+s)^2]^2], y_n+1 = y_n+δ s^3x_n[[(y_n-y_s+s)^2-x_n+1^2] / [x_n+1^2+(y_n-y_s+s)^2]^2] The good surface is x_0=0 and y_0=0.952873. We start with the strength of perturbation, δ=0, increase it slowly, and observe how the surface changes. The results will be presented. This project is supported by the QEM network NASA Sharp Plus program and by DE-FG02-02ER54673. The work is done on the FUSION1 server under Profs. Ali and Punjabi 1. Punjabi A., Verma A. and Boozer A., Phys. Rev. Lett., 69, 3322 (1992) 2. Ali H., Punjabi A. and Boozer A., Dipole Map for Single-null Divertor Tokamak, 2002 International Congress on Plasma Physics, Manly, Australia.

Calculation of the Area of Stochastic Layer for a Single-Null Divertor Tokamak with the Effects of Dipole Coil Using Method of Maps

NASA Astrophysics Data System (ADS)

Basemore, Alphonso; Ali, Halima; Watson, Michael; Punjabi, Alkesh

1996-11-01

We calculate the variation in area of the stochastic scrape-off layer of a single-null divertor tokamak resulting from the effects of an externally placed dipole coil using the Method of Maps (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The unperturbed magnetic topology is represented by the Symmetric Simple Map (Ali H, Watson M, Mayer C, Punjabi A and Boozer A, Bull Am Phys Soc), 40, 1855 (1995). The effects of the dipole coil are repesented by the Dipole Map (Ali H, Watson M, Punjabi A and Boozer A, Sherwood Mtg), paper 1C20 (1996). A single dipole coil is placed across from the X-point below the last good surface. The strength of the dipole perturbation and the distance of the coil from last good surface are varied. The area of the stochastic layer is calculated using the method of fractal dimension. This work is supported by US DOE OFES. Alphonso Basemore is a HU CFRT Summer Fusion High School Workshop scholar from Mount Tabor High School in North Carolina. He is supported by NASA under its NASA SharpPlus Program.

Statistical characterization of surface features from tungsten-coated divertor inserts in the DIII-D Metal Rings Campaign

NASA Astrophysics Data System (ADS)

Adams, Jacob; Unterberg, Ezekial; Chrobak, Christopher; Stahl, Brian; Abrams, Tyler

2017-10-01

Continuing analysis of tungsten-coated inserts from the recent DIII-D Metal Rings Campaign utilizes a statistical approach to study carbon migration and deposition on W surfaces and to characterize the pre- versus post-exposure surface morphology. A TZM base was coated with W using both CVD and PVD and allowed for comparison between the two coating methods. The W inserts were positioned in the lower DIII-D divertor in both the upper (shelf) region and lower (floor) region and subjected to multiple plasma shots, primarily in H-mode. Currently, the post-exposure W inserts are being characterized using SEM/EDX to qualify the surface morphology and to quantify the surface chemical composition. In addition, profilometry is being used to measure the surface roughness of the inserts both before and after plasma exposure. Preliminary results suggest a correlation between the pre-exposure surface roughness and the level of carbon deposited on the surface. Furthermore, ongoing in-depth analysis may reveal insights into the formation mechanism of nanoscale bumps found in the carbon-rich regions of the W surfaces that have not yet been explained. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

GITR Simulation of Helium Exposed Tungsten Erosion and Redistribution in PISCES-A

NASA Astrophysics Data System (ADS)

Younkin, T. R.; Green, D. L.; Doerner, R. P.; Nishijima, D.; Drobny, J.; Canik, J. M.; Wirth, B. D.

2017-10-01

The extreme heat, charged particle, and neutron flux / fluence to plasma facing materials in magnetically confined fusion devices has motivated research to understand, predict, and mitigate the associated detrimental effects. Of relevance to the ITER divertor is the helium interaction with the tungsten divertor, the resulting erosion and migration of impurities. The linear plasma device PISCES A has performed dedicated experiments for high (4x10-22 m-2s-1) and low (4x10-21 m-2s-1) flux, 250 eV He exposed tungsten targets to assess the net and gross erosion of tungsten and volumetric transport. The temperature of the target was held between 400 and 600 degrees C. We present results of the erosion / migration / re-deposition of W during the experiment from the GITR (Global Impurity Transport) code coupled to materials response models. In particular, the modeled and experimental W I emission spectroscopy data for the 429.4 nm wavelength and net erosion through target and collector mass difference measurements are compared. Overall, the predictions are in good agreement with experiments. This material is supported by the US DOE, Office of Science, Office of Fusion Energy Sciences and Office of Advanced Scientific Computing Research through the SciDAC program on Plasma-Surface Interactions.

Gyrokinetic simulation of edge blobs and divertor heat-load footprint

NASA Astrophysics Data System (ADS)

Chang, C. S.; Ku, S.; Hager, R.; Churchill, M.; D'Azevedo, E.; Worley, P.

2015-11-01

Gyrokinetic study of divertor heat-load width Lq has been performed using the edge gyrokinetic code XGC1. Both neoclassical and electrostatic turbulence physics are self-consistently included in the simulation with fully nonlinear Fokker-Planck collision operation and neutral recycling. Gyrokinetic ions and drift kinetic electrons constitute the plasma in realistic magnetic separatrix geometry. The electron density fluctuations from nonlinear turbulence form blobs, as similarly seen in the experiments. DIII-D and NSTX geometries have been used to represent today's conventional and tight aspect ratio tokamaks. XGC1 shows that the ion neoclassical orbit dynamics dominates over the blob physics in setting Lq in the sample DIII-D and NSTX plasmas, re-discovering the experimentally observed 1/Ip type scaling. Magnitude of Lq is in the right ballpark, too, in comparison with experimental data. However, in an ITER standard plasma, XGC1 shows that the negligible neoclassical orbit excursion effect makes the blob dynamics to dominate Lq. Differently from Lq 1mm (when mapped back to outboard midplane) as was predicted by simple-minded extrapolation from the present-day data, XGC1 shows that Lq in ITER is about 1 cm that is somewhat smaller than the average blob size. Supported by US DOE and the INCITE program.

DIII-D Upgrade to Prepare the Basis for Steady-State Burning Plasmas

NASA Astrophysics Data System (ADS)

Buttery, R. J.; Guo, H. Y.; Taylor, T. S.; Wade, M. R.; Hill, D. N.

2014-10-01

Future steady-state burning plasma facilities will access new physics regimes and modes of plasma behavior. It is vital to prepare for this both experimentally using existing facilities, and theoretically in order to develop the tools to project to and optimize these devices. An upgrade to DIII-D is proposed to address the three critical aspects where research must go beyond what we can do now: (i) torque free electron heating to address the energy, particle and momentum transport mechanisms of burning plasmas using electron cyclotron (EC) heating and full power balanced neutral beams; (ii) off-axis heating and current drive to develop the path to true fusion steady state by reorienting neutral beams and deploying EC and helicon current drive; (iii) a new divertor with hot walls and reactor relevant materials to develop the basis for benign detached divertor operation compatible with wall materials and a high performance fusion core. These elements with modest incremental cost and enacted as a user facility for the whole US program will enable the US to lead on ITER and take a decision to proceed with a Fusion Nuclear Science Facility. Work supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC52-07NA27344.

Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology [Effects of two- and three-dimensional magnetic fields on the Small Angle Slot divertor magnetic topology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trevisan, Gregorio L.; Lao, Lang L.; Evans, Todd E.

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic topology is presented in this paper. On one hand, a twodimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional “vacuum”more » analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the threedimensional I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Furthermore, such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.« less

Upgrades of edge, divertor and scrape-off layer diagnostics of W7-X for OP1.2

DOE PAGES

Hathiramani, D.; Ali, A.; Anda, G.; ...

2018-02-07

In this work, Wendelstein 7-X (W7-X) is the world’s largest superconducting nuclear fusion experiment of the optimized stellarator type. In the first Operation Phase (OP1.1) helium and hydrogen plasmas were studied in limiter configuration. The heating energy was limited to 4 MJ and the main purpose of that campaign was the integral commissioning of the machine and diagnostics, which was achieved very successfully. Already from the beginning a comprehensive set of diagnostics was available to study the plasma. On the path towards high-power, high-performance plasmas, W7-X will be stepwise upgraded from an inertially cooled (OP1.2, limited to 80 MJ) tomore » an actively cooled island divertor (OP2, 10 MW steady-state plasma operation). The machine is prepared for OP1.2 with 10 inertially cooled divertor units, and the experimental campaign has started recently.The paper describes a subset of diagnostics which will be available for OP1.2 to study the plasma edge, divertor and scrape-off layer physics including those already available for OP1.1, plus modifications, upgrades and new systems. In conclusion, the focus of this summary will be on technical and engineering aspects, like feasibility and assembly but also on reliability, thermal loads and shielding against magnetic fields.« less

A new scaling for divertor detachment

NASA Astrophysics Data System (ADS)

Goldston, R. J.; Reinke, M. L.; Schwartz, J. A.

2017-05-01

The ITER design, and future reactor designs, depend on divertor ‘detachment,’ whether partial, pronounced or complete, to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. It would be valuable to have a measure of the difficulty of achieving detachment as a function of machine parameters, such as input power, magnetic field, major radius, etc. Frequently the parallel heat flux, estimated typically as proportional to P sep/R or P sep B/R, is used as a proxy for this difficulty. Here we argue that impurity cooling is dependent on the upstream density, which itself must be limited by a Greenwald-like scaling. Taking this into account self-consistently, we find the impurity fraction required for detachment scales dominantly as power divided by poloidal magnetic field. The absence of any explicit scaling with machine size is concerning, as P sep surely must increase greatly for an economic fusion system, while increases in the poloidal field strength are limited by coil technology and plasma physics. This result should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. Nonetheless, it suggests that higher magnetic field, stronger shaping, double-null operation, ‘advanced’ divertor configurations, as well as alternate means to handle heat flux such as metallic liquid and/or vapor targets merit greater attention.

Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology [Effects of two- and three-dimensional magnetic fields on the Small Angle Slot divertor magnetic topology

DOE PAGES

Trevisan, Gregorio L.; Lao, Lang L.; Evans, Todd E.; ...

2018-01-04

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic topology is presented in this paper. On one hand, a twodimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional “vacuum”more » analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the threedimensional I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Furthermore, such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.« less

A new scaling for divertor detachment

DOE PAGES

Goldston, R. J.; Reinke, M. L.; Schwartz, J. A.

2017-03-29

The ITER design, and future reactor designs, depend on divertor `detachment,'whether partial, pronounced or complete, to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. It would be valuable to have a measure of the difficulty of achieving detachment as a function of machine parameters, such as input power, magnetic field, major radius, etc. Frequently the parallel heat flux, estimated typically as proportional to P-sep/R or PsepB/R, is used as a proxy for this difficulty. Here we argue that impurity cooling is dependent on the upstream density, which itself must be limited by a Greenwald-likemore » scaling. Taking this into account self-consistently, we find the impurity fraction required for detachment scales dominantly as power divided by poloidal magnetic field. The absence of any explicit scaling with machine size is concerning, as P-sep surely must increase greatly for an economic fusion system, while increases in the poloidal field strength are limited by coil technology and plasma physics. This result should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. Nonetheless, it suggests that higher magnetic field, stronger shaping, double-null operation, `advanced' divertor configurations, as well as alternate means to handle heat flux such as metallic liquid and/or vapor targets merit greater attention.« less

Fast wave power flow along SOL field lines in NSTX

NASA Astrophysics Data System (ADS)

Perkins, R. J.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; Leblanc, B. P.; Kramer, G. J.; Phillips, C. K.; Roquemore, L.; Taylor, G.; Wilson, J. R.; Ahn, J.-W.; Gray, T. K.; Green, D. L.; McLean, A.; Maingi, R.; Ryan, P. M.; Jaeger, E. F.; Sabbagh, S.

2012-10-01

On NSTX, a major loss of high-harmonic fast wave (HHFW) power can occur along open field lines passing in front of the antenna over the width of the scrape-off layer (SOL). Up to 60% of the RF power can be lost and at least partially deposited in bright spirals on the divertor floor and ceiling [1,2]. The flow of HHFW power from the antenna region to the divertor is mostly aligned along the SOL magnetic field [3], which explains the pattern of heat deposition as measured with infrared (IR) cameras. By tracing field lines from the divertor back to the midplane, the IR data can be used to estimate the profile of HHFW power coupled to SOL field lines. We hypothesize that surface waves are being excited in the SOL, and these results should benchmark advanced simulations of the RF power deposition in the SOL (e.g., [4]). Minimizing this loss is critical optimal high-power long-pulse ICRF heating on ITER while guarding against excessive divertor erosion.[4pt] [1] J.C. Hosea et al., AIP Conf Proceedings 1187 (2009) 105. [0pt] [2] G. Taylor et al., Phys. Plasmas 17 (2010) 056114. [0pt] [3] R.J. Perkins et al., to appear in Phys. Rev. Lett. [0pt] [4] D.L. Green et al., Phys. Rev. Lett. 107 (2011) 145001.

Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

DOE PAGES

Lore, Jeremy D.; Reinke, M. L.; Brunner, D.; ...

2015-04-28

We study experiments in Alcator C-Mod to assess the level of toroidal asymmetry in divertor conditions resulting from poloidally and toroidally localized extrinsic impurity gas seeding show a weak toroidal peaking (~1.1) in divertor electron temperatures for high-power enhanced D-alpha H-modeplasmas. This is in contrast to similar experiments in Ohmically heated L-modeplasmas, which showed a clear toroidal modulation in the divertor electron temperature. Modeling of these experiments using the 3D edge transport code EMC3-EIRENE [Y. Feng et al., J. Nucl. Mater. 241, 930 (1997)] qualitatively reproduces these trends, and indicates that the different response in the simulations is due tomore » the ionization location of the injected nitrogen. Low electron temperatures in the private flux region (PFR) in L-mode result in a PFR plasma that is nearly transparent to neutral nitrogen, while in H-mode the impurities are ionized in close proximity to the injection location, with this latter case yielding a largely axisymmetric radiation pattern in the scrape-off-layer. In conclusion, the consequences for the ITER gas injection system are discussed. Quantitative agreement with the experiment is lacking in some areas, suggesting potential areas for improving the physics model in EMC3-EIRENE.« less

TEMPEST simulations of the plasma transport in a single-null tokamak geometry

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.

2010-06-01

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.

Modeling Large-Scale Networks Using Virtual Machines and Physical Appliances

DTIC Science & Technology

2014-01-27

downloaded and run locally. The lab solution couldn’t be based on ActiveX because the military Report Documentation Page Form ApprovedOMB No. 0704-0188...unclassified b. ABSTRACT unclassified c. THIS PAGE unclassified Standard Form 298 (Rev. 8-98) Prescribed by ANSI Std Z39-18 disallowed ActiveX support on...its systems, which made running an RDP client over ActiveX not possible. The challenges the SEI encountered in delivering the instruction were

Social factors ameliorate psychiatric disorders in community-based asylum seekers independent of visa status.

PubMed

Hocking, Debbie C; Kennedy, Gerard A; Sundram, Suresh

2015-12-15

The impact of industrialised host nations' deterrent immigration policies on the mental health of forced migrants has not been well characterised. The present study investigated the impact of Australia's refugee determination process (RDP) on psychiatric morbidity in community-based asylum-seekers (AS) and refugees. Psychiatric morbidity was predicted to be greater in AS than refugees, and to persist or increase as a function of time in the RDP. The effect on mental health of demographic and socio-political factors such as health cover and work rights were also investigated. Psychiatric morbidity was measured prospectively on five mental health indices at baseline (T1, n=131) and an average of 15.7 months later (T2, n=56). Psychiatric morbidity in AS significantly decreased between time points such that it was no longer greater than that of refugees at T2. Caseness of PTSD and demoralisation reduced in AS who gained protection; however, those who maintained asylum-seeker status at T2 also had a significant reduction in PTS and depression symptom severity. Reduced PTS and demoralisation symptoms were associated with securing work rights and health cover. Living in the community with work rights and access to health cover significantly improves psychiatric symptoms in forced migrants irrespective of their protection status. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

The Trans-Pacific Partnership and pharmaceutical innovation.

PubMed

Freeman, Robert A

2016-01-01

Trade agreements are an overlooked area of research and policy analysis that affect market access, pricing and reimbursement decisions by pharmaceutical manufacturers, and research and development decisions in the long term. The Trans-Pacific Partnership (TPP) is the most recent multi-national agreement under considerations that may have profound implications in developed and developing countries in the Pacific Rim. As in the case of other trade arrangements, the TPP negotiations are not transparent, but a major leak of the most recent draft has been published in WikiLeaks. The leaked document has raised a number of concerns about intellectual property rights (IPR) and regulatory data protection (RDP) that have implications for public health and economic policy throughout the region. In particular, IPR and RDP go beyond the minimum standards set under the World Trade Organization (WTO) and may affect drug access negatively by delaying generic drug and biosimilar product availability and by raising prices by removing national regulations dealing with drug pricing and reimbursement. Of particular concern is the establishment of a litigation process where multi-national companies can sue individual countries before a panel of private attorneys who are appointed by the World Bank or United Nations. This paper addresses these concerns along with a commentary on the likelihood of occurring and the need for future research. Copyright © 2015 Elsevier Inc. All rights reserved.

Agroecology and the Sustainable Production of Food and ...

EPA Pesticide Factsheets

The silvopastoral, agricultural system of the montado in Southern Portugal is an example of the self-organization of an agroecological system adapted to the climate and soil conditions of the Mediterranean basin. This system with its consistent production of food, fiber, and ecosystem services along with its concomitant rural social organization has been sustained in the region for over 1000 years. However, the system has been gradually decreasing in extent since the 19th century and its rate of decline has accelerated since the 1980s. The causes of this decline have been traced in descending order of importance to land managment choices, spatial factors and environmental factors. In addition, past studies have shown that there is an optimum livestock support capacity for maintaining the health of the montado agroecosystem. In this study, we used the results of an emergy evaluation of a cattle farm as part of a montado agroecosystem to examine the effects of the European Union’s (EU) Common Agricultural Policy (CAP) on the viability of both cattle rearing and the long term regional sustainability of montado agroecosystems. We found that the CAP and its two pillars for providing subsidies, (1) Common Market Organization (CMO) and (2) Rural Development Policy (RDP) are complex and take into account many aspects of prices and markets for particular products, e.g., beef and veal (CMO) and sustainable rural development, e.g., silvopastoral agroecosystems (RDP). How

Relationships between in situ protein degradability and grass developmental morphology.

PubMed

Mitchell, R B; Redfearn, D D; Moser, L E; Grant, R J; Moore, K J; Kirch, B H

1997-06-01

The objective of this research was to determine the relationships between the morphological development and in situ ruminally degradable protein (RDP), ruminally undegradable protein (RUP), and microbial protein of two cool season grasses (intermediate wheatgrass and smooth bromegrass) and two warm season grasses (switchgrass and big bluestem). The initial growth of grass tillers grown near Mead, Nebraska was clipped at ground level six times during the 1992 growing season and morphologically classified. Mean stage was calculated. Forage was ground to pass a 2-mm screen and was incubated in ruminally fistulated steers for 16 h. The RUP was adjusted for microbial protein and acid detergent insoluble N. The mean stage of cool season grasses was higher than that of warm season grasses throughout the growing season. The RDP decreased as plant maturity increased for all species. The RUP expressed as a percentage of crude protein for the cool season grasses was lower than that for warm season grasses. The RUP for intermediate wheatgrass, smooth bromegrass, and switchgrass remained constant across maturities, but RUP for big bluestem decreased as maturity increased. Microbial augmentation of RUP decreased as crude protein decreased in all species. The RUP corrected for acid detergent insoluble N and microbial protein was relatively constant across plant maturities. The quantification of RUP across a range of plant maturities provided information for incorporating RUP content of forage grasses into the diets of animals.

Infection rates in tacrolimus versus cyclosporine-treated pediatric kidney transplant recipients on a rapid discontinuation of prednisone protocol: 1-year analysis.

PubMed

Kizilbash, Sarah J; Rheault, Michelle N; Bangdiwala, Ananta; Matas, Arthur; Chinnakotla, Srinath; Chavers, Blanche M

2017-06-01

AR is lower in pKTx recipients on Tac vs CsA. Data comparing infection outcomes for children treated with these agents are limited. We retrospectively studied infection outcomes in 96 pKTx recipients on a RDP. PS, DCGS, AR, and infection-free survival were assessed using Kaplan-Meier/log-rank tests and proportional hazards models. There were no differences in 1-year PS, DCGS, or AR between Tac and CsA recipients. After adjusting for AR, the hazard of CMV viremia was 4.0 times higher (95%CI: 1.04, 15.5; P = .044) and that of BK viremia was 3.8 times higher (95%CI: 1.5, 10.2; P = .007) in Tac recipients. The incidence of EBV viremia was similar between the groups (P = .56). PostTx lymphoproliferative disease was only observed in Tac recipients (3%). There was no difference in the incidence of pneumonia, urinary tract, or Clostridium difficile infections between Tac and CsA recipients. Among KTx recipients on RDP, the hazards of CMV and BK viremia within 1 year post-KTx were significantly higher in Tac recipients compared to CsA. Regular assessment for infections and lower Tac trough levels may be warranted in Tac recipients. © 2017 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

Plasma-surface interaction in the Be/W environment: Conclusions drawn from the JET-ILW for ITER

NASA Astrophysics Data System (ADS)

Brezinsek, S.; JET-EFDA contributors

2015-08-01

The JET ITER-Like Wall experiment (JET-ILW) provides an ideal test bed to investigate plasma-surface interaction (PSI) and plasma operation with the ITER plasma-facing material selection employing beryllium in the main chamber and tungsten in the divertor. The main PSI processes: material erosion and migration, (b) fuel recycling and retention, (c) impurity concentration and radiation have be1en studied and compared between JET-C and JET-ILW. The current physics understanding of these key processes in the JET-ILW revealed that both interpretation of previously obtained carbon results (JET-C) and predictions to ITER need to be revisited. The impact of the first-wall material on the plasma was underestimated. Main observations are: (a) low primary erosion source in H-mode plasmas and reduction of the material migration from the main chamber to the divertor (factor 7) as well as within the divertor from plasma-facing to remote areas (factor 30 - 50). The energetic threshold for beryllium sputtering minimises the primary erosion source and inhibits multi-step re-erosion in the divertor. The physical sputtering yield of tungsten is low as 10-5 and determined by beryllium ions. (b) Reduction of the long-term fuel retention (factor 10 - 20) in JET-ILW with respect to JET-C. The remaining retention is caused by implantation and co-deposition with beryllium and residual impurities. Outgassing has gained importance and impacts on the recycling properties of beryllium and tungsten. (c) The low effective plasma charge (Zeff = 1.2) and low radiation capability of beryllium reveal the bare deuterium plasma physics. Moderate nitrogen seeding, reaching Zeff = 1.6 , restores in particular the confinement and the L-H threshold behaviour. ITER-compatible divertor conditions with stable semi-detachment were obtained owing to a higher density limit with ILW. Overall JET demonstrated successful plasma operation in the Be/W material combination and confirms its advantageous PSI behaviour and gives strong support to the ITER material selection.

Dynamic ELM and divertor control using resonant toroidal multi-mode magnetic fields in DIII-D and EAST

NASA Astrophysics Data System (ADS)

Sun, Youwen

2017-10-01

A rotating n = 2 Resonant Magnetic Perturbation (RMP) field combined with a stationary n = 3 RMP field has validated predictions that access to ELM suppression can be improved, while divertor heat and particle flux can also be dynamically controlled in DIII-D. Recent observations in the EAST tokamak indicate that edge magnetic topology changes, due to nonlinear plasma response to magnetic perturbations, play a critical role in accessing ELM suppression. MARS-F code MHD simulations, which include the plasma response to the RMP, indicate the nonlinear transition to ELM suppression is optimized by configuring the RMP coils to drive maximal edge stochasticity. Consequently, mixed toroidal multi-mode RMP fields, which produce more densely packed islands over a range of additional rational surfaces, improve access to ELM suppression, and further spread heat loading on the divertor. Beneficial effects of this multi-harmonic spectrum on ELM suppression have been validated in DIII-D. Here, the threshold current required for ELM suppression with a mixed n spectrum, where part of the n = 3 RMP field is replaced by an n = 2 field, is smaller than the case with pure n = 3 field. An important further benefit of this multi-mode approach is that significant changes of 3D particle flux footprint profiles on the divertor are found in the experiment during the application of a rotating n = 2 RMP field superimposed on a static n = 3 RMP field. This result was predicted by modeling studies of the edge magnetic field structure using the TOP2D code which takes into account plasma response from MARS-F code. These results expand physics understanding and potential effectiveness of the technique for reliably controlling ELMs and divertor power/particle loading distributions in future burning plasma devices such as ITER. Work supported by USDOE under DE-FC02-04ER54698 and NNSF of China under 11475224.

The ARIES Advanced and Conservative Tokamak Power Plant Study

DOE PAGES

Kessel, C. E; Tillak, M. S; Najmabadi, F.; ...

2015-12-22

Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦ total N of 5.75, an H98 of 1.65,more » an n/n Gr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦ total N of 2.5, an H₉₈ of 1.25, an n/n Gr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less

The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kessel, C. E.; Poli, F. M.; Ghantous, K.

2014-03-05

Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a βN total of 5.75, Hmore » 98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m 2. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βN total of 2.5, H 98 of 1.25, n/n Gr of 1.3, and peak divertor heat flux of 10 MW/m 2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m 2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.« less

Divertor, scrape-off layer and pedestal particle dynamics in the ELM cycle on ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Laggner, F. M.; Keerl, S.; Gnilsen, J.; Wolfrum, E.; Bernert, M.; Carralero, D.; Guimarais, L.; Nikolaeva, V.; Potzel, S.; Cavedon, M.; Mink, F.; Dunne, M. G.; Birkenmeier, G.; Fischer, R.; Viezzer, E.; Willensdorfer, M.; Wischmeier, M.; Aumayr, F.; the EUROfusion MST1 Team; the ASDEX Upgrade Team

2018-02-01

In addition to the relaxation of the pedestal, edge localised modes (ELMs) introduce changes to the divertor and scrape-off layer (SOL) conditions. Their impact on the inter-ELM pedestal recovery is investigated, with emphasis on the electron density (n e) evolution. The typical ELM cycle occurring in an exemplary ASDEX Upgrade discharge interval at moderate applied gas puff and heating power is characterised, utilising several divertor, SOL and pedestal diagnostics. In the studied discharge interval the inner divertor target is detached before the ELM crash, while the outer target is attached. The particles and power expelled by the ELM crash lead to a re-attachment of the inner target plasma. After the ELM crash, the outer divertor target moves into a high recycling regime with large n e in front of the plate, which is accompanied by high main chamber neutral fluxes. On similar timescales, the inner target fully detaches and the high field side high density region (HFSHD) is formed reaching up to the high field side midplane. This state evolves again to the pre-ELM state, when the main chamber neutral fluxes are reduced later in the ELM cycle. Neither the timescale of the appearance of the HFSHD nor the increase of the main chamber neutral fluxes fit the timescale of the n e pedestal, which is faster. It is found that during the n e pedestal recovery, the magnetic activity at the low field side midplane is strongly reduced indicating a lower level of fluctuations. A rough estimation of the particle flux across the pedestal suggests that the particle flux is reduced in this period. In conclusion, the evolution of the n e pedestal is determined by a combination of neutral fluxes, HFSHD and reduced particle flux across the pedestal. A reduced particle flux explains the fast, experimentally observed re-establishment of the n e pedestal best, whereas neutrals and HFSHD impact on the evolution of the SOL and separatrix conditions.

Overview of innovative PMI research on NSTX-U and associated PMI facilities at PPPL

DOE PAGES

M. Ono; Jaworski, M.; Kaita, R.; ...

2013-05-01

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

Overview of the TCV tokamak program: scientific progress and facility upgrades

NASA Astrophysics Data System (ADS)

Coda, S.; Ahn, J.; Albanese, R.; Alberti, S.; Alessi, E.; Allan, S.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Ariola, M.; Bernert, M.; Beurskens, M.; Bin, W.; Blanchard, P.; Blanken, T. C.; Boedo, J. A.; Bolzonella, T.; Bouquey, F.; Braunmüller, F. H.; Bufferand, H.; Buratti, P.; Calabró, G.; Camenen, Y.; Carnevale, D.; Carpanese, F.; Causa, F.; Cesario, R.; Chapman, I. T.; Chellai, O.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Costea, S.; Crisanti, F.; Cruz, N.; Czarnecka, A.; Decker, J.; De Masi, G.; De Tommasi, G.; Douai, D.; Dunne, M.; Duval, B. P.; Eich, T.; Elmore, S.; Esposito, B.; Faitsch, M.; Fasoli, A.; Fedorczak, N.; Felici, F.; Février, O.; Ficker, O.; Fietz, S.; Fontana, M.; Frassinetti, L.; Furno, I.; Galeani, S.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garrido, I.; Geiger, B.; Giovannozzi, E.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Graves, J. P.; Guirlet, R.; Hakola, A.; Ham, C.; Harrison, J.; Hawke, J.; Hennequin, P.; Hnat, B.; Hogeweij, D.; Hogge, J.-Ph.; Honoré, C.; Hopf, C.; Horáček, J.; Huang, Z.; Igochine, V.; Innocente, P.; Ionita Schrittwieser, C.; Isliker, H.; Jacquier, R.; Jardin, A.; Kamleitner, J.; Karpushov, A.; Keeling, D. L.; Kirneva, N.; Kong, M.; Koubiti, M.; Kovacic, J.; Krämer-Flecken, A.; Krawczyk, N.; Kudlacek, O.; Labit, B.; Lazzaro, E.; Le, H. B.; Lipschultz, B.; Llobet, X.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Maget, P.; Maljaars, E.; Malygin, A.; Maraschek, M.; Marini, C.; Martin, P.; Martin, Y.; Mastrostefano, S.; Maurizio, R.; Mavridis, M.; Mazon, D.; McAdams, R.; McDermott, R.; Merle, A.; Meyer, H.; Militello, F.; Miron, I. G.; Molina Cabrera, P. A.; Moret, J.-M.; Moro, A.; Moulton, D.; Naulin, V.; Nespoli, F.; Nielsen, A. H.; Nocente, M.; Nouailletas, R.; Nowak, S.; Odstrčil, T.; Papp, G.; Papřok, R.; Pau, A.; Pautasso, G.; Pericoli Ridolfini, V.; Piovesan, P.; Piron, C.; Pisokas, T.; Porte, L.; Preynas, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Reich, M.; Reimerdes, H.; Reux, C.; Ricci, P.; Rittich, D.; Riva, F.; Robinson, T.; Saarelma, S.; Saint-Laurent, F.; Sauter, O.; Scannell, R.; Schlatter, Ch.; Schneider, B.; Schneider, P.; Schrittwieser, R.; Sciortino, F.; Sertoli, M.; Sheikh, U.; Sieglin, B.; Silva, M.; Sinha, J.; Sozzi, C.; Spolaore, M.; Stange, T.; Stoltzfus-Dueck, T.; Tamain, P.; Teplukhina, A.; Testa, D.; Theiler, C.; Thornton, A.; Tophøj, L.; Tran, M. Q.; Tsironis, C.; Tsui, C.; Uccello, A.; Vartanian, S.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vijvers, W. A. J.; Vlahos, L.; Vu, N. M. T.; Walkden, N.; Wauters, T.; Weisen, H.; Wischmeier, M.; Zestanakis, P.; Zuin, M.; the EUROfusion MST1 Team

2017-10-01

The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

Ideal plasma response to vacuum magnetic fields with resonant magnetic perturbations in non-axisymmetric tokamaks

DOE PAGES

Kim, Kimin; Ahn, J. -W.; Scotti, F.; ...

2015-09-03

Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifiesmore » the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Furthermore, amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.« less

Manufacturing and testing of a prototypical divertor vertical target for ITER

NASA Astrophysics Data System (ADS)

Merola, M.; Plöchl, L.; Chappuis, Ph; Escourbiac, F.; Grattarola, M.; Smid, I.; Tivey, R.; Vieider, G.

2000-12-01

After an extensive R&D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m 2, 1000 cycles at 10 MW/m 2 and more then 1000 cycles at 15-20 MW/m 2. The final critical heat flux test reached a value in excess of 30 MW/m 2.

A practical globalization of one-shot optimization for optimal design of tokamak divertors

NASA Astrophysics Data System (ADS)

Blommaert, Maarten; Dekeyser, Wouter; Baelmans, Martine; Gauger, Nicolas R.; Reiter, Detlev

2017-01-01

In past studies, nested optimization methods were successfully applied to design of the magnetic divertor configuration in nuclear fusion reactors. In this paper, so-called one-shot optimization methods are pursued. Due to convergence issues, a globalization strategy for the one-shot solver is sought. Whereas Griewank introduced a globalization strategy using a doubly augmented Lagrangian function that includes primal and adjoint residuals, its practical usability is limited by the necessity of second order derivatives and expensive line search iterations. In this paper, a practical alternative is offered that avoids these drawbacks by using a regular augmented Lagrangian merit function that penalizes only state residuals. Additionally, robust rank-two Hessian estimation is achieved by adaptation of Powell's damped BFGS update rule. The application of the novel one-shot approach to magnetic divertor design is considered in detail. For this purpose, the approach is adapted to be complementary with practical in parts adjoint sensitivities. Using the globalization strategy, stable convergence of the one-shot approach is achieved.

Characterized the pattern of the material deposition in the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Cai, Laizhong; Wang, Jianbao; Wu, Ting; Zeng, Xiaoxiao; Hai, Ran; Ding, Hongbin

2017-03-01

Since the divertor geometry of a tokamak has a strong impact on the material erosion and deposition on the wall and HL-2A has a unique divertor configuration, it is necessary to investigate the material deposition pattern in HL-2A although a few results on other tokamaks have already been published. In this paper, tiles retrieved from the vessel are analyzed ex-situ by SIMS, SEM and laser-induced breakdown spectroscopy (LIBS). And deposition behind the lower divertor is in-situ measured by a quartz crystal microbalance (QMB). The deposition in HL-2A displays a complex pattern and clear localization characteristic. The thickness of the deposition layer varies in the range of 0-4μm. And in-situ diagnostic of QMB indicates that the average thickness of the deposition layer per pulse is over ten nanometers. In addition, the results imply that Si, Fe and D have different behaviors during the material deposition in HL-2A.

Global transport of light elements boron and carbon in the full-W ASDEX Upgrade

NASA Astrophysics Data System (ADS)

ASDEX Upgrade Team; Hakola, A.; Likonen, J.; Koivuranta, S.; Krieger, K.; Mayer, M.; Neu, R.; Rohde, V.; Sugiyama, K.

2011-08-01

Transport of carbon and boron has been investigated in the full-W ASDEX Upgrade after experimental campaigns with (2008) and without (2007) boronizations. For this purpose, poloidal deposition profiles of the two elements on tungsten and graphite regions of lower-divertor tiles have been determined. Carbon is mainly deposited in the inner divertor - 80-90% of the determined 12C and 13C inventories on W - while boron shows a much more symmetric deposition profile. In the unboronized machine, the boron inventories are a factor of 10 smaller than in the boronized case and result from residual boron atoms left in the torus prior to the 2007 campaign. Both carbon and boron are deposited more efficiently and/or show less erosion on graphite than on tungsten, particularly in the outer divertor. For 13C, the difference is 10-100 in favor of graphite. This is most probably caused by a higher re-erosion from tungsten surfaces.

EMC3-EIRENE modelling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

DOE PAGES

Lore, Jeremy D.; Reinke, M. L.; LaBombard, Brian; ...

2014-09-30

Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ~50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modelling, with the simulation yielding a toroidal asymmetry in the heat flow to the outermore » strike point. Finally, toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.« less

Powder Injection Molding for mass production of He-cooled divertor parts

NASA Astrophysics Data System (ADS)

Antusch, S.; Norajitra, P.; Piotter, V.; Ritzhaupt-Kleissl, H.-J.

2011-10-01

A He-cooled divertor for future fusion power plants has been developed at KIT. Tungsten and tungsten alloys are presently considered the most promising materials for functional and structural divertor components. The advantages of tungsten materials lie, e.g. in the high melting point, and low activation, the disadvantages are high hardness and brittleness. The machinig of tungsten, e.g. milling, is very complex and cost-intensive. Powder Injection Molding (PIM) is a method for cost effective mass production of near-net-shape parts with high precision. The complete W-PIM process route is outlined and, results of product examination discussed. A binary tungsten powder feedstock with a grain size distribution in the range 0.7-1.7 μm FSSS, and a solid load of 50 vol.% was developed. After heat treatment, the successfully finished samples showed promising results, i.e. 97.6% theoretical density, a grain size of approximately 5 μm, and a hardness of 457 HV0.1.

Bohm criterion and plasma particle/power exhaust to and recycling at the wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, Xianzhu; Guo, Zehua

The plasma particle and power exhaust to the divertor surface drives both particle and power recycling at the surface, which in return constrains the plasma density and temperature at the target and their profile further upstream. Both particle and power exhaust fluxes are mediated by the plasma sheath next to the divertor surface. In particular, the Bohm criterion constrains the ion exit flow speed, which enters directly into the particle flux and the kinetic flow energy component of the ion power flux, and indirectly into the electron power flux through the sheath potential drop. Here we give an overview onmore » how the Bohm speed is set in a general plasma and how it enters power exhaust and power recycling at the divertor surface, and the implication on the correct implementation of sheath boundary conditions in numerical codes. The cases of ideal and non-ideal Bohm speed are distinguished as a result of the physics discussion.« less

Bohm criterion and plasma particle/power exhaust to and recycling at the wall

DOE PAGES

Tang, Xianzhu; Guo, Zehua

2017-06-07

The plasma particle and power exhaust to the divertor surface drives both particle and power recycling at the surface, which in return constrains the plasma density and temperature at the target and their profile further upstream. Both particle and power exhaust fluxes are mediated by the plasma sheath next to the divertor surface. In particular, the Bohm criterion constrains the ion exit flow speed, which enters directly into the particle flux and the kinetic flow energy component of the ion power flux, and indirectly into the electron power flux through the sheath potential drop. Here we give an overview onmore » how the Bohm speed is set in a general plasma and how it enters power exhaust and power recycling at the divertor surface, and the implication on the correct implementation of sheath boundary conditions in numerical codes. The cases of ideal and non-ideal Bohm speed are distinguished as a result of the physics discussion.« less

Operational limits on WEST inertial divertor sector during the early phase experiment

NASA Astrophysics Data System (ADS)

Firdaouss, M.; Corre, Y.; Languille, P.; Greuner, H.; Autissier, E.; Desgranges, C.; Guilhem, D.; Gunn, J. P.; Lipa, M.; Missirlian, M.; Pascal, J.-Y.; Pocheau, C.; Richou, M.; Tsitrone, E.

2016-02-01

The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m-2 for the required pulse length.

DDN (Defense Data Network) Protocol Handbook. Volume 3. Supplement

DTIC Science & Technology

1985-12-01

the physical configuration which results from this arrangement, Michael Padlipsky has described this as the " milking machine" approach to computer...NEIRJS-3 74 NETRJS-4 75 77 79 FINGER 81 HOSTS2-NS 83 MIT-ML-DEV 85 MIT-ML-DEV 87 89 SU-MIT-TG 91 MIT-DOV 93 DCP 95 SUPDUP 97 SWIFT- RVF ...PUP QUOTE RDP RJE RLP RTELNET RVD SAT-EXPAK SAT-MON SFTT? SMTP ST SU-MIT-TG SUNRPC SUPDUP SUR-MEAS SWIFT- RVF TACACS-DS TACNEWS

Baseline Tests of the Electra Van Model 1000 Electric Vehicle.

DTIC Science & Technology

1980-07-01

3RR ATALGNUMBER(. 21 0 IIACNTLING TFESSO NAME ND TR ADDRESS~eot MOrELe rDp o00 EERIy AEHICtan Srtr f, or Hybrd Veicle S. NC4"IXICtION NGRA DINGUMER...state-of-the-art with respect to electric and hybrid vehicles. The data so developed are to serve as a baseline to compare im- provements in electric and... hybrid vehicle technologies, to assist in establishing per- formance standards for electric and hybrid vehicles, and to guide future research and

Computer-aided fabrication of a zirconia 14-unit removable dental prosthesis: a technical report.

PubMed

Grösser, Julian; Sachs, Caroline; Stadelmann, Markus; Schweiger, Josef; Güthe, Jan-Frederik; Beuer, Florian

2014-01-01

Double crown systems with primary crowns made from zirconia are used to support removable dental prostheses (RDPs). However, the fabrication of RDPs is labor-intensive and costly. Manufacturing primary and secondary crowns from zirconia with a CAD/CAM system might simplify the fabrication protocol and reduce costs. Furthermore, only ceramic materials are used in this method, providing an RDP with the highest possible biocompatibility and greatest possible esthetics. This article describes the fabrication protocol step by step.

Time-to-burnout data for a prototypical ITER divertor tube during a simulated loss of flow accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, T.D.; Watson, R.D.; McDonald, J.M.

The Loss of Flow Accident (LOFA) is a serious safety concern for the International Thermonuclear Experimental Reactor (ITER) as it has been suggested that greater than 100 seconds are necessary to safely shutdown the plasma when ITER is operating at full power. In this experiment, the thermal response of a prototypical ITER divertor tube during a simulated LOFA was studied. The divertor tube was fabricated from oxygen-free high-conductivity copper to have a square geometry with a circular coolant channel. The coolant channel inner diameter was 0.77 cm, the heated length was 4.0 cm, and the heated width was 1.6 cm.more » The mockup did not feature any flow enhancement techniques, i.e., swirl tape, helical coils, or internal fins. One-sided surface heating of the mockup was accomplished through the use of the 30 kW Sandia Electron Beam Test System. After reaching steady state temperatures in the mockup, as determined by two Type-K thermocouples installed 0.5 mm beneath the heated surface, the coolant pump was manually tripped off and the coolant flow allowed to naturally coast down. Electron beam heating continued after the pump trip until the divertor tube`s heated surface exhibited the high temperature transient normally indicative of rapidly approaching burnout. Experimental data showed that time-to-burnout increases proportionally with increasing inlet velocity and decreases proportionally with increasing incident heat flux.« less

The lithium vapor box divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goldston, R. J.; Myers, R.; Schwartz, J.

It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Our recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m -2, implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et almore » as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. Furthermore, at the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required in order to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma.« less

The lithium vapor box divertor

NASA Astrophysics Data System (ADS)

Goldston, R. J.; Myers, R.; Schwartz, J.

2016-02-01

It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m-2, implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et al as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. At the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma.

Non-solenoidal Startup with High-Field-Side Local Helicity Injection on the Pegasus ST

NASA Astrophysics Data System (ADS)

Perry, J. M.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Pierren, C.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Reusch, J. A.; Weberski, J. D.

2017-10-01

Local Helicity Injection (LHI) is a non-solenoidal startup technique utilizing electron current injectors at the plasma edge to initiate a tokamak-like plasma at high Ip . Recent experiments on Pegasus explore the inherent tradeoffs between high-field-side (HFS) injection in the lower divertor region and low-field-side (LFS) injection at the outboard midplane. Trade-offs include the relative current drive contributions of HI and poloidal induction, and the magnetic geometry required for relaxation to a tokamak-like state. HFS injection using a set of two increased-area injectors (Ainj = 4 cm2, Vinj 1.5 kV, and Iinj 8 kA) in the lower divertor is demonstrated over the full range of toroidal field available on Pegasus (BT 0 <= 0.15 T). Increased PMI on both the injectors and the lower divertor plates was observed during HFS injection, and was substantively mitigated through optimization of injector geometry and placement of local limiters to reduce scrape-off density in the divertor region. Ip up to 200 kA is achieved with LHI as the dominant current drive, consistent with expectations from helicity balance. To date, experiments support Ip increasing linearly with helicity injection rate. The high normalized current (IN >= 10) attainable with LHI and the favorable stability of the ultra-low aspect ratio, low-li LHI-driven plasmas allow access to high βt-up to 100 % , as indicated by kinetically-constrained equilibrium reconstructions. Work supported by US DOE Grant DE-FG02-96ER54375.

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE PAGES

X. Q. Xu; Bodi, K.; Cohen, R. H.; ...

2010-05-28

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

X. Q. Xu; Bodi, K.; Cohen, R. H.

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

The lithium vapor box divertor

DOE PAGES

Goldston, R. J.; Myers, R.; Schwartz, J.

2016-01-13

It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Our recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m -2, implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et almore » as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. Furthermore, at the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required in order to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma.« less

Time-dependent modeling of dust injection in semi-detached ITER divertor plasma

NASA Astrophysics Data System (ADS)

Smirnov, Roman; Krasheninnikov, Sergei

2017-10-01

At present, it is generally understood that dust related issues will play important role in operation of the next step fusion devices, i.e. ITER, and in the development of future fusion reactors. Recent progress in research on dust in magnetic fusion devises has outlined several topics of particular concern: a) degradation of fusion plasma performance; b) impairment of in-vessel diagnostic instruments; and c) safety issues related to dust reactivity and tritium retention. In addition, observed dust events in fusion edge plasmas are highly irregular and require consideration of temporal evolution of both the dust and the fusion plasma. In order to address the dust-related fusion performance issues, we have coupled the dust transport code DUSTT and the edge plasma transport code UEDGE in time-dependent manner, allowing modeling of transient dust-induced phenomena in fusion edge plasmas. Using the coupled codes we simulate burst-like injection of tungsten dust into ITER divertor plasma in semi-detached regime, which is considered as preferable ITER divertor operational mode based on the plasma and heat load control restrictions. Analysis of transport of the dust and the dust-produced impurities, and of dynamics of the ITER divertor and edge plasma in response to the dust injection will be presented. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-06ER54852.

DECIPHER, a Search-Based Approach to Chimera Identification for 16S rRNA Sequences

PubMed Central

Wright, Erik S.; Yilmaz, L. Safak

2012-01-01

DECIPHER is a new method for finding 16S rRNA chimeric sequences by the use of a search-based approach. The method is based upon detecting short fragments that are uncommon in the phylogenetic group where a query sequence is classified but frequently found in another phylogenetic group. The algorithm was calibrated for full sequences (fs_DECIPHER) and short sequences (ss_DECIPHER) and benchmarked against WigeoN (Pintail), ChimeraSlayer, and Uchime using artificially generated chimeras. Overall, ss_DECIPHER and Uchime provided the highest chimera detection for sequences 100 to 600 nucleotides long (79% and 81%, respectively), but Uchime's performance deteriorated for longer sequences, while ss_DECIPHER maintained a high detection rate (89%). Both methods had low false-positive rates (1.3% and 1.6%). The more conservative fs_DECIPHER, benchmarked only for sequences longer than 600 nucleotides, had an overall detection rate lower than that of ss_DECIPHER (75%) but higher than those of the other programs. In addition, fs_DECIPHER had the lowest false-positive rate among all the benchmarked programs (<0.20%). DECIPHER was outperformed only by ChimeraSlayer and Uchime when chimeras were formed from closely related parents (less than 10% divergence). Given the differences in the programs, it was possible to detect over 89% of all chimeras with just the combination of ss_DECIPHER and Uchime. Using fs_DECIPHER, we detected between 1% and 2% additional chimeras in the RDP, SILVA, and Greengenes databases from which chimeras had already been removed with Pintail or Bellerophon. DECIPHER was implemented in the R programming language and is directly accessible through a webpage or by downloading the program as an R package (http://DECIPHER.cee.wisc.edu). PMID:22101057

Halo current diagnostic system of experimental advanced superconducting tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, D. L.; Shen, B.; Sun, Y.

2015-10-15

The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

Full toroidal imaging of non-axisymmetric plasma material interaction in the National Spherical Torus Experiment divertor.

PubMed

Scotti, Filippo; Roquemore, A L; Soukhanovskii, V A

2012-10-01

A pair of two dimensional fast cameras with a wide angle view (allowing a full radial and toroidal coverage of the lower divertor) was installed in the National Spherical Torus Experiment in order to monitor non-axisymmetric effects. A custom polar remapping procedure and an absolute photometric calibration enabled the easier visualization and quantitative analysis of non-axisymmetric plasma material interaction (e.g., strike point splitting due to application of 3D fields and effects of toroidally asymmetric plasma facing components).

Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

NASA Astrophysics Data System (ADS)

Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

2017-01-01

The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

Disruption Neutral Point Experiment on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Granetz, R. S.; Nakamura, Y.

2000-10-01

Disruptions of single-null elongated plasmas generally result in loss of vertical position control, leading to a current quench occurring at the top or bottom of the machine, with all the attendant problems of halo and eddy currents flowing in divertor structures. On JT-60U, it has been found that if the plasma is operated with its magnetic axis at a particular height, called the neutral point, the initial vertical drift after a thermal quench is significantly slower than usual, and sometimes can even be arrested, thereby avoiding a current quench in the divertor region entirely. In an ongoing collaboration between MIT and JAERI, the neutral point concept is being tested in Alcator C-Mod, which has a significantly higher plasma elongation than JT-60U (1.65 vs 1.3). Calculations using TSC predict a neutral point at z~=+1 cm above the midplane (a=22 cm). The existence of a neutral point has now been experimentally confirmed, albeit at a height of z=+2.7 cm. The plasma has remained vertically stable for up to 9 ms after the disruption thermal quench, which in principle, is long enough for the PF control system to respond, if programmed appropriately. In addition, the physics of the neutral point stability on C-Mod appears to be somewhat different than that on JT-60U.

Engineering Proteins for Thermostability with iRDP Web Server

PubMed Central

Ghanate, Avinash; Ramasamy, Sureshkumar; Suresh, C. G.

2015-01-01

Engineering protein molecules with desired structure and biological functions has been an elusive goal. Development of industrially viable proteins with improved properties such as stability, catalytic activity and altered specificity by modifying the structure of an existing protein has widely been targeted through rational protein engineering. Although a range of factors contributing to thermal stability have been identified and widely researched, the in silico implementation of these as strategies directed towards enhancement of protein stability has not yet been explored extensively. A wide range of structural analysis tools is currently available for in silico protein engineering. However these tools concentrate on only a limited number of factors or individual protein structures, resulting in cumbersome and time-consuming analysis. The iRDP web server presented here provides a unified platform comprising of iCAPS, iStability and iMutants modules. Each module addresses different facets of effective rational engineering of proteins aiming towards enhanced stability. While iCAPS aids in selection of target protein based on factors contributing to structural stability, iStability uniquely offers in silico implementation of known thermostabilization strategies in proteins for identification and stability prediction of potential stabilizing mutation sites. iMutants aims to assess mutants based on changes in local interaction network and degree of residue conservation at the mutation sites. Each module was validated using an extensively diverse dataset. The server is freely accessible at http://irdp.ncl.res.in and has no login requirements. PMID:26436543

Engineering Proteins for Thermostability with iRDP Web Server.

PubMed

Panigrahi, Priyabrata; Sule, Manas; Ghanate, Avinash; Ramasamy, Sureshkumar; Suresh, C G

2015-01-01

Engineering protein molecules with desired structure and biological functions has been an elusive goal. Development of industrially viable proteins with improved properties such as stability, catalytic activity and altered specificity by modifying the structure of an existing protein has widely been targeted through rational protein engineering. Although a range of factors contributing to thermal stability have been identified and widely researched, the in silico implementation of these as strategies directed towards enhancement of protein stability has not yet been explored extensively. A wide range of structural analysis tools is currently available for in silico protein engineering. However these tools concentrate on only a limited number of factors or individual protein structures, resulting in cumbersome and time-consuming analysis. The iRDP web server presented here provides a unified platform comprising of iCAPS, iStability and iMutants modules. Each module addresses different facets of effective rational engineering of proteins aiming towards enhanced stability. While iCAPS aids in selection of target protein based on factors contributing to structural stability, iStability uniquely offers in silico implementation of known thermostabilization strategies in proteins for identification and stability prediction of potential stabilizing mutation sites. iMutants aims to assess mutants based on changes in local interaction network and degree of residue conservation at the mutation sites. Each module was validated using an extensively diverse dataset. The server is freely accessible at http://irdp.ncl.res.in and has no login requirements.

Infection Rates in Tacrolimus versus Cyclosporine Treated Pediatric Kidney Transplant Recipients on a Rapid Discontinuation of Prednisone Protocol: 1-year Analysis

PubMed Central

Kizilbash, Sarah J; Rheault, Michelle N; Bangdiwala, Ananta; Matas, Arthur; Chinnakotla, Srinath; Chavers, Blanche M

2017-01-01

Introduction Acute rejection (AR) is lower in pediatric kidney transplant (pKTx) recipients on tacrolimus (Tac) versus cyclosporine (CsA). Data comparing infection outcomes for children treated with these agents are limited. Methods We retrospectively studied infection outcomes in 96 pKTx recipients on a rapid discontinuation of prednisone protocol (RDP). Patient survival (PS), death-censored graft survival (DCGS), AR and infection free survival were assessed using Kaplan-Meier/log-rank tests and proportional hazards models. Results There were no differences in 1-year PS, DCGS, or AR between Tac and CsA recipients. After adjusting for AR, the hazard of CMV viremia was 4.0 times higher (95%CI: 1.04, 15.5; p=0.044) and that of BK viremia was 3.8 times higher (95%CI: 1.5, 10.2; p=0.007) in Tac recipients. The incidence of EBV viremia was similar between the groups (p=0.56). Posttransplant lymphoproliferative disease was only observed in Tac recipients (3%). There was no difference in the incidence of pneumonia, urinary tract or clostridium difficile infections between Tac and CsA recipients. Conclusion Among KTx recipients on RDP, the hazards of CMV and BK viremia within 1-year post-KTx were significantly higher in Tac recipients compared to CsA. Regular assessment for infections and lower Tac trough levels may be warranted in Tac recipients. PMID:28371243

A New Scaling for Divertor Detachment

NASA Astrophysics Data System (ADS)

Goldston, Robert

2017-10-01

The ITER design and future fusion power plant designs depend on divertor detachment, whether partial, pronounced or complete, both to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. Generally the parallel heat flux, estimated as proportional to Psep / R or Psep B / R , is used as a proxy for the difficulty of achieving detachment. Here we argue that the impurity cooling required for detachment is strongly dependent on the upstream separatrix density, which is limited by Greenwald scaling. Taking this into account self-consistently, along with the Heuristic Drift (HD) model for the SOL width, and using a Lengyel radiation model that includes non-coronal effects, we find that the relative impurity concentration, cz ≡nz /ne , required for detachment scales dominantly as cz Psep /Bp(nsep /nGW) 2 . The absence of any explicit favorable size scaling is concerning, as Psep must increase by an order of magnitude from present experiments to an economic fusion power system, while increases in the poloidal magnetic field strength are limited by magnet technology and MHD stability. This result should not be surprising, as it follows from the simplest scaling, Psep czne2VSOL , taking into account the Greenwald density limit and the HD SOL volume scaling. Reinke has combined a similar approach with the requirement to maintain H-mode, which sets a lower limit on Psep, and also arrives at an incentive for high field and disincentive for large size. These results should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. In particular measurements are required for extrinsic divertor impurity concentration over a range of power and density conditions far from the regime where detachment can be achieved with deuterium puffing and intrinsic impurities alone. Nonetheless, these results suggest that higher magnetic field, stronger shaping, double-null operation, `advanced' divertor magnetic and baffle configurations, as well as lithium vapor targets merit greater attention. This work supported by the US DOE under contract DE-AC02-09CH11466.

Investigation into the formation of the scrape-off layer density shoulder in JET ITER-like wall L-mode and H-mode plasmas

NASA Astrophysics Data System (ADS)

Wynn, A.; Lipschultz, B.; Cziegler, I.; Harrison, J.; Jaervinen, A.; Matthews, G. F.; Schmitz, J.; Tal, B.; Brix, M.; Guillemaut, C.; Frigione, D.; Huber, A.; Joffrin, E.; Kruzei, U.; Militello, F.; Nielsen, A.; Walkden, N. R.; Wiesen, S.; Contributors, JET

2018-05-01

The low temperature boundary layer plasma (scrape-off layer or SOL) between the hot core and the surrounding vessel determines the level of power loading, erosion and implantation of material surfaces, and thus the viability of tokamak-based fusion as an energy source. This study explores mechanisms affecting the formation of flattened density profiles, so-called ‘density shoulders’, in the low-field side (LFS) SOL, which modify ion and neutral fluxes to surfaces—and subsequent erosion. We find that increases in SOL parallel resistivity, Λdiv (=[L || ν eiΩi]/c sΩe), postulated to lead to shoulder growth through changes in SOL turbulence characteristics, correlates with increases in SOL shoulder amplitude, A s, only under a subset of conditions (D2-fuelled L-mode density scans with outer strike point on the horizontal target). Λdiv fails to correlate with A s for cases of N2 seeding or during sweeping of the strike point across the horizontal target. The limited correlation of Λdiv and A s is also found for H-mode discharges. Thus, while it may be necessary for Λdiv to be above a threshold of ~1 for shoulder formation and/or growth, another mechanism is required. More significantly, we find that in contrast to parallel resistivity, outer divertor recycling, as quantified by the total outer divertor Balmer D α emission, I-D α , does scale with A s where Λdiv does and even where Λdiv does not. Divertor recycling could lead to SOL density shoulder formation through: (a) reducing the parallel to the field flow (loss) of ions out of the SOL to the divertor; and (b) changes in radial electric fields which lead to E  ×  B poloidal flows as well as potentially affecting SOL turbulence birth characteristics. Thus, changes in divertor recycling may be the sole process involved in bringing about SOL density shoulders or it may be that it acts in tandem with parallel resistivity.

Evidence and modeling of 3D divertor footprint induced by lower hybrid waves on EAST with tungsten divertor operations

NASA Astrophysics Data System (ADS)

Feng, W.; Wang, L.; Rack, M.; Liang, Y.; Guo, H. Y.; Xu, G. S.; Xu, J. C.; Liu, J. B.; Sun, Y. W.; Jia, M. N.; Yang, Q. Q.; Zhang, B.; Zou, X. L.; Liu, H.; Zhang, T.; Ding, F.; Chen, J. B.; Duan, Y. M.; Zheng, X. W.; Dai, S. Y.; Deng, G. Z.; Chen, R.; Hu, G. H.; Yan, N.; Si, H.; Liu, S. C.; Xu, S.; Wang, M.; Li, M. H.; Ding, B. J.; Wingen, A.; Huang, J.; Gao, X.; Luo, G. N.; Gong, X. Z.; Garofalo, A. M.; Li, J.; Wan, B. N.; the EAST Team

2017-12-01

Three dimensional (3D) divertor particle flux footprints induced by the lower hybrid wave (LHW) have been systematically investigated in the EAST superconducting tokamak during the recent experimental campaign. We find that the striated particle flux (SPF) peaks away from the strike point (SP) closely fit the pitch of the edge magnetic field line for different safety factors q 95, as predicted by a field line tracing code taking into account the helical current filaments (HCFs) in the scrape-off-layer (SOL). As LHW power increases, it requires the fuelling to be increased e.g. by super molecular beam injection (SMBI), to maintain a similar plasma density, which may be attributed to the pump-out effect due to LHW, and may thus be beneficial for EAST steady state operations. The 3D SPF structure is observed with a LHW power threshold (P LHW ~ 0.9 MW). The ratio of the particle fluxes between SPF and outer strike point (OSP), i.e. {{Γ }ion,SPF}/{{Γ }ion,OSP} , increases with the LHW power. Upon transition to divertor detachment, the particle flux at the main OSP decreases, as expected, however, the particle flux at SPF continues increasing, in contrast to the RMP-induced striations that vanish with increasing divertor density. In addition, we also find that the in-out asymmetry of the 3D particle flux footprint pattern exhibits a clear dependence on the toroidal field direction (B  ×    ∇   B  ↓  and B  ×    ∇   B↑). Experiments using neon impurity seeding show a promising capability in 3D particle and heat flux control on EAST. LHW-induced particle and heat flux striations are also present in the H-mode plasmas, reducing the peak heat flux and erosion at the main strike point, thus facilitating long-pulse operation with a new steady-state H-mode over 60 s being recently achieved in EAST.

Modelling of mitigation of the power divertor loading for the EU DEMO through Ar injection

NASA Astrophysics Data System (ADS)

Subba, Fabio; Aho-Mantila, Leena; Coster, David; Maddaluno, Giorgio; Nallo, Giuseppe F.; Sieglin, Bernard; Wenninger, Ronald; Zanino, Roberto

2018-03-01

In this paper we present a computational study on the divertor heat load mitigation through impurity injection for the EU DEMO. The study is performed by means of the SOLPS5.1 code. The power crossing the separatrix is considered fixed and corresponding to H-mode operation, whereas the machine operating condition is defined by the outboard mid-plane upstream electron density and the impurity level. The selected impurity for this study is Ar, based on its high radiation efficiency at SOL characteristic temperatures. We consider a conventional vertical target geometry for the EU DEMO and monitor target conditions for different operational points, considering as acceptability criteria the target electron temperature (≤5 eV to provide sufficiently low W sputtering rate) and the peak heat flux (below 5-10 MW m-2 to guarantee safe steady-state cooling conditions). Our simulations suggest that, neglecting the radiated power deposition on the plate, it is possible to satisfy the desired constraints. However, this requires an upstream density of the order of at least 50% of the Greenwald limit and a sufficiently high argon fraction. Furthermore, if the radiated power deposition is taken into account, the peak heat flux on the outer plate could not be reduced below 15 MW m-2 in these simulations. As these simulations do not take into account neutron loading, they strongly indicate that the vertical target divertor solution with a radiative front distributed along the divertor leg has a very marginal operational space in an EU DEMO sized reactor.

Compatibility of separatrix density scaling for divertor detachment with H-mode pedestal operation in DIII-D

NASA Astrophysics Data System (ADS)

Leonard, A. W.; McLean, A. G.; Makowski, M. A.; Stangeby, P. C.

2017-08-01

The midplane separatrix density is characterized in response to variations in upstream parallel heat flux density and central density through deuterium gas injection. The midplane density is determined from a high spatial resolution Thomson scattering diagnostic at the midplane with power balance analysis to determine the separatrix location. The heat flux density is varied by scans of three parameters, auxiliary heating, toroidal field with fixed plasma current, and plasma current with fixed safety factor, q 95. The separatrix density just before divertor detachment onset is found to scale consistent with the two-point model when radiative dissipation is taken into account. The ratio of separatrix to pedestal density, n e,sep/n e,ped varies from  ⩽30% to  ⩾60% over the dataset, helping to resolve the conflicting scaling of core plasma density limit and divertor detachment onset. The scaling of the separatrix density at detachment onset is combined with H-mode power threshold scaling to obtain a scaling ratio of minimum n e,sep/n e,ped expected in future devices.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Budaev, V. P., E-mail: budaev@mail.ru

2016-12-15

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 ofmore » 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.« less

Divertor tungsten tile melting and its effect on core plasma performance

NASA Astrophysics Data System (ADS)

Lipschultz, B.; Coenen, J. W.; Barnard, H. S.; Howard, N. T.; Reinke, M. L.; Whyte, D. G.; Wright, G. M.

2012-12-01

For the 2007 and 2008 run campaigns, Alcator C-Mod operated with a full toroidal row of tungsten tiles in the high heat flux region of the outer divertor; tungsten levels in the core plasma were below measurement limits. An accidental creation of a tungsten leading edge in the 2009 campaign led to this study of a melting tungsten source: H-mode operation with strike point in the region of the melting tile was immediately impossible due to some fraction of tungsten droplets reaching the main plasma. Approximately 15 g of tungsten was lost from the tile over ˜100 discharges. Less than 1% of the evaporated tungsten was found re-deposited on surfaces, the rest is assumed to have become dust. The strong discharge variability of the tungsten reaching the core implies that the melt layer topology is always varying. There is no evidence of healing of the surface with repeated melting. Forces on the melted tungsten tend to lead to prominences that extend further into the plasma. A discussion of the implications of melting a divertor tungsten monoblock on the ITER plasma is presented.

Overview of experimental preparation for the ITER-Like Wall at JET

NASA Astrophysics Data System (ADS)

Jet Efda Contributors Brezinsek, S.; Fundamenski, W.; Eich, T.; Coad, J. P.; Giroud, C.; Huber, A.; Jachmich, S.; Joffrin, E.; Krieger, K.; McCormick, K.; Lehnen, M.; Loarer, T.; de La Luna, E.; Maddison, G.; Matthews, G. F.; Mertens, Ph.; Nunes, I.; Philipps, V.; Riccardo, V.; Rubel, M.; Stamp, M. F.; Tsalas, M.

2011-08-01

Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 × 1021 D s-1 were obtained as references in accompanied gas balance studies.

Modelling the detachment dependence on strike point location in the small angle slot divertor (SAS) with SOLPS

NASA Astrophysics Data System (ADS)

Casali, Livia; Covele, Brent; Guo, Houyang

2017-10-01

The new Small Angle Slot (SAS) divertor in DIII-D is characterized by a shallow-angle target enclosed by a slot structure about the strike point (SP). SOLPS modelling results of SAS have demonstrated divertor closure's utility in widening the range of acceptable densities for adequate heat handling. An extensive database of runs has been built to study the detachment dependence on SP location in SAS. Density scans show that lower Te at lower upstream density occur when the SP is at the critical location in the slot. The cooling front spreads across the entire target at higher densities, in agreement with experimental Langmuir probe measurements. A localized increase of the atomic and molecular density takes place near the SP, which reduces the target incident power density and facilitates detachment at lower upstream density. Systematic scans of variables such as power, transport, and viscosity have been carried out to assess the detachment sensitivity. Therein, a positive role of the viscosity is found. This work supported by DOE Contract Number DE-FC02-04ER54698.

A practical globalization of one-shot optimization for optimal design of tokamak divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blommaert, Maarten, E-mail: maarten.blommaert@kuleuven.be; Dekeyser, Wouter; Baelmans, Martine

In past studies, nested optimization methods were successfully applied to design of the magnetic divertor configuration in nuclear fusion reactors. In this paper, so-called one-shot optimization methods are pursued. Due to convergence issues, a globalization strategy for the one-shot solver is sought. Whereas Griewank introduced a globalization strategy using a doubly augmented Lagrangian function that includes primal and adjoint residuals, its practical usability is limited by the necessity of second order derivatives and expensive line search iterations. In this paper, a practical alternative is offered that avoids these drawbacks by using a regular augmented Lagrangian merit function that penalizes onlymore » state residuals. Additionally, robust rank-two Hessian estimation is achieved by adaptation of Powell's damped BFGS update rule. The application of the novel one-shot approach to magnetic divertor design is considered in detail. For this purpose, the approach is adapted to be complementary with practical in parts adjoint sensitivities. Using the globalization strategy, stable convergence of the one-shot approach is achieved.« less

Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

NASA Astrophysics Data System (ADS)

Xue, L.; Duan, X. R.; Zheng, G. Y.; Liu, Y. Q.; Pan, Y. D.; Yan, S. L.; Dokuka, V. N.; Lukash, V. E.; Khayrutdinov, R. R.

2016-05-01

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.

Measurements of C V flows from thermal charge-exchange excitation in divertor plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zaniol, B.; Isler, R. C.; Brooks, N. H.

2001-10-01

Certain transitions of C IV (C{sup 3+}) from n=7 to n=6 ({approx}7226 {angstrom}) and from n=6 to n=5 ({approx}4660 {angstrom}) sometimes appear much brighter in tokamak divertors than expected for electron-impact excitation from the ground state. This situation occurs because of charge exchange between C V (C{sup 4+}) and recycling thermal deuterium atoms in the n=2 level. As a result, it is possible to extend parallel flow measurements of carbon, which have previously been performed on C II--C IV ions using Doppler shift spectroscopy, to include flows of the He-like C V ions. The work described here includes modeling ofmore » the spectral features, correlation of state populations with classical Monte Carlo trajectory (CTMC) predictions, and applications to flow measurements in the DIII-D divertor [Plasma Physics Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515].« less

Measurements of C V flows from thermal charge-exchange excitation in divertor plasmas

NASA Astrophysics Data System (ADS)

Zaniol, B.; Isler, R. C.; Brooks, N. H.; West, W. P.; Olson, R. E.

2001-10-01

Certain transitions of C IV (C3+) from n=7 to n=6 (≈7226 Å) and from n=6 to n=5 (≈4660 Å) sometimes appear much brighter in tokamak divertors than expected for electron-impact excitation from the ground state. This situation occurs because of charge exchange between C V (C4+) and recycling thermal deuterium atoms in the n=2 level. As a result, it is possible to extend parallel flow measurements of carbon, which have previously been performed on C II-C IV ions using Doppler shift spectroscopy, to include flows of the He-like C V ions. The work described here includes modeling of the spectral features, correlation of state populations with classical Monte Carlo trajectory (CTMC) predictions, and applications to flow measurements in the DIII-D divertor [Plasma Physics Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515].

Partial detachment of high power discharges in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Beurskens, M.; Casali, L.; Dunne, M.; Eich, T.; Giannone, L.; Herrmann, A.; Maraschek, M.; Potzel, S.; Reimold, F.; Rohde, V.; Schweinzer, J.; Viezzer, E.; Wischmeier, M.; the ASDEX Upgrade Team

2015-05-01

Detachment of high power discharges is obtained in ASDEX Upgrade by simultaneous feedback control of core radiation and divertor radiation or thermoelectric currents by the injection of radiating impurities. So far 2/3 of the ITER normalized heat flux Psep/R = 15 MW m-1 has been obtained in ASDEX Upgrade under partially detached conditions with a peak target heat flux well below 10 MW m-2. When the detachment is further pronounced towards lower peak heat flux at the target, substantial changes in edge localized mode (ELM) behaviour, density and radiation distribution occur. The time-averaged peak heat flux at both divertor targets can be reduced below 2 MW m-2, which offers an attractive DEMO divertor scenario with potential for simpler and cheaper technical solutions. Generally, pronounced detachment leads to a pedestal and core density rise by about 20-40%, moderate (<20%) confinement degradation and a reduction of ELM size. For AUG conditions, some operational challenges occur, like the density cut-off limit for X-2 electron cyclotron resonance heating, which is used for central tungsten control.

Calculation of ion distribution functions and neoclassical transport in the edge of single-null divertor tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Cohen, R. H.; Xu, X. Q.

2007-11-01

The ion distribution function in the H-mode pedestal region and outward across the magnetic separatrix is expected to have a substantial non-Maxwellian character owing to the large banana orbits and steep gradients in temperature and density. The 4D (2r,2v) version of the TEMPEST continuum gyrokinetic code is used with a Coulomb collision model to calculate the ion distribution in a single-null tokamak geometry throughout the pedestal/scrape-off-layer regions. The mean density, parallel velocity, and energy radial profiles are shown at various poloidal locations. The collisions cause neoclassical energy transport through the pedestal that is then lost to the divertor plates along the open field lines outside the separatrix. The resulting heat flux profiles at the inner and outer divertor plates are presented and discussed, including asymmetries that depend on the B-field direction. Of particular focus is the effect on ion profiles and fluxes of a radial electric field exhibiting a deep well just inside the separatrix, which reduces the width of the banana orbits by the well-known squeezing effect.

Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Raffray, A.R.; Chiocchio, S.

1995-12-31

This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC`s) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m{sup 2}) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM`s) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects themore » target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC`s clad with different PFM`s are discussed.« less

Creation of a Mouse with Stress-Induced Dystonia: Control of an ATPase Chaperone

DTIC Science & Technology

2012-10-01

Mice with mutations in genes known to cause dystonia in humans are so far virtually asymptomatic. Only mild motor deficiencies have been seen, such...identified the gene for one of the subunits of Na,K- ATPase, ATP1A3, as the site of mutations in RDP (de Carvalho Aguiar et al. 2004). Our prior work in...first paper, and to be able to give the mouse line an official name based on the mutated gene . Funding has been applied for from the following

Some problems of brazing technology for the divertor plate manufacturing

NASA Astrophysics Data System (ADS)

Prokofiev, Yu. G.; Barabash, V. R.; Khorunov, V. F.; Maksimova, S. V.; Gervash, A. A.; Fabritsiev, S. A.; Vinokurov, V. F.

1992-09-01

Among the different design options of the ITER reactor divertor, the joints of the carbon-based materials and molybdenum alloys and joints of tungsten and copper alloys are considered. High-temperature brazing is one of the most promising joining methods for the plasma facing and heat sink materials. The use of brazing for creation of W-Cu and graphite-Mo joints are given here. In addition, the investigation results of microstructure, microhardness and mechanical properties of the joints are presented. For W-Cu samples an influence of the neutron irradiation on the joining strength was studied.

Divertor for use in fusion reactors

DOEpatents

Christensen, Uffe R.

1979-01-01

A poloidal divertor for a toroidal plasma column ring having a set of poloidal coils co-axial with the plasma ring for providing a space for a thick shielding blanket close to the plasma along the entire length of the plasma ring cross section and all the way around the axis of rotation of the plasma ring. The poloidal coils of this invention also provide a stagnation point on the inside of the toroidal plasma column ring, gently curving field lines for vertical stability, an initial plasma current, and the shaping of the field lines of a separatrix up and around the shielding blanket.

An exploration of advanced X-divertor scenarios on ITER

NASA Astrophysics Data System (ADS)

Covele, B.; Valanju, P.; Kotschenreuther, M.; Mahajan, S.

2014-07-01

It is found that the X-divertor (XD) configuration (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA), CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502) can be made with the conventional poloidal field (PF) coil set on ITER (Tomabechi et al and Team 1991 Nucl. Fusion 31 1135), where all PF coils are outside the TF coils. Starting from the standard divertor, a sequence of desirable XD configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. This opens the possibility that the XD could be tested and used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the super-X-divertor (Kotschenreuther et al 2007 Bull. Am. Phys. Soc. 53 11, Valanju et al 2009 Phys. Plasmas 16 5, Kotschenreuther et al 2010 Nucl. Fusion 50 035003, Valanju et al 2010 Fusion Eng. Des. 85 46) is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO (Kim et al 2013 Fusion Eng. Des. 88 123) to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm (2012 Fusion Sci. Technol. 63 43) for the Snowflake (Ryutov 2007 Phys. Plasmas 14 064502, Ryutov et al 2008 Phys. Plasmas 15 092501), where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard divertor SOL (Kotschenreuther 2013 Phys. Plasmas 20 102507) in the recently created XD configurations reproduces what was presented in the earlier XD papers (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA) CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502). Consequently, the same advantages accrue, but no close-in PF coils are employed.

Interaction of plasmas with lithium and tungsten fusion plasma facing components

NASA Astrophysics Data System (ADS)

Fiflis, Peter Robert

One of the largest outstanding issues in magnetic confinement fusion is the interaction of the fusion plasma with the first wall of the device; an interaction which is strongest in the divertor region. Erosion, melting, sputtering, and deformation are all concerns which inform choices of divertor material. Of the many materials proposed for use in the divertor, only a few remain as promising choices. Tungsten has been chosen as the material for the ITER divertor, and liquid lithium stands poised as its replacement in higher heat flux devices. As a refractory metal, tungsten's large melting point and thermal conductivity as well as its low sputtering yield have led to its selection as the material of choice of the ITER divertor. Experiments have reinforced this choice demonstrating tungsten's ability to withstand large heat fluxes when adequately cooled. However, tungsten has shown a propensity to nanostructure under exposure within a certain temperature range to large fluxes of helium ions. These nanostructures if disrupted into the plasma as dust by an off-normal event would cause quenching of the plasma from the generated dust. Liquid lithium, meanwhile, has gathered growing interest within the fusion community in recent years as a divertor, limiter, and alternative first wall material. Liquid lithium is attractive as a low-Z material replacement for refractory metals due to its ability to getter impurities, while also being self-healing in nature. However, concerns exist about the stability of a liquid metal surface at the edge of a fusion device. Liquid metal pools, such as the Li-DiMes probe, have shown evidence of macroscopic lithium displacement as well as droplet formation and ejection into the plasma. These issues must be mitigated in future implementations of liquid lithium divertor concepts. Rayleigh-Taylor-like (RT) and Kelvin-Helmholtz-like (KH) instabilities have been claimed as the initiators of droplet ejection, yet not enough data exists to delineate a stability boundary. The influences of plasma pressure and current driven instabilities on lithium surfaces that lead to droplet ejection are investigated to determine which of the two effects is dominant for a given set of plasma conditions. This work studies the influence of large plasma fluxes on these two materials to better inform the selection and design of plasma facing components (PFCs). The nanostructuring of tungsten was investigated to determine the mechanisms by which tungsten nanostructures so that its formation may be mitigated. Experiments investigated the dependence of nanostructuring on temperature, looked at the morphological evolution, and grew nanostructures on a variety of metals to examine their similarity to tungsten. Additionally, a computational model is presented for the initial stages of fuzz formation showing good quantitative and qualitative agreement with experimental observations. The influences of RT and KH instabilities on the surface of liquid lithium were experimentally observed and quantified on the ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) chamber at the University of Illinois at Urbana-Champaign and the stabilizing effect of surface tension, an effect employed by the LiMIT concept as well as other liquid lithium concepts, was studied, and the stability boundary afforded by surface tension was compared between experiment, computational simulation, and theory.

Structures for handling high heat fluxes

NASA Astrophysics Data System (ADS)

Watson, R. D.

1990-12-01

The divertor is reconized as one of the main performance limiting components for ITER. This paper reviews the critical issues for structures that are designed to withstand heat fluxes > 5 MW/m 2. High velocity, sub-cooled water with twisted tape inserts for enhanced heat transfer provides a critical heat flux limit of 40-60 MW/m 2. Uncertainties in physics and engineering heat flux peaking factors require that the design heat flux not exceed 10 MW/m 2 to maintain an adequate burnout safety margin. Armor tiles and heat sink materials must have a well matched thermal expansion coefficient to minimize stresses. The divertor lifetime from sputtering erosion is highly uncertain. The number of disruptions specified for ITER must be reduced to achieve a credible design. In-situ plasma spray repair with thick metallic coatings may reduce the problems of erosion. Runaway electrons in ITER have the potential to melt actively cooled components in a single event. A water leak is a serious accident because of steam reactions with hot carbon, beryllium, or tungsten that can mobilize large amounts of tritium and radioactive elements. If the plasma does not shutdown immediately, the divertor can melt in 1-10 s after a loss of coolant accident. Very high reliability of carbon tile braze joints will be required to achieve adequate safety and performance goals. Most of these critical issues will be addressed in the near future by operation of the Tore Supra pump limiters and the JET pumped divertor. An accurate understanding of the power flow out of edge of a DT burning plasma is essential to successful design of high heat flux components.

Creating Hybrid Plasmas With Off-Axis ECCD for Radiating Divertor Studies in DIII-D

NASA Astrophysics Data System (ADS)

Petty, C. C.; Ferron, J. R.; Luce, T. C.; Osborne, T. H.; Petrie, T. W.; Turco, F.; Holcomb, C. T.; Thome, K. E.

2017-10-01

A long duration, high density, high power hybrid scenario has been developed for use in radiative divertor studies in DIII-D. Using 11.2 MW of co-NBI power and 3.4 MW of ECCD, with a total injected energy of up to 56 MJ, high performance hybrid plasmas with βN = 3.7 and H98y2 = 1.5 were created. The hybrid plasmas were fully non-inductive at densities of n 4.2 ×1019 m-3 with central ECCD, but the EC deposition needed to be moved to ρ = 0.45 to avoid the right-hand cutoff when the density was raised to n 5.8 ×1019 m-3 for radiative divertor studies. Although moving the EC deposition to ρ = 0.45 had the effect of dropping τE by 10%, the energy confinement time increased with higher density like τE n0.4, allowing high beta to be maintained. While the plasma current profile displays the usual self-organizing properties of hybrids - an anomalously broad profile with qmin >1 - local current drive can still have a measurable effect on stability, either positively or negatively. For example, hybrid discharges with radial ECH deposited at ρ = 0.45 proved to be more robustly stable to n = 1 modes (can be either a 1/1 or 2/1 mode) than similar discharges with co-ECCD at the same location. Interestingly, the large 1/1 mode had almost no effect on energy confinement but strongly degraded particle confinement; thus this mode needed to be suppressed to achieve the high pedestal densities required for radiative divertor studies. Work supported by USDOE under DE-FC02-04ER54698.

Tungsten: an option for divertor and main chamber plasma facing components in future fusion devices

NASA Astrophysics Data System (ADS)

Neu, R.; Dux, R.; Kallenbach, A.; Pütterich, T.; Balden, M.; Fuchs, J. C.; Herrmann, A.; Maggi, C. F.; O'Mullane, M.; Pugno, R.; Radivojevic, I.; Rohde, V.; Sips, A. C. C.; Suttrop, W.; Whiteford, A.; ASDEX Upgrade Team

2005-03-01

The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic of W has been extended and refined and the cooling factor of W has been re-evaluated. The W coated surfaces now represent a fraction of 65% of all plasma facing components (24.8 m2). The only two major components that are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. One very successful remedy is the addition of central RF heating at the 20-30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however, at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10-3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper W coated divertor do not show higher W concentrations than comparable discharges in the lower C based divertor. According to impurity transport calculations no strong high-Z accumulation is expected for the ITER standard scenario as long as the anomalous transport is at least as high as the neoclassical one.

Long-term evolution of the impurity composition and impurity events with the ITER-like wall at JET

NASA Astrophysics Data System (ADS)

Coenen, J. W.; Sertoli, M.; Brezinsek, S.; Coffey, I.; Dux, R.; Giroud, C.; Groth, M.; Huber, A.; Ivanova, D.; Krieger, K.; Lawson, K.; Marsen, S.; Meigs, A.; Neu, R.; Puetterich, T.; van Rooij, G. J.; Stamp, M. F.; Contributors, JET-EFDA

2013-07-01

This paper covers aspects of long-term evolution of intrinsic impurities in the JET tokamak with respect to the newly installed ITER-like wall (ILW). At first the changes related to the change over from the JET-C to the JET-ILW with beryllium (Be) as the main wall material and tungsten (W) in the divertor are discussed. The evolution of impurity fluxes in the newly installed W divertor with respect to studying material migration is described. In addition, a statistical analysis of transient impurity events causing significant plasma contamination and radiation losses is shown. The main findings comprise a drop in carbon content (×20) (see also Brezinsek et al (2013 J. Nucl. Mater. 438 S303)), low oxygen content (×10) due to the Be first wall (Douai et al 2013 J. Nucl. Mater. 438 S1172-6) as well as the evolution of the material mix in the divertor. Initially, a short period of repetitive ohmic plasmas was carried out to study material migration (Krieger et al 2013 J. Nucl. Mater. 438 S262). After the initial 1600 plasma seconds the material surface composition is, however, still evolving. With operational time, the levels of recycled C are increasing slightly by 20% while the Be levels in the deposition-dominated inner divertor are dropping, hinting at changes in the surface layer material mix made of Be, C and W. A steady number of transient impurity events, consisting of W and constituents of inconel, is observed despite the increase in variation in machine operation and changes in magnetic configuration as well as the auxiliary power increase.

Analysis of a Multi-Machine Database on Divertor Heat Fluxes

NASA Astrophysics Data System (ADS)

Makowski, M. A.

2011-10-01

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Corresponding plasma parameters were systematically varied in each tokamak, resulting in a combined data set in which Ip varies by a factor 3, Bt varies by a factor of 14.5, and major radius varies by a factor of 2.6. The derived scaling relation consistently predicts narrower heat flux widths than relations currently in use. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip. All three tokamaks independently demonstrate this dependence. The midplane SOL profiles in DIII-D are also found to steepen with higher Ip, similar to the divertor heat flux profiles. Weaker dependencies on the toroidal field and normalized Greenwald density, fGW, are also found, but vary across devices and with the measure of the heat flux width used, either FWHM or integral width. In the combined data set, the strongest size scaling is with minor radius resulting in an approximately linear dependence on a /Ip . This suggests a scaling correlated with the inverse of the poloidal field, as would be expected for critical gradient or drift-based transport. Supported by the US DOE under DE-AC52-07NA27344 and DE-FC02-04ER54698.

Access to edge scenarios for testing a scraper element in early operation phases of Wendelstein 7-X

DOE PAGES

Holbe, H.; Pedersen, T. Sunn; Geiger, J.; ...

2016-01-29

The edge topology of magnetic fusion devices is decisive for the control of the plasma exhaust. In Wendelstein 7-X, the island divertor concept will be used, for which the edge topology can change significantly as the internal currents in a plasma discharge evolve towards steady-state. Consequently, the device has been optimized to minimize such internal currents, in particular the bootstrap current [1]. Nonetheless, there are predicted pulse scenarios where effects of the remaining internal currents could potentially lead to overload of plasma-facing components. These internal currents are predicted to evolve on long time scales (tens of seconds) so their effectsmore » on the edge topology and the divertor heat loads may not be experimentally accessible in the first years of W7-X operation, where only relatively short pulses are possible. However, we show here that for at least one important long-pulse divertor operation issue, relevant physics experiments can be performed already in short-pulse operation, through judicious adjustment of the edge topology by the use of the existing coil sets. The specific issue studied here is a potential overload of the divertor element edges. This overload might be mitigated by the installation of an extra set of plasma-facing components, so-called scraper elements, as suggested in earlier publications. It is shown here that by a targeted control of edge topology, the effectiveness of such scraper elements can be tested already with uncooled test-scraper elements in short-pulse operation. Furthermore, this will allow an early and well-informed decision on whether long-pulse-capable (actively cooled) scraper elements should be built and installed.« less

Activation, decay heat, and waste classification studies of the European DEMO concept

NASA Astrophysics Data System (ADS)

Gilbert, M. R.; Eade, T.; Bachmann, C.; Fischer, U.; Taylor, N. P.

2017-04-01

Inventory calculations have a key role to play in designing future fusion power plants because, for a given irradiation field and material, they can predict the time evolution in chemical composition, activation, decay heat, gamma-dose, gas production, and even damage (dpa) dose. For conceptual designs of the European DEMO fusion reactor such calculations provide information about the neutron shielding requirements, maintenance schedules, and waste disposal prospects; thereby guiding future development. Extensive neutron-transport and inventory calculations have been performed for a reference DEMO reactor model with four different tritium-breeding blanket concepts. The results have been used to chart the post-operation variation in activity and decay heat from different vessel components, demonstrating that the shielding performance of the different blanket concepts—for a given blanket thickness—varies significantly. Detailed analyses of the simulated nuclide inventories for the vacuum vessel (VV) and divertor highlight the most dominant radionuclides, potentially suggesting how changes in material composition could help to reduce activity. Minor impurities in the raw composition of W used in divertor tiles, for example, are shown to produce undesirable long-lived radionuclides. Finally, waste classifications, based on UK regulations, and a recycling potential limit, have been applied to estimate the time-evolution in waste masses for both the entire vessel (including blanket modules, VV, divertor, and some ex-vessel components) and individual components, and also to suggest when a particular component might be suitable for recycling. The results indicate that the large mass of the VV will not be classifiable as low level waste on the 100 year timescale, but the majority of the divertor will be, and that both components will be potentially recyclable within that time.

Time-resolved deposition in the remote region of the JET-ILW divertor: measurements and modelling

NASA Astrophysics Data System (ADS)

Catarino, N.; Widdowson, A.; Baron-Wiechec, A.; Coad, J. P.; Heinola, K.; Rubel, M.; Alves, E.; Contributors, JET

2017-12-01

One crucial requirement for the development of fusion power is to know where, and how much, impurities collect in the machine, and how much of the fuelling isotope tritium will be trapped therein. The most relevant information on this issue comes from the operation of the Joint European Tokamak (JET), which is the world’s largest operating tokamak and has the same interior plasma-facing materials as the next step machine, ITER. Much of the information gained so far has been from post-mortem analysis of samples collected after whole campaigns involving varied types of operation. This paper describes time-resolved measurements of the deposition rate using rotating collectors (RC) placed in remote areas of the JET divertor during the 2013-2014 campaign with the ITER-like Wall (ILW). These techniques allow the effects of different types of operation to be distinguished. Rotating collectors made of silicon discs housed behind an aperture are exposed to the plasma. Each time the magnetic field coils are ramped up for a discharge the disc rotates, providing a linear relationship between the exposed region and the discharge number. Post-mortem ion beam analyses provide information on the deposit composition as a function of the discharge number. The results show that the Be deposition average for the RC in the corners of the inner and outer divertor are 4.9 × 1016 cm-2 and 1.8 × 1017 cm-2, respectively, accumulated over an average of ˜25 pulses. Data from the rotating collector below Tile 5 in the central region of divertor indicate a Be deposition rate of 9.3 × 1015 cm-2, per ˜25 pulses.

Rumen Degradability and Post-ruminal Digestion of Dry Matter, Nitrogen and Amino Acids of Three Protein Supplements.

PubMed

Gao, Wei; Chen, Aodong; Zhang, Bowen; Kong, Ping; Liu, Chenli; Zhao, Jie

2015-04-01

This study evaluated the in situ ruminal degradability, and subsequent small intestinal digestibility (SID) of dry matter, crude protein (CP), and amino acids (AA) of cottonseed meal (CSM), sunflower seed meal (SFSM) and distillers dried grains with solubles (DDGS) by using the modified three-step in vitro procedure. The ruminal degradability and subsequent SID of AA in rumen-undegradable protein (RUP-AA) varied among three protein supplements. The result show that the effective degradability of DM for SFSM, CSM, and DDGS was 60.8%, 56.4%, and 41.0% and their ruminal fermentable organic matter was 60.0%, 55.9%, and 39.9%, respectively. The ruminal degradable protein (RDP) content in CP for SFSM, CSM, and DDGS was 68.3%, 39.0%, and 32.9%, respectively, at the ruminal solid passage rate of 1.84%/h. The SFSM is a good source of RDP for rumen micro-organisms; however, the SID of RUP of SFSM was lower. The DDGS and CSM are good sources of RUP for lambs to digest in the small intestine to complement ruminal microbial AA of growing lambs. Individual RUP-AA from each protein source was selectively removed by the rumen micro-organisms, especially for Trp, Arg, His, and Lys (p<0.01). The SID of individual RUP-AA was different within specific RUP origin (p<0.01). Limiting amino acid was Leu for RUP of CSM and Lys for both RUP of SFSM and DDGS, respectively. Therefore, different protein supplements with specific limitations should be selected and combined carefully in growing lambs ration to optimize AA balance.

A knowledge brokerage approach for assessing the impacts of the setting up young farmers policy measure in Greece

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bournaris, Th.; Moulogianni, Ch.; Arampatzis, S.

This study explores Knowledge Brokerage (KB) aspects of an ex-post Impact Assessment (IA) for the Rural Development Programme (RDP) measure of setting up young farmers, under the Common Agricultural Policy (CAP), at the regional level in Northern Greece. The measure supports the entry of young farmers in agriculture by moving land from older to younger farmers. The aim of the study was to test a set of KB tools for improving the interaction between researchers and policy makers. Our analysis mainly focused on a suite of IA Support Modules to guide practitioners, and on a technical tool kit, a web-basedmore » contextualisation platform, to support the IA of the specific test case. Offering a structured approach towards IA, both the Support Modules and LIAISE-KIT allow framing the context, organisation, scheduling and method selection in the light of KB objectives. The evaluation of how IA Support Modules influence the Science Policy Interface (SPI), in the case of the ex-post assessment, demonstrated the high relevance of KB activities for facilitating the interaction between researchers and regional policy makers. The assessment bridges the gap between knowledge producers developing scientific output to be applied in a specific context, and knowledge users, who want clear messages regarding the policy challenges they face. Other conclusions include the need for specific guidelines and training for knowledge users, especially with regard to the use of tools. According to our findings, a consequent application of KB activities is a crucial pre-condition for successfully implementing IAs in future RDP measures.« less

Alteration of biomacromolecule in corn by steam flaking in relation to biodegradation kinetics in ruminant, revealed with vibrational molecular spectroscopy

NASA Astrophysics Data System (ADS)

Xu, Ningning; Liu, Jianxin; Yu, Peiqiang

2018-02-01

Large scale of steam flaked corn has been used in dairy ration to maintain high milk production level. This study aimed to determine effects of steam flaking on processing-induced intrinsic molecular structure changes that were associated with rumen degradation kinetics and nutrients supply. The advanced vibrational molecular spectroscopy was applied to reveal the processing-induced intrinsic structure changes on a molecular basis. The rumen degradation kinetics and nutrient supply were determined using in situ approach in ruminant livestock system. Raw corn grain (RC) and steam flaked corn grain (SFC) were obtained from two different processing plants. The results showed that (1) Compared to RC, SFC had greater truly digestible non-fiber carbohydrate [tdNFC: 86.8 versus 78.0% dry matter (DM)], but lower truly digestible crude protein [tdCP: 7.7 versus 9.0% DM]. (2) The steam flaking increased (P < 0.01) rumen degradable DM (RDDM) and starch (RDSt), but decreased (P < 0.01) rumen degradable protein (RDP). (3) Molecular absorbance intensities of most carbohydrate biopolymers were greater in SFC (P < 0.01), but protein amides associated molecular spectral intensities were lower (P < 0.01) in SFC. (4). The molecular structure and nutrient interactive study showed that carbohydrate spectral intensities were positively (P < 0.10) associated with RDDM and RDSt and protein amide spectral intensities were positively (P < 0.10) associated with RDP. This results indicated that the steam flaking induced molecular structure changes had an interactive relationship with rumen degradation kinetics.