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

Maienschein, J.L.; Garcia, F.; Garza, R.G.

Tritium-handling apparatus has been decontaminated as part of the downsizing of the LLNL Tritium Facility. Two stainless-steel glove boxes that had been used to process lithium deuteride-tritide (LiDT) slat were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. In this paper the details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium, in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculationalmore » method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given.« less

Report on Analyses of WAC Samples of Evaporator Overheads - 2004

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

OJI, LAWRENCE

2004-08-16

All water received into ETF requires characterization versus the defined Waste Acceptance Criteria. Currently much of the water received by ETF comes from the F and H Evaporator Overheads. Concentration, Storage and Transfer Engineering issued a modified list of species to be determined. In March of 2004, the Tank Farm submitted annual samples from 2F, 2H and 3H Evaporator Overhead streams for characterization to verify compliance with the Effluent Treatment Facility (ETF) Waste Acceptance Criteria (WAC) and to look for organic species. With the exception of high silicon in the 2H and slightly high tritium in 2F evaporator overheads, allmore » the overheads samples were found to be in compliance with the Effluent Treatment Facility WAC. The silicon concentration in the 2H-evaporator overhead, at 44 mg/L, was above the ETF WAC limit of 5 mg/L and tritium at 2.11E+05 dpm/mL in 2F overhead sample was above the ETF WAC limit of 1.2E+05 dpm/mL.« less

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

Ene, D.; Andersson, K.; Jensen, M.

The European Spallation Source (ESS) will produce tritium via spallation and activation processes during operational activities. Within the location of ESS facility in Lund, Sweden site it is mandatory to demonstrate that the management strategy of the produced tritium ensures the compliance with the country regulation criteria. The aim of this paper is to give an overview of the different aspects of the tritium management in ESS facility. Besides the design parameter study of the helium coolant purification system of the target the consequences of the tritium releasing into the environment were also analyzed. Calculations show that the annual releasemore » of tritium during the normal operations represents a small fraction from the estimated total dose. However, more refined calculations of migration of activated-groundwater should be performed for higher hydraulic conductivities, with the availability of the results on soil examinations. With the assumption of 100% release of tritium to the atmosphere during the occurring of the extreme accidents, it was found as well that the total dose complies with the constraint. (authors)« less

EFFECTS OF GAMMA IRRADIATION ON EPDM ELASTOMERS

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

Clark, E.

Two formulations of EPDM elastomer, one substituting a UV stabilizer for the normal antioxidant in this polymer, and the other the normal formulation, were synthesized and samples of each were exposed to gamma irradiation in initially pure deuterium gas to compare their radiation stability. Stainless steel containers having rupture disks were designed for this task. After 130 MRad dose of cobalt-60 radiation in the SRNL Gamma Irradiation Facility, a significant amount of gas was created by radiolysis; however the composition indicated by mass spectroscopy indicated an unexpected increase in the total amount deuterium in both formulations. The irradiated samples retainedmore » their ductility in a bend test. No change of sample weight, dimensions, or density was observed. No change of the glass transition temperature as measured by dynamic mechanical analysis was observed, and most of the other dynamic mechanical properties remained unchanged. There appeared to be an increase in the storage modulus of the irradiated samples containing the UV stabilizer above the glass transition, which may indicate hardening of the material by radiation damage. Polymeric materials become damaged by exposure over time to ionizing radiation. Despite the limited lifetime, polymers have unique engineering material properties and polymers continue to be used in tritium handling systems. In tritium handling systems, polymers are employed mainly in joining applications such as valve sealing surfaces (eg. Stem tips, valve packing, and O-rings). Because of the continued need to employ polymers in tritium systems, over the past several years, programs at the Savannah River National Laboratory have been studying the effect of tritium on various polymers of interest. In these studies, samples of materials of interest to the SRS Tritium Facilities (ultra-high molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, Teflon{reg_sign}), Vespel{reg_sign} polyimide, and the elastomer ethylene propylene diene monomer (EPDM)) have been exposed in closed containers to tritium gas initially at 1 atmosphere pressure. These studies have demonstrated the degradation of properties when exposed to tritium gas. Also, the radiolytic production of significant amounts of hydrogen has been observed for UHMW-PE and EPDM. The study documented in this report exposes two similar formulations of EPDM elastomer to gamma irradiation in a closed container backfilled with deuterium. Deuterium is chemically identical to protium and tritium, but allows the identification of protium that is radiolytically produced from the samples. The goal of this program is to compare and contrast the response of EPDM exposure to two different types of ionizing radiation in a similar chemical environment.« less

Tritium Plasma Experiment Upgrade and Improvement of Surface Diagnostic Capabilities at STAR Facility for Enhancing Tritium and Nuclear PMI Sciences

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

Shimada, M.; Taylor, C. N.; Pawelko, R. J.

2016-04-01

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to directly study tritium retention and permeation in neutron-irradiated materials with tritium [M. Shimada et.al., Rev. Sci. Instru. 82 (2011) 083503 and and M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. The plasma-material-interaction (PMI) determines a boundary condition for diffusing tritium into bulk PFCs, and the tritium PMI is crucial for enhancing fundamental sciences that dictate tritium fuel cycles and safety and are high importance to an FNSF and DEMO. Recentlymore » the TPE has undergone major upgrades in its electrical and control systems. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium. We discuss the electrical upgrade, enhanced operational safety, improved plasma performance, and development of optical spectrometer system. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, Fusion Nuclear Science Facility (FNSF), and Demonstration reactor (DEMO). This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.« less

2009 EVALUATION OF TRITIUM REMOVAL AND MITIGATION TECHNOLOGIES FOR WASTEWATER TREATMENT

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

LUECK KJ; GENESSE DJ; STEGEN GE

2009-02-26

Since 1995, a state-approved land disposal site (SALDS) has received tritium contaminated effluents from the Hanford Site Effluent Treatment Facility (ETF). Tritium in this effluent is mitigated by storage in slow moving groundwater to allow extended time for decay before the water reaches the site boundary. By this method, tritium in the SALDS is isolated from the general environment and human contact until it has decayed to acceptable levels. This report contains the 2009 update evaluation of alternative tritium mitigation techniques to control tritium in liquid effluents and groundwater at the Hanford site. A thorough literature review was completed andmore » updated information is provided on state-of-the-art technologies for control of tritium in wastewaters. This report was prepared to satisfy the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-026-07B (Ecology, EPA, and DOE 2007). Tritium separation and isolation technologies are evaluated periodically to determine their feasibility for implementation to control Hanford site liquid effluents and groundwaters to meet the Us. Code of Federal Regulations (CFR), Title 40 CFR 141.16, drinking water maximum contaminant level (MCL) for tritium of 20,000 pOll and/or DOE Order 5400.5 as low as reasonably achievable (ALARA) policy. Since the 2004 evaluation, there have been a number of developments related to tritium separation and control with potential application in mitigating tritium contaminated wastewater. These are primarily focused in the areas of: (1) tritium recycling at a commercial facility in Cardiff, UK using integrated tritium separation technologies (water distillation, palladium membrane reactor, liquid phase catalytic exchange, thermal diffusion), (2) development and demonstration of Combined Electrolysis Catalytic Exchange (CECE) using hydrogen/water exchange to separate tritium from water, (3) evaporation of tritium contaminated water for dispersion in the atmosphere, and (4) use of barriers to minimize the transport of tritium in groundwater. Continuing development efforts for tritium separations processes are primarily to support the International Thermonuclear Experimental Reactor (ITER) program, the nuclear power industry, and the production of radiochemicals. While these applications are significantly different than the Hanford application, the technology could potentially be adapted for Hanford wastewater treatment. Separations based processes to reduce tritium levels below the drinking water MCL have not been demonstrated for the scale and conditions required for treating Hanford wastewater. In addition, available cost information indicates treatment costs for such processes will be substantially higher than for discharge to SALDS or other typical pump and treat projects at Hanford. Actual mitigation projects for groundwater with very low tritium contamination similar to that found at Hanford have focused mainly on controlling migration and on evaporation for dispersion in the atmosphere.« less

Health risk assessment of potable water containing small amount of tritium oxide

NASA Astrophysics Data System (ADS)

Momot, O. A.; Synzynys, B. I.; Oudalova, A. A.

2017-01-01

The problem of groundwater pollution with tritium in a vicinity of radiation-dangerous facilities in Obninsk is considered. The information on the specific activity of tritium in Obninsk water sources is provided. The formula for the calculation of the β-radiation absorbed dose from tritium ingestion is proposed, reflecting the biological behavior of tritium in a human body. To establish the extent of tritium effects on human, the health risk is assessed. It is shown that if the specific activity of tritium in drinking water amounts to 10 Bq/l, the risk of stochastic effects of radiation will not exceed the limit of the individual lifetime risk.

Study on the temperature control mechanism of the tritium breeding blanket for CFETR

NASA Astrophysics Data System (ADS)

Liu, Changle; Qiu, Yang; Zhang, Jie; Zhang, Jianzhong; Li, Lei; Yao, Damao; Li, Guoqiang; Gao, Xiang; Wu, Songtao; Wan, Yuanxi

2017-12-01

The Chinese fusion engineering testing reactor (CFETR) will demonstrate tritium self- sufficiency using a tritium breeding blanket for the tritium fuel cycle. The temperature control mechanism (TCM) involves the tritium production of the breeding blanket and has an impact on tritium self-sufficiency. In this letter, the CFETR tritium target is addressed according to its missions. TCM research on the neutronics and thermal hydraulics issues for the CFETR blanket is presented. The key concerns regarding the blanket design for tritium production under temperature field control are depicted. A systematic theory on the TCM is established based on a multiplier blanket model. In particular, a closed-loop method is developed for the mechanism with universal function solutions, which is employed in the CFETR blanket design activity for tritium production. A tritium accumulation phenomenon is found close to the coolant in the blanket interior, which has a very important impact on current blanket concepts using water coolant inside the blanket. In addition, an optimal tritium breeding ratio (TBR) method based on the TCM is proposed, combined with thermal hydraulics and finite element technology. Meanwhile, the energy gain factor is adopted to estimate neutron heat deposition, which is a key parameter relating to the blanket TBR calculations, considering the structural factors. This work will benefit breeding blanket engineering for the CFETR reactor in the future.

TSTA Piping and Flame Arrestor Operating Experience Data

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

Cadwallader, Lee C.; Willms, R. Scott

The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility operated from 1984 to 2001, running a prototype fusion fuel processing loop with ~100 grams of tritium as well as small experiments. There have been several operating experience reports written on this facility’s operation and maintenance experience. This paper describes analysis of two additional components from TSTA, small diameter gas piping that handled small amounts of tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The operating experiences andmore » the component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.« less

Development of a tritium monitor combined with an electrochemical tritium pump using a proton conducting oxide

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

Tanaka, M.; Sugiyama, T.

2015-03-15

The detection of low level tritium is one of the key issues for tritium management in tritium handling facilities. Such a detection can be performed by tritium monitors based on proton conducting oxide technique. We tested a tritium monitoring system composed of a commercial proportional counter combined with an electrochemical hydrogen pump equipped with CaZr{sub 0.9}In{sub 0.1}O{sub 3-α} as proton conducting oxide. The hydrogen pump operated at 973 K under electrolysis conditions using tritiated water vapor (HTO). The proton conducting oxide extracts tritium molecules (HT) from HTO and tritium concentration is measured by the proportional counter. The advantage of themore » proposed tritium monitoring system is that it is able to convert HTO into molecular hydrogen.« less

Classification methodology for tritiated waste requiring interim storage

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

Cana, D.; Dall'ava, D.; Decanis, C.

2015-03-15

Fusion machines like the ITER experimental research facility will use tritium as fuel. Therefore, most of the solid radioactive waste will result not only from activation by 14 MeV neutrons, but also from contamination by tritium. As a consequence, optimizing the treatment process for waste containing tritium (tritiated waste) is a major challenge. This paper summarizes the studies conducted in France within the framework of the French national plan for the management of radioactive materials and waste. The paper recommends a reference program for managing this waste based on its sorting, treatment and packaging by the producer. It also recommendsmore » setting up a 50-year temporary storage facility to allow for tritium decay and designing future disposal facilities using tritiated radwaste characteristics as input data. This paper first describes this waste program and then details an optimized classification methodology which takes into account tritium decay over a 50-year storage period. The paper also describes a specific application for purely tritiated waste and discusses the set-up expected to be implemented for ITER decommissioning waste (current assumption). Comparison between this optimized approach and other viable detritiation techniques will be drawn. (authors)« less

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

NASA Astrophysics Data System (ADS)

Calderoni, P.; Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G.; Hatano, Y.; Hara, M.; Oya, Y.; Otsuka, T.; Katayama, K.; Konishi, S.; Noborio, K.; Yamamoto, Y.

2011-10-01

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

Sandia, California Tritium Research Laboratory transition and reutilization project

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

Garcia, T.B.

1997-02-01

This paper describes a project within Sandia National Laboratory to convert the shut down Tritium Research Laboratory into a facility which could be reused within the laboratory complex. In the process of decommissioning and decontaminating the facility, the laboratory was able to save substantial financial resources by transferring much existing equipment to other DOE facilities, and then expeditiously implementing a decontamination program which has resulted in the building being converted into laboratory space for new lab programs. This project of facility reuse has been a significant financial benefit to the laboratory.

Tritium handling experience at Atomic Energy of Canada Limited

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

Suppiah, S.; McCrimmon, K.; Lalonde, S.

2015-03-15

Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritiummore » powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.« less

RESULTS OF TRITIUM TRACKING AND GROUNDWATER MONITORING AT THE HANFORD SITE 200 AREA STATE APPROVED LAND DISPOSAL SITE FY2008

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

ERB DB

2008-11-19

The Hanford Site's 200 Area Effluent Treatment Facility (ETF) processes contaminated aqueous wastes derived from Hanford Site facilities. The treated wastewater occasionally contains tritium, which cannot be removed by the ETF prior to the wastewater being discharged to the 200 Area State-Approved Land Disposal Site (SALDS). During the first 11 months of fiscal year 2008 (FY08) (September 1, 2007, to July 31, 2008), approximately 75.15 million L (19.85 million gal) of water were discharged to the SALDS. Groundwater monitoring for tritium and other constituents, as well as water-level measurements, is required for the SALDS by State Waste Discharge Permit Numbermore » ST-4500 (Ecology 2000). The current monitoring network consists of three proximal (compliance) monitoring wells and nine tritium-tracking wells. Quarterly sampling of the proximal wells occurred in October 2007 and in January/February 2008, April 2008, and August 2008. The nine tritium-tracking wells, including groundwater monitoring wells located upgradient and downgradient of the SALDS, were sampled in January through April 2008. Water-level measurements taken in the three proximal SALDS wells indicate that a small groundwater mound is present beneath the facility, which is a result of operational discharges. The mound increased in FY08 due to increased ETF discharges from treating groundwater from extraction wells at the 200-UP-l Operable Unit and the 241-T Tank Farm. Maximum tritium activities increased by an order of magnitude at well 699-48-77A (to 820,000 pCi/L in April 2008) but remained unchanged in the other two proximal wells. The increase was due to higher quantities of tritium in wastewaters that were treated and discharged in FY07 beginning to appear at the proximal wells. The FY08 tritium activities for the other two proximal wells were 68,000 pCi/L at well 699-48-77C (October 2007) and 120,000 pCi/L at well 699-48-77D (October 2007). To date, no indications of a tritium incursion from the SALDS have been detected in the tritium-tracking wells. Concentrations of all chemical constituents were within Permit limits or were below method detection limits when sampled during FY08. A summary of the chemical constituent concentrations or method detection limits is provided in Table 3-2 in the main text discussion. This report presents the results of groundwater monitoring and tritium-tracking samples from the SALDS facility during FY08. Due to the 30-day laboratory turnaround for analysis of proximal well groundwater samples, this report addresses available date extending from August 1, 2007, through September 30, 2008 (August 2007 data were not included in the FY07 report). Updated background information, which is necessary to understand the results of the groundwater analyses, is also provided on facility operations. Interpretive discussions and recommendations for future monitoring are also provided, where possible.« less

A study of tritium in municipal solid waste leachate and gas

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

Mutch Jr, R. D.; Manhattan College, Riverdale, NY; Columbia Univ., New York, NY

2008-07-15

It has become increasingly clear in the last few years that the vast majority of municipal solid waste landfills produce leachate that contains elevated levels of tritium. The authors recently conducted a study of landfills in New York and New Jersey and found that the mean concentration of tritium in the leachate from ten municipal solid waste (MSW) landfills was 33,800 pCi/L with a peak value of 192,000 pCi/L. A 2003 study in California reported a mean tritium concentration of 99,000 pCi/L with a peak value of 304,000 pCi/L. Studies in Pennsylvania and the UK produced similar results. The USEPAmore » MCL for tritium is 20,000 pCi/L. Tritium is also manifesting itself as landfill gas and landfill gas condensate. Landfill gas condensate samples from landfills in the UK and California were found to have tritium concentrations as high as 54,400 and 513,000 pCi/L, respectively. The tritium found in MSW leachate is believed to derive principally from gaseous tritium lighting devices used in some emergency exit signs, compasses, watches, and even novelty items, such as 'glow stick' key chains. This study reports the findings of recent surveys of leachate from a number of municipal solid waste landfills, both open and closed, from throughout the United States and Europe. The study evaluates the human health and ecological risks posed by elevated tritium levels in municipal solid waste leachate and landfill gas and the implications to their safe management. We also assess the potential risks posed to solid waste management facility workers exposed to tritium-containing waste materials in transfer stations and other solid waste management facilities. (authors)« less

The use of dynamic modeling in assessing tritium phytoremediation

Treesearch

Karin T. Rebel; Susan J. Riha; John C. Seaman; Clinton d. Barton

2005-01-01

To minimize movement of tritium into surface waters at the Mixed Waste Management Facility at the Savannah River Site, tritiumcontaminated groundwater released to the surface along seeps in the hillside is being retained in a constructed pond and used to irrigate forest acreage that lies over the contaminated groundwater. Management of the application of tritium-...

An Overview of INEL Fusion Safety R&D Facilities

NASA Astrophysics Data System (ADS)

McCarthy, K. A.; Smolik, G. R.; Anderl, R. A.; Carmack, W. J.; Longhurst, G. R.

1997-06-01

The Fusion Safety Program at the Idaho National Engineering Laboratory has the lead for fusion safety work in the United States. Over the years, we have developed several experimental facilities to provide data for fusion reactor safety analyses. We now have four major experimental facilities that provide data for use in safety assessments. The Steam-Reactivity Measurement System measures hydrogen generation rates and tritium mobilization rates in high-temperature (up to 1200°C) fusion relevant materials exposed to steam. The Volatilization of Activation Product Oxides Reactor Facility provides information on mobilization and transport and chemical reactivity of fusion relevant materials at high temperature (up to 1200°C) in an oxidizing environment (air or steam). The Fusion Aerosol Source Test Facility is a scaled-up version of VAPOR. The ion-implanta-tion/thermal-desorption system is dedicated to research into processes and phenomena associated with the interaction of hydrogen isotopes with fusion materials. In this paper we describe the capabilities of these facilities.

2001 Evaluation of Tritium Removal & Mitigation Technologies for Waste Water Treatment

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

PENWELL, D.L.

2001-06-01

This report contains the 2001 biennial update evaluation of separation technologies and other mitigation techniques to control tritium in liquid effluents and groundwater at the Hanford site. A thorough literature review was completed, and national and international experts in the field of tritium separation and mitigation techniques were consulted. Current state-of-the-art technologies to address the control of tritium in wastewaters were identified and are described. This report was prepared to satisfy the Hanford Federal Facility Agreement and Consent Order Tri-Party Agreement, Milestone M-29-O5H (Ecology, EPA, and DOE 1996). Tritium separation and isolation technologies are evaluated on a biennial basis tomore » determine their feasibility for implementation for the control of Hanford site liquid effluents and groundwater to meet the US. Code of Federal Regulations (CFR), Title 40 CFR 141.16, drinking water maximum contaminant level (MCL) for tritium of 0.02 {mu} Ci/l ({approx}2 parts per quadrillion [10{sup -15}]) and/or DOE Order 5400.5 as low as reasonably achievable (ALARA) policy The objectives of this evaluation were to (1) status the development of potentially viable tritium separations technologies with regard to reducing tritium concentrations in current Hanford site process waters and existing groundwater to MCL levels and (2) status control methods to prevent the flow of tritiated water at concentrations greater than the MCL to the environment. Current tritium releases are in compliance with applicable US Environmental Protection Agency, Washington State Department of Ecology, and U.S. Department of Energy requirements under the Tri-Party Agreement. Advances in technologies for the separation of tritium from wastewater since the 1999 Hanford Site evaluation report include: (1) construction and testing of the Combined Industrial Reforming and Catalytic Exchange (CIRCE) Prototype Plant by Atomic Energy Canada Limited (AECL). The plant has a stage that uses the combined electrolysis catalytic exchange (CECE) and a stage that uses the bithermal hydrogen-waterprocess. The testing is still ongoing at the time of the development of this evaluation report, therefore, final results of the testing are not available; (2) further testing and a DOE sponsored American Society of Mechanical Engineers (ASME) peer review of a tritium resin separations process to remove tritium from wastewaters; and (3) completion of the design of the water detritiation system for the International Thermonuclear Experimental Reactor (ITER). The system uses a variation of the CECE process, and is designed to process 20 Whr of feed. The primary advance in technologies to control tritium migration in groundwater are the implementation of phytoremediation as a method of reducing the amount of tritium contaminated groundwater reaching the surface waters at Argonne National Laboratory, and initiation of a project for phytoremediation at the Savannah River Site.« less

Role of soil-to-leaf tritium transfer in controlling leaf tritium dynamics: Comparison of experimental garden and tritium-transfer model results.

PubMed

Ota, Masakazu; Kwamena, Nana-Owusua A; Mihok, Steve; Korolevych, Volodymyr

2017-11-01

Environmental transfer models assume that organically-bound tritium (OBT) is formed directly from tissue-free water tritium (TFWT) in environmental compartments. Nevertheless, studies in the literature have shown that measured OBT/HTO ratios in environmental samples are variable and generally higher than expected. The importance of soil-to-leaf HTO transfer pathway in controlling the leaf tritium dynamics is not well understood. A model inter-comparison of two tritium transfer models (CTEM-CLASS-TT and SOLVEG-II) was carried out with measured environmental samples from an experimental garden plot set up next to a tritium-processing facility. The garden plot received one of three different irrigation treatments - no external irrigation, irrigation with low tritium water and irrigation with high tritium water. The contrast between the results obtained with the different irrigation treatments provided insights into the impact of soil-to-leaf HTO transfer on the leaf tritium dynamics. Concentrations of TFWT and OBT in the garden plots that were not irrigated or irrigated with low tritium water were variable, responding to the arrival of the HTO-plume from the tritium-processing facility. In contrast, for the plants irrigated with high tritium water, the TFWT concentration remained elevated during the entire experimental period due to a continuous source of high HTO in the soil. Calculated concentrations of OBT in the leaves showed an initial increase followed by quasi-equilibration with the TFWT concentration. In this quasi-equilibrium state, concentrations of OBT remained elevated and unchanged despite the arrivals of the plume. These results from the model inter-comparison demonstrate that soil-to-leaf HTO transfer significantly affects tritium dynamics in leaves and thereby OBT/HTO ratio in the leaf regardless of the atmospheric HTO concentration, only if there is elevated HTO concentrations in the soil. The results of this work indicate that assessment models should be refined to consider the importance of soil-to-leaf HTO transfer to ensure that dose estimates are accurate and conservative. Copyright © 2017 Elsevier Ltd. All rights reserved.

Tritium dynamics in soils and plants grown under three irrigation regimes at a tritium processing facility in Canada.

PubMed

Mihok, S; Wilk, M; Lapp, A; St-Amant, N; Kwamena, N-O A; Clark, I D

2016-03-01

The dynamics of tritium released from nuclear facilities as tritiated water (HTO) have been studied extensively with results incorporated into regulatory assessment models. These models typically estimate organically bound tritium (OBT) for calculating public dose as OBT itself is rarely measured. Higher than expected OBT/HTO ratios in plants and soils are an emerging issue that is not well understood. To support the improvement of models, an experimental garden was set up in 2012 at a tritium processing facility in Pembroke, Ontario to characterize the circumstances under which high OBT/HTO ratios may arise. Soils and plants were sampled weekly to coincide with detailed air and stack monitoring. The design included a plot of native grass/soil, contrasted with sod and vegetables grown in barrels with commercial topsoil under natural rain and either low or high tritium irrigation water. Air monitoring indicated that the plume was present infrequently at concentrations of up to about 100 Bq/m(3) (the garden was not in a major wind sector). Mean air concentrations during the day on workdays (HTO 10.3 Bq/m(3), HT 5.8 Bq/m(3)) were higher than at other times (0.7-2.6 Bq/m(3)). Mean Tissue Free Water Tritium (TFWT) in plants and soils and OBT/HTO ratios were only very weakly or not at all correlated with releases on a weekly basis. TFWT was equal in soils and plants and in above and below ground parts of vegetables. OBT/HTO ratios in above ground parts of vegetables were above one when the main source of tritium was from high tritium irrigation water (1.5-1.8). Ratios were below one in below ground parts of vegetables when irrigated with high tritium water (0.4-0.6) and above one in vegetables rain-fed or irrigated with low tritium water (1.3-2.8). In contrast, OBT/HTO ratios were very high (9.0-13.5) when the source of tritium was mainly from the atmosphere. TFWT varied considerably through time as a result of SRBT's operations; OBT/HTO ratios showed no clear temporal pattern in above or below ground plant parts. Native soil after ∼20 years of operations at SRBT had high initial OBT that persisted through the growing season; little OBT formed in garden plot soil during experiments. High OBT in native soil appeared to be a signature of higher past releases at SRBT. This phenomenon was confirmed in soils obtained at another processing facility in Canada with a similar history. The insights into variation in OBT/HTO ratios found here are of regulatory interest and should be incorporated in assessment models to aid in the design of relevant environmental monitoring programs for OBT. Crown Copyright © 2016. Published by Elsevier Ltd. All rights reserved.

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.

Tritium contamination at EG&G/EM in North Las Vegas, Nevada

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

Sowell, C.V.; Arent, L.J.

1996-06-01

The tritium contamination discovered at the EG&G Energy Measurements (EG&G/EM) facility in North Las Vegas, Nevada, on 20 April 1995, could have been averted by good health physics practices and/or adequate management oversight. Scandium tritide (ScT{sub 3}) targets were installed for use in sealed tube neutron generators at EG&G/EM. In addition, EG&G/EM was also storing zirconium tritide (ZrT{sub 3}) and titanium tritide (TiT{sub 3}) foils. Since the targets were classified as sealed sources, the appropriate administrative and engineering control measures such as relocating targets/sources, air monitoring, bioassay, waste stream management, labeling/posting and training were not implemented. In all there weremore » six unreported incidents of tritium contamination from March 1994 to July 1995. Swipe surveys revealed areas exceeding the action level of 10,000 dpm/100 cm{sup 2} by up to three orders of magnitude. After reclassifying the targets as unsealed sources, a bioassay program was instituted, and the results were higher than expected for three employees. The doses assigned to the three individuals working in the contaminated area were 35, 58, and 61 mrem committed effective dose equivalent. Though the doses were low, the decontamination costs were in excess of $350,000.00. An investigation, was initiated by the U.S. Department of Energy Nevada Operations Office to analyze the events that led to the tritium contamination and recommend actions to prevent recurrence. Event and causal factor charting, Project Evaluation Tree (PET) analysis techniques, and root cause analysis, were used to evaluate management systems, causal sequences, and systems factors contributing to the tritium release.« less

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

Maienschein, J.L.; Garcia, F.; Garza, R.G.

Tritium-handling apparatus has been decontaminated as part of the shutdown of the LLNL Tritium Facility. Two stainless-steel gloveboxes that had been used to process lithium deuteride-tritide (LiDT) salt were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. Further surface decontamination was performed by scrubbing the interior with paper towels and ethyl alcohol or Swish{trademark}. The surface contamination, as shown by swipe surveys, was reduced from 4{times}10{sup 4}--10{sup 6} disintegrations per minute (dpm)/cm{sup 2} to 2{times}10{sup 2}--4{times}10{sup 4} dpm/cm{sup 2}. Details on the decontamination operation are provided. A seriesmore » of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given.« less

Performance testing of a prototype Pd-Ag diffuser

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

Morgan, G. A.; Hodge, B. J.

The fusion fuel cycle has gained significant attention over the last decade as interest in fusion programs has increased. One of the critical components of the fusion process is the tritium fuel cycle. The tritium fuel cycle is designed to supply and recycle process tritium at a specific throughput rate. One of the most important processes within the tritium fuel cycle is the clean-up of the of the process tritium. This step will initially separate the hydrogen isotopes (H2, D2, and T2) from the rest of the process gas using Pd-Ag diffusers or permeators. The Pd-Ag diffuser is an integralmore » component for any tritium purification system; whether part of the United States’ defense mission or fusion programs. Domestic manufacturers of Pd-Ag diffusers are extremely limited and only a few manufacturers exist. Johnson-Matthey (JM) Pd-Ag diffusers (permeators) have previously been evaluated for the separation of hydrogen isotopes from non-hydrogen gas species in the process. JM is no longer manufacturing Pd-Ag diffusers and a replacement vendor needs to be identified to support future needs. A prototype Pd-Ag diffuser has been manufactured by Power and Energy, and is considered a potential replacement for the JM diffuser for tritium service. New diffuser designs for a tritium facility for any fusion energy applications must be characterized by evaluating their operating envelope prior to installation in a tritium processing facility. The prototype Pd-Ag diffuser was characterized to determine the overall performance as a function of the permeation of hydrogen through the membrane. The tests described in this report consider the effects of feed gas compositions, feed flow rates, pump configuration and internal tube pressure on the permeation of H2 through the Pd-Ag tubes.« less

Techniques for tritium recovery from carbon flakes and dust at the JET active gas handling system

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

Gruenhagen, S.; Perevezentsev, A.; Brennan, P. D.

2008-07-15

Detritiation of highly tritium contaminated carbon and metal material used as first wall armour is a key issue for fusion machines like JET and ITER. Re-deposited carbon and hydrogen in the form of flakes and dust can lead to a build-up of the tritium inventory and therefore this material must be removed and processed. The high tritium concentration of the flake and dust material collected from the JET vacuum vessel makes it unsuitable for direct waste disposal without detritiation. A dedicated facility to process the tritiated carbon flake material and recover the tritium has been designed and built. In severalmore » test runs active material was successfully processed and de-tritiated in the new facility. Samples containing only carbon and hydrogen isotopes have been completely oxidized without any residue. Samples containing metallic impurities, e.g. beryllium, require longer processing times, adjusted processing parameters and yield an oxide residue. The detritiation factor was 2x10{sup 4}. In order to simulate in-vessel and ex-vessel detritiation techniques, the detritiation of a carbon flake sample by isotopic exchange in a hydrogen atmosphere was investigated. 2.8% of tritium was recovered by this means. (authors)« less

Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

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

Jolodosky, A.; Fratoni, M.

2014-11-20

Pre-conceptual fusion blanket designs require research and development to reflect important proposed changes in the design of essential systems, and the new challenges they impose on related fuel cycle systems. One attractive feature of using liquid lithium as the breeder and coolant is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. If the chemical reactivity of lithium could be overcome, the result would have a profound impact on fusion energy and associated safety basis.more » The overriding goal of this project is to develop a lithium-based alloy that maintains beneficial properties of lithium (e.g. high tritium breeding and solubility) while reducing overall flammability concerns. To minimize the number of alloy combinations that must be explored, only those alloys that meet certain nuclear performance metrics will be considered for subsequent thermodynamic study. The specific scope of this study is to evaluate the neutronics performance of lithium-based alloys in the blanket of an inertial confinement fusion (ICF) engine. The results of this study will inform the development of lithium alloys that would guarantee acceptable neutronics performance while mitigating the chemical reactivity issues of pure lithium.« less

Dismantling of the PETRA glove box: tritium contamination and inventory assessment

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

Wagner, R.

2015-03-15

The PETRA facility is the first installation in which experiments with tritium were carried out at the Tritium Laboratory Karlsruhe. After completion of two main experimental programs, the decommissioning of PETRA was initiated with the aim to reuse the glove box and its main still valuable components. A decommissioning plan was engaged to: -) identify the source of tritium release in the glove box, -) clarify the status of the main components, -) assess residual tritium inventories, and -) de-tritiate the components to be disposed of as waste. Several analytical techniques - calorimetry on small solid samples, wipe test followedmore » by liquid scintillation counting for surface contamination assessment, gas chromatography on gaseous samples - were deployed and cross-checked to assess the remaining tritium inventories and initiate the decommissioning process. The methodology and the main outcomes of the numerous different tritium measurements are presented and discussed. (authors)« less

Use of system code to estimate equilibrium tritium inventory in fusion DT machines, such as ARIES-AT and components testing facilities

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

C.P.C. Wong; B. Merrill

2014-10-01

ITER is under construction and will begin operation in 2020. This is the first 500 MWfusion class DT device, and since it is not going to breed tritium, it will consume most of the limited supply of tritium resources in the world. Yet, in parallel, DT fusion nuclear component testing machines will be needed to provide technical data for the design of DEMO. It becomes necessary to estimate the tritium burn-up fraction and corresponding initial tritium inventory and the doubling time of these machines for the planning of future supply and utilization of tritium. With the use of a systemmore » code, tritium burn-up fraction and initial tritium inventory for steady state DT machines can be estimated. Estimated tritium burn-up fractions of FNSF-AT, CFETR-R and ARIES-AT are in the range of 1–2.8%. Corresponding total equilibrium tritium inventories of the plasma flow and tritium processing system, and with the DCLL blanket option are 7.6 kg, 6.1 kg, and 5.2 kg for ARIES-AT, CFETR-R and FNSF-AT, respectively.« less

Evaluation of tritium release properties of advanced tritium breeders

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

Hoshino, T.; Ochiai, K.; Edao, Y.

2015-03-15

Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li{sub 2}TiO{sub 3}) advanced tritium breeders with excess Li (Li{sub 2+x}TiO{sub 3+y}) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Consideringmore » the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li{sub 2+x}TiO{sub 3+y} pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties similar to the Li{sub 2}TiO{sub 3} pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)« less

First results of radiation-driven, layered deuterium-tritium implosions with a 3-shock adiabat-shaped drive at the National Ignition Facility

DOE PAGES

Smalyuk, V. A.; Robey, H. F.; Döppner, T.; ...

2015-08-27

Radiation-driven, layered deuterium-tritium plastic capsule implosions were carried out using a new, 3-shock “adiabat-shaped” drive on the National Ignition Facility. The purpose of adiabat shaping is to use a stronger first shock, reducing hydrodynamic instability growth in the ablator. The shock can decay before reaching the deuterium-tritium fuel leaving it on a low adiabat and allowing higher fuel compression. The fuel areal density was improved by ~25% with this new drive compared to similar “high-foot” implosions, while neutron yield was improved by more than 4 times, compared to “low-foot” implosions driven at the same compression and implosion velocity.

Progress in the development of the MARBLE platform for studying thermonuclear burn in the presence of heterogeneous mix on OMEGA and the National Ignition Facility

NASA Astrophysics Data System (ADS)

Murphy, T. J.; Douglas, M. R.; Fincke, J. R.; Olson, R. E.; Cobble, J. A.; Haines, B. M.; Hamilton, C. E.; Lee, M. N.; Oertel, J. A.; Parra-Vasquez, N. A. G.; Randolph, R. B.; Schmidt, D. W.; Shah, R. C.; Smidt, J. M.; Tregillis, I. L.

2016-05-01

Mix of ablator material into fuel of an ICF capsule adds non-burning material, diluting the fuel and reducing burn. The amount of the reduction is dependent in part on the morphology of the mix. A probability distribution function (PDF) burn model has been developed [6] that utilizes the average concentration of mixed materials as well as the variance in this quantity across cells provided by the BHR turbulent transport model [3] and its revisions [4] to describe the mix in terms of a PDF of concentrations of fuel and ablator material, and provides the burn rate in mixed material. Work is underway to develop the MARBLE ICF platform for use on the National Ignition Facility in experiments to quantify the influence of heterogeneous mix on fusion burn. This platform consists of a plastic (CH) capsule filled with a deuterated plastic foam (CD) with a density of a few tens of milligrams per cubic centimeter, with tritium gas filling the voids in the foam. This capsule will be driven using x-ray drive on NIF, and the resulting shocks will induce turbulent mix that will result in the mixing of deuterium from the foam with the tritium gas. In order to affect the morphology of the mix, engineered foams with voids of diameter up to 100 microns will be utilized. The degree of mix will be determined from the ratio of DT to DD neutron yield. As the mix increases, the yield from reactions between the deuterium of the CD foam with tritium from the gas will increase. The ratio of DT to DD neutrons will be compared to a variation of the PDF burn model that quantifies reactions from initially separated reactants.

Glovebox stripper system tritium capture efficiency-literature review

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

James, D. W.; Poore, A. S.

2015-09-28

Glovebox Stripper Systems (GBSS) are intended to minimize tritium emissions from glovebox confinement systems in Tritium facilities. A question was raised to determine if an assumed 99% stripping (decontamination) efficiency in the design of a GBBS was appropriate. A literature review showed the stated 99% tritium capture efficiency used for design of the GBSS is reasonable. Four scenarios were indicated for GBSSs. These include release with a single or dual stage setup which utilizes either single-pass or recirculation for stripping purposes. Examples of single-pass as well as recirculation stripper systems are presented and reviewed in this document.

DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE

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

Blanton, P.

A new Bulk Tritium Shipping Package (BTSP) was designed by the Savannah River National Laboratory to be a replacement for a package that has been used to ship tritium in a variety of content configurations and forms since the early 1970s. The BTSP was certified by the National Nuclear Safety Administration in 2011 for shipments of up to 150 grams of Tritium. Thirty packages were procured and are being delivered to various DOE sites for operational use. This paper summarizes the design features of the BTSP, as well as associated engineered material improvements. Fabrication challenges encountered during production are discussedmore » as well as fielding requirements. Current approved tritium content forms (gas and tritium hydrides), are reviewed, as well as, a new content, tritium contaminated water on molecular sieves. Issues associated with gas generation will also be discussed.« less

Production of highly tritiated water for tritium exposure studies

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

Muirhead, C.; Pilatzke, K.; Tripple, A.

2015-03-15

Tritium Facility staff at Chalk River Laboratories (CRL) have successfully prepared highly tritiated water for use in radiation resistance of PEM (Proton Exchange Membrane-based)electrolyser membrane. The goal of System A was to convert a known amount of elemental tritium (HT) into tritiated water vapour using a copper(II) oxide bed, and to condense the tritiated water vapour into a known amount of chilled heavy water (D{sub 2}O). The conversion and capture of tritium using this system is close to 100%. The goal of System B was to transfer tritiated water from the containment vessel to an exposure vessel (experiment) in amore » controlled and safe manner. System B is based on the pushing of D{sub 2}0 with low-pressure argon carrier gas to a calibrated volume and then to the exposure vessel. A method for delivering a known and controlled amount of tritiated water has been successfully demonstrated at CRL. Using both systems Tritium Facility staff have made and distributed highly tritiated water in a safe and controlled manner. This paper focuses on how the tritiated water was produced and dispensed to the experiment.« less

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

Hayashi, T.; Nakamura, H.; Kawamura, Y.

JAEA (Japan Atomic Energy Agency) manages 2 tritium handling laboratories: Tritium Processing Laboratory (TPL) in Tokai and DEMO-RD building in Rokkasho. TPL has been accumulating a gram level tritium safety handling experiences without any accidental tritium release to the environment for more than 25 years. Recently, our activities have focused on 3 categories, as follows. First, the development of a detritiation system for ITER. This task is the demonstration test of a wet Scrubber Column (SC) as a pilot scale (a few hundreds m{sup 3}/h of processing capacity). Secondly, DEMO-RD tasks are focused on investigating the general issues required formore » DEMO-RD design, such as structural materials like RAFM (Reduced Activity Ferritic/Martensitic steels) and SiC/SiC, functional materials like tritium breeder and neutron multiplier, and tritium. For the last 4 years, we have spent a lot of time and means to the construction of the DEMO-RD facility and to its licensing, so we have just started the actual research program with tritium and other radioisotopes. This tritium task includes tritium accountancy, tritium basic safety research such as tritium interactions with various materials, which will be used for DEMO-RD and durability. The third category is the recovery work from the Great East Japan earthquake (2011 earthquake). It is worth noting that despite the high magnitude of the earthquake, TPL was able to confine tritium properly without any accidental tritium release.« less

Tritium

DTIC Science & Technology

2011-11-01

fusion energy -production processes of the particular type of reactor using a lithium (Li) blanket or related alloys such as the Pb-17Li eutectic. As such, tritium breeding is intimately connected with energy production, thermal management, radioactivity management, materials properties, and mechanical structures of any plausible future large-scale fusion power reactor. JASON is asked to examine the current state of scientific knowledge and engineering practice on the physical and chemical bases for large-scale tritium

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

NASA Astrophysics Data System (ADS)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

EFFECT OF TRITIUM AND DECAY HELIUM ON WELDMENT FRACTURE TOUGHNESS

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

Morgan, M; Scott West, S; Michael Tosten, M

2006-09-26

The fracture toughness data collected in this study are needed to assess the long-term effects of tritium and its decay product on tritium reservoirs. The results show that tritium and decay helium have negative effects on the fracture toughness properties of stainless steel and its weldments. The data and report from this study has been included in a material property database for use in tritium reservoir modeling efforts like the Technology Investment Program ''Lifecycle Engineering for Tritium Reservoirs''. A number of conclusions can be drawn from the data: (1) For unexposed Type 304L stainless steel, the fracture toughness of weldmentsmore » was two to three times higher than the base metal toughness. (2) Tritium exposure lowered the fracture toughness properties of both base metals and weldments. This was characterized by lower J{sub Q} values and lower J-da curves. (3) Tritium-exposed-and-aged base metals and weldments had lower fracture toughness values than unexposed ones but still retained good toughness properties.« less

Feasibility study of a magnetic fusion production reactor

NASA Astrophysics Data System (ADS)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.

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

Flach, G. P.; Whiteside, T. S.

The E-Area Vadose Zone Monitoring System (VZMS) includes lysimeter sampling points at many locations alongside and angling beneath the Engineered Trench #1 (ET1) disposal unit footprint. The sampling points for ET1 were selected for this study because collectively they showed consistently higher tritium (H-3) concentrations than lysimeters associated with other trench units. The VZMS tritium dataset for ET1 from 2001 through 2015 comprises concentrations at or near background levels at approximately half of locations through time, concentrations up to about 600 pCi/mL at a few locations, and concentrations at two locations that have exceeded 1000 pCi/mL. The highest three valuesmore » through 2015 were 6472 pCi/mL in 2014 and 4533 pCi/mL in 2013 at location VL-17, and 3152 pCi/mL in 2007 at location VL-15. As a point of reference, the drinking water standard for tritium and a DOE Order 435.1 performance objective in the saturated zone at the distant 100-meter facility perimeter is 20 pCi/mL. The purpose of this study is to assess whether these elevated concentrations are indicative of a general trend that could challenge 2008 E-Area Performance Assessment (PA) conclusions, or are isolated perturbations that when considered in the context of an entire disposal unit would support PA conclusions.« less

Letter Report for Analytical Results for five Swipe Samples from the Northern Biomedical Research Facility, Muskegon Michigan

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

Ivey, Wade

Oak Ridge Associated Universities (ORAU), under the Oak Ridge Institute for Science and Education (ORISE) contract, received five swipe samples on December 10, 2013 from the Northern Biomedical Research Facility in Norton Shores, Michigan. The samples were analyzed for tritium and carbon-14 according to the NRC Form 303 supplied with the samples. The sample identification numbers are presented in Table 1 and the tritium and carbon-14 results are provided in Table 2. The pertinent procedure references are included with the data tables.

Concentrations of tritium and strontium-90 in water from selected wells at the Idaho National Engineering Laboratory after purging one, two, and three borehole volumes

USGS Publications Warehouse

Bartholomay, R.C.

1993-01-01

Water from 11 wells completed in the Snake River Plain aquifer at the Idaho National Engineering Laboratory was sampled as part of the U.S. Geological Survey's quality assurance program to determine the effect of purging different borehole volumes on tritium and strontium-90 concentrations. Wells were selected for sampling on the basis of the length of time it took to purge a borehole volume of water. Samples were collected after purging one, two, and three borehole volumes. The U.S. Department of Energy's Radiological and Environmental Sciences Laboratory provided analytical services. Statistics were used to determine the reproducibility of analytical results. The comparison between tritium and strontium-90 concentrations after purging one and three borehole volumes and two and three borehole volumes showed that all but two sample pairs with defined numbers were in statistical agreement. Results indicate that concentrations of tritium and strontium-90 are not affected measurably by the number of borehole volumes purged.

Tritium Concentrations in Environmental Samples and Transpiration Rates from the Vicinity of Mary's Branch Creek and Background Areas, Barnwell, South Carolina, 2007-2009

USGS Publications Warehouse

Vroblesky, Don A.; Canova, Judy L.; Bradley, Paul M.; Landmeyer, James E.

2009-01-01

Tritium in groundwater from a low-level radioactive waste disposal facility near Barnwell, South Carolina, is discharging to Mary's Branch Creek. The U.S. Geological Survey conducted an investigation from 2007 to 2009 to examine the tritium concentration in trees and air samples near the creek and in background areas, in groundwater near the creek, and in surface water from the creek. Tritium was found in trees near the creek, but not in trees from background areas or from sites unlikely to be in direct root contact with tritium-contaminated groundwater. Tritium was found in groundwater near the creek and in the surface water of the creek. Analysis of tree material has the potential to be a useful tool in locating shallow tritium-contaminated groundwater. A tritium concentration of 1.4 million picocuries per liter was measured in shallow groundwater collected near a tulip poplar located in an area of tritium-contaminated groundwater discharge. Evapotranspiration rates from the tree and tritium concentrations in water extracted from tree cores indicate that during the summer, this tulip poplar may remove more than 17.1 million picocuries of tritium per day from the groundwater that otherwise would discharge to Mary's Branch Creek. Analysis of air samples near the tree showed no evidence that the transpirative release of tritium to the air created a vapor hazard in the forest.

Low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic

DOE PAGES

Pawelko, R. J.; Shimada, M.; Katayama, K.; ...

2015-11-28

This paper describes a new experimental system designed to investigate tritium mass transfer properties in materials important to fusion technology. Experimental activities were carried out at the Safety and Tritium Applied Research (STAR) facility located at the Idaho National Laboratory (INL). The tritium permeation measurement system was developed as part of the Japan/US TITAN collaboration to investigate tritium mass transfer properties in liquid lead lithium eutectic (LLE) alloy. The experimental system is configured to measure tritium mass transfer properties at low tritium partial pressures. Initial tritium permeation scoping tests were conducted on a 1 mm thick α-Fe plate to determinemore » operating parameters and to validate the experimental technique. A second series of permeation tests was then conducted with the α-Fe plate covered with an approximately 8.5 mm layer of liquid lead lithium eutectic alloy (α-Fe/LLE). We present preliminary tritium permeation data for α-Fe and α-Fe/LLE at temperatures between 400 and 600°C and at tritium partial pressures between 1.7E-3 and 2.5 Pa in helium. Preliminary results for the α-Fe plate and α-Fe/LLE indicate that the data spans a transition region between the diffusion-limited regime and the surface-limited regime. In conclusion, additional data is required to determine the existence and range of a surface-limited regime.« less

Movement of a tritium plume in shallow groundwater at a legacy low-level radioactive waste disposal site in eastern Australia.

PubMed

Hughes, C E; Cendón, D I; Harrison, J J; Hankin, S I; Johansen, M P; Payne, T E; Vine, M; Collins, R N; Hoffmann, E L; Loosz, T

2011-10-01

Between 1960 and 1968 low-level radioactive waste was buried in a series of shallow trenches near the Lucas Heights facility, south of Sydney, Australia. Groundwater monitoring carried out since the mid 1970s indicates that with the exception of tritium, no radioactivity above typical background levels has been detected outside the immediate vicinity of the trenches. The maximum tritium level detected in ground water was 390 kBq/L and the median value was 5400 Bq/L, decay corrected to the time of disposal. Since 1968, a plume of tritiated water has migrated from the disposal trenches and extends at least 100 m from the source area. Tritium in rainfall is negligible, however leachate from an adjacent and fill represents a significant additional tritium source. Study data indicate variation in concentration levels and plume distribution in response to wet and dry climatic periods and have been used to determine pathways for tritium migration through the subsurface.

Normetex Pump Alternatives Study

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

Clark, Elliot A.

2013-04-25

A mainstay pump for tritium systems, the Normetex scroll pump, is currently unavailable because the Normetex company went out of business. This pump was an all-metal scroll pump that served tritium processing facilities very well. Current tritium system operators are evaluating replacement pumps for the Normetex pump and for general used in tritium service. An all-metal equivalent alternative to the Normetex pump has not yet been identified. 1. The ideal replacement tritium pump would be hermetically sealed and contain no polymer components or oils. Polymers and oils degrade over time when they contact ionizing radiation. 2. Halogenated polymers (containing fluorine,more » chlorine, or both) and oils are commonly found in pumps. These materials have many properties that surpass those of hydrocarbon-based polymers and oils, including thermal stability (higher operating temperature) and better chemical resistance. Unfortunately, they are less resistant to degradation from ionizing radiation than hydrocarbon-based materials (in general). 3. Polymers and oils can form gaseous, condensable (HF, TF), liquid, and solid species when exposed to ionizing radiation. For example, halogenated polymers form HF and HCl, which are extremely corrosive upon reaction with water. If a pump containing polymers or oils must be used in a tritium system, the system must be designed to be able to process the unwanted by-products. Design features to mitigate degradation products include filters and chemical or physical traps (eg. cold traps, oil traps). 4. Polymer components can work in tritium systems, but must be replaced regularly. Polymer components performance should be monitored or be regularly tested, and regular replacement of components should be viewed as an expected normal event. A radioactive waste stream must be established to dispose of used polymer components and oil with an approved disposal plan developed based on the facility location and its regulators. Polymers have varying resistances to ionizing radiation - aromatic polymers such as polyimide Vespel (TM) and the elastomer EPDM (ethylene propylene diene monomer) have been found to be more resistant to degradation in tritium than other polymers. This report presents information to help select replacement pumps for Normetex pumps in tritium systems. Several pumps being considered as Normetex replacement pumps are discussed.« less

Progress in the development of the MARBLE platform for studying thermonuclear burn in the presence of heterogeneous mix on OMEGA and the National Ignition Facility

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

Murphy, T. J.; Douglas, M. R.; Fincke, J. R.

Mix of ablator material into fuel of an ICF capsule adds non-burning material, diluting the fuel and reducing burn. The amount of the reduction is dependent in part on the morphology of the mix. A probability distribution function (PDF) burn model has been developed [6] that utilizes the average concentration of mixed materials as well as the variance in this quantity across cells provided by the BHR turbulent transport model [3] and its revisions [4] to describe the mix in terms of a PDF of concentrations of fuel and ablator material, and provides the burn rate in mixed material. Workmore » is underway to develop the MARBLE ICF platform for use on the National Ignition Facility in experiments to quantify the influence of heterogeneous mix on fusion burn. This platform consists of a plastic (CH) capsule filled with a deuterated plastic foam (CD) with a density of a few tens of milligrams per cubic centimeter, with tritium gas filling the voids in the foam. This capsule will be driven using x-ray drive on NIF, and the resulting shocks will induce turbulent mix that will result in the mixing of deuterium from the foam with the tritium gas. In order to affect the morphology of the mix, engineered foams with voids of diameter up to 100 microns will be utilized. The degree of mix will be determined from the ratio of DT to DD neutron yield. As the mix increases, the yield from reactions between the deuterium of the CD foam with tritium from the gas will increase. Lastly, the ratio of DT to DD neutrons will be compared to a variation of the PDF burn model that quantifies reactions from initially separated reactants.« less

Progress in the development of the MARBLE platform for studying thermonuclear burn in the presence of heterogeneous mix on OMEGA and the National Ignition Facility

DOE PAGES

Murphy, T. J.; Douglas, M. R.; Fincke, J. R.; ...

2016-05-26

Mix of ablator material into fuel of an ICF capsule adds non-burning material, diluting the fuel and reducing burn. The amount of the reduction is dependent in part on the morphology of the mix. A probability distribution function (PDF) burn model has been developed [6] that utilizes the average concentration of mixed materials as well as the variance in this quantity across cells provided by the BHR turbulent transport model [3] and its revisions [4] to describe the mix in terms of a PDF of concentrations of fuel and ablator material, and provides the burn rate in mixed material. Workmore » is underway to develop the MARBLE ICF platform for use on the National Ignition Facility in experiments to quantify the influence of heterogeneous mix on fusion burn. This platform consists of a plastic (CH) capsule filled with a deuterated plastic foam (CD) with a density of a few tens of milligrams per cubic centimeter, with tritium gas filling the voids in the foam. This capsule will be driven using x-ray drive on NIF, and the resulting shocks will induce turbulent mix that will result in the mixing of deuterium from the foam with the tritium gas. In order to affect the morphology of the mix, engineered foams with voids of diameter up to 100 microns will be utilized. The degree of mix will be determined from the ratio of DT to DD neutron yield. As the mix increases, the yield from reactions between the deuterium of the CD foam with tritium from the gas will increase. Lastly, the ratio of DT to DD neutrons will be compared to a variation of the PDF burn model that quantifies reactions from initially separated reactants.« less

Concentrations of tritium and strontium-90 in water from selected wells at the Idaho National Engineering Laboratory after purging one, two, and three borehole volumes

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

Bartholomay, R.C.

1993-12-31

Water from 11 wells completed in the Snake River Plain aquifer at the Idaho National Engineering Laboratory was sampled as Part of the US. Geological Survey`s quality assurance program to determine the effect of Purging different borehole volumes on tritium and strontium-90 concentrations. Wells were selected for sampling on the basis of the length of time it took to purge a borehole volume of water. Samples were collected after purging one, two, and three borehole volumes. The US Department of Energy`s Radiological and Environmental Sciences Laboratory provided analytical services. Statistics were used to determine the reproducibility of analytical results. Themore » comparison between tritium and strontium-90 concentrations after purging one and three borehole volumes and two and three borehole volumes showed that all but two sample pairs with defined numbers were in statistical agreement. Results indicate that concentrations of tritium and strontium-90 are not affected measurably by the number of borehole volumes purged.« less

A Next Generation Digital Counting System For Low-Level Tritium Studies (Project Report)

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

Bowman, P.

2016-10-03

Since the early seventies, SRNL has pioneered low-level tritium analysis using various nuclear counting technologies and techniques. Since 1999, SRNL has successfully performed routine low-level tritium analyses with counting systems based on digital signal processor (DSP) modules developed in the late 1990s. Each of these counting systems are complex, unique to SRNL, and fully dedicated to performing routine tritium analyses of low-level environmental samples. It is time to modernize these systems due to a variety of issues including (1) age, (2) lack of direct replacement electronics modules and (3) advances in digital signal processing and computer technology. There has beenmore » considerable development in many areas associated with the enterprise of performing low-level tritium analyses. The objective of this LDRD project was to design, build, and demonstrate a Next Generation Tritium Counting System (NGTCS), while not disrupting the routine low-level tritium analyses underway in the facility on the legacy counting systems. The work involved (1) developing a test bed for building and testing new counting system hardware that does not interfere with our routine analyses, (2) testing a new counting system based on a modern state of the art DSP module, and (3) evolving the low-level tritium counter design to reflect the state of the science.« less

Tritium Management Loop Design Status

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

Rader, Jordan D.; Felde, David K.; McFarlane, Joanna

This report summarizes physical, chemical, and engineering analyses that have been done to support the development of a test loop to study tritium migration in 2LiF-BeF2 salts. The loop will operate under turbulent flow and a schematic of the apparatus has been used to develop a model in Mathcad to suggest flow parameters that should be targeted in loop operation. The introduction of tritium into the loop has been discussed as well as various means to capture or divert the tritium from egress through a test assembly. Permeation was calculated starting with a Modelica model for a transport through amore » nickel window into a vacuum, and modifying it for a FLiBe system with an argon sweep gas on the downstream side of the permeation interface. Results suggest that tritium removal with a simple tubular permeation device will occur readily. Although this system is idealized, it suggests that rapid measurement capability in the loop may be necessary to study and understand tritium removal from the system.« less

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

BENNETT,D.B.; PAQUETTE,D.E.; KLAUS,K.

The BNL water supply system meets all water quality standards and has sufficient pumping and storage capacity to meet current and anticipated future operational demands. Because BNL's water supply is drawn from the shallow Upper Glacial aquifer, BNL's source water is susceptible to contamination. The quality of the water supply is being protected through (1) a comprehensive program of engineered and operational controls of existing aquifer contamination and potential sources of new contamination, (2) groundwater monitoring, and (3) potable water treatment. The BNL Source Water Assessment found that the source water for BNL's Western Well Field (comprised of Supply Wellsmore » 4, 6, and 7) has relatively few threats of contamination and identified potential sources are already being carefully managed. The source water for BNL's Eastern Well Field (comprised of Supply Wells 10, 11, and 12) has a moderate number of threats to water quality, primarily from several existing volatile organic compound and tritium plumes. The g-2 Tritium Plume and portions of the Operable Unit III VOC plume fall within the delineated source water area for the Eastern Well Field. In addition, portions of the much slower migrating strontium-90 plumes associated with the Brookhaven Graphite Research Reactor, Waste Concentration Facility and Building 650 lie within the Eastern source water area. However, the rate of travel in the aquifer for strontium-90 is about one-twentieth of that for tritium and volatile organic compounds. The Laboratory has been carefully monitoring plume migration, and has made adjustments to water supply operations. Although a number of BNL's water supply wells were impacted by VOC contamination in the late 1980s, recent routine analysis of water samples from BNL's supply wells indicate that no drinking water standards have been reached or exceeded. The high quality of the water supply strongly indicates that the operational and engineered controls implemented over the past ten years have effectively protected the quality of the water supply.« less

Tritium resources available for fusion reactors

NASA Astrophysics Data System (ADS)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future fusion reactors.

MFTF-. cap alpha. + T progress report

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

Nelson, W.D.

1985-04-01

Early in FY 1983, several upgrades of the Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory (LLNL) were proposed to the fusion community. The one most favorably received was designated MFTF-..cap alpha..+T. The engineering design of this device, guided by LLNL, has been a principal activity of the Fusion Engineering Design Center during FY 1983. This interim progress report represents a snapshot of the device design, which was begun in FY 1983 and will continue for several years. The report is organized as a complete design description. Because it is an interim report, some parts are incomplete; theymore » will be supplied as the design study proceeds. As described in this report, MFTF-..cap alpha..+T uses existing facilities, many MFTF-B components, and a number of innovations to improve on the physics parameters of MFTF-B. It burns deuterium-tritium and has a central-cell Q of 2, a wall loading GAMMA/sub n/ of 2 MW/m/sup 2/ (with a central-cell insert module), and an availability of 10%. The machine is fully shielded, allows hands-on maintenance of components outside the vacuum vessel 24 h after shutdown, and has provisions for repair of all operating components.« less

A neutron activation spectrometer and neutronic experimental platform for the National Ignition Facility (invited)

NASA Astrophysics Data System (ADS)

Yeamans, C. B.; Gharibyan, N.

2016-11-01

At the National Ignition Facility, the diagnostic instrument manipulator-based neutron activation spectrometer is used as a diagnostic of implosion performance for inertial confinement fusion experiments. Additionally, it serves as a platform for independent neutronic experiments and may be connected to fast recording systems for neutron effect tests on active electronics. As an implosion diagnostic, the neutron activation spectrometers are used to quantify fluence of primary DT neutrons, downscattered neutrons, and neutrons above the primary DT neutron energy created by reactions of upscattered D and T in flight. At a primary neutron yield of 1015 and a downscattered fraction of neutrons in the 10-12 MeV energy range of 0.04, the downscattered neutron fraction can be measured to a relative uncertainty of 8%. Significant asymmetries in downscattered neutrons have been observed. Spectrometers have been designed and fielded to measure the tritium-tritium and deuterium-tritium neutron outputs simultaneously in experiments using DT/TT fusion ratio as a direct measure of mix of ablator into the gas.

A neutron activation spectrometer and neutronic experimental platform for the National Ignition Facility (invited).

PubMed

Yeamans, C B; Gharibyan, N

2016-11-01

At the National Ignition Facility, the diagnostic instrument manipulator-based neutron activation spectrometer is used as a diagnostic of implosion performance for inertial confinement fusion experiments. Additionally, it serves as a platform for independent neutronic experiments and may be connected to fast recording systems for neutron effect tests on active electronics. As an implosion diagnostic, the neutron activation spectrometers are used to quantify fluence of primary DT neutrons, downscattered neutrons, and neutrons above the primary DT neutron energy created by reactions of upscattered D and T in flight. At a primary neutron yield of 10 15 and a downscattered fraction of neutrons in the 10-12 MeV energy range of 0.04, the downscattered neutron fraction can be measured to a relative uncertainty of 8%. Significant asymmetries in downscattered neutrons have been observed. Spectrometers have been designed and fielded to measure the tritium-tritium and deuterium-tritium neutron outputs simultaneously in experiments using DT/TT fusion ratio as a direct measure of mix of ablator into the gas.

The AGHS at JET and preparations for a future DT campaign

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

Smith, R.; JET-EFDA, Culham Science Centre, Abingdon

2015-03-15

The Active Gas Handling System (AGHS) at JET is a unique facility enabling JET to perform reactor like, DT operations. As a future DT experimental campaign (DTE2) is scheduled for 2017 this paper provides a brief overview of the AGHS and a summary of ongoing work supporting the currently JET experimental campaign. In order to improve tritium accountancy a solid state based detector for tritium is being developed. Another important upgrade concerns tritium injection, 4 existing GIMs (Tritium Gas Introduction Module) will inject a mix of D and T rather than T{sub 2} in the divertor region rather than inmore » the torus mid plane enabling a far better control and variability of the introduction of tritium into the plasma. An overview of the scale of DTE2 is included as well as an example of some of the upgrades currently being undertaken to fully exploit the learning opportunities for ITER and DEMO DTE2 provides. (authors)« less

Thermal Release of 3He from Tritium Aged LaNi 4.25Al 0.75 Hydride

DOE PAGES

Staack, Gregory C.; Crowder, Mark L.; Klein, James E.

2015-02-01

Recently, the demand for He-3 has increased dramatically due to widespread use in nuclear nonproliferation, cryogenic, and medical applications. Essentially all of the world’s supply of He-3 is created by the radiolytic decay of tritium. The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 in the tritium process to store hydrogen isotopes. The vast majority of He-3 “born” from tritium stored in LANA.75 is trapped in the hydride metal matrix. The SRS-TF has multiple LANA.75 tritium storage beds that have been retired from service with significant quantities of He-3 trapped in the metal. To support He-3 recovery, the Savannah Rivermore » National Laboratory (SRNL) conducted thermogravimetric analysis coupled with mass spectrometry (TGA-MS) on a tritium aged LANA.75 sample. TGA-MS testing was performed in an argon environment. Prior to testing, the sample was isotopically exchanged with deuterium to reduce residual tritium and passivated with air to alleviate pyrophoric concerns associated with handling the material outside of an inert glovebox. Analyses indicated that gas release from this sample was bimodal, with peaks near 220 and 490°C. The first peak consisted of both He-3 and residual hydrogen isotopes, the second was primarily He-3. The bulk of the gas was released by 600 °C« less

Tritium Plume Dynamics in the Shallow Unsaturated Zone Adjacent to an Arid Waste Disposal Facility

NASA Astrophysics Data System (ADS)

Maples, S.; Andraski, B. J.; Stonestrom, D. A.; Cooper, C. A.; Michel, R. L.; Pohll, G. M.

2012-12-01

Previous studies at the U.S. Geological Survey's Amargosa Desert Research Site (ADRS) in southern Nevada have documented two plumes of tritiated water-vapor (3HHOg) adjacent to a closed, commercial low-level radioactive waste disposal facility. Wastes were disposed on-site from 1962-92. Tritium has moved long distances (> 400 m) through a shallow (1-2-m depth) dry gravelly layer—orders of magnitude further than anticipated by standard transport models. Geostatistical methods, spatial moment analyses and tritium flux calculations were applied to assess shallow plume dynamics. A grid-based plant-water sampling method was utilized to infer detailed, field-scale 3HHOg concentrations at 5-yr intervals during 2001-11. Results indicate that gravel-layer 3HHOg mass diminished faster than would be expected from radioactive decay (~70% in 10 yr). Both plumes exhibited center-of-mass stability, suggesting that bulk-plume movement is minimal during the period of study. Nonetheless, evidence of localized lateral advancement along some margins, combined with increases in the spatial covariance of concentration distribution, indicates intra-plume mass redistribution is ongoing. Previous studies have recognized that vertical movement of tritiated water from sub-root-zone gravel into the root-zone contributes to atmospheric release via evapotranspiration. Estimates of lateral and vertical tritium fluxes during the study period indicate (1) vertical tritiated water fluxes were dominated by diffusive-vapor fluxes (> 90%), and (2) vertical diffusive-vapor fluxes were roughly an order of magnitude greater than lateral diffusive-vapor fluxes. This behavior highlights the importance of the atmosphere as a tritium sink. Estimates of cumulative vertical diffusive-vapor flux and radioactive decay with time were comparable to observed declines in total shallow plume mass with time. This suggests observed changes in plume mass may (1) be attributed, in considerable part, to these removal mechanisms, and (2) appreciable input from the adjacent disposal facility is not occurring at this time.

Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

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

Jolodosky, A.; Fratoni, M.

Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within amore » low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding, low electrical conductivity and therefore low MHD pressure drop, low chemical reactivity, and extremely low tritium inventory; the addition of sodium (FLiNaBe) has been considered because it retains the properties of FliBe but also lowers the melting point. Although many of these blanket concepts are promising, challenges still remain. The limited amount of beryllium available poses a problem for ceramic breeders such as the HCPB. FLiBe and FLiNaBe are highly viscous and have a low thermal conductivity. Lithium lead possesses a poor thermal conductivity which can cause problems in both DCLL and LiPb blankets. Additionally, the tritium permeation from these two blankets into plant components can be a problem and must be reduced. Consequently, Lawrence Livermore National Laboratory (LLNL) is attempting to develop a lithium-based alloy—most likely a ternary alloy—which maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) while reducing overall flammability concerns for use in the blanket of an inertial fusion energy (IFE) power plant. The LLNL concept employs inertial confinement fusion (ICF) through the use of lasers aimed at an indirect-driven target composed of deuterium-tritium fuel. The fusion driver/target design implements the same physics currently experimented at the National Ignition Facility (NIF). The plant uses lithium in both the primary coolant and blanket; therefore, lithium-related hazards are of primary concern. Although reducing chemical reactivity is the primary motivation for the development of new lithium alloys, the successful candidates will have to guarantee acceptable performance in all their functions. The scope of this study is to evaluate the neutronics performance of a large number of lithium-based alloys in the blanket of the IFE engine and assess their properties upon activation. This manuscript is organized as follows: Section 12 presents the models and methodologies used for the analysis; Section 3 discusses the results; Section 4 summarizes findings and future work.« less

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

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

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of themore » important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.« less

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

NASA Astrophysics Data System (ADS)

Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

2016-08-01

Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

Tritium concentrations in flow from selected springs that discharge to the Snake River, Twin Falls-Hagerman area, Idaho

USGS Publications Warehouse

Mann, L.J.

1989-01-01

Concern has been expressed that some of the approximately 30,900 curies of tritium disposed to the Snake River Plain aquifer from 1952 to 1988 at the INEL (Idaho National Engineering Laboratory) have migrated to springs discharging to the Snake River in the Twin Falls-Hagerman area. To document tritium concentrations in springflow, 17 springs were sampled in November 1988 and 19 springs were sampled in March 1989. Tritium concentrations were less than the minimum detectable concentration of 0.5 pCi/mL (picocuries/mL) in November 1988 and less than the minimum detectable concentration of 0.2 pCi/mL in March 1989; the minimum detectable concentration was smaller in March 1989 owing to a longer counting time in the liquid scintillation system. The maximum contaminant level of tritium in drinking water as established by the U.S. Environmental Protection Agency is 20 pCi/mL. U.S. Environmental Protection Agency sample analyses indicate that the tritium concentration has decreased in the Snake River near Buhl since the 1970's. In 1974-79, tritium concentrations were less than 0.3 +/-0.2 pCi/mL in 3 of 20 samples; in 1983-88, 17 of 23 samples contained less than 0.3 +/-0.2 pCi/mL of tritium; the minimum detectable concentration is 0.2 pCi/mL. On the basis of decreasing tritium concentrations in the Snake River, their correlation to cessation of atmospheric weapons tests tritium concentrations in springflow less than the minimum detectable concentration, and the distribution of tritium in groundwater at the INEL, aqueous disposal of tritium at the INEL has had no measurable effect on tritium concentrations in springflow from the Snake River Plain aquifer and in the Snake River near Buhl. (USGS)

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

Takeishi, T.; Kotoh, K.; Kawabata, Y.

The existence of tritium-contaminated oils from vacuum pumps used in tritium facilities, is becoming an important issue since there is no disposal way for tritiated waste oils. On recovery of tritiated water vapor in gas streams, it is well-known that the isotope exchange reaction between the gas phase and the liquid phase occurs effectively at room temperature. We have carried out experiments using bubbles to examine the tritium contamination and decontamination of a volume of rotary-vacuum-pump oil. The contamination of the pump oil was made by bubbling tritiated water vapor and tritiated hydrogen gas into the oil. Subsequently the decontaminationmore » was processed by bubbling pure water vapor and dry argon gas into the tritiated oil. Results show that the water vapor bubbling was more effective than dry argon gas. The experiment also shows that the water vapor bubbling in an oil bottle can remove and transfer tritium efficiently from the tritiated oil into another water-bubbling bottle.« less

Percolation behavior of tritiated water into a soil packed bed

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

Honda, T.; Katayama, K.; Uehara, K.

2015-03-15

A large amount of cooling water is used in a D-T fusion reactor. The cooling water will contain tritium with high concentration because tritium can permeate metal walls at high temperature easily. A development of tritium handling technology for confining tritiated water in the fusion facility is an important issue. In addition, it is also important to understand tritium behavior in environment assuming severe accidents. In this study, percolation experiments of tritiated water in soil packed bed were carried out and tritium behavior in soil was discussed. Six soil samples were collected in Hakozaki campus of Kyushu University. These particlemore » densities were of the same degree as that of general soils and moisture contents were related to BET surface area. For two soil samples used in the percolation experiment of tritiated water, saturated hydraulic conductivity agreed well with the estimating value by Creager. Tritium retention ratio in the soil packed bed was larger than water retention. This is considered to be due to an effect of tritium sorption on the surface of soil particles. The isotope exchange capacity estimated by assuming that H/T ratio of supplied tritiated water and H/T ratio of surface water of soil particle was equal was comparable to that on cement paste and mortar which were obtained by exposure of tritiated water vapor. (authors)« less

Detection and Monitoring of Airborne Nuclear Waste Materials. Annual Report to Department of Energy.

DTIC Science & Technology

1979-12-04

an active core , its detection by counting techniques is often slow and impractical. For these reasons NRL under contract with DoE undertook develop ...Protection and Measurements, Tritium Measurement Techniques NCRP Report No. 47 (1976). 2. " Development of a Continuous Tritium Monitor for Fuel Reprocessing...Trans. Am. Nucl. Soc. 21, 91 (1975). 146. "Process Behavior of and Environmental Assessments of C Releases from an HTGR Fuel Reprocessing Facility" J. W

Pellet injector development at ORNL (Oak Ridge National Laboratory)

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Argo, B. E.; Baylor, L. R.; Combs, S. K.; Fehling, D. T.; Fisher, P. W.; Foster, C. A.; Foust, C. R.; Milora, S. L.; Qualls, A. L.

1990-09-01

Advanced plasma fueling systems for magnetic confinement experiments are under development at Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets to speeds in the kilometer-per-second range by either pneumatic (light-gas gun) or mechanical (centrifugal force) techniques. ORNL has recently provided a centrifugal pellet injector for the Tore Supra tokamak and a new, simplified, eight-shot pneumatic injector for the Advanced Toroidal Facility stellarator at ORNL. Hundreds of tritium and DT pellets were accelerated at the Tritium Systems Test Assembly facility at Los Alamos in 1988 to 1989. These experiments, done in a single-shot pipe-gun system, demonstrated the feasibility of forming and accelerating tritium pellets at low (sup 3)He levels. A new, tritium-compatible extruder mechanism is being designed for longer-pulse DT applications. Two-stage light-gas guns and electron beam rocket accelerators for speeds of the order of 2 to 10 km/s are also under development. Recently, a repeating, two-stage light-gas gun accelerated 10 surrogate pellets at a 1-Hz repetition rate to speeds in the range of 2 to 3 km/s; and the electron beam rocket accelerator completed initial feasibility and scaling experiments. ORNL has also developed conceptual designs of advanced plasma fueling systems for the Compact Ignition Tokamak and the International Thermonuclear Experimental Reactor.

Technical/ administrative options for managing tritium MCL exceedances in P-area groundwater and Steel Creek

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

Ross, J.

2017-04-01

This white paper was requested by the Core Team (United States Department of Energy [USDOE], United States Environmental Protection Agency [USEPA], and South Carolina Department of Health and Environmental Control [SCDHEC]) at the P-Area Groundwater (PAGW) Operable Unit (OU) Scoping Meeting held in January 2017 to discuss recent data and potential alternatives in support of a focused Corrective Measures Study/Feasibility Study (CMS/FS). This white paper presents an overview of the problem, and a range of technical and administrative options for addressing the tritium contamination in groundwater and Steel Creek. As tritium cannot be treated practicably, alternatives are limited to mediamore » transfer, containment and natural attenuation principally relying on radioactive decay. Using other groundwater OU decisions involving tritium as precedent, Savannah River Nuclear Solutions (SRNS) recommends that final tritium alternatives be evaluated in a CMS/FS, understanding that the likely preferred remedy will include natural attenuation with land use controls (LUCs). This is based on the inability to significantly reduce tritium impact to Steel Creek using an engineered solution as compared to natural attenuation. The timing of this evaluation could be conducted concurrently with the final remedy evaluation for volatile organic compounds (VOCs).« less

Transport of tritium contamination to the atmosphere in an arid environment

USGS Publications Warehouse

Garcia, C. Amanda; Andraski, Brian J.; Johnson, Michael J.; Stonestrom, David A.; Michel, Robert L.; Cooper, C.A.; Wheatcraft, S.W.

2009-01-01

Soil–plant–atmosphere interactions strongly influence water movement in desert unsaturated zones, but little is known about how such interactions affect atmospheric release of subsurface water-borne contaminants. This 2-yr study, performed at the U.S. Geological Survey's Amargosa Desert Research Site in southern Nevada, quantified the magnitude and spatiotemporal variability of tritium (3H) transport from the shallow unsaturated zone to the atmosphere adjacent to a low-level radioactive waste (LLRW) facility. Tritium fluxes were calculated as the product of 3H concentrations in water vapor and respective evaporation and transpiration water-vapor fluxes. Quarterly measured 3H concentrations in soil water vapor and in leaf water of the dominant creosote-bush [Larrea tridentata (DC.) Coville] were spatially extrapolated and temporally interpolated to develop daily maps of contamination across the 0.76-km2 study area. Maximum plant and root-zone soil concentrations (4200 and 8700 Bq L−1, respectively) were measured 25 m from the LLRW facility boundary. Continuous evaporation was estimated using a Priestley–Taylor model and transpiration was computed as the difference between measured eddy-covariance evapotranspiration and estimated evaporation. The mean evaporation/transpiration ratio was 3:1. Tritium released from the study area ranged from 0.12 to 12 μg d−1 and totaled 1.5 mg (8.2 × 1010 Bq) over 2 yr. Tritium flux variability was driven spatially by proximity to 3H source areas and temporally by changes in 3H concentrations and in the partitioning between evaporation and transpiration. Evapotranspiration removed and limited penetration of precipitation beneath native vegetation and fostered upward movement and release of 3H from below the root zone.

Savannah River Site nuclear materials management plan FY 2017-2031

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

Magoulas, V.

The purpose of the Nuclear Materials Management Plan (herein referred to as “this Plan”) is to integrate and document the activities required to disposition the legacy and/or surplus Enriched Uranium (EU) and Plutonium (Pu) and other nuclear materials already stored or anticipated to be received by facilities at the Department of Energy (DOE) Savannah River Site (SRS) as well as the activities to support the DOE Tritium mission. It establishes a planning basis for EU and Pu processing operations in Environmental Management Operations (EMO) facilities through the end of their program missions and for the tritium through the National Nuclearmore » Security Administration (NNSA) Defense Programs (DP) facilities. Its development is a joint effort among the Department of Energy - Savannah River (DOE-SR), DOE – Environmental Management (EM), NNSA Office of Material Management and Minimization (M3), NNSA Savannah River Field Office (SRFO), and the Management and Operations (M&O) contractor, Savannah River Nuclear Solutions, LLC (SRNS). Life-cycle program planning for Nuclear Materials Stabilization and Disposition and the Tritium Enterprise may use this Plan as a basis for the development of the nuclear materials disposition scope and schedule. This Plan assumes full funding to accomplish the required project and operations activities. It is recognized that some aspects of this Plan are pre decisional with regard to National Environmental Policy Act (NEPA); in such cases new NEPA actions will be required.« less

User's manual for COAST 4: a code for costing and sizing tokamaks

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

Sink, D. A.; Iwinski, E. M.

1979-09-01

The purpose of this report is to document the computer program COAST 4 for the user/analyst. COAST, COst And Size Tokamak reactors, provides complete and self-consistent size models for the engineering features of D-T burning tokamak reactors and associated facilities involving a continuum of performance including highly beam driven through ignited plasma devices. TNS (The Next Step) devices with no tritium breeding or electrical power production are handled as well as power producing and fissile producing fusion-fission hybrid reactors. The code has been normalized with a TFTR calculation which is consistent with cost, size, and performance data published in themore » conceptual design report for that device. Information on code development, computer implementation and detailed user instructions are included in the text.« less

Tritium in Australian Precipitation: a 40 Year Record

NASA Astrophysics Data System (ADS)

Tadros, C. V.; Stone, D. J.; Hill, D. M.; Henderson-Sellers, A.

2004-12-01

Tritium, the radioisotope of hydrogen, directly incorporated into water molecules in the global hydrological system, is the most commonly used radioisotope indicator of groundwater recharge. Tritium in precipitation has been measured in Australia over the past 40 years, as an essential research tool in hydro-climate studies and to contribute to the Global Network for Isotopes in Precipitation (GNIP). Tritium, which as tritiated water (3H 1H O) is very mobile in the environment, delivers the benefit of tracing groundwater systems in a 10 - 20 year timeframe as a result of last century's atmospheric thermonuclear testing. The concentration of tritium in Australian precipitation reached a maximum level of 160 TU in 1963, during one of the most intense periods of nuclear testing. Our data reveal Australia experienced a `minor' bomb pulse compared to the Northern Hemisphere eg. in Ottawa, Canada a value of 6000 TU was recorded in 1963 for tritium in precipitation. From 1963 to 1980 we observe a rapid drop in the concentration of tritium, more than expected from natural decay, mainly due to the wash out of tritium into the oceans and groundwater. Since 1990 the levels of tritium have stabilised globally and regionally. Currently the levels of tritium in Australia have stabilised to 2 to 3 TU latitudinally across the continent, a factor of 10 lower than values observed at stations in the Northern Hemisphere. At present, levels of tritium in Australia appear to have ceased declining and our analyses suggest that today the tritium in precipitation is predominantly natural. We believe that it may be possible that the increased levels observed in the Northern Hemisphere, due to nuclear power generation [1] could `leak' into the Southern Hemisphere. This is important for research in Australia because it could hinder the exploitation of tritium in providing information on the origin and mechanism of recharge of shallow groundwaters and rivers [2]. 1. J.D. Happell, et al. A history of atmospheric tritium gas (HT) 1950-2002. Tellus(2004) 56B, 183-193. 2. D.J.M. Stone, et al. Investigation of Groundwater-Streamflow interactions in the Bega alluvial aquifer using tritium and stable isotopic ratios. ANA 2001, 4th Conference on Nuclear Science and Engineering in Australasia, pp 191-197, Sydney, NSW.

Tritium assay of Li sub 2 O pellets in the LBM/LOTUS experiments

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

Quanci, J.; Azam, S.; Bertone, P.

1986-01-01

One of the objectives of the Lithium Blanket Module (LBM) program is to test the ability of advanced neutronics codes to model the tritium breeding characteristics of a fusion blanket exposed to a toroidal fusion neutron source. The LBM consists of over 20,000 cylindrical lithium oxide pellets and numerous diagnostic pellets and wafers. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a Haefely sealed neutron generator that gives a point deuterium-tritium neutron source up to 5 {times} 10{sup 12} 14-MeV n/s. Both Princeton Plasma Physics Laboratory (PPL) and EPFL assayed the tritiummore » bred at various positions in the LBM. EPFL employed a dissolution technique while PPL recovered the tritium by a thermal extraction method. EPFL uses 0.38-g, 75% TD, lithium oxide diagnostic wafers to evaluate the tritium bred in the LBM. PPPL employs a thermal extraction method to determine the tritium bred in lithium oxide samples. In the initial experiments, diagnostic pellets and wafers were placed at five locations in the LBM central removable test rod at distances of 3, 9, 21, 36, and 48 cm from the front face of the module. The two sets of data for the tritium bred in the LBM along its centerline as a function of distance from the front face of the module were compared with each other, and with the predictions of two-dimensional neutronics codes. 1 ref.« less

Report of the Defense Science Board task force on tritium production technology options. Final technical report

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

Welch, L.

1999-05-01

The long-standing national security policy of the US to maintain a robust nuclear deterrent continues to be supported by the Congress and the President. The President has stated that ``...the nuclear deterrent posture is one of the most visible and important examples of how US military capabilities can be used effectively to deter aggression and coercion. Nuclear weapons serve as a hedge against an uncertain future, a guarantee of our security commitments to allies, and a disincentive to those who would contemplate developing or otherwise acquiring their own nuclear weapons.`` US nuclear weapons designs require tritium, an isotope of hydrogen,more » which has not been produced in the US since 1988, when the last tritium production facility (the K-Reactor at the Savannah River Site) was shut down. This long period without tritium production in the US has been possible because arms control agreements reached in the early 1990s reduced the size of the US nuclear weapons stockpile and because the Department of Energy (DOE) met stockpile tritium requirements by recycling the tritium removed from dismantled nuclear weapons. However, since tritium decays at a rate of 5.5% each year, a dependable source of tritium is required to continue to sustain the US nuclear weapons stockpile to underwrite national security policy and to support arms control goals. The US does maintain a five-year reserve supply of tritium, but this reserve is to be used only in an emergency. Current guidance states the reserve must be restored to its original level within five years of being used. To sustain the START I level, tritium production needs to begin around 2005 at a production capacity of about 3.0 kg/ year. START II levels could be sustained with production of about 1.5 kg/year beginning around 2011.« less

Remediation of ground water containing volatile organic compounds and tritium

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

Shukla, S.N.; Folsom, E.N.

1994-03-01

The Trailer 5475 (T-5475) East Taxi Strip Area at Lawrence Livermore National Laboratory (LLNL), Livermore, California was used as a taxi strip by the US Navy to taxi airplanes to the runway from 1942 to 1947. Solvents were used in some unpaved areas adjacent to the East Taxi Strip for cleaning airplanes. From 1953 through 1976, the area was used to store and treat liquid waste. From 1962 to 1976 ponds were constructed and used for evaporation of liquid waste. As a result, the ground water in this area contains volatile organic compounds (VOCs) and tritium. The ground water inmore » this area is also known to contain hexavalent chromium that is probably naturally occurring. Therefore, LLNL has proposed ``pump-and-treat`` technology above grade in a completely closed loop system. The facility will be designed to remove the VOCs and hexavalent chromium, if any, from the ground water, and the treated ground water containing tritium will be reinjected where it will decay naturally in the subsurface. Ground water containing tritium will be reinjected into areas with equal or higher tritium concentrations to comply with California regulations.« less

Technical and Scientific Aspects of the JET Trace-Tritium Experimental Campaign

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

Jones, T.T.C.; Brennan, D; Pearce, R.J.H.

The JET Trace Tritium (TTE) programme marked the first use of tritium in experiments under the managerial control of UKAEA, which operates the JET Facility on behalf of EFDA. The introduction of tritium into the plasma by gas fuelling and neutral beam injection, even in trace quantities, required the mobilisation of gram-quantities of tritium gas from the Active Gas Handling System (AGHS) product storage units into the supply lines connected to the torus gas valve and the neutral beam injectors. All systems for DT gas handling, recovery and reprocessing were therefore recommissioned and operating procedures re-established, involving extensive operations staffmore » training. The validation of Key Safety Related Equipment (KSRE) is described with reference to specific examples. The differences between requirements for TTE and full DT operations are shown to be relatively small. The scientific motivation for TTE, such as the possibility to obtain high-quality measurements in key areas such as fuel-ion transport and fast ion dynamics, is described, and the re-establishment and development of JET's 14MeV neutron diagnostic capability for TTE and future DT campaigns are outlined. Some scientific highlights from the TTE campaign are presented.« less

Modeling the Removal of Xenon from Lithium Hydrate with Aspen HYSYS

NASA Astrophysics Data System (ADS)

Efthimion, Phillip; Gentile, Charles

2011-10-01

The Laser Inertial Fusion Engine (LIFE) project mission is to provide a long-term, carbon-free source of sustainable energy, in the form of electricity. A conceptual xenon removal system has been modeled with the aid of Aspen HYSYS, a chemical process simulator. Aspen HYSYS provides excellent capability to model chemical flow processes, which generates outputs which includes specific variables such as temperature, pressure, and molar flow. The system is designed to strip out hydrogen isotopes deuterium and tritium. The base design bubbles plasma exhaust laden with x filled with liquid helium. The system separates the xenon from the hydrogen, deuterium, and tritium with a lithium hydrate and a lithium bubbler. After the removal of the hydrogen and its isotopes, the xenon is then purified by way of the process of cryogenic distillation. The pure hydrogen, deuterium, and tritium are then sent to the isotope separation system (ISS). The removal of xenon is an integral part of the laser inertial fusion engine and Aspen HYSYS is an excellent tool to calculate how to create pure xenon.

EDITORIAL: Safety aspects of fusion power plants

NASA Astrophysics Data System (ADS)

Kolbasov, B. N.

2007-07-01

This special issue of Nuclear Fusion contains 13 informative papers that were initially presented at the 8th IAEA Technical Meeting on Fusion Power Plant Safety held in Vienna, Austria, 10-13 July 2006. Following recommendation from the International Fusion Research Council, the IAEA organizes Technical Meetings on Fusion Safety with the aim to bring together experts to discuss the ongoing work, share new ideas and outline general guidance and recommendations on different issues related to safety and environmental (S&E) aspects of fusion research and power facilities. Previous meetings in this series were held in Vienna, Austria (1980), Ispra, Italy (1983), Culham, UK (1986), Jackson Hole, USA (1989), Toronto, Canada (1993), Naka, Japan (1996) and Cannes, France (2000). The recognized progress in fusion research and technology over the last quarter of a century has boosted the awareness of the potential of fusion to be a practically inexhaustible and clean source of energy. The decision to construct the International Thermonuclear Experimental Reactor (ITER) represents a landmark in the path to fusion power engineering. Ongoing activities to license ITER in France look for an adequate balance between technological and scientific deliverables and complying with safety requirements. Actually, this is the first instance of licensing a representative fusion machine, and it will very likely shape the way in which a more common basis for establishing safety standards and policies for licensing future fusion power plants will be developed. Now that ITER licensing activities are underway, it is becoming clear that the international fusion community should strengthen its efforts in the area of designing the next generations of fusion power plants—demonstrational and commercial. Therefore, the 8th IAEA Technical Meeting on Fusion Safety focused on the safety aspects of power facilities. Some ITER-related safety issues were reported and discussed owing to their potential importance for the fusion power plant research programmes. The objective of this Technical Meeting was to examine in an integrated way all the safety aspects anticipated to be relevant to the first fusion power plant prototype expected to become operational by the middle of the century, leading to the first generation of economically viable fusion power plants with attractive S&E features. After screening by guest editors and consideration by referees, 13 (out of 28) papers were accepted for publication. They are devoted to the following safety topics: power plant safety; fusion specific operational safety approaches; test blanket modules; accident analysis; tritium safety and inventories; decommissioning and waste. The paper `Main safety issues at the transition from ITER to fusion power plants' by W. Gulden et al (EU) highlights the differences between ITER and future fusion power plants with magnetic confinement (off-site dose acceptance criteria, consequences of accidents inside and outside the design basis, occupational radiation exposure, and waste management, including recycling and/or final disposal in repositories) on the basis of the most recent European fusion power plant conceptual study. Ongoing S&E studies within the US inertial fusion energy (IFE) community are focusing on two design concepts. These are the high average power laser (HAPL) programme for development of a dry-wall, laser-driven IFE power plant, and the Z-pinch IFE programme for the production of an economically-attractive power plant using high-yield Z-pinch-driven targets. The main safety issues related to these programmes are reviewed in the paper `Status of IFE safety and environmental activities in the US' by S. Reyes et al (USA). The authors propose future directions of research in the IFE S&E area. In the paper `Recent accomplishments and future directions in the US Fusion Safety & Environmental Program' D. Petti et al (USA) state that the US fusion programme has long recognized that the S&E potential of fusion can be attained by prudent materials selection, judicious design choices, and integration of safety requirements into the design of the facility. To achieve this goal, S&E research is focused on understanding the behaviour of the largest sources of radioactive and hazardous materials in a fusion facility, understanding how energy sources in a fusion facility could mobilize those materials, developing integrated state-of-the-art S&E computer codes and risk tools for safety assessment, and evaluating and improving fusion facility design in terms of accident safety, worker safety, and waste disposal. There are three papers considering safety issues of the test blanket modules (TBM) producing tritium to be installed in ITER. These modules represent different concepts of demonstration fusion power facilities (DEMO). L. Boccaccini et al (Germany) analyses the possibility of jeopardizing the ITER safety under specific accidents in the European helium-cooled pebble-bed TBM, e.g. pressurization of the vacuum vessel (VV), hydrogen production from the Be-steam reaction, the possible interconnection between the port cell and VV causing air ingress. Safety analysis is also presented for Chinese TBM with a helium-cooled solid breeder to be tested in ITER by Z. Chen et al (China). Radiological inventories, afterheat, waste disposal ratings, electromagnetic characteristics, LOCA and tritium safety management are considered. An overview of a preliminary safety analysis performed for a US proposed TBM is presented by B. Merrill et al (USA). This DEMO relevant dual coolant liquid lead-lithium TBM has been explored both in the USA and EU. T. Pinna et al (Italy) summarize the six-year development of a failure rate database for fusion specific components on the basis of data coming from operating experience gained in various fusion laboratories. The activity began in 2001 with the study of the Joint European Torus vacuum and active gas handling systems. Two years later the neutral beam injectors and the power supply systems were considered. This year the ion cyclotron resonant heating system is under evaluation. I. Cristescu et al (Germany) present the paper `Tritium inventories and tritium safety design principles for the fuel cycle of ITER'. She and her colleagues developed the dynamic mathematical model (TRIMO) for tritium inventory evaluation within each system of the ITER fuel cycle in various operational scenarios. TRIMO is used as a tool for trade-off studies within the fuel cycle systems with the final goal of global tritium inventory minimization. M. Matsuyama et al (Japan) describes a new technique for in situ quantitative measurements of high-level tritium inventory and its distribution in the VV and tritium systems of ITER and future fusion reactors. This technique is based on utilization of x-rays induced by beta-rays emitting from tritium species. It was applied to three physical states of high-level tritium: to gaseous, aqueous and solid tritium retained on/in various materials. Finally, there are four papers devoted to safety issues in fusion reactor decommissioning and waste management. A paper by R. Pampin et al (UK) provides the revised radioactive waste analysis of two models in the PPCS. Another paper by M. Zucchetti (Italy), S.A. Bartenev (Russia) et al describes a radiochemical extraction technology for purification of V-Cr-Ti alloy components from activation products to the dose rate of 10 µSv/h allowing their clearance or hands-on recycling which has been developed and tested in laboratory stationary conditions. L. El-Guebaly (USA) and her colleagues submitted two papers. In the first paper she optimistically considers the possibility of replacing the disposal of fusion power reactor waste with recycling and clearance. Her second paper considers the implications of new clearance guidelines for nuclear applications, particularly for slightly irradiated fusion materials.

Long-Term Assessment of Critical Radionuclides and Associated Environmental Media at the Savannah River Site - 13038

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

Jannik, G.T.; Baker, R.A.; Lee, P.L.

2013-07-01

During the operational history of the Savannah River Site (SRS), many different radionuclides have been released from site facilities. However, only a relatively small number of the released radionuclides have been significant contributors to doses and risks to the public. At SRS dose and risk assessments indicate tritium oxide in air and surface water, and Cs-137 in fish and deer have been, and continue to be, the critical radionuclides and pathways. In this assessment, statistical analyses of the long-term trends of tritium oxide in atmospheric and surface water releases and Cs-137 concentrations in fish and deer are provided. Correlations alsomore » are provided with 1) operational changes and improvements, 2) geopolitical events (Cold War cessation), and 3) recent environmental remediation projects and decommissioning of excess facilities. For example, environmental remediation of the F- and H-Area Seepage Basins and the Solid Waste Disposal Facility have resulted in a measurable impact on the tritium oxide flux to the onsite Fourmile Branch stream. Airborne releases of tritium oxide have been greatly affected by operational improvements and the end of the Cold War in 1991. However, the effects of SRS environmental remediation activities and ongoing tritium operations on tritium concentrations in the environment are measurable and documented in this assessment. Controlled hunts of deer and feral hogs are conducted at SRS for approximately six weeks each year. Before any harvested animal is released to a hunter, SRS personnel perform a field analysis for Cs-137 concentrations to ensure the Hunter's dose does not exceed the SRS administrative game limit of 0.22 milli-sievert (22 mrem). However, most of the Cs-137 found in SRS onsite deer is not from site operations but is from nuclear weapons testing fallout from the 1950's and early 1960's. This legacy source term is trended in the SRS deer, and an assessment of the 'effective' half-life of Cs-137 in deer (including the physical decay half-life and the environmental dispersion half-life) is provided. The 'creek mouth' fisherman is the next most critical pathway at SRS. On an annual basis, three species of fish (panfish, catfish, and bass) are sampled from the mouths of the five SRS streams. Three composites of up to five fish of each species are analyzed from each sampling location. Long-term trending of the Cs-137 concentrations in fish and the subsequent doses from consumption of SRS fish is provided. (authors)« less

Long-Term Assessment of Critical Radionuclides and Associated Environmental Media at the Savannah River Site

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

Jannik, G. T.; Baker, R. A.; Lee, P. L.

2012-11-06

During the operational history of the Savannah River Site (SRS), many different radionuclides have been released from site facilities. However, only a relatively small number of the released radionuclides have been significant contributors to doses and risks to the public. At SRS dose and risk assessments indicate tritium oxide in air and surface water, and Cs-137 in fish and deer have been, and continue to be, the critical radionuclides and pathways. In this assessment, indepth statistical analyses of the long-term trends of tritium oxide in atmospheric and surface water releases and Cs-137 concentrations in fish and deer are provided. Correlationsmore » also are provided with 1) operational changes and improvements, 2) geopolitical events (Cold War cessation), and 3) recent environmental remediation projects and decommissioning of excess facilities. For example, environmental remediation of the F- and H-Area Seepage Basins and the Solid Waste Disposal Facility have resulted in a measurable impact on the tritium oxide flux to the onsite Fourmile Branch stream. Airborne releases of tritium oxide have been greatly affected by operational improvements and the end of the Cold War in 1991. However, the effects of SRS environmental remediation activities and ongoing tritium operations on tritium concentrations in the environment are measurable and documented in this assessment. Controlled hunts of deer and feral hogs are conducted at SRS for approximately six weeks each year. Before any harvested animal is released to a hunter, SRS personnel perform a field analysis for Cs-137 concentrations to ensure the hunter's dose does not exceed the SRS administrative game limit of 0.22 millisievert (22 mrem). However, most of the Cs-137 found in SRS onsite deer is not from site operations but is from nuclear weapons testing fallout from the 1950's and early 1960's. This legacy source term is trended in the SRS deer, and an assessment of the ''effective'' half-life of Cs-137 in deer (including the physical decay half-life and the environmental dispersion half-life) is provided. The ''creek mouth'' fisherman is the next most critical pathway at SRS. On an annual basis, three species of fish (panfish, catfish, and bass) are sampled from the mouths of the five SRS streams. Three composites of up to five fish of each species are analyzed from each sampling location. Long-term trending of the Cs-137 concentrations in fish and the subsequent doses from consumption of SRS fish is provided.« less

Tritium plume dynamics in the shallow unsaturated zone in an arid environment

USGS Publications Warehouse

Maples, S.R.; Andraski, Brian J.; Stonestrom, David A.; Cooper, C.A.; Pohll, G.; Michel, R.L.

2014-01-01

The spatiotemporal variability of a tritium plume in the shallow unsaturated zone and the mechanisms controlling its transport were evaluated during a 10-yr study. Plume movement was minimal and its mass declined by 68%. Upward-directed diffusive-vapor tritium fluxes and radioactive decay accounted for most of the observed plume-mass declines.Effective isolation of tritium (3H) and other contaminants at waste-burial facilities requires improved understanding of transport processes and pathways. Previous studies documented an anomalously widespread (i.e., theoretically unexpected) distribution of 3H (>400 m from burial trenches) in a dry, sub-root-zone gravelly layer (1–2-m depth) adjacent to a low-level radioactive waste (LLRW) burial facility in the Amargosa Desert, Nevada, that closed in 1992. The objectives of this study were to: (i) characterize long-term, spatiotemporal variability of 3H plumes; and (ii) quantify the processes controlling 3H behavior in the sub-root-zone gravelly layer beneath native vegetation adjacent to the facility. Geostatistical methods, spatial moment analyses, and mass flux calculations were applied to a spatiotemporally comprehensive, 10-yr data set (2001–2011). Results showed minimal bulk-plume advancement during the study period and limited Fickian spreading of mass. Observed spreading rates were generally consistent with theoretical vapor-phase dispersion. The plume mass diminished more rapidly than would be expected from radioactive decay alone, indicating net efflux from the plume. Estimates of upward 3H efflux via diffusive-vapor movement were >10× greater than by dispersive-vapor or total-liquid movement. Total vertical fluxes were >20× greater than lateral diffusive-vapor fluxes, highlighting the importance of upward migration toward the land surface. Mass-balance calculations showed that radioactive decay and upward diffusive-vapor fluxes contributed the majority of plume loss. Results indicate that plume losses substantially exceeded any continuing 3H contribution to the plume from the LLRW facility during 2001 to 2011 and suggest that the widespread 3H distribution resulted from transport before 2001.

High power neutron production targets

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

Wender, S.

1996-06-01

The author describes issues of concern in the design of targets and associated systems for high power neutron production facilities. The facilities include uses for neutron scattering, accelerator driven transmutation, accelerator production of tritium, short pulse spallation sources, and long pulse spallation sources. Each of these applications requires a source with different design needs and consequently different implementation in practise.

Diagnosing radiative shocks from deuterium and tritium implosions on NIF.

PubMed

Pak, A; Divol, L; Weber, S; Döppner, T; Kyrala, G A; Kilne, J; Izumi, N; Glenn, S; Ma, T; Town, R P; Bradley, D K; Glenzer, S H

2012-10-01

During the recent ignition tuning campaign at the National Ignition Facility, layered cryogenic deuterium and tritium capsules were imploded via x-ray driven ablation. The hardened gated x-ray imager diagnostic temporally and spatially resolves the x-ray emission from the core of the capsule implosion at energies above ~8 keV. On multiple implosions, ~200-400 ps after peak compression a spherically expanding radiative shock has been observed. This paper describes the methods used to characterize the radial profile and rate of expansion of the shock induced x-ray emission.

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

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

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

Tritium proof-of-principle pellet injector: Phase 2

NASA Astrophysics Data System (ADS)

Fisher, P. W.; Gouge, M. J.

1995-03-01

As part of the International Thermonuclear Engineering Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. This repeating, single-stage, pneumatic injector, called the Tritium-Proof-of-Principle Phase-2 (TPOP-2) Pellet Injector, has a piston-driven mechanical extruder and is designed to extrude hydrogenic pellets sized for the ITER device. The TPOP-II program has the following development goals: evaluate the feasibility of extruding tritium and DT mixtures for use in future pellet injection systems; determine the mechanical and thermal properties of tritium and DT extrusions; integrate, test and evaluate the extruder in a repeating, single-stage light gas gun sized for the ITER application (pellet diameter approximately 7-8 mm); evaluate options for recycling propellant and extruder exhaust gas; evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory requiring secondary and room containment systems. In initial tests with deuterium feed at ORNL, up to thirteen pellets have been extruded at rates up to 1 Hz and accelerated to speeds of order 1.0-1.1 km/s using hydrogen propellant gas at a supply pressure of 65 bar. The pellets are typically 7.4 mm in diameter and up to 11 mm in length and are the largest cryogenic pellets produced by the fusion program to date. These pellets represent about a 11% density perturbation to ITER. Hydrogenic pellets will be used in ITER to sustain the fusion power in the plasma core and may be crucial in reducing first wall tritium inventories by a process called isotopic fueling where tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge.

Thermal enhancement cartridge heater modified (TECH Mod) tritium hydride bed development, Part 1 - Design and fabrication

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

Klein, J.E.; Estochen, E.G.

The Savannah River Site (SRS) tritium facilities have used first generation (Gen1) LaNi{sub 4.25}Al{sub 0.75} (LANA0.75) metal hydride storage beds for tritium absorption, storage, and desorption. The Gen1 design utilizes hot and cold nitrogen supplies to thermally cycle these beds. Second and third generation (Gen2 and Gen3) storage bed designs include heat conducting foam and divider plates to spatially fix the hydride within the bed. For thermal cycling, the Gen2 and Gen3 beds utilize internal electric heaters and glovebox atmosphere flow over the bed inside the bed external jacket for cooling. The currently installed Gen1 beds require replacement due tomore » tritium aging effects on the LANA0.75 material, and cannot be replaced with Gen2 or Gen3 beds due to different designs of these beds. At the end of service life, Gen1 bed desorption efficiencies are limited by the upper temperature of hot nitrogen supply. To increase end-of-life desorption efficiency, the Gen1 bed design was modified, and a Thermal Enhancement Cartridge Heater Modified (TECH Mod) bed was developed. Internal electric cartridge heaters in the new design to improve end-of-life desorption, and also permit in-bed tritium accountability (IBA) calibration measurements to be made without the use of process tritium. Additional enhancements implemented into the TECH Mod design are also discussed. (authors)« less

THERMAL ENHANCEMENT CARTRIDGE HEATER MODIFIED TECH MOD TRITIUM HYDRIDE BED DEVELOPMENT PART I DESIGN AND FABRICATION

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

Klein, J.; Estochen, E.

The Savannah River Site (SRS) tritium facilities have used 1{sup st} generation (Gen1) LaNi{sub 4.25}Al{sub 0.75} (LANA0.75) metal hydride storage beds for tritium absorption, storage, and desorption. The Gen1 design utilizes hot and cold nitrogen supplies to thermally cycle these beds. Second and 3{sup rd} generation (Gen2 and Gen3) storage bed designs include heat conducting foam and divider plates to spatially fix the hydride within the bed. For thermal cycling, the Gen2 and Gen 3 beds utilize internal electric heaters and glovebox atmosphere flow over the bed inside the bed external jacket for cooling. The currently installed Gen1 beds requiremore » replacement due to tritium aging effects on the LANA0.75 material, and cannot be replaced with Gen2 or Gen3 beds due to different designs of these beds. At the end of service life, Gen1 bed desorption efficiencies are limited by the upper temperature of hot nitrogen supply. To increase end-of-life desorption efficiency, the Gen1 bed design was modified, and a Thermal Enhancement Cartridge Heater Modified (TECH Mod) bed was developed. Internal electric cartridge heaters in the new design to improve end-of-life desorption, and also permit in-bed tritium accountability (IBA) calibration measurements to be made without the use of process tritium. Additional enhancements implemented into the TECH Mod design are also discussed.« less

Pathways for Disposal of Commercially-Generated Tritiated Waste

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

Halverson, Nancy V.

From a waste disposal standpoint, tritium is a major challenge. Because it behaves like hydrogen, tritium exchanges readily with hydrogen in the ground water and moves easily through the ground. Land disposal sites must control the tritium activity and mobility of incoming wastes to protect human health and the environment. Consequently, disposal of tritiated low-level wastes is highly regulated and disposal options are limited. The United States has had eight operating commercial facilities licensed for low-level radioactive waste disposal, only four of which are currently receiving waste. Each of these is licensed and regulated by its state. Only two ofmore » these sites accept waste from states outside of their specified regional compact. For waste streams that cannot be disposed directly at one of the four active commercial low-level waste disposal facilities, processing facilities offer various forms of tritiated low-level waste processing and treatment, and then transport and dispose of the residuals at a disposal facility. These processing facilities may remove and recycle tritium, reduce waste volume, solidify liquid waste, remove hazardous constituents, or perform a number of additional treatments. Waste brokers also offer many low-level and mixed waste management and transportation services. These services can be especially helpful for small-quantity tritiated-waste generators, such as universities, research institutions, medical facilities, and some industries. The information contained in this report covers general capabilities and requirements for the various disposal/processing facilities and brokerage companies, but is not considered exhaustive. Typically, each facility has extensive waste acceptance criteria and will require a generator to thoroughly characterize their wastes. Then a contractual agreement between the waste generator and the disposal/processing/broker entity must be in place before waste is accepted. Costs for tritiated waste transportation, processing and disposal vary based a number of factors. In many cases, wastes with very low radioactivity are priced primarily based on weight or volume. For higher activities, costs are based on both volume and activity, with the activity-based charges usually being much larger than volume-based charges. Other factors affecting cost include location, waste classification and form, other hazards in the waste, etc. Costs may be based on general guidelines used by an individual disposal or processing site, but final costs are established by specific contract with each generator. For this report, seven hypothetical waste streams intended to represent commercially-generated tritiated waste were defined in order to calculate comparative costs. Ballpark costs for disposition of these hypothetical waste streams were calculated. These costs ranged from thousands to millions of dollars. Due to the complexity of the cost-determining factors mentioned above, the costs calculated in this report should be understood to represent very rough cost estimates for the various hypothetical wastes. Actual costs could be higher or could be lower due to quantity discounts or other factors.« less

In-vessel tritium retention and removal in ITER

NASA Astrophysics Data System (ADS)

Federici, G.; Anderl, R. A.; Andrew, P.; Brooks, J. N.; Causey, R. A.; Coad, J. P.; Cowgill, D.; Doerner, R. P.; Haasz, A. A.; Janeschitz, G.; Jacob, W.; Longhurst, G. R.; Nygren, R.; Peacock, A.; Pick, M. A.; Philipps, V.; Roth, J.; Skinner, C. H.; Wampler, W. R.

Tritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER. We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation. We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER - and more generically in any other next-step experimental fusion facility fuelled with tritium - the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

Tritium assay of Li/sub 2/O in the LBM/LOTUS experiments

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

Quanci, J.; Azam, S.; Bertone, P.

1986-11-01

The Lithium Blanket Module (LBM) is an assembly of over 20,000 cylindrical lithium oxide pellets in an array representative of a limited-coverage breeding zone for a toroidal fusion device. A principal objective of the LBM program is to test the ability of advanced neutronics coding to model the tritium breeding characteristics of a fusion device blanket. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a 14 MeV point-neutron source. Princeton Plasma Physics Laboratory (PPPL) and EPFL assayed the tritium bred in lithium oxide diagnostic samples placed at various positions in the LBM.more » PPPL employed a thermal extraction technique while EPFL used a dissolution method. The results for the assay are reported and compared to MCNP Monte Carlo neutronics calculations for the LBM/LOTUS system.« less

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

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

Spencer, Barry B.; Walker, T. B.; Bruffey, S. H.

2016-08-31

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when themore » solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

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

Spencer, Barry B.; Walker, T. B.; Bruffey, Stephanie H.

2016-08-31

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-basedmore » cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less

Liquid surfaces for fusion plasma facing components—A critical review. Part I: Physics and PSI

DOE PAGES

Nygren, R. E.; Tabares, F. L.

2016-12-01

This review of the potential of robust plasma facing components (PFCs) with liquid surfaces for applications in future D/T fusion device summarizes the critical issues for liquid surfaces and research being done worldwide in confinement facilities, and supporting R&D in plasma surface interactions. In the paper are a set of questions and related criteria by which we will judge the progress and readiness of liquid surface PFCs. Part-II (separate paper) will cover R&D on the technology-oriented aspects of liquid surfaces including the liquid surfaces as integrated first walls in tritium breeding blankets, tritium retention and recovery, and safety.

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

Soter, J.; Bhike, M.; Finch, S. W.

Measurements of the 169Tm(n,2n) 168Tm cross section have been performed via the activation technique at 13 energies between 8.5 and 15.0 MeV. The purpose of this comprehensive data set is to provide an alternative diagnostic tool for obtaining subtle information on the neutron energy distribution produced in inertial confinement deuterium-tritium fusion experiments at the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory. In conclusion, the 169Tm(n,2n) 168Tm reaction not only provides the primary 14-MeV neutron fluence, but also the important down-scattered neutron fluence, the latter providing information on the density achieved in the deuterium-tritium plasma during a laser shot.

The Marble Experiment: Overview and Simulations

NASA Astrophysics Data System (ADS)

Douglas, M. R.; Murphy, T. J.; Cobble, J. A.; Fincke, J. R.; Haines, B. M.; Hamilton, C. E.; Lee, M. N.; Oertel, J. A.; Olson, R. E.; Randolph, R. B.; Schmidt, D. W.; Shah, R. C.; Smidt, J. M.; Tregillis, I. L.

2015-11-01

The Marble ICF platform has recently been launched on both OMEGA and NIF with the goal to investigate the influence of heterogeneous mix on fusion burn. The unique separated reactant capsule design consists of an ``engineered'' CH capsule filled with deuterated plastic foam that contains pores or voids that are filled with tritium gas. Initially the deuterium and tritium are separated, but as the implosion proceeds, the D and T mix, producing a DT signature. The results of these experiments will be used to inform a probability density function (PDF) burn modelling approach for un-resolved cell morphology. Initial targets for platform development have consisted of either fine-pore foams or gas mixtures, with the goal to field the engineered foams in 2016. An overview of the Marble experimental campaign will be presented and simulations will be discussed. This work is supported by US DOE/NNSA, performed at LANL, operated by LANS LLC under contract DE-AC52-06NA25396.

Development of the Los Alamos National Laboratory Cryogenic Pressure Loader

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

Ebey, Peter S.; Dole, James M.; Hoffer, James K.

2003-05-15

Targets for inertial fusion research and ignition at OMEGA, the National Ignition Facility, LMJ, and future facilities rely on beta-radiation-driven layering of spherical cryogenic DT ice layers contained within plastic or metal shells. Plastic shells will be permeation filled at room temperature then cooled to cryogenic temperatures before removal of the overpressure. The cryogenic pressure loader (CPL) was recently developed at Los Alamos National Laboratory as a testbed for studying the filling and layering of plastic target shells with DT. A technical description of the CPL is provided. The CPL consists of a cryostat, which contains a high-pressure permeation cell,more » and has optical access for investigating beta layering. The cryostat is housed within a tritium glovebox that contains manifolds for supplying high-pressure DT. The CPL shares some design elements with the cryogenic target handling system at the OMEGA facility to allow testing of tritium issues related to that system. The CPL has the capability to fill plastic targets by permeation to pressures up to 100 MPa and to cool them to 15 K. The CPL will accommodate a range of targets and may be modified for future experiments.« less

Type A Accident Investigation Board report on the January 17, 1996, electrical accident with injury in Technical Area 21 Tritium Science and Fabrication Facility Los Alamos National Laboratory. Final report

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

NONE

1996-04-01

An electrical accident was investigated in which a crafts person received serious injuries as a result of coming into contact with a 13.2 kilovolt (kV) electrical cable in the basement of Building 209 in Technical Area 21 (TA-21-209) in the Tritium Science and Fabrication Facility (TSFF) at Los Alamos National Laboratory (LANL). In conducting its investigation, the Accident Investigation Board used various analytical techniques, including events and causal factor analysis, barrier analysis, change analysis, fault tree analysis, materials analysis, and root cause analysis. The board inspected the accident site, reviewed events surrounding the accident, conducted extensive interviews and document reviews,more » and performed causation analyses to determine the factors that contributed to the accident, including any management system deficiencies. Relevant management systems and factors that could have contributed to the accident were evaluated in accordance with the guiding principles of safety management identified by the Secretary of Energy in an October 1994 letter to the Defense Nuclear Facilities Safety Board and subsequently to Congress.« less

Development of a polar direct drive platform for mix and burn experiments on the National Ignition Facility

NASA Astrophysics Data System (ADS)

Murphy, T. J.; Kyrala, G. A.; Krasheninnikova, N. S.; Bradley, P. A.; Cobble, J. A.; Tregillis, I. L.; Obrey, K. A. D.; Baumgaertel, J. A.; Hsu, S. C.; Shah, R. C.; Hakel, P.; Kline, J. L.; Schmitt, M. J.; Kanzleiter, R. J.; Batha, S. H.; Wallace, R. J.; Bhandarkar, S.; Fitzsimmons, P.; Hoppe, M.; Nikroo, A.; McKenty, P.

2016-03-01

Capsules driven with polar drive [1, 2] on the National Ignition Facility [3] are being used [4] to study mix in convergent geometry. In preparation for experiments that will utilize deuterated plastic shells with a pure tritium fill, hydrogen-filled capsules with copper- doped deuterated layers have been imploded on NIF to provide spectroscopic and nuclear measurements of capsule performance. Experiments have shown that the mix region, when composed of shell material doped with about 1% copper (by atom), reaches temperatures of about 2 keV, while undoped mixed regions reach about 3 keV. Based on the yield from these implosions, we estimate the thickness of CD that mixed into the gas as between about 0.25 and 0.43 μm of the inner capsule surface, corresponding to about 5 to 9 μg of material. Using 5 atm of tritium as the fill gas should result in over 1013 DT neutrons being produced, which is sufficient for neutron imaging [5].

EFFECTS OF TRITIUM EXPOSURE ON UHMW-PE, PTFE, AND VESPEL

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

Clark, E; Kirk Shanahan, K

2006-05-31

Samples of three polymers, Ultra-High Molecular Weight Polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, also known as Teflon{reg_sign}), and Vespel{reg_sign} polyimide were exposed to 1 atmosphere of tritium gas at ambient temperature for varying times up to 2.3 years in closed containers. Sample mass and size measurements (to calculate density), spectra-colorimetry, dynamic mechanical analysis (DMA), and Fourier-transform infrared spectroscopy (FT-IR) were employed to characterize the effects of tritium exposure on these samples. Changes of the tritium exposure gas itself were characterized at the end of exposure by measuring total pressure and by mass spectroscopic analysis of the gas composition. None of the polymersmore » exhibited significant changes of density. The color of initially white UHMW-PE and PTFE dramatically darkened to the eye and the color also significantly changed as measured by colorimetry. The bulk of UHMW-PE darkened just like the external surfaces, however the fracture surface of PTFE appeared white compared to the PTFE external surfaces. The white interior could have been formed while the sample was breaking or could reflect the extra tritium dose at the surface directly from the gas. The dynamic mechanical response of UHMW-PE was typical of radiation effects on polymers- an initial stiffening (increased storage modulus) and reduction of viscous behavior after three months exposure, followed by lowering of the storage modulus after one year exposure and longer. The storage modulus of PTFE increased through about nine months tritium exposure, then the samples became too weak to handle or test using DMA. Characterization of Vespel{reg_sign} using DMA was problematic--sample-to-sample variations were significant and no systematic change with tritium exposure could be discerned. Isotopic exchange and incorporation of tritium into UHMW-PE (exchanging for protium) and into PTFE (exchanging for fluorine) was observed by FT-IR using an attenuated total reflectance method. No significant change in the Vespel{reg_sign} infrared spectrum was observed after three months exposure. Protium significantly pressurized the UHMW-PE containers during exposure to about nine atmospheres (the initial pressure was one atmosphere of tritium). This is consistent with the well-known production of hydrogen by irradiation of polyethylene by ionizing radiation. The total pressure in the PTFE containers decreased, and a mass balance reveals that the observed decrease is consistent with the formation of small amounts of {sup 3}HF, which is condensed at ambient temperature. No significant change of pressure occurred in the Vespel{reg_sign} containers; however the composition of the gas became about 50% protium, showing that Vespel{reg_sign} interacted with the tritium gas atmosphere to some degree. The relative resistance to degradation from tritium exposure is least for PTFE, more for UHMW-PE, and the most for Vespel{reg_sign}, which is consistent with the known relative resistance of these polymers to gamma irradiation. This qualitatively agrees with the concept of equivalent effects for equivalent absorbed doses of radiation damage of polymers. Some of the changes of different polymers are qualitatively similar; however each polymer exhibited unique property changes when exposed to tritium. Information from this study that can be applied to a tritium facility is: (1) the relative resistance to tritium degradation of the three polymers studied is the same as the relative resistance to gamma irradiation in air (so relative rankings of polymer resistance to ionizing radiation can be used as a relative ranking for assessing tritium compatibility and polymer selection); and (2) all three polymers changed the gas atmosphere during tritium exposure--UHMW-PE and Vespel{reg_sign} exposed to tritium formed H{sub 2} gas (UHMW-PE much more so), and PTFE exposed to tritium formed {sup 3}HF. This observation of forming {sup 3}HF supports the general concept of minimizing chlorofluorocarbon polymers in tritium systems.« less

Tritium and plutonium in waters from the Bering and Chukchi Seas

USGS Publications Warehouse

Landa, E.R.; Beals, D.M.; Halverson, J.E.; Michel, R.L.; Cefus, G.R.

1999-01-01

During the summer of 1993, seawater in the Bering and Chukchi Seas was sampled on a joint Russian-American cruise [BERPAC] of the RV Okean to determine concentrations of tritium, 239Pu and 240Pu. Concentrations of tritium were determined by electrolytic enrichment and liquid scintilation counting. Tritium levels ranged up to 420 mBq L-1 showed no evidence of inputs other than those attribute atmospheric nuclear weapons testing. Plutonium was recovered from water samples by ferric hydroxide precipitation, and concentrations were determined by thermal ionization mass spectrometry. 239+240Pu concentrations ranged from <1 to 5.5 [mu]Bq L-1. These concentrations are lower than those measured in water samples from other parts of the ocean during the mid-1960's to the late 1980's. The 240Pu:239Pu ratios, although associated with large uncertainties, suggest that most of the plutonium is derived from world-wide fallout. As points of comparison, the highest concentrations of tritium and plutonium observed here were about five orders of magnitude lower than the maximum permissible concentrations allowed in water released to the off-site environs from licensed nuclear facilities in the United States. This study and others sponsored by the International Atomic Energy Agency and the Office of Naval Research's Arctic Nuclear Waste Assessment Program are providing data for the assessment of potential radiological impacts in the Arctic regions associated with nuclear waste disposal by the former Soviet Union.

First calibration measurements of an FTIR absorption spectroscopy system for liquid hydrogen isotopologues for the isotope separation system of fusion power plants

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

Groessle, R.; Beck, A.; Bornschein, B.

2015-03-15

Fusion facilities like ITER and DEMO will circulate huge amounts of deuterium and tritium in their fuel cycle with an estimated throughput of kg per hour. One important capability of these fuel cycles is to separate the hydrogen isotopologues (H{sub 2}, D{sub 2}, T{sub 2}, HD, HT, DT). For this purpose the Isotope Separation System (ISS), using cryogenic distillation, as part of the Tritium Enrichment Test Assembly (TRENTA) is under development at Tritium Laboratory Karlsruhe. Fourier transform infrared absorption spectroscopy (FTIR) has been selected to prove its capability for online monitoring of the tritium concentration in the liquid phase atmore » the bottom of the distillation column of the ISS. The actual research-development work is focusing on the calibration of such a system. Two major issues are the identification of appropriate absorption lines and their dependence on the isotopic concentrations and composition. For this purpose the Tritium Absorption IR spectroscopy experiment has been set up as an extension of TRENTA. For calibration a Raman spectroscopy system is used. First measurements, with equilibrated mixtures of H{sub 2}, D{sub 2} and HD demonstrate that FTIR can be used for quantitative analysis of liquid hydro-gen isotopologues and reveal a nonlinear dependence of the integrated absorbance from the D{sub 2} concentration in the second vibrational branch of D{sub 2} FTIR spectra. (authors)« less

Parameter Study of the LIFE Engine Nuclear Design

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

Kramer, K J; Meier, W R; Latkowski, J F

2009-07-10

LLNL is developing the nuclear fusion based Laser Inertial Fusion Energy (LIFE) power plant concept. The baseline design uses a depleted uranium (DU) fission fuel blanket with a flowing molten salt coolant (flibe) that also breeds the tritium needed to sustain the fusion energy source. Indirect drive targets, similar to those that will be demonstrated on the National Ignition Facility (NIF), are ignited at {approx}13 Hz providing a 500 MW fusion source. The DU is in the form of a uranium oxycarbide kernel in modified TRISO-like fuel particles distributed in a carbon matrix forming 2-cm-diameter pebbles. The thermal power ismore » held at 2000 MW by continuously varying the 6Li enrichment in the coolants. There are many options to be considered in the engine design including target yield, U-to-C ratio in the fuel, fission blanket thickness, etc. Here we report results of design variations and compare them in terms of various figures of merit such as time to reach a desired burnup, full-power years of operation, time and maximum burnup at power ramp down and the overall balance of plant utilization.« less

Fuel gain exceeding unity in an inertially confined fusion implosion.

PubMed

Hurricane, O A; Callahan, D A; Casey, D T; Celliers, P M; Cerjan, C; Dewald, E L; Dittrich, T R; Döppner, T; Hinkel, D E; Berzak Hopkins, L F; Kline, J L; Le Pape, S; Ma, T; MacPhee, A G; Milovich, J L; Pak, A; Park, H-S; Patel, P K; Remington, B A; Salmonson, J D; Springer, P T; Tommasini, R

2014-02-20

Ignition is needed to make fusion energy a viable alternative energy source, but has yet to be achieved. A key step on the way to ignition is to have the energy generated through fusion reactions in an inertially confined fusion plasma exceed the amount of energy deposited into the deuterium-tritium fusion fuel and hotspot during the implosion process, resulting in a fuel gain greater than unity. Here we report the achievement of fusion fuel gains exceeding unity on the US National Ignition Facility using a 'high-foot' implosion method, which is a manipulation of the laser pulse shape in a way that reduces instability in the implosion. These experiments show an order-of-magnitude improvement in yield performance over past deuterium-tritium implosion experiments. We also see a significant contribution to the yield from α-particle self-heating and evidence for the 'bootstrapping' required to accelerate the deuterium-tritium fusion burn to eventually 'run away' and ignite.

Apparent enrichment of organically bound tritium in rivers explained by the heritage of our past.

PubMed

Eyrolle-Boyer, Frédérique; Boyer, Patrick; Claval, David; Charmasson, Sabine; Cossonnet, Catherine

2014-10-01

The global inventory of naturally produced tritium (3H) is estimated at 2.65 kg, whereas more than 600 kg have been released during atmospheric nuclear tests (NCRP, 1979; UNSCEAR, 2000) constituting the main source of artificial tritium throughout the Anthropocene. The behaviour of this radioactive isotope in the environment has been widely studied since the 1950s, both through laboratory experiments and, more recently, through field observations (e.g., Cline, 1953; Kirchmann et al., 1979; Daillant et al., 2004; McCubbin et al., 2001; Kim et al., 2012). In its "free" forms, [i.e. 3H gas or 3H hydride (HT); methyl 3H gas (CH3T); tritiated H2O or 3H-oxide (HTO); and Tissue Free Water 3H (TFWT)], tritium closely follows the water cycle. However, 3H bound with organic compounds, mainly during the basic stages of photosynthesis or through weak hydrogen links, is less exchangeable with water, which explains its persistence in the carbon cycle as re underlined recently by Baglan et al. (2013), Jean-Batiste and Fourré (2013), Kim et al. (2013a,b). In this paper, we demonstrate that terrestrial biomass pools, historically contaminated by global atmospheric fallout from nuclear testing, have constituted a significant delayed source of organically bound tritium (OBT) for aquatic systems, resulting in an apparent enrichment of OBT as compared to HTO. This finding helps to explain concentration factors (tritium concentration in biota/concentration in water) greater than 1 observed in areas that are not directly affected by industrial radioactive wastes, and thus sheds light on the controversies regarding tritium 'bioaccumulation'. Such apparent enrichment of OBT is expected to be more pronounced in the Northern Hemisphere where fallout was most significant, depending on the nature and biodegradability of terrestrial biomass at the regional scale. We further believe that OBT transfers from the continent to oceans have been sufficient to affect tritium concentrations in coastal marine biota (i.e., near river inputs). Our findings demonstrate that the persistence of terrestrial organic (3)H explains imbalances between organically bound tritium and free (3)H in most river systems in particular those not impacted by releases from nuclear facilities. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

Brown, Thomas

The design features developed for the Spherical Tokamak (ST) in the PPPL pilot plant study was used as the starting point in developing designs to meet the mission of a Fusion Nuclear Science Facility (FNSF) considering a range of machine sizes based on the influence of tritium consumption and maintenance strategies. The compact nature of a steady state operated ST device for this mission pushes operating conditions and places challenges in the design of components, device maintenance and the integration of supports and services. This paper reviews the general arrangement, design details and maintenance strategy of the ST-FNSF device coremore » for a 1.6-m and 1.0-m device; operating points which bracket the region between purchasing and breeding tritium.« less

Cryogenic tritium-hydrogen-deuterium and deuterium-tritium layer implosions with high density carbon ablators in near-vacuum hohlraums

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

Meezan, N. B., E-mail: meezan1@llnl.gov; Hopkins, L. F. Berzak; Pape, S. Le

2015-06-15

High Density Carbon (or diamond) is a promising ablator material for use in near-vacuum hohlraums, as its high density allows for ignition designs with laser pulse durations of <10 ns. A series of Inertial Confinement Fusion (ICF) experiments in 2013 on the National Ignition Facility [Moses et al., Phys. Plasmas 16, 041006 (2009)] culminated in a deuterium-tritium (DT) layered implosion driven by a 6.8 ns, 2-shock laser pulse. This paper describes these experiments and comparisons with ICF design code simulations. Backlit radiography of a tritium-hydrogen-deuterium (THD) layered capsule demonstrated an ablator implosion velocity of 385 km/s with a slightly oblate hot spot shape.more » Other diagnostics suggested an asymmetric compressed fuel layer. A streak camera-based hot spot self-emission diagnostic (SPIDER) showed a double-peaked history of the capsule self-emission. Simulations suggest that this is a signature of low quality hot spot formation. Changes to the laser pulse and pointing for a subsequent DT implosion resulted in a higher temperature, prolate hot spot and a thermonuclear yield of 1.8 × 10{sup 15} neutrons, 40% of the 1D simulated yield.« less

A new tritium monitor design based on plasma source ion implantation technique

NASA Astrophysics Data System (ADS)

Nassar, Rafat Mohammad

Tritium is an important isotope of hydrogen. The availability of tritium in our environment is manifest through both natural and artificial sources. Consequently, the requirement for tritium handling and usage will continue to increase in the future. An important future contributor is nuclear fusion power plants and facilities. Essential safety regulations and procedures require effective monitoring and measurements of tritium concentrations in workplaces. The unique characteristics of tritium impose an important role on the criteria for its detection and measurement. As tritium decays by the emission of soft beta particles, maximum 18 keV, it cannot be readily detected by commonly used detectors. Specially built monitors are required. Additional complications occur due to the presence of other radioactive isotopes or ambient radiation fields and because of the high diffusivity of tritium. When it is in oxidized form it is 25000 times more hazardous biologically than when in elemental form. Therefore, contamination of the monitor is expected and compound specific monitors are important. A summary is given of the various well known methods of detecting tritium-in-air. This covers the direct as well as the indirect measuring techniques, although each has been continually improved and further developed, nevertheless, each has its own limitations. Ionization chambers cannot discriminate against airborne P emitters. Proportional counters have a narrow operating range, 3-4 decades, and have poor performance in relatively high humid environments and require a dry counting gas. Liquid scintillation counters are sensitive, but inspection of the sample is slow and they produce chemical liquid waste. A new way to improve the sensitivity of detecting tritium with plastic scintillators has been developed. The technique is based on a non-line-of-sight implantation of tritium ions into a 20 mum plastic scintillator using a plasma source ion implantation (PSII) technique, This type of source is different, superior to the line-of-sight implantation and requires no additional beam handling. It is capable of implanting ion species in a broad beam configuration into the entire surface of a target. The technique requires a special ion source with special characteristics of the type obtained from a surfatron plasma source. This ion source has a large high ion density plasma with minimum contamination and produces ions of low temperature. It was constructed to ionize the sampled air and to produce a plasma over a wide range of pressure, 4-0.1 mTorr. A plasma source ion implantation cell was designed and constructed using mathematical modeling with personal computer, to optimize the essential variables of the design and to estimate the implantation rate under different operation conditions. Also, a high voltage pulse modulator was designed and constructed to produce a series of 10 musec pulses (up to 2 MHz) with a maximum magnitude of -60 kV. The developed device was capable of ionizing air samples and implanting the resulting ions into a plastic scintillator. Two different methods to enhance the collection and deposition of the tritium ions, have been proposed and assessed. A movable prototype device for monitoring environmental tritium in air has been designed and constructed. Although this prototype was not fully tested, the primary calculations have shown that measurable concentrations of tritium ions can be collected from an air sample, with tritium activity ranging from 0.3 Bq/cm3 down to 0.03 mBq/cm3, in a short time, to the order of seconds, on-line. This sensitivity fulfills the requirement for environmental monitoring.

Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

NASA Astrophysics Data System (ADS)

Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

2014-04-01

CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

An update of the distribution of selected radiochemical and chemical constituents in perched ground water, Idaho National Laboratory, Idaho, Emphasis 1999-2001

USGS Publications Warehouse

Davis, Linda C.

2006-01-01

Radiochemical and chemical wastes generated at facilities at the Idaho National Laboratory (INL) were discharged since 1952 to infiltration ponds at the Reactor Technology Complex (RTC) (known as the Test Reactor Area [TRA] until 2005), and the Idaho Nuclear Technology and Engineering Center (INTEC) and buried at the Radioactive Waste Management Complex (RWMC). Disposal of wastewater to infiltration ponds and infiltration of surface water at waste burial sites resulted in formation of perched ground water in basalts and in sedimentary interbeds above the Snake River Plain aquifer. Perched ground water is an integral part of the pathway for waste-constituent migration to the aquifer. The U.S. Geological Survey (USGS), in cooperation with the U.S. Department of Energy, maintains ground-water monitoring networks at the INL to determine hydrologic trends, and to monitor the movement of radiochemical and chemical constituents in wastewater discharged from facilities to both perched ground water and the aquifer. This report presents an analysis of water-quality and water-level data collected from wells completed in perched ground water at the INL during 1999-2001, and summarizes historical disposal data and water-level-and water-quality trends. At the RTC, tritium, strontium-90, cesium-137, dissolved chromium, chloride, sodium, and sulfate were monitored in shallow and deep perched ground water. In shallow perched ground water, no tritium was detected above the reporting level. In deep perched ground water, tritium concentrations generally decreased or varied randomly during 1999-2001. During October 2001, tritium concentrations ranged from less than the reporting level to 39.4?1.4 picocuries per milliliter (pCi/mL). Reportable concentrations of tritium during July-October 2001 were smaller than the reported concentrations measured during July-December 1998. Tritium concentrations in water from wells at the RTC were likely affected by: well's distance from the radioactive-waste infiltration ponds (commonly referred to as the warm-waste ponds); water depth below the ponds; the amount of tritium discharged to radioactive-waste infiltration ponds in the past; discontinued use of radioactive-waste infiltration ponds; radioactive decay; and dilution from disposal of nonradioactive water. During 1999-2001, the strontium-90 concentrations in two wells completed in shallow perched water near the RTC exceeded the reporting level. Strontium-90 concentrations in water from wells completed in deep perched ground water at the RTC varied randomly with time. During October 2001, concentrations in water from five wells exceeded the reporting level and ranged from 2.8?0.7 picocuries per liter (pCi/L) in well USGS 63 to 83.8?2.1 pCi/L in well USGS 54. No reportable concentrations of cesium-137, chromium-51, or cobalt-60 were present in water samples from any of the shallow or deep wells at the RTC during 1999-2001. Dissolved chromium was not detected in shallow perched ground water at the RTC during 1999-2001. Concentrations of dissolved chromium during July-October 2001 in deep perched ground water near the RTC ranged from 10 micrograms per liter (?g/L) in well USGS 61 to 82 ?g/L in well USGS 55. The largest concentrations were in water from wells north and west of the radioactive-waste infiltration ponds. During July-October 2001, dissolved sodium concentrations ranged from 7 milligrams per liter (mg/L) in well USGS 78 to 20 mg/L in all wells except well USGS 68 (413 mg/L). Dissolved chloride concentrations in shallow perched ground water ranged from 10 mg/L in wells CWP 1, 3, and 4 to 53 mg/L in well TRA A 13 during 1999-2001. Dissolved chloride concentrations in deep perched ground water ranged from 5 mg/L in well USGS 78 to 91 mg/L in well USGS 73. The maximum dissolved sulfate concentration in shallow perched ground water was 419 mg/L in well CWP 1 during July 2000. Concentrations of dissolved sulfate in water from wells USGS 54, 60

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

NASA Astrophysics Data System (ADS)

Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

2016-02-01

China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

Precision shock tuning on the national ignition facility.

PubMed

Robey, H F; Celliers, P M; Kline, J L; Mackinnon, A J; Boehly, T R; Landen, O L; Eggert, J H; Hicks, D; Le Pape, S; Farley, D R; Bowers, M W; Krauter, K G; Munro, D H; Jones, O S; Milovich, J L; Clark, D; Spears, B K; Town, R P J; Haan, S W; Dixit, S; Schneider, M B; Dewald, E L; Widmann, K; Moody, J D; Döppner, T D; Radousky, H B; Nikroo, A; Kroll, J J; Hamza, A V; Horner, J B; Bhandarkar, S D; Dzenitis, E; Alger, E; Giraldez, E; Castro, C; Moreno, K; Haynam, C; LaFortune, K N; Widmayer, C; Shaw, M; Jancaitis, K; Parham, T; Holunga, D M; Walters, C F; Haid, B; Malsbury, T; Trummer, D; Coffee, K R; Burr, B; Berzins, L V; Choate, C; Brereton, S J; Azevedo, S; Chandrasekaran, H; Glenzer, S; Caggiano, J A; Knauer, J P; Frenje, J A; Casey, D T; Johnson, M Gatu; Séguin, F H; Young, B K; Edwards, M J; Van Wonterghem, B M; Kilkenny, J; MacGowan, B J; Atherton, J; Lindl, J D; Meyerhofer, D D; Moses, E

2012-05-25

Ignition implosions on the National Ignition Facility [J. D. Lindl et al., Phys. Plasmas 11, 339 (2004)] are underway with the goal of compressing deuterium-tritium fuel to a sufficiently high areal density (ρR) to sustain a self-propagating burn wave required for fusion power gain greater than unity. These implosions are driven with a very carefully tailored sequence of four shock waves that must be timed to very high precision to keep the fuel entropy and adiabat low and ρR high. The first series of precision tuning experiments on the National Ignition Facility, which use optical diagnostics to directly measure the strength and timing of all four shocks inside a hohlraum-driven, cryogenic liquid-deuterium-filled capsule interior have now been performed. The results of these experiments are presented demonstrating a significant decrease in adiabat over previously untuned implosions. The impact of the improved shock timing is confirmed in related deuterium-tritium layered capsule implosions, which show the highest fuel compression (ρR~1.0 g/cm(2)) measured to date, exceeding the previous record [V. Goncharov et al., Phys. Rev. Lett. 104, 165001 (2010)] by more than a factor of 3. The experiments also clearly reveal an issue with the 4th shock velocity, which is observed to be 20% slower than predictions from numerical simulation.

A near one-dimensional indirectly driven implosion at convergence ratio 30

NASA Astrophysics Data System (ADS)

MacLaren, S. A.; Masse, L. P.; Czajka, C. E.; Khan, S. F.; Kyrala, G. A.; Ma, T.; Ralph, J. E.; Salmonson, J. D.; Bachmann, B.; Benedetti, L. R.; Bhandarkar, S. D.; Bradley, P. A.; Hatarik, R.; Herrmann, H. W.; Mariscal, D. A.; Millot, M.; Patel, P. K.; Pino, J. E.; Ratledge, M.; Rice, N. G.; Tipton, R. E.; Tommasini, R.; Yeamans, C. B.

2018-05-01

Inertial confinement fusion cryogenic-layered implosions at the National Ignition Facility, while successfully demonstrating self-heating due to alpha-particle deposition, have fallen short of the performance predicted by one-dimensional (1D) multi-physics implosion simulations. The current understanding, from experimental evidence as well as simulations, suggests that engineering features such as the capsule tent and fill tube, as well as time-dependent low-mode asymmetry, are to blame for the lack of agreement. A short series of experiments designed specifically to avoid these degradations to the implosion are described here in order to understand if, once they are removed, a high-convergence cryogenic-layered deuterium-tritium implosion can achieve the 1D simulated performance. The result is a cryogenic layered implosion, round at stagnation, that matches closely the performance predicted by 1D simulations. This agreement can then be exploited to examine the sensitivity of approximations in the model to the constraints imposed by the data.

Measurements of the 169Tm(n ,2 n )168Tm cross section from threshold to 15 MeV

NASA Astrophysics Data System (ADS)

Soter, J.; Bhike, M.; Finch, S. W.; Krishichayan, Tornow, W.

2017-12-01

Measurements of the 169Tm(n ,2 n )168Tm cross section have been performed via the activation technique at 13 energies between 8.5 and 15.0 MeV. The purpose of this comprehensive data set is to provide an alternative diagnostic tool for obtaining subtle information on the neutron energy distribution produced in inertial confinement deuterium-tritium fusion experiments at the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory. The 169Tm(n ,2 n )168Tm reaction not only provides the primary 14-MeV neutron fluence, but also the important down-scattered neutron fluence, the latter providing information on the density achieved in the deuterium-tritium plasma during a laser shot.

Evidence for Stratification of Deuterium-Tritium Fuel in Inertial Confinement Fusion Implosions

NASA Astrophysics Data System (ADS)

Casey, D. T.; Frenje, J. A.; Gatu Johnson, M.; Manuel, M. J.-E.; Rinderknecht, H. G.; Sinenian, N.; Séguin, F. H.; Li, C. K.; Petrasso, R. D.; Radha, P. B.; Delettrez, J. A.; Glebov, V. Yu; Meyerhofer, D. D.; Sangster, T. C.; McNabb, D. P.; Amendt, P. A.; Boyd, R. N.; Rygg, J. R.; Herrmann, H. W.; Kim, Y. H.; Bacher, A. D.

2012-02-01

Measurements of the D(d,p)T (dd) and T(t,2n)He4 (tt) reaction yields have been compared with those of the D(t,n)He4 (dt) reaction yield, using deuterium-tritium gas-filled inertial confinement fusion capsule implosions. In these experiments, carried out on the OMEGA laser, absolute spectral measurements of dd protons and tt neutrons were obtained. From these measurements, it was concluded that the dd yield is anomalously low and the tt yield is anomalously high relative to the dt yield, an observation that we conjecture to be caused by a stratification of the fuel in the implosion core. This effect may be present in ignition experiments planned on the National Ignition Facility.

Evidence for stratification of deuterium-tritium fuel in inertial confinement fusion implosions.

PubMed

Casey, D T; Frenje, J A; Johnson, M Gatu; Manuel, M J-E; Rinderknecht, H G; Sinenian, N; Séguin, F H; Li, C K; Petrasso, R D; Radha, P B; Delettrez, J A; Glebov, V Yu; Meyerhofer, D D; Sangster, T C; McNabb, D P; Amendt, P A; Boyd, R N; Rygg, J R; Herrmann, H W; Kim, Y H; Bacher, A D

2012-02-17

Measurements of the D(d,p)T (dd) and T(t,2n)(4)He (tt) reaction yields have been compared with those of the D(t,n)(4)He (dt) reaction yield, using deuterium-tritium gas-filled inertial confinement fusion capsule implosions. In these experiments, carried out on the OMEGA laser, absolute spectral measurements of dd protons and tt neutrons were obtained. From these measurements, it was concluded that the dd yield is anomalously low and the tt yield is anomalously high relative to the dt yield, an observation that we conjecture to be caused by a stratification of the fuel in the implosion core. This effect may be present in ignition experiments planned on the National Ignition Facility.

RAMI modeling of plant systems for proposed tritium production and extraction facilities

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

Blanchard, A.

2000-04-05

The control of life-cycle cost is a primary concern during the development, construction, operation, and decommissioning of DOE systems and facilities. An effective tool that can be used to control these costs, beginning with the design stage, is called a reliability, availability, maintainability, and inspectability analysis or, simply, RAMI for short. In 1997, RAMI technology was introduced to the Savannah River Site with applications at the conceptual design stage beginning with the Accelerator Production of Tritium (APT) Project and later extended to the Commercial Light Water Reactor (CLWR) Tritium Extraction Facility (TEF) Project. More recently it has been applied tomore » the as-build Water Treatment Facilities designed for ground water environmental restoration. This new technology and database was applied to the assessment of balance-of-plant systems for the APT Conceptual Design Report. Initial results from the Heat Removal System Assessment revealed that the system conceptual design would cause the APT to fall short of its annual production goal. Using RAM technology to immediately assess this situation, it was demonstrated that the product loss could be gained back by upgrading the system's chiller unit capacity at a cost of less than $1.3 million. The reclaimed production is worth approximately $100 million. The RAM technology has now been extended to assess the conceptual design for the CLWR-TEF Project. More specifically, this technology and database is being used to translate high level availability goals into lower level system design requirements that will ensure the TEF meets its production goal. Results, from the limited number of system assessments performed to date, have already been used to modify the conceptual design for a remote handling system, improving its availability to the point that a redundant system, with its associated costs of installation and operation may no longer be required. RAMI results were also used to justify the elimination of a metal uranium bed in the design of a water cracker system, producing a significant reduction in the estimated construction and operating costs.« less

Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility.

PubMed

Döppner, T; Callahan, D A; Hurricane, O A; Hinkel, D E; Ma, T; Park, H-S; Berzak Hopkins, L F; Casey, D T; Celliers, P; Dewald, E L; Dittrich, T R; Haan, S W; Kritcher, A L; MacPhee, A; Le Pape, S; Pak, A; Patel, P K; Springer, P T; Salmonson, J D; Tommasini, R; Benedetti, L R; Bond, E; Bradley, D K; Caggiano, J; Church, J; Dixit, S; Edgell, D; Edwards, M J; Fittinghoff, D N; Frenje, J; Gatu Johnson, M; Grim, G; Hatarik, R; Havre, M; Herrmann, H; Izumi, N; Khan, S F; Kline, J L; Knauer, J; Kyrala, G A; Landen, O L; Merrill, F E; Moody, J; Moore, A S; Nikroo, A; Ralph, J E; Remington, B A; Robey, H F; Sayre, D; Schneider, M; Streckert, H; Town, R; Turnbull, D; Volegov, P L; Wan, A; Widmann, K; Wilde, C H; Yeamans, C

2015-07-31

We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a "high-foot" laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shape closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 10^{16} neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.

Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility

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

Döppner, T.; Callahan, D. A.; Hurricane, O. A.

We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a “highfoot” laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shapemore » closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 10 16 neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.« less

Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility

DOE PAGES

Döppner, T.; Callahan, D. A.; Hurricane, O. A.; ...

2015-07-28

We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a “highfoot” laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shapemore » closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 10 16 neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.« less

Annual INTEC Groundwater Monitoring Report for Group 5 - Snake River Plain Aquifer (2001)

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

Roddy, Michael Scott

2002-02-01

This report describes the monitoring activities conducted and presents the results of groundwater sampling and water-level measurements from October 2000 to September 2001. Groundwater samples were initially collected from 41 wells from the Idaho Nuclear Technology and Engineering Center and the Central Facilities Area and analyzed for iodine-129, strontium-90, tritium, gross alpha, gross beta, technetium-99, uranium isotopes, plutonium isotopes, neptunium-237, americium-241, gamma spectrometry, and mercury. Samples from 41 wells were collected in April and May 2001. Additional sampling was conducted in August 2001 and included the two CFA production wells, the CFA point of compliance for the production wells, onemore » well that was previously sampled and five additional monitoring wells. Iodine-129 and strontium-90 were the only analytes above their respective maximum contaminant levels. Iodine-129 was detected just above its maximum contaminant level of 1 pCi/L at two of the Central Facilities Area landfill wells. Iodine-129 was detected in the CFA production wells at 0.35±0.083 pCi/L in CFA-1, but was below detectable activity in CFA-2. Strontium-90 was above its maximum contaminant level of 8 pCi/L in several wells near the Idaho Nuclear Technology and Engineering Center but was below its maximum contaminant level in the downgradient wells at the Central Facilities Area landfills. Sr-90 was not detected in the CFA production wells. Gross beta results generally mirrored the results for strontium-90 and technetium-99. Plutonium isotopes and neptunium-237 were not detected. Uranium-233/234 and uranium-238 isotopes were detected in all samples. Concentrations of background and site wells were similar and are within background limits for total uranium determined by the USGS, suggesting that the concentrations are background. Uranium-235/236 was detected in 11 samples, but all the detected concentrations were similar and near the minimum detectable activity. Americium-241 was detected at three locations near the minimum detectable activity of approximately 0.07 pCi/L. The gamma spectrometry results detected cesium-137 in three samples, potassium-40 at eight locations, and radium-226 at one location. Mercury was below its maximum contaminant level of 2 µg/L in all samples. Gamma spectrometry results for the CFA production wells did not detect any analytes. Water-level measurements were taken from wells in the Idaho Nuclear Technology and Engineering Center, Central Facilities Area, and the area south of Central Facilities Area to evaluate groundwater flow directions. Water-level measurements indicated groundwater flow to the south-southwest from the Idaho Nuclear Technology and Engineering Center.« less

Current status of tritium calorimetry at TLK

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

Buekki-Deme, A.; Alecu, C.G.; Kloppe, B.

2015-03-15

Inside a tritium facility, calorimetry is an important analytical method as it is the only reference method for accountancy (it is based on the measurement of the heat generated by the radioactive decay). Presently, at Tritium Laboratory Karlsruhe (TLK), 4 calorimeters are in operation, one of isothermal type and three of inertial guidance control type (IGC). The volume of the calorimeters varies between 0.5 and 20.6 liters. About two years ago we started an extensive work to improve our calorimeters with regard to reliability and precision. We were forced to upgrade 3 of our 4 calorimeters due to the outdatedmore » interfaces and software. This work involved creating new LabView programs driving the devices, re-tuning control loops and replacing obsolete hardware components. In this paper we give a review on the current performance of our calorimeters, comparing it to recently available devices from the market and in the literature. We also show some ideas for a next generation calorimeter based on experiences with our IGC calorimeters and other devices reported in the literature. (authors)« less

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

Hitchcock, Daniel; Barton, Christopher, D.; Rebel, Karin, T.

Hitchcock, Daniel R., C.D. Barton, K.T. Rebel, J. Singer, J.C. Seaman, J.D. Strawbridge, S.J. Riha, and J.I. Blake. 2005. A containment and disposition strategy for tritium-contaminated groundwater at the Savannah River Site, South Carolina, United States.. Env. Geosci. 12(1): 17-28. Abstract - A containment and disposition water management strategy has been implemented at the Savannah River Site to minimize the discharge of tritiated groundwater from the Old Radioactive Waste Burial Ground to Four Mile Branch, a tributary of the Savannah River. This paper presents a general overview of the water management strategy, which includes a two-component (pond and irrigation) system,more » and a summary of operations and effectiveness for the first 3 yr of operations. Tritiated groundwater seep discharge was impounded by a dam and distributed via irrigation to a 22-ac (8.9-ha) upland forested area comprised of mixed pines (loblolly and slash) and hardwoods(primarily sweetgum and laurel oak). As of March 2004, the system has irrigated approximately 133.2 million L (35.2 million gal) and prevented approximately 1880 Ci of tritium from entering Four Mile Branch via forest evapotranspiration, as well as via pond storage and evaporation. Prior to installation of the containment and disposition strategy, tritium activity in Four Mile Branch downstream of the seep averaged approximately 500 pCi mL_1. Six months after installation, tritium activity averaged approximately 200 pCi mL_1 in Fourmile Branch. After 1 yr of operations, tritium activity averaged below 100 pCi mL_1 in Fourmile Branch, and a range of 100-200 pCi mL_1 tritium activity has been maintained as of March 2004. Complex hydrological factors and operational strategies influence remediation system success. Analyses may assist in developing groundwater management and remediation strategies for future projects at the Savannah River Site and other facilities located on similar landscapes.« less

Correlation of rates of tritium migration through porous concrete

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

Fukada, S.; Katayama, K.; Takeishi, T.

In a nuclear facility when tritium leaks from a glovebox to room accidentally, an atmosphere detritiation system (ADS) starts operating, and HTO released is recovered by ADS. ADS starts when tritium activity in air becomes higher than its controlled level. Before ADS operates, the laboratory walls are the final enclosure facing tritium and are usually made of porous concrete coated with a hydrophobic paint. In the present study, previous data on the diffusivity and adsorption coefficient of concrete and paints are reviewed. Tritium penetrates and migrates into concrete by following 3 ways. First, gaseous HT or T{sub 2} easily penetratesmore » into porous concrete. Its diffusivity is almost equal to that of H{sub 2}. When a gaseous molecule diffuses through pores with a smaller diameter than a mean free path, its migration rate is described by the Knudsen diffusion formula. The second mechanism is H{sub 2}O vapor diffusion in pores. Concrete holds a lot of structural water. Therefore, H{sub 2}O or HTO vapor can diffuse inside concrete pores along with adsorption-desorption and isotopic exchange with structural water, which is the third mechanism. Literature shows that the diffusivity of HTO through the epoxy-resin paint is determined as D(HTO)=1.0*10{sup -16} m{sup 2}/s. We have used this data to set a model and we have applied it to estimate residual tritium in laboratory walls. We have considered 2 accidental cases and a normal case: first, ADS starts operating 1 hour after 100 Ci HTO is released in the room, secondly, ADS starts 24 hours after 100 Ci HTO release and thirdly, when the walls are exposed to HTO for 10 years of normal operation. It appears that the immediate start up of ADS is indispensable for safety.« less

Experimental studies on tungsten-armour impact on nuclear responses of solid breeding blanket

NASA Astrophysics Data System (ADS)

Sato, Satoshi; Nakao, Makoto; Verzilov, Yury; Ochiai, Kentaro; Wada, Masayuki; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori; Nishitani, Takeo

2005-07-01

In order to experimentally evaluate the tungsten armour impact on tritium production of the solid breeding blanket being developed by JAERI for tokamak-type DEMO reactors, neutronics integral experiments have been performed using DT neutrons at the Fusion Neutron Source facility of JAERI. Solid breeding blanket mockups relevant to the DEMO blanket have been applied in this study. The mockups are made of a set of layers consisting of 0-25.2 mm thick tungsten, 16 mm thick F82H, 12 mm thick Li2TiO3 and 100-200 mm thick beryllium with a cross-section of 660 × 660 mm in maximum. Pellets of Li2CO3 are embedded in the Li2TiO3 layers to measure the tritium production rate. By installing the 5 mm, 12.6 mm and 25.2 mm thick tungsten armours, the sum of the integrated tritium productions at the pellets are reduced by about 2.1%, 2.5% and 6.1% relative to the case without the armour, respectively. Numerical calculations have been conducted using the Monte Carlo code. In the case of the mockups with the tungsten armour, calculation results for the sum of the integrated tritium productions agree well with the experimental data within 4% and 19% in the experiments without and with a neutron reflector, respectively.

Human Health and the Biological Effects of Tritium in Drinking Water: Prudent Policy Through Science – Addressing the ODWAC New Recommendation

PubMed Central

Dingwall, S.; Mills, C.E.; Phan, N.; Taylor, K.; Boreham, D.R.

2011-01-01

Tritium is a radioactive form of hydrogen and is a by-product of energy production in Canadian Deuterium Uranium (CANDU) reactors. The release of this radioisotope into the environment is carefully managed at CANDU facilities in order to minimize radiation exposure to the public. However, under some circumstances, small accidental releases to the environment can occur. The radiation doses to humans and non-human biota from these releases are low and orders of magnitude less than doses received from naturally occurring radioisotopes or from manmade activities, such as medical imaging and air travel. There is however a renewed interest in the biological consequences of low dose tritium exposures and a new limit for tritium levels in Ontario drinking water has been proposed. The Ontario Drinking Water Advisory Council (ODWAC) issued a formal report in May 2009 in response to a request by the Minister of the Environment, concluding that the Ontario Drinking Water Quality Standard for tritium should be revised from the current 7,000 Bq/L level to a new, lower 20 Bq/L level. In response to this recommendation, an international scientific symposium was held at McMaster University to address the issues surrounding this change in direction and the validity of a new policy. Scientists, regulators, government officials, and industrial stakeholders were present to discuss the potential health risks associated with low level radiation exposure from tritium. The regulatory, economic, and social implications of the new proposed limit were also considered. The new recommendation assumed a linear-no-threshold model to calculate carcinogenic risk associated with tritium exposure, and considered tritium as a non-threshold chemical carcinogen. Both of these assumptions are highly controversial given that recent research suggests that low dose exposures have thresholds below which there are no observable detrimental effects. Furthermore, mutagenic and carcinogenic risk calculated from tritium exposure at 20 Bq/L would be orders of magnitude less than that from exposure to natural background sources of radiation. The new proposed standard would set the radiation dose limit for drinking water to 0.0003 mSv/year, which is equivalent to approximately three times the dose from naturally occurring tritium in drinking water. This new standard is incongruent with national and international standards for safe levels of radiation exposure, currently set at 1 mSv/year for the general public. Scientific research from leading authorities on the carcinogenic health effects of tritium exposure supports the notion that the current standard of 7,000 Bq/L (annual dose of 0.1 mSv) is a safe standard for human health. Policy-making for the purpose of regulating tritium levels in drinking water is a dynamic multi-stage process that is influenced by more than science alone. Ethics, economics, and public perception also play important roles in policy development; however, these factors sometimes undermine the scientific evidence that should form the basis of informed decision making. Consequently, implementing a new standard without a scientific basis may lead the public to perceive that risks from tritium have been historically underestimated. It was concluded that the new recommendation is not supported by any new scientific insight regarding negative consequences of low dose effects, and may be contrary to new data on the potential benefits of low dose effects. Given the lack of cost versus benefit analysis, this type of dramatic policy change could have detrimental effects to society from an ethical, economical, and public perception perspective. PMID:21431084

Human Health and the Biological Effects of Tritium in Drinking Water: Prudent Policy Through Science - Addressing the ODWAC New Recommendation.

PubMed

Dingwall, S; Mills, C E; Phan, N; Taylor, K; Boreham, D R

2011-02-22

Tritium is a radioactive form of hydrogen and is a by-product of energy production in Canadian Deuterium Uranium (CANDU) reactors. The release of this radioisotope into the environment is carefully managed at CANDU facilities in order to minimize radiation exposure to the public. However, under some circumstances, small accidental releases to the environment can occur. The radiation doses to humans and non-human biota from these releases are low and orders of magnitude less than doses received from naturally occurring radioisotopes or from manmade activities, such as medical imaging and air travel. There is however a renewed interest in the biological consequences of low dose tritium exposures and a new limit for tritium levels in Ontario drinking water has been proposed. The Ontario Drinking Water Advisory Council (ODWAC) issued a formal report in May 2009 in response to a request by the Minister of the Environment, concluding that the Ontario Drinking Water Quality Standard for tritium should be revised from the current 7,000 Bq/L level to a new, lower 20 Bq/L level. In response to this recommendation, an international scientific symposium was held at McMaster University to address the issues surrounding this change in direction and the validity of a new policy. Scientists, regulators, government officials, and industrial stakeholders were present to discuss the potential health risks associated with low level radiation exposure from tritium. The regulatory, economic, and social implications of the new proposed limit were also considered.The new recommendation assumed a linear-no-threshold model to calculate carcinogenic risk associated with tritium exposure, and considered tritium as a non-threshold chemical carcinogen. Both of these assumptions are highly controversial given that recent research suggests that low dose exposures have thresholds below which there are no observable detrimental effects. Furthermore, mutagenic and carcinogenic risk calculated from tritium exposure at 20 Bq/L would be orders of magnitude less than that from exposure to natural background sources of radiation. The new proposed standard would set the radiation dose limit for drinking water to 0.0003 mSv/year, which is equivalent to approximately three times the dose from naturally occurring tritium in drinking water. This new standard is incongruent with national and international standards for safe levels of radiation exposure, currently set at 1 mSv/year for the general public. Scientific research from leading authorities on the carcinogenic health effects of tritium exposure supports the notion that the current standard of 7,000 Bq/L (annual dose of 0.1 mSv) is a safe standard for human health.Policy-making for the purpose of regulating tritium levels in drinking water is a dynamic multi-stage process that is influenced by more than science alone. Ethics, economics, and public perception also play important roles in policy development; however, these factors sometimes undermine the scientific evidence that should form the basis of informed decision making. Consequently, implementing a new standard without a scientific basis may lead the public to perceive that risks from tritium have been historically underestimated. It was concluded that the new recommendation is not supported by any new scientific insight regarding negative consequences of low dose effects, and may be contrary to new data on the potential benefits of low dose effects. Given the lack of cost versus benefit analysis, this type of dramatic policy change could have detrimental effects to society from an ethical, economical, and public perception perspective.

LIFE: a sustainable solution for developing safe, clean fusion power.

PubMed

Reyes, Susana; Dunne, Mike; Kramer, Kevin; Anklam, Tom; Havstad, Mark; Mazuecos, Antonio Lafuente; Miles, Robin; Martinez-Frias, Joel; Deri, Bob

2013-06-01

The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in California is currently in operation with the goal to demonstrate fusion energy gain for the first time in the laboratory-also referred to as "ignition." Based on these demonstration experiments, the Laser Inertial Fusion Energy (LIFE) power plant is being designed at LLNL in partnership with other institutions with the goal to deliver baseload electricity from safe, secure, sustainable fusion power in a time scale that is consistent with the energy market needs. For this purpose, the LIFE design takes advantage of recent advances in diode-pumped, solid-state laser technology and adopts the paradigm of Line Replaceable Units used on the NIF to provide high levels of availability and maintainability and mitigate the need for advanced materials development. The LIFE market entry plant will demonstrate the feasibility of a closed fusion fuel cycle, including tritium breeding, extraction, processing, refueling, accountability, and safety, in a steady-state power-producing device. While many fusion plant designs require large quantities of tritium for startup and operations, a range of design choices made for the LIFE fuel cycle act to reduce the in-process tritium inventory. This paper presents an overview of the delivery plan and the preconceptual design of the LIFE facility with emphasis on the key safety design principles being adopted. In order to illustrate the favorable safety characteristics of the LIFE design, some initial accident analysis results are presented that indicate potential for a more attractive licensing regime than that of current fission reactors.

124Xe(n,γ)125Xe and 124Xe(n,2n)123Xe measurements for National Ignition Facility

NASA Astrophysics Data System (ADS)

Bhike, Megha; Ludin, Nurin; Tornow, Werner

2015-05-01

The cross section for the 124Xe(n,γ)125Xe reaction has been measured for the first time for neutron energies above 100 keV. In addition, the 124Xe(n,2n)123Xe reaction has been studied between threshold and 14.8 MeV. The results of these measurements provide sensitive diagnostic tools for investigating properties of the inertial confinement fusion plasma in Deuterium-Tritium (DT) capsules at the National Ignition Facility (NIF) located at Lawrence Livermore National Laboratory.

Measurements of the Tm 169 ( n , 2 n ) Tm 168 cross section from threshold to 15 MeV

DOE PAGES

Soter, J.; Bhike, M.; Finch, S. W.; ...

2017-12-27

Measurements of the 169Tm(n,2n) 168Tm cross section have been performed via the activation technique at 13 energies between 8.5 and 15.0 MeV. The purpose of this comprehensive data set is to provide an alternative diagnostic tool for obtaining subtle information on the neutron energy distribution produced in inertial confinement deuterium-tritium fusion experiments at the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory. In conclusion, the 169Tm(n,2n) 168Tm reaction not only provides the primary 14-MeV neutron fluence, but also the important down-scattered neutron fluence, the latter providing information on the density achieved in the deuterium-tritium plasma during a laser shot.

HTO and OBT activity concentrations in soil at the historical atmospheric HT release site (Chalk River Laboratories).

PubMed

Kim, S B; Bredlaw, M; Korolevych, V Y

2012-01-01

Tritium is routinely released by the Chalk River Laboratories (CRL) nuclear facilities. Three International HT release experiments have been conducted at the CRL site in the past. The site has not been disturbed since the last historical atmospheric testing in 1994 and presents an opportunity to assess the retention of tritium in soil. This study is devoted to the measurement of HTO and OBT activity concentration profiles in the subsurface 25 cm of soil. In terms of soil HTO, there is no evidence from the past HT release experiments that HTO was retained. The HTO activity concentration in the soil pore water appears similar to concentrations found in background areas in Ontario. In contrast, OBT activity concentrations in soil at the same site were significantly higher than HTO activity concentrations in soil. Elevated OBT appears to reside in the top layer of the soil (0-5 cm). In addition, OBT activity concentrations in the top soil layer did not fluctuate much with season, again, quite in contrast with soil HTO. This result suggests that OBT activity concentrations retained the signature of the historical tritium releases. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

Indications of flow near maximum compression in layered deuterium-tritium implosions at the National Ignition Facility

DOE PAGES

Gatu Johnson, M.; Knauer, J. P.; Cerjan, C. J.; ...

2016-08-15

Here, an accurate understanding of burn dynamics in implosions of cryogenically layered deuterium (D) and tritium (T) filled capsules, obtained partly through precision diagnosis of these experiments, is essential for assessing the impediments to achieving ignition at the National Ignition Facility. We present measurements of neutrons from such implosions. The apparent ion temperatures T ion are inferred from the variance of the primary neutron spectrum. Consistently higher DT than DD T ion are observed and the difference is seen to increase with increasing apparent DT T ion. The line-of-sight rms variations of both DD and DT T ion are small,more » ~150eV, indicating an isotropic source. The DD neutron yields are consistently high relative to the DT neutron yields given the observed T ion. Spatial and temporal variations of the DT temperature and density, DD-DT differential attenuation in the surrounding DT fuel, and fluid motion variations contribute to a DT Tion greater than the DD T ion, but are in a one-dimensional model insufficient to explain the data. We hypothesize that in a three-dimensional interpretation, these effects combined could explain the results.« less

Indications of flow near maximum compression in layered deuterium-tritium implosions at the National Ignition Facility

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

Gatu Johnson, M.; Knauer, J. P.; Cerjan, C. J.

Here, an accurate understanding of burn dynamics in implosions of cryogenically layered deuterium (D) and tritium (T) filled capsules, obtained partly through precision diagnosis of these experiments, is essential for assessing the impediments to achieving ignition at the National Ignition Facility. We present measurements of neutrons from such implosions. The apparent ion temperatures T ion are inferred from the variance of the primary neutron spectrum. Consistently higher DT than DD T ion are observed and the difference is seen to increase with increasing apparent DT T ion. The line-of-sight rms variations of both DD and DT T ion are small,more » ~150eV, indicating an isotropic source. The DD neutron yields are consistently high relative to the DT neutron yields given the observed T ion. Spatial and temporal variations of the DT temperature and density, DD-DT differential attenuation in the surrounding DT fuel, and fluid motion variations contribute to a DT Tion greater than the DD T ion, but are in a one-dimensional model insufficient to explain the data. We hypothesize that in a three-dimensional interpretation, these effects combined could explain the results.« less

Efficiencies of Tritium (3H) bubbling systems.

PubMed

Duda, Jean-Marie; Le Goff, Pierre; Leblois, Yoan; Ponsard, Samuel

2018-09-01

Bubbling systems are among the devices most used by nuclear operators to measure atmospheric tritium activity in their facilities or the neighbouring environment. However, information about trapping efficiency and bubbling system oxidation is not accessible and/or, at best, only minimally supported by demonstrations in actual operating conditions. In order to evaluate easily these parameters and thereby meet actual normative and regulatory requirements, a statistical study was carried out over 2000 monitoring records from the CEA Valduc site. From this data collection obtained over recent years of monitoring the CEA Valduc facilities and environment, a direct relation was highlighted between the 3H-samplers trapping efficiency of tritium as tritiated water and the sampling time and conditions of use: temperature and atmospheric moisture. It was thus demonstrated that this efficiency originated from two sources. The first one is intrinsic to the bubbling system operating parameters and the sampling time. That part applies equally to all four bubblers. The second part, however, is specific to the first bubbler. In essence, it depends on the sampling time and the sampled air characteristics. It was also highlighted that the water volume variation in the first bubbler, between the beginning and the end of the sampling process, is directly related to the average water concentration of the sampled air. In this way, it was possible to model the variations in trapping efficiency of the 3H-samplers relative to the sampling time and the water volume variation in the first bubbler. This model makes it possible to obtain the quantities required to comply with the current standards governing the monitoring of radionuclides in the environment and to associate an uncertainty concerning the measurements as well as the sampling parameters. Copyright © 2018 Elsevier Ltd. All rights reserved.

First measurements of deuterium-tritium and deuterium-deuterium fusion reaction yields in ignition-scalable direct-drive implosions

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

Forrest, C. J.; Radha, P. B.; Knauer, J. P.

In this study, the deuterium-tritium (D-T) and deuterium-deuterium neutron yield ratio in cryogenic inertial confinement fusion (ICF) experiments is used to examine multifluid effects, traditionally not included in ICF modeling. This ratio has been measured for ignition-scalable direct-drive cryogenic DT implosions at the Omega Laser Facility using a high-dynamic-range neutron time-of-flight spectrometer. The experimentally inferred yield ratio is consistent with both the calculated values of the nuclear reaction rates and the measured preshot target-fuel composition. These observations indicate that the physical mechanisms that have been proposed to alter the fuel composition, such as species separation of the hydrogen isotopes, aremore » not significant during the period of peak neutron production in ignition-scalable cryogenic direct-drive DT implosions.« less

First measurements of deuterium-tritium and deuterium-deuterium fusion reaction yields in ignition-scalable direct-drive implosions

DOE PAGES

Forrest, C. J.; Radha, P. B.; Knauer, J. P.; ...

2017-03-03

In this study, the deuterium-tritium (D-T) and deuterium-deuterium neutron yield ratio in cryogenic inertial confinement fusion (ICF) experiments is used to examine multifluid effects, traditionally not included in ICF modeling. This ratio has been measured for ignition-scalable direct-drive cryogenic DT implosions at the Omega Laser Facility using a high-dynamic-range neutron time-of-flight spectrometer. The experimentally inferred yield ratio is consistent with both the calculated values of the nuclear reaction rates and the measured preshot target-fuel composition. These observations indicate that the physical mechanisms that have been proposed to alter the fuel composition, such as species separation of the hydrogen isotopes, aremore » not significant during the period of peak neutron production in ignition-scalable cryogenic direct-drive DT implosions.« less

Measurement of high-pressure shock waves in cryogenic deuterium-tritium ice layered capsule implosions on NIF.

PubMed

Robey, H F; Moody, J D; Celliers, P M; Ross, J S; Ralph, J; Le Pape, S; Berzak Hopkins, L; Parham, T; Sater, J; Mapoles, E R; Holunga, D M; Walters, C F; Haid, B J; Kozioziemski, B J; Dylla-Spears, R J; Krauter, K G; Frieders, G; Ross, G; Bowers, M W; Strozzi, D J; Yoxall, B E; Hamza, A V; Dzenitis, B; Bhandarkar, S D; Young, B; Van Wonterghem, B M; Atherton, L J; Landen, O L; Edwards, M J; Boehly, T R

2013-08-09

The first measurements of multiple, high-pressure shock waves in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility have been performed. The strength and relative timing of these shocks must be adjusted to very high precision in order to keep the DT fuel entropy low and compressibility high. All previous measurements of shock timing in inertial confinement fusion implosions [T. R. Boehly et al., Phys. Rev. Lett. 106, 195005 (2011), H. F. Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] have been performed in surrogate targets, where the solid DT ice shell and central DT gas regions were replaced with a continuous liquid deuterium (D2) fill. This report presents the first experimental validation of the assumptions underlying this surrogate technique.

The LBM program at the EPFL/LOTUS Facility

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

File, J.; Jassby, D.L.; Tsang, F.Y.

1986-11-01

An experimental program of neutron transport studies of the Lithium Blanket Module (LBM) is being carried out with the LOTUS point-neutron source facility at Ecole Polytechnique Federale de Lausanne (EPFL), Switzerland. Preliminary experiments use passive neutron dosimetry within the fuel rods in the LBM central zone, as well as, both thermal extraction and dissolution methods to assay tritium bred in Li/sub 2/O diagnostic wafers and LBM pellets. These measurements are being compared and reconciled with each other and with the predictions of two-dimensional discrete-ordinates and continuous-energy Monte-Carlo analyses of the Lotus/LBM system.

Commercial D-T FRC Power Plant Systems Analysis

NASA Astrophysics Data System (ADS)

Nguyen, Canh; Santarius, John; Emmert, Gilbert; Steinhauer, Loren; Stubna, Michael

1998-11-01

Results of an engineering issues scoping study of a Field-Reversed Configuration (FRC) burning D-T fuel will be presented. The study primarily focuses on engineering issues, such as tritium-breeding blanket design, radiation shielding, neutron damage, activation, safety, and environment. This presentation will concentrate on plasma physics, current drive, economics, and systems integration, which are important for the overall systems analysis. A systems code serves as the key tool in defining a reference point for detailed physics and engineering calculations plus parametric variations, and typical cases will be presented. Advantages of the cylindrical geometry and high beta (plasma pressure/magnetic-field pressure) are evident.

25 Years Of Environmental Remediation In The General Separations Area Of The Savannah River Site: Lessons Learned About What Worked And What Did Not Work In Soil And Groundwater Cleanup

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

Blount, Gerald; Thibault, Jeffrey; Millings, Margaret

The Savannah River Site (SRS) is owned and administered by the US Department of Energy (DOE). SRS covers an area of approximately 900 square kilometers. The General Separation Area (GSA) is located roughly in the center of the SRS and includes: radioactive material chemical separations facilities, radioactive waste tank farms, a variety of radioactive seepage basins, and the radioactive waste burial grounds. Radioactive wastes were disposed in the GSA from the mid-1950s through the mid-1990s. Radioactive operations at the F Canyon began in 1954; radioactive operations at H Canyon began in 1955. Waste water disposition to the F and Hmore » Seepage Basins began soon after operations started in the canyons. The Old Radioactive Waste Burial Ground (ORWBG) began operations in 1952 to manage solid waste that could be radioactive from all the site operations, and ceased receiving waste in 1972. The Mixed Waste Management Facility (MWMF) and Low Level Radioactive Waste Disposal Facility (LLRWDF) received radioactive solid waste from 1969 until 1995. Environmental legislation enacted in the 1970s, 1980s, and 1990s led to changes in waste management and environmental cleanup practices at SRS. The US Congress passed the Clean Air Act in 1970, and the Clean Water Act in 1972; the Resource Conservation and Recovery Act (RCRA) was enacted in 1976; the Comprehensive Environmental Response Compensation, and Liability Act (CERCLA) was enacted by Congress in 1980; the Federal Facilities Compliance Act (FFCA) was signed into law in 1992. Environmental remediation at the SRS essentially began with a 1987 Settlement Agreement between the SRS and the State of South Carolina (under the South Carolina Department of Health and Environmental Control - SCDHEC), which recognized linkage between many SRS waste management facilities and RCRA. The SRS manages several of the larger groundwater remedial activities under RCRA for facilities recognized early on as environmental problems. All subsequent environmental remediation projects tend to be managed under tri-party agreement (DOE, Environmental Protection Agency, and SCDHEC) through the Federal Facilities Agreement. During 25 years of environmental remediation SRS has stabilized and capped seepage basins, and consolidated and capped waste units and burial grounds in the GSA. Groundwater activities include: pump and treat systems in the groundwater, installation of deep subsurface barrier systems to manage groundwater flow, in situ chemical treatments in the groundwater, and captured contaminated groundwater discharges at the surface for management in a forest irrigation system. Over the last 25 years concentrations of contaminants in the aquifers beneath the GSA and in surface water streams in the GSA have dropped significantly. Closure of 65 waste sites and 4 RCRA facilities has been successfully accomplished. Wastes have been successfully isolated in place beneath a variety of caps and cover systems. Environmental clean-up has progressed to the stage where most of the work involves monitoring, optimization, and maintenance of existing remedial systems. Many lessons have been learned in the process. Geotextile covers outperform low permeability clay caps, especially with respect to the amount of repairs required to upkeep the drainage layers as the caps age. Passive, enhanced natural processes to address groundwater contamination are much more cost effective than pump and treat systems. SRS operated two very large pump and treat systems at the F and H Seepage Basins to attempt to limit the release of tritium to Fourmile Branch, a tributary of the Savannah River. The systems were designed to extract contaminated acidic groundwater, remove all contamination except tritium (not possible to remove the tritium from the water), and inject the tritiated groundwater up-gradient of the source area and the plume. The concept was to increase the travel time of the injected water for radioactive decay of the tritium. The two systems were found to be non-effective and potentially mobilizing more contamination. SRS invested approximately $50 million in construction and approximately $100 million in 6 years of operation. The H Seepage Basin pump and treat system was replaced by a series of subsurface barriers that alters the groundwater velocity; the F Seepage Basin pump and treat system was replaced by subsurface barriers forming a funnel and gate augmented by chemical treatment within the gates. These replacement systems are mostly passive and cost approximately $13 million to construct, and have reduced the tritium flux to Fourmile Branch, in these plumes, by over 70%. SRS manages non-acidic tritiated groundwater releases to Fourmile Branch from the southwest plume of the MWMF with a forest irrigation system. Tritiated water is captured with a sheetpile dam below the springs that caused releases to Fourmile Branch. Water from the irrigation pond is pumped to a filter plant prior to irrigation of approximately 26 hectares of mixed forest and developing pine plantation. SRS has almost achieved a 70% reduction in tritium flux to the Branch from this plume. The system cost approximately $5 million to construct with operation cost of approximately $500K per year. In conclusion, many lessons have been learned in 25 years of relatively aggressive remedial activities in the GSA. Geotextile covers outperform low permeability clay caps, especially with respect to the amount of repairs required to upkeep the drainage layers as the caps age. Passive, enhanced natural processes to address groundwater contamination are much more cost effective than pump and treat systems. In water management situations with non-accumulative contaminants (tritium, VOCs, etc.) irrigation in a forest setting can be very effective.« less

Results from and Plans for the Development of the MARBLE Platform for Studying Thermonuclear Burn in the Presence of Heterogeneous Mix on OMEGA and the National Ignition Facility

NASA Astrophysics Data System (ADS)

Murphy, T. J.; Douglas, M. R.; Fincke, J. R.; Cobble, J. A.; Haines, B. M.; Hamilton, C. E.; Lee, M. N.; Oertel, J. A.; Olson, R. E.; Randolph, R. B.; Schmidt, D. W.; Shah, R. C.; Smidt, J. M.; Tregillis, I. L.

2015-11-01

Work is underway to develop the MARBLE ICF platform for use on OMEGA and NIF in experiments to quantify the influence of heterogeneous mix on fusion burn. This platform consists of a plastic (CH) capsule filled with a deuterated plastic foam (CD) with a density of a few tens of milligrams per cubic centimeter, with tritium gas filling the voids in the foam. In order to affect the morphology of the mix, engineered foams with voids of diameter up to 100 microns will be utilized. The degree of mix will be determined from the ratio of DT to DD neutron yield. Experiments have been performed on OMEGA and are planned for NIF to develop techniques and verify that with uniform fine-pore foam, these implosions behave like atomically mixed plastic and gas. Results will be reviewed and future experiments discussed. This work is supported by US DOE/NNSA, performed at LANL, operated by LANS LLC under contract DE-AC52-06NA25396.

Report on the oversight assessment of the operational readiness review of the Replacement Tritium Facility at Savannah River Site

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

Lee, B.T.

1993-03-01

This report presents the results of an oversight assessment (OA) conducted by the US Department of Energy's (DOE) Office of Environment, Safety and Health (EH) of operational readiness review (ORR) activities for the Replacement Tritium Facility (RTF) located at Savannah River Site (SRS). The EH OA of this facility took place concurrently with an ORR conducted by the DOE Office of Defense Programs (DP). The DP ORR was conducted from January 19 through February 5, 1993. The EH OA was performed in accordance with the protocol and procedures specified in EH Program for Oversight Assessment of Operational Readiness Evaluations formore » Startups and Restarts,'' dated September 15, 1992. The EH OA Team evaluated the DP ORR to determine whether it was thorough and demonstrated sufficient inquisitiveness to verify that the implementation of programs and procedures adequately ensures the protection of worker safety and health. The EH OA Team performed its evaluation of the DP ORR in the following technical areas: occupational safety, industrial hygiene, and respiratory protection; fire protection; and chemical safety. In the areas of fire protection and chemical safety, the EH OA Team conducted independent vertical-slice reviews to confirm DP ORR results. Within each technical area, the EH OA Team reviewed the DP ORR Plan, including the Criteria Review and Approach Documents (CRADs); the qualifications of individual DP ORR team members; the performance of planned DP ORR activities; and the results of the DP ORR.« less

Report on the oversight assessment of the operational readiness review of the Replacement Tritium Facility at Savannah River Site

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

Lee, B.T.

1993-03-01

This report presents the results of an oversight assessment (OA) conducted by the US Department of Energy`s (DOE) Office of Environment, Safety and Health (EH) of operational readiness review (ORR) activities for the Replacement Tritium Facility (RTF) located at Savannah River Site (SRS). The EH OA of this facility took place concurrently with an ORR conducted by the DOE Office of Defense Programs (DP). The DP ORR was conducted from January 19 through February 5, 1993. The EH OA was performed in accordance with the protocol and procedures specified in ``EH Program for Oversight Assessment of Operational Readiness Evaluations formore » Startups and Restarts,`` dated September 15, 1992. The EH OA Team evaluated the DP ORR to determine whether it was thorough and demonstrated sufficient inquisitiveness to verify that the implementation of programs and procedures adequately ensures the protection of worker safety and health. The EH OA Team performed its evaluation of the DP ORR in the following technical areas: occupational safety, industrial hygiene, and respiratory protection; fire protection; and chemical safety. In the areas of fire protection and chemical safety, the EH OA Team conducted independent vertical-slice reviews to confirm DP ORR results. Within each technical area, the EH OA Team reviewed the DP ORR Plan, including the Criteria Review and Approach Documents (CRADs); the qualifications of individual DP ORR team members; the performance of planned DP ORR activities; and the results of the DP ORR.« less

EFFECTS OF GAMMA IRRADIATION ON EPDM ELASTOMERS (REVISION 1)

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

Clark, E.

Two formulations of EPDM elastomer, one substituting a UV stabilizer for the normal antioxidant in this polymer, and the other the normal formulation, were synthesized and samples of each were exposed to gamma irradiation in initially pure deuterium gas to compare their radiation stability. Stainless steel containers having rupture disks were designed for this task. After 130 MRad dose of cobalt-60 radiation in the SRNL Gamma Irradiation Facility, a significant amount of gas was created by radiolysis; however the composition indicated by mass spectroscopy indicated an unexpected increase in the total amount deuterium in both formulations. The irradiated samples retainedmore » their ductility in a bend test. No change of sample weight, dimensions, or density was observed. No change of the glass transition temperature as measured by dynamic mechanical analysis was observed, and most of the other dynamic mechanical properties remained unchanged. There appeared to be an increase in the storage modulus of the irradiated samples containing the UV stabilizer above the glass transition, which may indicate hardening of the material by radiation damage. Revision 1 adds a comparison with results of a study of tritium exposed EPDM. The amount of gas produced by the gamma irradiation was found to be equivalent to about 280 days exposure to initially pure tritium gas at one atmosphere. The glass transition temperature of the tritium exposed EPDM rose about 10°C. over 280 days, while no glass transition temperature change was observed for gamma irradiated EPDM. This means that gamma irradiation in deuterium cannot be used as a surrogate for tritium exposure.« less

Reemission of Tritium from Tritium-Sorbed Molecular Sieve

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

Cao Xiaohua; Cheng Guijun

2005-07-15

In handling of tritium-containing waste gas, tritium is oxidized to tritiated water and immobilized in a molecular sieve (MS), which is then disposed of as solid radioactive waste. So reemission of tritium from tritium-sorbed molecular sieve is concerned for tritium waste disposal. 4A, 5A and 10X MS were chosen for the tritium reemission test. The tritium-containing MS samples with specific activity of 3 GBq/g were prepared and the reemission coefficients of tritium from the three types of MS were determined. The effects of storage conditions of the MS on the reemission of tritium were examined. The results show that duringmore » two months of storage period, the reemission coefficients of 4A, 5A and 10X MS are (1.9{approx}5.5) x 10{sup -6} d{sup -1}.g{sup -1}. Among them, 5A MS has the largest reemission coefficient and 4A MS the smallest. The tritium released from tritium-sorbed MS is mostly in the form of HTO, only less than 1.2% of the tritium is in the form of HT. The atmosphere for storing tritium-sorbed MS has rather effect on reemission of tritium. The reemission coefficient in argon is lower than that in Ar+2%H{sub 2}.« less

Demonstration of Fuel Hot-Spot Pressure in Excess of 50 Gbar for Direct-Drive, Layered Deuterium-Tritium Implosions on OMEGA

NASA Astrophysics Data System (ADS)

Regan, S. P.; Goncharov, V. N.; Igumenshchev, I. V.; Sangster, T. C.; Betti, R.; Bose, A.; Boehly, T. R.; Bonino, M. J.; Campbell, E. M.; Cao, D.; Collins, T. J. B.; Craxton, R. S.; Davis, A. K.; Delettrez, J. A.; Edgell, D. H.; Epstein, R.; Forrest, C. J.; Frenje, J. A.; Froula, D. H.; Gatu Johnson, M.; Glebov, V. Yu.; Harding, D. R.; Hohenberger, M.; Hu, S. X.; Jacobs-Perkins, D.; Janezic, R.; Karasik, M.; Keck, R. L.; Kelly, J. H.; Kessler, T. J.; Knauer, J. P.; Kosc, T. Z.; Loucks, S. J.; Marozas, J. A.; Marshall, F. J.; McCrory, R. L.; McKenty, P. W.; Meyerhofer, D. D.; Michel, D. T.; Myatt, J. F.; Obenschain, S. P.; Petrasso, R. D.; Radha, P. B.; Rice, B.; Rosenberg, M. J.; Schmitt, A. J.; Schmitt, M. J.; Seka, W.; Shmayda, W. T.; Shoup, M. J.; Shvydky, A.; Skupsky, S.; Solodov, A. A.; Stoeckl, C.; Theobald, W.; Ulreich, J.; Wittman, M. D.; Woo, K. M.; Yaakobi, B.; Zuegel, J. D.

2016-07-01

A record fuel hot-spot pressure Phs=56 ±7 Gbar was inferred from x-ray and nuclear diagnostics for direct-drive inertial confinement fusion cryogenic, layered deuterium-tritium implosions on the 60-beam, 30-kJ, 351-nm OMEGA Laser System. When hydrodynamically scaled to the energy of the National Ignition Facility, these implosions achieved a Lawson parameter ˜60 % of the value required for ignition [A. Bose et al., Phys. Rev. E 93, LM15119ER (2016)], similar to indirect-drive implosions [R. Betti et al., Phys. Rev. Lett. 114, 255003 (2015)], and nearly half of the direct-drive ignition-threshold pressure. Relative to symmetric, one-dimensional simulations, the inferred hot-spot pressure is approximately 40% lower. Three-dimensional simulations suggest that low-mode distortion of the hot spot seeded by laser-drive nonuniformity and target-positioning error reduces target performance.

Demonstration of fuel hot-spot pressure in excess of 50 Gbar for direct-drive, layered deuterium-tritium implosions on OMEGA

DOE PAGES

Regan, S. P.; Goncharov, V. N.; Igumenshchev, I. V.; ...

2016-07-07

A record fuel hot-spot pressure P hs = 56±7 Gbar was inferred from x-ray and nuclear diagnostics for direct-drive inertial confinement fusion cryogenic, layered deuterium–tritium implosions on the 60-beam, 30-kJ, 351-nm OMEGA Laser System. When hydrodynamically scaled to the energy of the National Ignition Facility (NIF), these implosions achieved a Lawson parameter ~60% of the value required for ignition [A. Bose et al., Phys. Rev. E (in press)], similar to indirect-drive implosions [R. Betti et al., Phys. Rev. Lett. 114, 255003 (2015)], and nearly half of the direct-drive ignition-threshold pressure. Relative to symmetric, one-dimensional simulations, the inferred hot-spot pressure ismore » ~40% lower. Furthermore, three-dimensional simulations suggest that low-mode distortion of the hot spot seeded by laser-drive nonuniformity and target-positioning error reduces target performance.« less

Demonstration of fuel hot-spot pressure in excess of 50 Gbar for direct-drive, layered deuterium-tritium implosions on OMEGA

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

Regan, S. P.; Goncharov, V. N.; Igumenshchev, I. V.

A record fuel hot-spot pressure P hs = 56±7 Gbar was inferred from x-ray and nuclear diagnostics for direct-drive inertial confinement fusion cryogenic, layered deuterium–tritium implosions on the 60-beam, 30-kJ, 351-nm OMEGA Laser System. When hydrodynamically scaled to the energy of the National Ignition Facility (NIF), these implosions achieved a Lawson parameter ~60% of the value required for ignition [A. Bose et al., Phys. Rev. E (in press)], similar to indirect-drive implosions [R. Betti et al., Phys. Rev. Lett. 114, 255003 (2015)], and nearly half of the direct-drive ignition-threshold pressure. Relative to symmetric, one-dimensional simulations, the inferred hot-spot pressure ismore » ~40% lower. Furthermore, three-dimensional simulations suggest that low-mode distortion of the hot spot seeded by laser-drive nonuniformity and target-positioning error reduces target performance.« less

Results from MARBLE DT Experiments on the National Ignition Facility: Implosion of Foam-Filled Capsules for Studying Thermonuclear Burn in the Presence of Heterogeneous Mix

NASA Astrophysics Data System (ADS)

Murphy, T. J.; Douglas, M. R.; Cardenas, T.; Cooley, J. H.; Gunderson, M. A.; Haines, B. M.; Hamilton, C. E.; Kim, Y.; Lee, M. N.; Oertel, J. A.; Olson, R. E.; Randolph, R. B.; Shah, R. C.; Smidt, J. M.

2017-10-01

The MARBLE campaign on NIF investigates the effect of heterogeneous mix on thermonuclear burn for comparison to a probability distribution function (PDF) burn model. MARBLE utilizes plastic capsules filled with deuterated plastic foam and tritium gas. The ratio of DT to DD neutron yield is indicative of the degree to which the foam and the gas atomically mix. Platform development experiments have been performed to understand the behavior of the foam and of the gas separately using two types of capsule. The first experiments using deuterated foam and tritium gas have been performed. Results of these experiments, and the implications for our understanding of thermonuclear burn in heterogeneously mixed separated reactant experiments will be discussed. This work is supported by US DOE/NNSA, performed at LANL, operated by LANS LLC under contract DE-AC52-06NA25396.

US Department of Energy Nevada Operations Office annual site environmental report: 1993. Volume 2: Appendices

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

Black, S.C.; Glines, W.M.; Townsend, Y.E.

1994-09-01

This report is comprised of appendices which support monitoring and surveillance on and around the Nevada Test Site (NTS) during 1993. Appendix A contains onsite Pu-238, gross beta, and gamma-emitting radionuclides in air. Appendix B contains onsite tritium in air. Appendix C contains onsite Pu-238, Sr-90, gross alpha and beta, gamma-emitting radionuclides, Ra-226, Ra-228 and tritium in water. A summary of 1993 results of offsite radiological monitoring is included in Appendix D. Appendix E contains radioactive noble gases in air onsite. Appendix F contains onsite thermoluminescent dosimeter data. Historical trends in onsite thermoluminescent dosimeter data are contained in Appendix G.more » Appendix H summarizes 1993 compliance at the DOE/NV NTS and non-NTS facilities. Appendix I summarizes the 1993 results of non radiological monitoring.« less

Measurements of Deuterium-Tritium Fuel Fractionation from Kinetic Effects in Ignition-Relevant Direct-Drive Cryogenic Implosions

NASA Astrophysics Data System (ADS)

Forrest, C.; Glebov, V. Yu.; Knauer, J. P.; Radha, P. B.; Regan, S. P.; Sangster, T. C.; Stoeckl, C.

2016-10-01

Measurements of DT and DD reaction yields have been studied using ignition-relevant, cryogenically cooled deuterium-tritium gas-filled cryogenic DT targets in inertial confinement fusion (ICF) implosions. In these experiments, carried out at the Omega Laser Facility, highresolution time-of-flight spectroscopy was used to measure the primary neutron peak distribution required to infer the DT and DD reaction yields. From these measurements, it will be shown that the yield ratio has a χ2/per degree of freedom of 0.67 as compared with the measured fraction of the target fuel composition. This observation indicates that kinetic effects leading to species separation are insignificant in ICF ignition-relevant DT implosions on OMEGA. This material is based upon work supported by the Department Of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

Evaluation of a predictive ground-water solute-transport model at the Idaho National Engineering Laboratory, Idaho

USGS Publications Warehouse

Lewis, Barney D.; Goldstein, Flora J.

1982-01-01

Aqueous chemical and radioactive wastes discharged to shallow ponds and to shallow or deep wells on the Idaho National Engineering Laboratory (INEL) since 1952 have affected the quality of the ground water in the underlying Snake River Plain aquifer. The aqueous wastes have created large and laterally dispersed concentration plumes within the aquifer. The waste plumes with the largest areal distribution are those of chloride , tritium, and with high specific conductance values. The data from eight wells drilled near the southern INEL boundary during the summer of 1980 were used to evaluate the accuracy of a predictive modeling study completed in 1973, and to simulate 1980 positions of chloride and tritium plumes. Data interpretation from the drilling program indicates that the hydrogeologic characteristics of the subsurface rocks have marked effects on the regional ground-water flow regimen and, therefore, the movement of aqueous wastes. As expected, the waste plumes projected by the computer model for 1980, extended somewhat further downgradient than indicated by well data due to conservative worst-case assumptions in the model input and inacurate approximations of subsequent waste discharge and aquifer recharge conditions. (USGS)

Engineering directorate technical facilities catalog

NASA Technical Reports Server (NTRS)

Maloy, Joseph E.

1993-01-01

The Engineering Directorate Technical Facilities Catalog is designed to provide an overview of the technical facilities available within the Engineering Directorate at the National Aeronautics and Space Administration (NASA), Lyndon B. Johnson Space Center (JSC) in Houston, Texas. The combined capabilities of these engineering facilities are essential elements of overall JSC capabilities required to manage and perform major NASA engineering programs. The facilities are grouped in the text by chapter according to the JSC division responsible for operation of the facility. This catalog updates the facility descriptions for the JSC Engineering Directorate Technical Facilities Catalog, JSC 19295 (August 1989), and supersedes the Engineering Directorate, Principle test and Development Facilities, JSC, 19962 (November 1984).

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

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

Hosea, J.; Adler, J.H.; Alling, P.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning;more » possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.« less

Vacuum system transient simulator and its application to TFTR

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

Sredniawski, J.

The vacuum system transient simulator (VSTS) models transient gas transport throughout complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. VSTS is capable of treating gas models of up to 10 species, for all flow regimes from pure molecular to continuum. Viscous interactions between species are considered as well as non-uniform temperature of a system. Although this program was specifically developed for use on the Tokamak Fusion Test Reactor (TFTR) project at Princeton, it is a generalized tool capable of handling a broad range of vacuum system problems. During the TFTR engineering design phase, VSTSmore » has been used in many applications. Two applications selected for presentation are: torus vacuum pumping system performance between 400 Ci tritium pulses and tritium backstreaming to neutral beams during pulses.« less

The accomplishment of the Engineering Design Activities of IFMIF/EVEDA: The European-Japanese project towards a Li(d,xn) fusion relevant neutron source

NASA Astrophysics Data System (ADS)

Knaster, J.; Ibarra, A.; Abal, J.; Abou-Sena, A.; Arbeiter, F.; Arranz, F.; Arroyo, J. M.; Bargallo, E.; Beauvais, P.-Y.; Bernardi, D.; Casal, N.; Carmona, J. M.; Chauvin, N.; Comunian, M.; Delferriere, O.; Delgado, A.; Diaz-Arocas, P.; Fischer, U.; Frisoni, M.; Garcia, A.; Garin, P.; Gobin, R.; Gouat, P.; Groeschel, F.; Heidinger, R.; Ida, M.; Kondo, K.; Kikuchi, T.; Kubo, T.; Le Tonqueze, Y.; Leysen, W.; Mas, A.; Massaut, V.; Matsumoto, H.; Micciche, G.; Mittwollen, M.; Mora, J. C.; Mota, F.; Nghiem, P. A. P.; Nitti, F.; Nishiyama, K.; Ogando, F.; O'hira, S.; Oliver, C.; Orsini, F.; Perez, D.; Perez, M.; Pinna, T.; Pisent, A.; Podadera, I.; Porfiri, M.; Pruneri, G.; Queral, V.; Rapisarda, D.; Roman, R.; Shingala, M.; Soldaini, M.; Sugimoto, M.; Theile, J.; Tian, K.; Umeno, H.; Uriot, D.; Wakai, E.; Watanabe, K.; Weber, M.; Yamamoto, M.; Yokomine, T.

2015-08-01

The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept, the re-location of the quench tank outside the test cell (TC) with a reduction of tritium inventory and a simplification on its replacement in case of failure, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction, and the water cooling of the liner and biological shielding of the TC, enhancing the efficiency and economy of the related sub-systems. In addition, the maintenance strategy has been modified to allow a shorter yearly stop of the irradiation operations and a more careful management of the irradiated samples. The design of the IFMIF plant is intimately linked with the EVA phase carried out since the entry into force of IFMIF/EVEDA in June 2007. These last activities and their on-going accomplishment have been thoroughly described elsewhere (Knaster J et al [19]), which, combined with the present paper, allows a clear understanding of the maturity of the European-Japanese international efforts. This released IFMIF Intermediate Engineering Design Report (IIEDR), which could be complemented if required concurrently with the outcome of the on-going EVA, will allow decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community.

Radiochemical Processing Laboratory (RPL) at PNNL

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

Peurrung, Tony; Clark, Sue; Bryan, Sam

2017-03-23

Nuclear research is one of the core components of PNNL's mission. The centerpiece of PNNL's nuclear research is the Radiochemical Processing Laboratory (RPL), a Category 2 nuclear facility with state-of-the-art instrumentation, scientific expertise, and specialized capabilities that enable research with significant quantities of fissionable materials and other radionuclides—from tritium to plutonium. High impact radiological research has been conducted in the RPL since the 1950's, when nuclear weapons and energy production at Hanford were at the forefront of national defense. Since then, significant investments have been made in the RPL to maintain it as a premier nuclear science research facility supportingmore » multiple programs. Most recently, PNNL is developing a world-class analytical electron microscopy facility dedicated to the characterization of radiological materials.« less

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

Jordan, Jacob A.; Jubin, Robert Thomas

US regulations could require the removal of both iodine and tritium from the off-gas stream of a used nuclear fuel (UNF) reprocessing facility. Advanced tritium pretreatment is a pretreatment step that uses high concentrations of NOR2R in a gas stream to volatilize tritium and iodine from UNF prior to traditional dissolution. The gaseous effluent from this process would then require abatement to remove tritium and iodine, but high levels of NOR2R could have a detrimental effect on the ability of various solid sorbents to remove the volatile radionuclides. For tritium and iodine, the sorbents of interest are 3Å molecular sievemore » (3AMS) for tritium and reduced silver mordenite (AgP 0 PZ), silver-functionalized silica-aerogel (AgAerogel), and silver-nitrate-impregnated alumina (AgA) for iodine. Prior research has demonstrated that exposure to high concentrations of NOR2R can reduce the iodine loading capacity of AgP 0 PZ by > 90% when exposed for 1 week. Research in Japan has demonstrated that AgA is more robust to NOR2R exposure than AgZ. The testing described here was intended to assess the effects of high concentrations of NOR2R on the iodine capture capacity of AgA and the water adsorption capacity of 3AMS. To determine the effect of extended exposure of the sorbents to NOR2R, both 3AMS and AgA were aged in a 75% NOR2R environment prior to loading. The 3AMS samples were aged for 1, 4, and 5.5 weeks at 40°C. They were then loaded with water in a 10°C dew point stream (corresponding to a water concentration of ~12,000 ppmv) at 40°C. There was no significant change in the water adsorption capacity of the 3AMS upon exposure to 75% NOR2R. The AgA samples were aged for 1, 2, and 4 weeks at 150°C and were loaded with 50 ppmv IR2R at 150°C. The results show that the iodine capture capacity of AgA is reduced by exposure to high concentrations of NOR2R. The iodine capacity reductions were 16%, 36%, and 76% for 1, 2, and 4 week exposures, respectively. This is less of a capacity loss than that seen in similar testing with the AgP 0 PZ sorbent.« less

Tritium well depth, tritium well time and sponge mechanism for reducing tritium retention

NASA Astrophysics Data System (ADS)

Deng, B. Q.; Li, Z. X.; Li, C. Y.; Feng, K. M.

2011-07-01

New simulation results are predicted in a fusion reactor operation process. They are somewhat similar to, but quite different from, the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor. We obtained completely new results of tritium well depth and tritium well time in magnetic confinement fusion energy research area. This study is carried out to investigate the following: what will be the least amount of tritium storage required to start up a fusion reactor and how long the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium-breeding time, which is dependent on the tritium breeder, specific structure of the breeding zone, layout of the coolant flow pipes, tritium recovery scheme and applied extraction process, the tritium retention of plasma facing component (PFC) and other reactor components, unrecoverable tritium fraction in the breeder, leakage to the inertial gas container and the natural radioactive decay time constant. We describe these new issues and answer these problems by setting up and solving a set of equations, which are described by a dynamic subsystem model of tritium inventory evolution in a fusion experimental breeder (FEB). Reasonable results are obtained using our simulation model. It is found that the tritium well depth is about 0.319 kg and the tritium well time is approximately 235 full power operation days for the reference case of the designed FEB configuration, and it is also found that after one-year operation the tritium storage reaches 0.767 kg, which is more than the least amount of tritium storage required to start up another FEB-like fusion reactor. The results show that the tritium retention in the PFC is equivalent to 11.9% of tritium well depth that is fairly consistent with the result of 10-20% deduced from the integrated particle balance of European tokamaks. Based on our experimental and theoretical studies, some new mechanisms are proposed for reducing the tritium retention in PFC and structure materials of tritium-breeding blanket. In this paper, a qualitative analysis of the 'sponge effect' is carried out. The 'sponge effect' may help us to reduce tritium retention by ~20% in the PFC.

Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

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

Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph

2015-09-01

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T 2O. In a standard processing flowsheet, tritium management would bemore » accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480–600ºC expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 µCi/g, respectively. Heating the Surry-2 cladding at 500°C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700°C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700°C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700°C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600°C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.« less

Draft environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R

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

None

1991-04-01

This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build newmore » production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.« less

Future use of tritium in mapping pre-bomb groundwater volumes.

PubMed

Eastoe, C J; Watts, C J; Ploughe, M; Wright, W E

2012-01-01

The tritium input to groundwater, represented as volume-weighted mean tritium concentrations in precipitation, has been close to constant in Tucson and Albuquerque since 1992, and the decrease in tritium concentrations at the tail end of the bomb tritium pulse has ceased. To determine the future usefulness of tritium measurements in southwestern North America, volume-weighted mean tritium levels in seasonal aggregate precipitation samples have been gathered from 26 sites. The averages range from 2 to 9 tritium units (TU). Tritium concentrations increase with site latitude, and possibly with distance from the coast and with site altitude, reflecting local ratios of combination of low-tritium moisture advected from the oceans with high-tritium moisture originating near the tropopause. Tritium used alone as a tool for mapping aquifer volumes containing only pre-bomb recharge to groundwater will become ambiguous when the tritium in precipitation at the end of the bomb tritium pulse decays to levels close to the analytical detection limit. At such a time, tritium in precipitation from the last one to two decades of the bomb pulse will become indistinguishable from pre-bomb recharge. The threshold of ambiguity has already arrived in coastal areas with a mean of 2 TU in precipitation and will follow in the next three decades throughout the study region. Where the mean tritium level is near 5 TU, the threshold will occur between 2025 and 2030, given a detection limit of 0.6 TU. Similar thresholds of ambiguity, with different local timing possible, apply globally. © 2011, The Author(s). Ground Water © 2011, National Ground Water Association.

Improving tritium exposure reconstructions using accelerator mass spectrometry

PubMed Central

Hunt, J. R.; Vogel, J. S.; Knezovich, J. P.

2010-01-01

Direct measurement of tritium atoms by accelerator mass spectrometry (AMS) enables rapid low-activity tritium measurements from milligram-sized samples and permits greater ease of sample collection, faster throughput, and increased spatial and/or temporal resolution. Because existing methodologies for quantifying tritium have some significant limitations, the development of tritium AMS has allowed improvements in reconstructing tritium exposure concentrations from environmental measurements and provides an important additional tool in assessing the temporal and spatial distribution of chronic exposure. Tritium exposure reconstructions using AMS were previously demonstrated for a tree growing on known levels of tritiated water and for trees exposed to atmospheric releases of tritiated water vapor. In these analyses, tritium levels were measured from milligram-sized samples with sample preparation times of a few days. Hundreds of samples were analyzed within a few months of sample collection and resulted in the reconstruction of spatial and temporal exposure from tritium releases. Although the current quantification limit of tritium AMS is not adequate to determine natural environmental variations in tritium concentrations, it is expected to be sufficient for studies assessing possible health effects from chronic environmental tritium exposure. PMID:14735274

In-reactor performance of LWR-type tritium target rods

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

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

Tritium monitor with improved gamma-ray discrimination

DOEpatents

Cox, S.A.; Bennett, E.F.; Yule, T.J.

1982-10-21

Apparatus and method are presented for selective measurement of tritium oxide in an environment which may include other radioactive components and gamma radiation, the measurement including the selective separation of tritium oxide from a sample gas through a membrane into a counting gas, the generation of electrical pulses individually representative by rise times of tritium oxide and other radioactivity in the counting gas, separation of the pulses by rise times, and counting of those pulses representative of tritium oxide. The invention further includes the separate measurement of any tritium in the sample gas by oxidizing the tritium to tritium oxide and carrying out a second separation and analysis procedure as described above.

Tritium monitor with improved gamma-ray discrimination

DOEpatents

Cox, Samson A.; Bennett, Edgar F.; Yule, Thomas J.

1985-01-01

Apparatus and method for selective measurement of tritium oxide in an environment which may include other radioactive components and gamma radiation, the measurement including the selective separation of tritium oxide from a sample gas through a membrane into a counting gas, the generation of electrical pulses individually representative by rise times of tritium oxide and other radioactivity in the counting gas, separation of the pulses by rise times, and counting of those pulses representative of tritium oxide. The invention further includes the separate measurement of any tritium in the sample gas by oxidizing the tritium to tritium oxide and carrying out a second separation and analysis procedure as described above.

Water - Isotope - Map (δ 18O, δ 2H, 3H) of Austria: Applications, Extremes and Trends

NASA Astrophysics Data System (ADS)

Wyhlidal, Stefan; Kralik, Martin; Benischke, Ralf; Leis, Albrecht; Philippitsch, Rudolf

2016-04-01

The isotopic ratios of oxygen and hydrogen in water (2H/1H and 18O/16O) are important tools to characterise waters and their cycles. This starts in the atmosphere as rain or snow and continues in surface water and ends in shallow groundwater as well as in deep groundwater. Tritium formed by natural cosmic radiation in the upper atmosphere and in the last century by tests of thermonuclear bombs in the atmosphere, is characterised by its radioactive decay with a half-life of 12.32 years and is an ideal age-marker during the last 60 years. To determine the origin and mean age of waters in many projects concerning water supply, engineering and scientific projects in the last 45 years on more than 1,350 sites, more than 40,000 isotope measurements were performed in Austria. The median value of all sites of oxygen-18 is δ 18O -10.7 ‰ and for hydrogen-2 δ 2H -75 ‰. As the fractionation is mainly temperature dependent the lowest negative values are observed in winter precipitation (oxygen-18 as low as δ 18O -23 ‰) and in springs in the mountain regions (δ 18O -15.1 ‰). In contrast the highest values were observed in summer precipitation (up to δ 18O - 0.5 ‰) and in shallow lakes in the Seewinkel (up to δ 18O + 5 ‰). The isotopic ratios of the Austrian waters are also influenced by the origin of the evaporated water masses. Therefore the precipitation in the region south of the main Alpine crest (East-Tyrol, Carinthia and South-East Styria) is approximately 1 ‰ higher in δ 18O-values than sites at the same altitude in the northern part. This is most probably caused by the stronger influence of precipitation from the mediterranean area. The median value of all 1,120 sampling sites of decay corrected (2015) tritium measurements is 6.2 tritium units (TU). This is somewhat smaller than the median value of all precipitation stations with 7.2 TU. This can be explained by the fact that in most cases in groundwater the median value has been reduced by decay according to the residence time underground. The tritium concentration increases in the summer up to 10 - 11 TU and decreases in winter down to 3 - 4 TU. This is due to the better circulation in the atmosphere in spring which brings the tritium formed by cosmic radiation down to the lower atmosphere and precipitation. A smaller mean tritium concentration in aquifers than approximately 3.5 TU indicates large amounts of water older than 60 years. Waters with approximately more than 12 TU contain still tritium from the 1960s and 1970s formed originally by thermonuclear bomb experiments. In Austria the highest Tritium values can be observed in the rivers Danube and March which show periodic or permanent tritium contamination up to 70 TU coming from nuclear power plants in the neighbouring countries.

Fusion Power—A Chemical Engineering View of the Integrated Enterprise

NASA Astrophysics Data System (ADS)

Manganaro, James L.

2003-03-01

The purpose of this article was to achieve the beginning of an understanding of the integrated fusion enterprise from raw materials through power generation to decommissioning and waste disposal. The particular view point is that of a technically trained person who is only casually acquainted with the field. Emphasis is given to the chemical engineering aspects of controlled fusion power. It is concluded that there are indeed many areas in which the discipline of chemical engineering may contribute to the fusion effort. These areas include separation technology by physical and chemical means, heat and mass transfer in a packed bed blanket, tritium removal from molten coolants, distillation technology for isotope separation, and preparation of deuterium and lithium feed materials.

Use of Tritium Accelerator Mass Spectrometry for Tree Ring Analysis

PubMed Central

LOVE, ADAM H.; HUNT, JAMES R.; ROBERTS, MARK L.; SOUTHON, JOHN R.; CHIARAPPA - ZUCCA, MARINA L.; DINGLEY, KAREN H.

2010-01-01

Public concerns over the health effects associated with low-level and long-term exposure to tritium released from industrial point sources have generated the demand for better methods to evaluate historical tritium exposure levels for these communities. The cellulose of trees accurately reflects the tritium concentration in the source water and may contain the only historical record of tritium exposure. The tritium activity in the annual rings of a tree was measured using accelerator mass spectrometry to reconstruct historical annual averages of tritium exposure. Milligram-sized samples of the annual tree rings from a Tamarix located at the Nevada Test Site are used for validation of this methodology. The salt cedar was chosen since it had a single source of tritiated water that was well-characterized as it varied over time. The decay-corrected tritium activity of the water in which the salt cedar grew closely agrees with the organically bound tritium activity in its annual rings. This demonstrates that the milligram-sized samples used in tritium accelerator mass spectrometry are suited for reconstructing anthropogenic tritium levels in the environment. PMID:12144257

Regeneration and tritium recovery from the large JET neutral injection cryopump system after the FTE. [First Tritium Experiment (FTE)

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

Obert, W.; Bell, A.; Davies, J.

1992-01-01

Neutral Beam Injection (NBI) was used to introduce tritium into the plasma for the First Tritium Experiment In addition to the decisive advantage of depositing the tritium into the centre of the plasma, the use of NBI also minimized the total quantity of tritium introduced into the Torus and the contamination of the vacuum vessel. However, because of the relatively low gas efficiency of the positive ion injection system approximately 95% of the total quantity of tritium introduced was pumped by the large condensation cryopumps which form an integral part of the injector. Several hardware and associated software changes weremore » implemented in order to making provision for possible fault scenarios during operation with tritium and to ensure complete regeneration of the tritium from the cryopumps. The tritium released after all subsequent regeneration's has been monitored carefully in order to determine the amount of tritium retained by the black anodized liquid nitrogen panel surfaces of the cryopump and to compare it with experiments at TSTA on JET samples before the FTE.« less

Behaviour of tritium in the vacuum vessel of JT-60U

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

Kobayashi, K.; Miya, N.; Ikeda, Y.

2015-03-15

The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D{sub 2} (92.8 %) - T{sub 2} (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating upmore » to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily.« less

Laser-assisted isotope separation of tritium

DOEpatents

Herman, Irving P.; Marling, Jack B.

1983-01-01

Methods for laser-assisted isotope separation of tritium, using infrared multiple photon dissociation of tritium-bearing products in the gas phase. One such process involves the steps of (1) catalytic exchange of a deuterium-bearing molecule XYD with tritiated water DTO from sources such as a heavy water fission reactor, to produce the tritium-bearing working molecules XYT and (2) photoselective dissociation of XYT to form a tritium-rich product. By an analogous procedure, tritium is separated from tritium-bearing materials that contain predominately hydrogen such as a light water coolant from fission or fusion reactors.

How to improve the irradiation conditions for the International Fusion Materials Irradiation Facility

NASA Astrophysics Data System (ADS)

Daum, Eric

2000-12-01

The accelerator-based intense D-Li neutron source International Fusion Materials Irradiation Facility (IFMIF) provides very suitable irradiation conditions for fusion materials development with the attractive option of accelerated irradiations. Investigations show that a neutron moderator made of tungsten and placed in the IFMIF test cell can further improve the irradiation conditions. The moderator softens the IFMIF neutron spectrum by enhancing the fraction of low energy neutrons. For displacement damage, the ratio of point defects to cascades is more DEMO relevant and for tritium production in Li-based breeding ceramic materials it leads to a preferred production via the 6Li(n,t) 4He channel as it occurs in a DEMO breeding blanket.

Note: Radiochemical measurement of fuel and ablator areal densities in cryogenic implosions at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Hagmann, C.; Shaughnessy, D. A.; Moody, K. J.; Grant, P. M.; Gharibyan, N.; Gostic, J. M.; Wooddy, P. T.; Torretto, P. C.; Bandong, B. B.; Bionta, R.; Cerjan, C. J.; Bernstein, L. A.; Caggiano, J. A.; Herrmann, H. W.; Knauer, J. P.; Sayre, D. B.; Schneider, D. H.; Henry, E. A.; Fortner, R. J.

2015-07-01

A new radiochemical method for determining deuterium-tritium (DT) fuel and plastic ablator (CH) areal densities (ρR) in high-convergence, cryogenic inertial confinement fusion implosions at the National Ignition Facility is described. It is based on measuring the 198Au/196Au activation ratio using the collected post-shot debris of the Au hohlraum. The Au ratio combined with the independently measured neutron down scatter ratio uniquely determines the areal densities ρR(DT) and ρR(CH) during burn in the context of a simple 1-dimensional capsule model. The results show larger than expected ρR(CH) values, hinting at the presence of cold fuel-ablator mix.

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

Wyatt, Douglas

Planning for ultimate Decontamination and Decommissioning (D and D) of a nuclear facility is as much a part of a successful nuclear strategy as is the ultimate disposal of radioactive waste. As facilities, in this case radioactive waste disposal trenches, are closed and abandoned leading to ultimate decommissioning, long term monitoring may be required. However, preplanning by characterizing, modeling, and monitoring the environment around the facility prior to and during operations will allow a performance assessment to be made and future behavior predicted. In the radioactive waste burial grounds of the Savannah River Site new slit trenches were constructed tomore » receive demolition debris associated with site foot print reduction. Some of the construction debris and associated process waste contained small amounts of tritium. Since the trenches were constructed over an existing tritium groundwater plume the monitoring and performance assessment of the trench, particularly with respect to tritium contributions to the vadose zone and groundwater, were important. These disposal trenches vary in length and width but are typically constructed within the upper 7 to 8 meters (21 to 24 feet) of the local sediments. The unconfined aquifer (water table) typically underlies the area at depths varying from 20 to 24 meters (60 to 72 feet), depending on elevation. Therefore, with downward flow and 13 to 16 meters (40 to 48 feet) of unsaturated sediments separating the base of the waste trenches from the unconfined aquifer, there was potential for an environmental impact to the sediments within the vadose zone and to the underlying groundwater. Monitoring and predicting this impact can support ultimate D and D activities and future performance assessment evaluation. From this work several key observations were made that will support long term monitoring and subsequent D and D: - The observed lateral variation of thinly bedded sands and clays may be less than 20 meters particularly if lenticular sands are present. Ultimate D and D should consider monitoring and remedial activities that consider sampling on scales to address this issue. - The detailed modeling, when compared with the modeled depositional patterns, indicates flow paths for vadose zone fluids, therefore a plan should allow for these flow paths. - Detailed lithostratigraphic modeling, when based on correlations between soil properties, CPT soundings and borehole geophysical logs, can aid in precision placement of subsurface sensors and sample points for performance monitoring and D and D assessment.« less

Task toward a Realization of Commercial Tokamak Fusion Plants in 2050 -The Role of ITER and the Succeeding Developments- 4.Technology and Material Research in Fusion Power Plant Development

NASA Astrophysics Data System (ADS)

Akiba, Masato; Matsui, Hideki; Takatsu, Hideyuki; Konishi, Satoshi

Technical issues regarding the fusion power plant that are required to be developed in the period of ITER construction and operation, both with ITER and with other facilities that complement ITER are described in this section. Three major fields are considered to be important in fusion technology. Section 4.1 summarizes blanket study, and ITER Test Blanket Module (TBM) development that focuses its effort on the first generation power blanket to be installed in DEMO. ITER will be equipped with 6 TBMs which are developed under each party's fusion program. In Japan, the solid breeder using water as a coolant is the primary candidate, and He-cooled pebble bed is the alternative. Other liquid options such as LiPb, Li or molten salt are developed by other parties' initiatives. The Test Blanket Working Group (TBWG) is coordinating these efforts. Japanese universities are investigating advanced concepts and fundamental crosscutting technologies. Section 4.2 introduces material development and particularly, the international irradiation facility, IFMIF. Reduced activation ferritic/martensitic steels are identified as promising candidates for the structural material of the first generation fusion blanket, while and vanadium alloy and SiC/SiC composite are pursued as advanced options. The IFMIF is currently planning the next phase of joint activity, EVEDA (Engineering Validation and Engineering Design Activity) that encompasses construction. Material studies together with the ITER TBM will provide essential technical information for development of the fusion power plant. Other technical issues to be addressed regarding the first generation fusion power plant are summarized in section 4.3. Development of components for ITER made remarkable progress for the major essential technology also necessary for future fusion plants, however many still need further improvements toward power plant. Such areas includes; the divertor, plasma heating/current drive, magnets, tritium, and remote handling. There remain many other technical issues for power plant which require integrated efforts.

Measurement of tritium penetration through concrete material covered by various paints coating

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

Edao, Y.; Kawamura, Y.; Kurata, R.

The present study aims at obtaining fundamental data on tritium migration in porous materials, which include soaking effect, interaction between tritium and cement paste coated with paints and transient tritium sorption in porous cement. The amounts of tritium penetrated into or released from cement paste with epoxy and urethane paint coatings were measured. The tritium penetration amounts were increased with the HTO (tritiated water) exposure time. Time to achieve a saturated value of tritium sorption was more than 60 days for cement paste coated with epoxy paint and with urethane paint, while that for cement paste without any paint coatingmore » took 2 days to achieve it. The effect of tritium permeation reduction by the epoxy paint was higher than that of the urethane. Although their paint coatings were effective for reduction of tritium penetration through the cement paste which was exposed to HTO for a short period, it was found that the amount of tritium trapped in the paints became large for a long period. Tritium penetration rates were estimated by an analysis of one-dimensional diffusion in the axial direction of a thickness of a sample. Obtained data were helpful for evaluation of tritium contamination and decontamination. (authors)« less

APT radionuclide production experiment

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

Ullmann, J.L.; Gavron, A.; King, J.D.

1994-07-02

Tritium ({sup 3}H, a heavy isotope of hydrogen) is produced by low energy neutron-induced reactions on various elements. One such reaction is n+{sup 3}He {yields}>{sup 3}H+{sup 1}H in which {sup 3}He is transmuted to tritium. Another reaction, which has been used in reactor production of tritium, is the n+{sup 6}Li {yields}> {sup 3}H+{sup 4}He reaction. Accelerator Production of Tritium relies on a high-energy proton beam to produce these neutrons using the spallation reaction, in which high-energy proton beam to produce these neutrons using the spallation reaction, in which high-energy protons reacting with a heavy nucleus produce a shower of low-energymore » neutrons and a lower-mass residual nucleus. It is important to quantify the residual radionuclides produced in the spallation target for two reasons. From an engineering point of view, one must understand short-lived isotopes that may contribute to decay heat. From a safety viewpoint, one must understand what nuclei and decay gammas are produced in order to design adequate shielding, to estimate ultimate waste disposal problems, and to predict possible effects due to accidental dispersion during operation. The authors have performed an experiment to measure the production of radioisotopes in stopping-length W and Pb targets irradiated by a 800 MeV proton beam, and are comparing the results to values obtained from calculations using LAHET and MCNP. The experiment was designed to pay particular attention to the short half-life radionuclides, which have not been previously measured. In the following, they present details of the experiment, explain how they analyzed the data and obtain the results, how they perform the calculations, and finally, how the experimental data agree with the calculations.« less

Engine component instrumentation development facility at NASA Lewis Research Center

NASA Technical Reports Server (NTRS)

Bruckner, Robert J.; Buggele, Alvin E.; Lepicovsky, Jan

1992-01-01

The Engine Components Instrumentation Development Facility at NASA Lewis is a unique aeronautics facility dedicated to the development of innovative instrumentation for turbine engine component testing. Containing two separate wind tunnels, the facility is capable of simulating many flow conditions found in most turbine engine components. This facility's broad range of capabilities as well as its versatility provide an excellent location for the development of novel testing techniques. These capabilities thus allow a more efficient use of larger and more complex engine component test facilities.

[Mechanism of tritium persistence in porous media like clay minerals].

PubMed

Wu, Dong-Jie; Wang, Jin-Sheng; Teng, Yan-Guo; Zhang, Ke-Ni

2011-03-01

To investigate the mechanisms of tritium persistence in clay minerals, three types of clay soils (montmorillonite, kaolinite and illite) and tritiated water were used in this study to conduct the tritium sorption tests and the other related tests. Firstly, the ingredients, metal elements and heat properties of clay minerals were studied with some instrumental analysis methods, such as ICP and TG. Secondly, with a specially designed fractionation and condensation experiment, the adsorbed water, the interlayer water and the structural water in the clay minerals separated from the tritium sorption tests were fractionated for investigating the tritium distributions in the different types of adsorptive waters. Thirdly, the location and configuration of tritium adsorbed into the structure of clay minerals were studied with infrared spectrometry (IR) tests. And finally, the forces and mechanisms for driving tritium into the clay minerals were analyzed on the basis of the isotope effect of tritium and the above tests. Following conclusions have been reached: (1) The main reason for tritium persistence in clay minerals is the entrance of tritium into the adsorbed water, the interlayer water and the structural water in clay minerals. The percentage of tritium distributed in these three types of adsorptive water are in the range of 13.65% - 38.71%, 0.32% - 5.96%, 1.28% - 4.37% of the total tritium used in the corresponding test, respectively. The percentages are different for different types of clay minerals. (2) Tritium adsorbed onto clay minerals are existed in the forms of the tritiated hydroxyl radical (OT) and the tritiated water molecule (HTO). Tritium mainly exists in tritiated water molecule for adsorbed water and interlayer water, and in tritiated hydroxyl radical for structural water. (3) The forces and effects driving tritium into the clay minerals may include molecular dispersion, electric charge sorption, isotope exchange and tritium isotope effect.

Monitoring radionuclide contamination in the unsaturated zone - Lessons learned at the Amargosa Desert Research Site, Nye County, Nevada

USGS Publications Warehouse

Stonestrom, David A.; Abraham, Jared D.; Andraski, Brian J.; Baker, Ronald J.; Mayers, C. Justin; Michel, Robert L.; Prudic, David E.; Striegl, Robert G.; Walvoord, Michelle Ann

2004-01-01

Contaminant-transport processes are being investigated at the U.S. Geological Survey’s Amargosa Desert Research Site (A DRS), adjacent to the Nation’s first commercial disposal facility for low-level radioactive waste. Gases containing tritium and radiocarbon are migrating through a 110-m thick unsaturated zone from unlined trenches that received waste from 1962 to 1992. Results relevant to long- term monitoring of radionuclides are summarized as follows. Contaminant plumes have unexpected histories and spatial configurations due to uncertainties in the: (1) geologic framework, (2) biochemical reactions involving waste components, (3) interactions between plume components and unsaturated-zone materials, (4) disposal practices, and (5) physical transport processes. Information on plume dynamics depends on ex-situ wet-chemical techniques because in-situ sensors for the radionuclides of interest do not exist. As at other radioactive-waste disposal facilities, radionuclides at the ADRS are mixed with varying amounts of volatile organic compounds (VOCs). Carbon-dioxide and VOC anomalies provide proxies for radioactive contamination. Contaminants in the unsaturated zone migrate along preferential pathways. Effective monitoring thus requires accurate geologic characterization. Direct- current electrical-resistivity imaging successfully mapped geologic units controlling preferential transport at the ADRS. Direct sampling of water from the unsaturated zone is complex and time consuming. Sampling plant water is an efficient alternative for mapping shallow tritium contamination.

Tennessee Valley region study: potential year 2000 radiological dose to population resulting from nuclear facility operations. [Includes glossary

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

None

A companion report, DOE/ET-0064/1, presents a geographic, cultural, and demographic profile of the Tennessee Valley Region study area. This report describes the calculations of radionuclide release and transport and of the resultant dose to the regional population, assuming a projected installed capacity of 220,000 MW in the year 2000, of which 144,000 MW would be nuclear. All elements of the fuel cycle were assumed to be in operation. The radiological dose was calculated as a one-year dose based on ingestion of 35 different food types as well as for nine non-food pathways, and was reported as dose to the totalmore » body and for six specific organs for each of four age groups (infant, child, teen, and adult). Results indicate that the average individual would receive an incremental dose of 7 x 10/sup -4/ millirems in the year 2000 from the operation of nuclear facilities within and adjacent to the region, five orders of magnitude smaller than the dose from naturally occurring radiation in the area. The major contributor to dose was found to be tritium, and the most significant pathways were immersion in air, inhalation of air, transpiration of tritium (absorption through the skin), and exposure radionuclide-containing soil. 60 references.« less

Deuterium-Tritium Beta-Layering Within a National Ignition Facility Scale Polymer Target in the LANL Cryogenic Pressure Loader

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

Ebey, Peter S.; Dole, James M.; Geller, Drew A.

2005-11-15

Beta-layering, the process of beta-decay heat-driven mass redistribution, has been demonstrated in a deuterium-tritium (D-T)-filled polymer sphere of the type required for fusion ignition experiments at the National Ignition Facility. This is the first report, to the best of the authors' knowledge, of a D-T layer formed in a permeation-filled sphere. The 2-mm-diam sphere was filled with D-T by permeation; cooled to cryogenic temperatures while in the high-pressure permeation vessel; and, while cold, removed to an optical axis where the D-T was frozen, melted, and beta-layered in a series of experiments over several weeks' time. This work was performed inmore » the Los Alamos National Laboratory cryogenic pressure loader system. The beta-layering time constant was 24.0 {+-} 2.5 min, less than the theoretical value of 26.8 min, and not showing the significant increase due to build-up of {sup 3}He often observed in beta-layered samples. Supercooling of the liquid D-T was observed. Neither the polymer target nor its tenting material showed visual signs of degradation after 5 weeks of exposure to D-T. Small external thermal gradients were used to shift the D-T material back and forth within the sphere.« less

Steady-State and Transient Groundwater Flow and Advective Transport, Eastern Snake River Plain Aquifer, Idaho National Laboratory and Vicinity, Idaho

NASA Astrophysics Data System (ADS)

Fisher, J. C.; Ackerman, D. J.; Rousseau, J. P.; Rattray, G. W.

2009-12-01

Three-dimensional steady-state and transient models of groundwater flow and advective transport through the fractured basalts and interbedded sediments of the Eastern Snake River Plain (ESRP) aquifer were developed by the U.S. Geological Survey in cooperation with the U.S. Department of Energy. The model domain covers an area of 1,940 square miles that includes most of the Idaho National Laboratory (INL). A 50-year history of waste disposal at the INL has resulted in measurable concentrations of waste contaminants in the aquifer. Numerical models simulated 1980 steady-state conditions and transient flow for 1980-95. In the transient model, streamflow infiltration was the major stress. The models were calibrated using the parameter-estimation program incorporated in MODFLOW-2000. The steady-state model reasonably simulated the observed water-table altitude and gradients. Simulation of transient conditions reproduced changes in the flow system resulting from episodic infiltration from the Big Lost River. Analysis of simulations shows that flow is (1) dominantly horizontal through interflow zones in basalt, vertical anisotropy resulting from contrasts in hydraulic conductivity of different types of basalt and the interbedded sediments, (2) temporally variable due to streamflow infiltration from the Big Lost River, and (3) moving downward downgradient of the INL. Particle-tracking simulations were used to evaluate how simulated groundwater flow paths and travel times differ between the steady-state and transient flow models, and how well model-derived groundwater flow directions and velocities compare to independently-derived estimates. Particle tracking also was used to simulate the growth of tritium plumes originating at two INL facilities over a 16 year period under steady-state and transient flow conditions (1953-68). The shape, dimensions, and areal extent of these plumes were compared to a map of the plumes for 1968 from tritium releases beginning in 1952. Collectively, the particle-tracking simulations indicate that groundwater flow paths and velocities, based on uncalibrated estimates of porosity, are influenced by the dynamic character of the water table and the large contrasts in the hydraulic properties of the media, primarily hydraulic conductivity. Simulation results also indicate that temporal changes in the local hydraulic gradient can account for some of the observed dispersion of contaminants in the aquifer near the major sources of contamination and perhaps the majority of the observed dispersion several miles downgradient of these facilities. The distance downgradient of the facilities where simulated particle plumes were able to reasonably reproduce the 1968 tritium plume extended only to the boundary separating sediment-rich from sediment-poor aquifer layers about 4 mi downgradient of the contaminant source. Particle plumes simulated beyond this boundary were narrow and long, and did not reasonably reproduce the shape, dimensions, or position of the leading edge of the tritium plume; however, few data were available to characterize its true areal extent and shape.

Composition containing aerogel substrate loaded with tritium

DOEpatents

Ashley, Carol S.; Brinker, C. Jeffrey; Ellefson, Robert E.; Gill, John T.; Reed, Scott; Walko, Robert J.

1992-01-01

The invention provides a process for loading an aerogel substrate with tritium and the resultant compositions. According to the process, an aerogel substrate is hydrolyzed so that surface OH groups are formed. The hydrolyzed aerogel is then subjected to tritium exchange employing, for example, a tritium-containing gas, whereby tritium atoms replace H atoms of surface OH groups. OH and/or CH groups of residual alcohol present in the aerogel may also undergo tritium exchange.

Monitoring and management of tritium from the nuclear power plant effluent

NASA Astrophysics Data System (ADS)

Zhang, Qiaoe; Liu, Ting; Yang, Lili; Meng, De; Song, Dahu

2018-01-01

It is important to regulate tritium nuclides from the nuclear power plant effluent, the paper briefly analyzes the main source of tritium, and the regulatory requirements associated with tritium in our country and the United States. The monitoring methods of tritium from the nuclear power plant effluent are described, and the purpose to give some advice to our national nuclear power plant about the effluent of tritium monitoring and management.

Method and apparatus for controlling accidental releases of tritium

DOEpatents

Galloway, T.R.

1980-04-01

An improvement is described in a tritium control system based on a catalytic oxidation reactor wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release. 1 fig.

Method and apparatus for controlling accidental releases of tritium

DOEpatents

Galloway, Terry R. [Berkeley, CA

1980-04-01

An improvement in a tritium control system based on a catalytic oxidation reactor wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release.

Tritium labeling of amino acids and peptides with liquid and solid tritium

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

Souers, P.C.; Coronado, P.R.; Peng, C.T.

Amino acids and peptides were labeled with liquid and solid tritium at 21/degree/K and 9/degree/K. At these low temperatures radiation degradation is minimal, and tritium incorporation increases with tritium concentration and exposure time. Ring saturation in L-phenylalanine does not occur. Peptide linkage in oligopeptides is stable toward tritium. Deiodination in 3-iodotyrosine and 3,5-diiodotyrosine occurs readily and proceeds in steps by losing one iodine atom at a time. Nickel and noble metal supported catalysts when used as supports for dispersion of the substrate promote tritium labeling at 21 K. Our study shows that both liquid and solid tritiums are potentially usefulmore » agents for labeling peptides and proteins.« less

Contamination mechanisms of air basin with tritium in venues of underground nuclear explosions at the former Semipalatinsk test site.

PubMed

Lyakhova, O N; Lukashenko, S N; Larionova, N V; Tur, Y S

2012-11-01

During the period of testing from 1945 to 1962 at the territory of Semipalatinsk test site (STS) within the Degelen Mountains in tunnels, 209 underground nuclear explosions were produced. Many of the tunnels have seasonal water seepage in the form of streams, through which tritium migrates from the underground nuclear explosion (UNE) venues towards the surface. The issue of tritium contamination occupies a special place in the radioactive contamination of the environment. In this paper we assess the level and distribution of tritium in the atmospheric air of ecosystems with water seepage at tunnels № 176 and № 177, located on "Degelen" site. There has been presented general nature of tritium distribution in the atmosphere relative to surface of a watercourse which has been contaminated with tritium. The basic mechanisms were studied for tritium distribution in the air of studied ecosystems, namely, the distribution of tritium in the systems: water-atmosphere, tunnel air-atmosphere, soil water-atmosphere, vegetation-atmosphere. An analytical calculation of tritium concentration in the atmosphere by the concentration of tritium in water has been performed. There has experimentally obtained the dependence for predictive assessment of tritium concentrations in air as a function of tritium concentration in one of the inlet sources such as water, tunnel air, soil water, vegetation, etc.. The paper also describes the general nature of tritium distribution in the air in the area "Degelen". Copyright © 2012 Elsevier Ltd. All rights reserved.

Determination of in vitro lung solubility and intake-to-dose conversion factor for tritiated lanthanum nickel aluminum alloy.

PubMed

Farfán, Eduardo B; Labone, Thomas R; Staack, Gregory C; Cheng, Yung-Sung; Zhou, Yue; Varallo, Thomas P

2012-09-01

A sample of tritiated lanthanum nickel aluminum alloy (LaNi4.25Al0.75 or LANA.75) similar to that used at the Savannah River Site Tritium Facilities was analyzed to estimate the particle size distribution of this metal tritide powder and the rate at which this material dissolves in the human respiratory tract after it is inhaled. This information is used to calculate the committed effective dose received by a worker after inhaling the material. These doses, which were calculated using the same methodology given in the U.S. Department of Energy Tritium Handbook, are presented as inhalation intake-to-dose conversion factors (DCF). The DCF for this metal tritide was determined to be 9.4 × 10 Sv Bq, which is less than the DCF for tritiated water. Therefore, the radiation worker bioassay programs designed for tritiated water are adequate to monitor for intakes of this material.

First Measurements of Deuterium-Tritium and Deuterium-Deuterium Fusion Reaction Yields in Ignition-Scalable Direct-Drive Implosions

NASA Astrophysics Data System (ADS)

Forrest, C. J.; Radha, P. B.; Knauer, J. P.; Glebov, V. Yu.; Goncharov, V. N.; Regan, S. P.; Rosenberg, M. J.; Sangster, T. C.; Shmayda, W. T.; Stoeckl, C.; Gatu Johnson, M.

2017-03-01

The deuterium-tritium (D-T) and deuterium-deuterium neutron yield ratio in cryogenic inertial confinement fusion (ICF) experiments is used to examine multifluid effects, traditionally not included in ICF modeling. This ratio has been measured for ignition-scalable direct-drive cryogenic DT implosions at the Omega Laser Facility [T. R. Boehly et al., Opt. Commun. 133, 495 (1997), 10.1016/S0030-4018(96)00325-2] using a high-dynamic-range neutron time-of-flight spectrometer. The experimentally inferred yield ratio is consistent with both the calculated values of the nuclear reaction rates and the measured preshot target-fuel composition. These observations indicate that the physical mechanisms that have been proposed to alter the fuel composition, such as species separation of the hydrogen isotopes [D. T. Casey et al., Phys. Rev. Lett. 108, 075002 (2012), 10.1103/PhysRevLett.108.075002], are not significant during the period of peak neutron production in ignition-scalable cryogenic direct-drive DT implosions.

Demonstration of Fuel Hot-Spot Pressure in Excess of 50 Gbar for Direct-Drive, Layered Deuterium-Tritium Implosions on OMEGA.

PubMed

Regan, S P; Goncharov, V N; Igumenshchev, I V; Sangster, T C; Betti, R; Bose, A; Boehly, T R; Bonino, M J; Campbell, E M; Cao, D; Collins, T J B; Craxton, R S; Davis, A K; Delettrez, J A; Edgell, D H; Epstein, R; Forrest, C J; Frenje, J A; Froula, D H; Gatu Johnson, M; Glebov, V Yu; Harding, D R; Hohenberger, M; Hu, S X; Jacobs-Perkins, D; Janezic, R; Karasik, M; Keck, R L; Kelly, J H; Kessler, T J; Knauer, J P; Kosc, T Z; Loucks, S J; Marozas, J A; Marshall, F J; McCrory, R L; McKenty, P W; Meyerhofer, D D; Michel, D T; Myatt, J F; Obenschain, S P; Petrasso, R D; Radha, P B; Rice, B; Rosenberg, M J; Schmitt, A J; Schmitt, M J; Seka, W; Shmayda, W T; Shoup, M J; Shvydky, A; Skupsky, S; Solodov, A A; Stoeckl, C; Theobald, W; Ulreich, J; Wittman, M D; Woo, K M; Yaakobi, B; Zuegel, J D

2016-07-08

A record fuel hot-spot pressure P_{hs}=56±7  Gbar was inferred from x-ray and nuclear diagnostics for direct-drive inertial confinement fusion cryogenic, layered deuterium-tritium implosions on the 60-beam, 30-kJ, 351-nm OMEGA Laser System. When hydrodynamically scaled to the energy of the National Ignition Facility, these implosions achieved a Lawson parameter ∼60% of the value required for ignition [A. Bose et al., Phys. Rev. E 93, 011201(R) (2016)], similar to indirect-drive implosions [R. Betti et al., Phys. Rev. Lett. 114, 255003 (2015)], and nearly half of the direct-drive ignition-threshold pressure. Relative to symmetric, one-dimensional simulations, the inferred hot-spot pressure is approximately 40% lower. Three-dimensional simulations suggest that low-mode distortion of the hot spot seeded by laser-drive nonuniformity and target-positioning error reduces target performance.

D 2 and DT Liquid-Layer Target Shots on NIF

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

Walters, Curtis; Alger, Ethan; Bhandarkar, Suhas

Experiments at the National Ignition Facility (NIF) using targets containing a Deuterium-Tritium (DT) fuel layer have, until recently, required that a high-quality layer of solid deuterium-tritium (herein referred to as an "ice-layer") be formed in the capsule. The development of a process to line the inner surface of a target capsule with a foam layer of a thickness that is typical of icelayers has resulted in the ability to field targets with liquid layers wetting the foam. Successful fielding of liquid-layer targets on NIF required not only a foam lined capsule, but also changes to the capsule filling process andmore » the manner with which the inventory is maintained in the capsule. Additionally, changes to target heater power and the temperature drops across target components were required in order to achieve the desired range of shot temperatures. These changes, and the target's performance during four target shots on NIF will be discussed.« less

Tritium laboratory with multiple purposes at NIPNE Magurele Romania

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

Matei, L.; Postolache, C.

2008-07-15

The Tritium Laboratory from NIPNE (Romania)) is part of Radioisotope Research and Production Center. The Tritium Laboratory has been in operation since 1960, and carries out R and D activities involving tritium sources in gaseous, liquids and solid state, provides specialized service to CANDU NPP Cernavoda (Romania)), and provides tritium assay services to internal and external customers. The paper presents the activities and perspectives of Tritium Laboratory and its performances in accordance with Quality System Management. (authors)

10 CFR 39.55 - Tritium neutron generator target sources.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Tritium neutron generator target sources. 39.55 Section 39... Equipment § 39.55 Tritium neutron generator target sources. (a) Use of a tritium neutron generator target....77. (b) Use of a tritium neutron generator target source, containing quantities exceeding 1,110 GBg...

10 CFR 39.55 - Tritium neutron generator target sources.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Tritium neutron generator target sources. 39.55 Section 39... Equipment § 39.55 Tritium neutron generator target sources. (a) Use of a tritium neutron generator target....77. (b) Use of a tritium neutron generator target source, containing quantities exceeding 1,110 GBg...

10 CFR 39.55 - Tritium neutron generator target sources.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Tritium neutron generator target sources. 39.55 Section 39... Equipment § 39.55 Tritium neutron generator target sources. (a) Use of a tritium neutron generator target....77. (b) Use of a tritium neutron generator target source, containing quantities exceeding 1,110 GBg...

10 CFR 39.55 - Tritium neutron generator target sources.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Tritium neutron generator target sources. 39.55 Section 39... Equipment § 39.55 Tritium neutron generator target sources. (a) Use of a tritium neutron generator target....77. (b) Use of a tritium neutron generator target source, containing quantities exceeding 1,110 GBg...

10 CFR 39.55 - Tritium neutron generator target sources.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Tritium neutron generator target sources. 39.55 Section 39... Equipment § 39.55 Tritium neutron generator target sources. (a) Use of a tritium neutron generator target....77. (b) Use of a tritium neutron generator target source, containing quantities exceeding 1,110 GBg...

Regeneration and tritium recovery from the large JET neutral injection cryopump system after the FTE

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

Obert, W.; Bell, A.; Davies, J.

1992-12-01

Neutral Beam Injection (NBI) was used to introduce tritium into the plasma for the First Tritium Experiment In addition to the decisive advantage of depositing the tritium into the centre of the plasma, the use of NBI also minimized the total quantity of tritium introduced into the Torus and the contamination of the vacuum vessel. However, because of the relatively low gas efficiency of the positive ion injection system approximately 95% of the total quantity of tritium introduced was pumped by the large condensation cryopumps which form an integral part of the injector. Several hardware and associated software changes weremore » implemented in order to making provision for possible fault scenarios during operation with tritium and to ensure complete regeneration of the tritium from the cryopumps. The tritium released after all subsequent regeneration`s has been monitored carefully in order to determine the amount of tritium retained by the black anodized liquid nitrogen panel surfaces of the cryopump and to compare it with experiments at TSTA on JET samples before the FTE.« less

A Fusion Nuclear Science Facility for a fast-track path to DEMO

DOE PAGES

Garofalo, Andrea M.; Abdou, M.; Canik, John M.; ...

2014-10-01

An accelerated fusion energy development program, a “fast-track” approach, requires developing an understanding of fusion nuclear science (FNS) in parallel with research on ITER to study burning plasmas. A Fusion Nuclear Science Facility (FNSF) in parallel with ITER provides the capability to resolve FNS feasibility issues related to power extraction, tritium fuel sustainability, and reliability, and to begin construction of DEMO upon the achievement of Q~10 in ITER. Fusion nuclear components, including the first wall (FW)/blanket, divertor, heating/fueling systems, etc. are complex systems with many inter-related functions and different materials, fluids, and physical interfaces. These in-vessel nuclear components must operatemore » continuously and reliably with: (a) Plasma exposure, surface particle & radiation loads, (b) High energy 2 neutron fluxes and their interactions in materials (e.g. peaked volumetric heating with steep gradients, tritium production, activation, atomic displacements, gas production, etc.), (c) Strong magnetic fields with temporal and spatial variations (electromagnetic coupling to the plasma including off-normal events like disruptions), and (d) a High temperature, high vacuum, chemically active environment. While many of these conditions and effects are being studied with separate and multiple effect experimental test stands and modeling, fusion nuclear conditions cannot be completely simulated outside the fusion environment. This means there are many new multi-physics, multi-scale phenomena and synergistic effects yet to be discovered and accounted for in the understanding, design and operation of fusion as a self-sustaining, energy producing system, and significant experimentation and operational experience in a true fusion environment is an essential requirement. In the following sections we discuss the FNSF objectives, describe the facility requirements and a facility concept and operation approach that can accomplish those objectives, and assess the readiness to construct with respect to several key FNSF issues: materials, steady-state operation, disruptions, power exhaust, and breeding blanket. Finally we present our conclusions.« less

Apparatus and method for stripping tritium from molten salt

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

Holcomb, David E.; Wilson, Dane F.

A method of stripping tritium from flowing stream of molten salt includes providing a tritium-separating membrane structure having a porous support, a nanoporous structural metal-ion diffusion barrier layer, and a gas-tight, nonporous palladium-bearing separative layer, directing the flowing stream of molten salt into contact with the palladium-bearing layer so that tritium contained within the molten salt is transported through the tritium-separating membrane structure, and contacting a sweep gas with the porous support for collecting the tritium.

Continuous aqueous tritium monitor

DOEpatents

McManus, Gary J.; Weesner, Forrest J.

1989-05-30

An apparatus for a selective on-line determination of aqueous tritium concentration is disclosed. A moist air stream of the liquid solution being analyzed is passed through a permeation dryer where the tritium and moisture and selectively removed to a purge air stream. The purge air stream is then analyzed for tritium concentration, humidity, and temperature, which allows computation of liquid tritium concentration.

Estimation of the tritium retention in ITER tungsten divertor target using macroscopic rate equations simulations

NASA Astrophysics Data System (ADS)

Hodille, E. A.; Bernard, E.; Markelj, S.; Mougenot, J.; Becquart, C. S.; Bisson, R.; Grisolia, C.

2017-12-01

Based on macroscopic rate equation simulations of tritium migration in an actively cooled tungsten (W) plasma facing component (PFC) using the code MHIMS (migration of hydrogen isotopes in metals), an estimation has been made of the tritium retention in ITER W divertor target during a non-uniform exponential distribution of particle fluxes. Two grades of materials are considered to be exposed to tritium ions: an undamaged W and a damaged W exposed to fast fusion neutrons. Due to strong temperature gradient in the PFC, Soret effect’s impacts on tritium retention is also evaluated for both cases. Thanks to the simulation, the evolutions of the tritium retention and the tritium migration depth are obtained as a function of the implanted flux and the number of cycles. From these evolutions, extrapolation laws are built to estimate the number of cycles needed for tritium to permeate from the implantation zone to the cooled surface and to quantify the corresponding retention of tritium throughout the W PFC.

On the conversion of tritium units to mass fractions for hydrologic applications

USGS Publications Warehouse

Stonestrom, David A.; Andraski, Brian J.; Cooper, Clay A.; Mayers, Charles J.; Michel, Robert L.

2013-01-01

We develop a general equation for converting laboratory-reported tritium levels, expressed either as concentrations (tritium isotope number fractions) or mass-based specific activities, to mass fractions in aqueous systems. Assuming that all tritium is in the form of monotritiated water simplifies the derivation and is shown to be reasonable for most environmental settings encountered in practice. The general equation is nonlinear. For tritium concentrations c less than 4.5×1012 tritium units (TU) - i.e. specific tritium activities11 Bq kg-1 - the mass fraction w of tritiated water is approximated to within 1 part per million by w ≈ c×2.22293×10-18, i.e. the conversion is linear for all practical purposes. Terrestrial abundances serve as a proxy for non-tritium isotopes in the absence of sample-specific data. Variation in the relative abundances of non-tritium isotopes in the terrestrial hydrosphere produces a minimum range for the mantissa of the conversion factor of [2.22287; 2.22300].

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

Koval, G.N.; Kuzmina, A.I.; Kolomiets, N.F.

In this paper results of the long term of control of tritium concentration in the water fractions in the region close to the tritium laboratories of INR NAS of Ukraine are presented. The regular observations for the tritium concentration in the water fractions (thawed water of the snow cover, birch juice and sewer water) in the influence region of tritium laboratories shows small amount of tritium concentration in all kinds of investigated water fractions in comparison with the tritium concentration in the reper points. The proper connection of the levels of tritium concentration of the water samples with the quantitymore » of the technology production is observed. In common, the tritium pollution on the territory of INR shows the tendency for a considerable decrease of the environmental pollution levels from year to year. It can be explained by the perfection of the production technology of tritium structures and targets as well as the rising of the qualification of the personnel. 3 refs., 4 figs.« less

In-pile tritium-permeation measurements on T91 tubes with double walls or a Fe-Al/Al 2O 3 coating

NASA Astrophysics Data System (ADS)

Conrad, R.; Bakker, K.; Chabrol, C.; Fütterer, M. A.; van der Laan, J. G.; Rigal, E.; Stijkel, M. P.

2000-12-01

Two new irradiation projects are being performed at the HFR Petten, named EXOTIC-8.9 and EXOTIC-8.10. Issues such as tritium release from candidate ceramic breeder pebbles for the HCPB blanket and tritium permeation through cooling tubes of the WCLL blanket are investigated simultaneously. In EXOTIC-8.9, the tritium release behaviour of a Li 2TiO 3 pebble bed is measured along with the tritium-permeation rate through a double-wall tube (DWT) of T91 with a Cu interlayer. In EXOTIC-8.10, the tritium release behaviour of a Li 4SiO 4 pebble bed is measured along with the tritium permeation rate through a T91 tube with a Fe-Al/Al 2O 3 coating as tritium permeation barrier (TPB). Tritium permeation phenomena are studied by variations of temperatures and purge gas conditions. This paper reports on the results of the first 100 irradiation days.

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

Morgan, Michael J.

The Hydrogen Fracture Toughness Tester (HFTT) is a mechanical testing machine designed for conducting fracture mechanics tests on materials in high-pressure hydrogen gas. The tester is needed for evaluating the effects of hydrogen on the cracking properties of tritium reservoir materials. It consists of an Instron Model 8862 Electromechanical Test Frame; an Autoclave Engineering Pressure Vessel, an Electric Potential Drop Crack Length Measurement System, associated computer control and data acquisition systems, and a high-pressure hydrogen gas manifold and handling system.

Radionuclides in ground water at the Idaho National Engineering Laboratory, Idaho

USGS Publications Warehouse

Knobel, LeRoy L.; Mann, Larry J.

1988-01-01

Sampling for radionuclides in groundwater was conducted at the Idaho National Engineering Laboratory during September to November 5 1987. Water samples from 80 wells that obtain water from the Snake River Plain aquifer and 1 well that obtains water from a shallow, discontinuous perched-water body at the Radioactive Waste Management Complex were collected and analyzed for tritium, strontium-90, plutonium-238, plutonium-239, -240 (undivided), americium-241, cesium-137, cobalt-60, and potassium-40--a naturally occurring radionuclide. The groundwater samples were analyzed at the Idaho National Engineering Laboratory in Idaho. Tritium and strontium-90 concentrations ranged from below the reporting level to 80.6 +/-0.000005 and 193 +/-5x10 to the minus eight micrograms Ci/ml, respectively. Water from a disposal well at Test Area North--which has not been used to dispose of waste water since September 1972--contained 122 +/-9x10 to the minus eleven micrograms Ci/ml of plutonium-238, 500 +/-20x10 to the minus eleven of plutonium-239, -240 (undivided), 21 +/-4x10 to the minus eleven micrograms Ci/ml of americium-241, and 750 +/-20x10 to the minus eight micrograms Ci/ml cesium-137; the presence of these radionuclides was verified by resampling and reanalysis. The disposal well had 8.9 +/-0.0000009 micrograms Ci/ml of cobalt-60 on October 28, 1987, but cobalt-60 was not detected when the well was resampled on January 11, 1988. Potassium-40 concentrations were less than the reporting level in all wells. (USGS)

Fermilab | Tritium at Fermilab | Ferry Creek Results

Science.gov Websites

newsletter Ferry Creek Results chart This chart (click chart for larger version) shows the levels of tritium following the detection of low levels of tritium in Indian Creek in November 2005. The levels of tritium in . Fermilab continues to monitor the ponds and creeks on its site and take steps to keep the levels of tritium

Continuous aqueous tritium monitor

DOEpatents

McManus, G.J.; Weesner, F.J.

1987-10-19

An apparatus for a selective on-line determination of aqueous tritium concentration is disclosed. A moist air stream of the liquid solution being analyzed is passed through a permeation dryer where the tritium and moisture are selectively removed to a purge air stream. The purge air stream is then analyzed for tritium concentration, humidity, and temperature, which allows computation of liquid tritium concentration. 2 figs.

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

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

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium monitor and collection system

DOEpatents

Bourne, G.L.; Meikrantz, D.H.; Ely, W.E.; Tuggle, D.G.; Grafwallner, E.G.; Wickham, K.L.; Maltrud, H.R.; Baker, J.D.

1992-01-14

This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber. Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next on-line getter is valved on. Simultaneously, the first getter is heated and a pure helium purge is employed to carry the tritium from the getter. The tritium loaded gas stream is then routed through an ion chamber which measures the tritium activity. The ion chamber effluent passes through a collection getter that readsorbs the tritium and is removable from the system once it is loaded and is then replaced with a clean getter. Prior to removal of the collection getter, the system switches to a parallel collection getter. The effluent from the collection getter passes through a guard getter to remove traces of tritium prior to exiting the system. The tritium loaded collection getter, once removed, is analyzed by liquid scintillation techniques. The entire sequence is under computer control except for the removal and analysis of the collection getter. 7 figs.

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Drum bubbler tritium processing system

DOEpatents

Rule, K.; Gettelfinger, G.; Kivler, P.

1999-08-17

A method is described for separating tritium oxide from a gas stream containing tritium oxide. The gas stream containing tritium oxide is fed into a container of water having a head space above the water. The tritium oxide is separated by bubbling the gas stream containing tritium oxide through the container of water and removing gas from the container head space above the water. Thereafter, the gas from the head space is dried to remove water vapor from the gas, and the water vapor is recycled to the container of water. 2 figs.

Estimation of Biological Effects of Tritium.

PubMed

Umata, Toshiyuki

2017-01-01

Nuclear fusion technology is expected to create new energy in the future. However, nuclear fusion requires a large amount of tritium as a fuel, leading to concern about the exposure of radiation workers to tritium beta radiation. Furthermore, countermeasures for tritium-polluted water produced in decommissioning of the reactor at Fukushima Daiichi Nuclear Power Station may potentially cause health problems in radiation workers. Although, internal exposure to tritium at a low dose/low dose rate can be assumed, biological effect of tritium exposure is not negligible, because tritiated water (HTO) intake to the body via the mouth/inhalation/skin would lead to homogeneous distribution throughout the whole body. Furthermore, organically-bound tritium (OBT) stays in the body as parts of the molecules that comprise living organisms resulting in long-term exposure, and the chemical form of tritium should be considered. To evaluate the biological effect of tritium, the effect should be compared with that of other radiation types. Many studies have examined the relative biological effectiveness (RBE) of tritium. Hence, we report the RBE, which was obtained with radiation carcinogenesis classified as a stochastic effect, and serves as a reference for cancer risk. We also introduce the outline of the tritium experiment and the principle of a recently developed animal experimental system using transgenic mouse to detect the biological influence of radiation exposure at a low dose/low dose rate.

2017 Status report-Tritium aging studies on stainless steel: Effect of hydrogen, tritium and decay helium on the fracture-toughness properties of stem, cup and block forgings

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

Morgan, Michael J.

The materials of construction of tritium reservoirs are forged stainless steels. During service, the structural properties of the stainless steel change over time because of the diffusion of tritium into the reservoir wall and its radioactive decay to helium-3. This aging effect can cause cracks to initiate and grow which could result in a tritium leak or delayed failure of a tritium reservoir. Numerous factors affect the tendency for crack formation and propagation and are being investigated in this program. The goal of the research is to provide relevant fracture mechanics data that can be used by the design agenciesmore » in their assessments of tritium reservoir structural integrity. In this status report, new experimental results are presented on the effects of tritium and decay helium on the cracking properties of specimens taken from actual tritium reservoir forgings instead of the experimental forgings of past programs. The properties measured are more representative of actual reservoir properties because the microstructure of the specimens tested are more like that of the actual tritium reservoirs. The program was designed to measure the effects of material variables on tritium compatibility and includes two stainless steels (Type 304L and 316L stainless steel), multiple yield strengths (360-500 MPa), and multiple forging shapes (Stem, Cup, and Block).« less

High-density carbon capsule experiments on the national ignition facility

DOE PAGES

Ross, J. S.; Ho, D.; Milovich, J.; ...

2015-02-25

Indirect-drive implosions with a high-density carbon (HDC) capsule were conducted on the National Ignition Facility (NIF) to test HDC properties as an ablator material for inertial confinement fusion. In this study, a series of five experiments were completed with 76-μm-thick HDC capsules using a four-shock laser pulse optimized for HDC. The pulse delivered a total energy of 1.3 MJ with a peak power of 360 TW. The experiment demonstrated good laser to target coupling (~90 %) and excellent nuclear performance. Lastly, a deuterium and tritium gas-filled HDC capsule implosion produced a neutron yield of 1.6×10 15 ± 3×10 13, amore » yield over simulated in one dimension of 70%.« less

Calibration of a transient transport model to tritium data in streams and simulation of groundwater ages in the western Lake Taupo catchment, New Zealand

NASA Astrophysics Data System (ADS)

Gusyev, M. A.; Toews, M.; Morgenstern, U.; Stewart, M.; White, P.; Daughney, C.; Hadfield, J.

2013-03-01

Here we present a general approach of calibrating transient transport models to tritium concentrations in river waters developed for the MT3DMS/MODFLOW model of the western Lake Taupo catchment, New Zealand. Tritium has a known pulse-shaped input to groundwater systems due to the bomb tritium in the early 1960s and, with its radioactive half-life of 12.32 yr, allows for the determination of the groundwater age. In the transport model, the tritium input (measured in rainfall) passes through the groundwater system, and the simulated tritium concentrations are matched to the measured tritium concentrations in the river and stream outlets for the Waihaha, Whanganui, Whareroa, Kuratau and Omori catchments from 2000-2007. For the Kuratau River, tritium was also measured between 1960 and 1970, which allowed us to fine-tune the transport model for the simulated bomb-peak tritium concentrations. In order to incorporate small surface water features in detail, an 80 m uniform grid cell size was selected in the steady-state MODFLOW model for the model area of 1072 km2. The groundwater flow model was first calibrated to groundwater levels and stream baseflow observations. Then, the transient tritium transport MT3DMS model was matched to the measured tritium concentrations in streams and rivers, which are the natural discharge of the groundwater system. The tritium concentrations in the rivers and streams correspond to the residence time of the water in the groundwater system (groundwater age) and mixing of water with different age. The transport model output showed a good agreement with the measured tritium values. Finally, the tritium-calibrated MT3DMS model is applied to simulate groundwater ages, which are used to obtain groundwater age distributions with mean residence times (MRTs) in streams and rivers for the five catchments. The effect of regional and local hydrogeology on the simulated groundwater ages is investigated by demonstrating groundwater ages at five model cross-sections to better understand MRTs simulated with tritium-calibrated MT3DMS and lumped parameter models.

Tritium monitor

DOEpatents

Chastagner, Philippe

1994-01-01

A system for continuously monitoring the concentration of tritium in an aqueous stream. The system pumps a sample of the stream to magnesium-filled combustion tube which reduces the sample to extract hydrogen gas. The hydrogen gas is then sent to an isotope separation device where it is separated into two groups of isotopes: a first group of isotopes containing concentrations of deuterium and tritium, and a second group of isotopes having substantially no deuterium and tritium. The first group of isotopes containing concentrations of deuterium and tritium is then passed through a tritium detector that produces an output proportional to the concentration of tritium detected. Preferably, the detection system also includes the necessary automation and data collection equipment and instrumentation for continuously monitoring an aqueous stream.

Tritium monitor

DOEpatents

Chastagner, P.

1994-06-14

A system is described for continuously monitoring the concentration of tritium in an aqueous stream. The system pumps a sample of the stream to magnesium-filled combustion tube which reduces the sample to extract hydrogen gas. The hydrogen gas is then sent to an isotope separation device where it is separated into two groups of isotopes: a first group of isotopes containing concentrations of deuterium and tritium, and a second group of isotopes having substantially no deuterium and tritium. The first group of isotopes containing concentrations of deuterium and tritium is then passed through a tritium detector that produces an output proportional to the concentration of tritium detected. Preferably, the detection system also includes the necessary automation and data collection equipment and instrumentation for continuously monitoring an aqueous stream. 1 fig.

Studies on the behaviour of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

NASA Astrophysics Data System (ADS)

Kobayashi, K.; Isobe, K.; Iwai, Y.; Hayashi, T.; Shu, W.; Nakamura, H.; Kawamura, Y.; Yamada, M.; Suzuki, T.; Miura, H.; Uzawa, M.; Nishikawa, M.; Yamanishi, T.

2007-12-01

Confinement and the removal of tritium are key subjects for the safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell, tritium is often released as vapour and is in contact with the inner walls. The inner walls of the ITER tritium plant building will also be exposed to tritium in an accident. The tritium released in the buildings is removed by the atmosphere detritiation systems (ADS), where the tritium is oxidized by catalysts and is removed as water. A special gas of SF6 is used in ITER and is expected to be released in an accident such as a fire. Although the SF6 gas has potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed as yet. Tritiated water is produced in the regeneration process of ADS and is subsequently processed by the ITER water detritiation system (WDS). One of the key components of the WDS is an electrolysis cell. To overcome the issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R&D task: (1) tritium behaviour in concrete; (2) the effect of SF6 on the performance of ADS and (3) tritium durability of the electrolysis cell of the ITER-WDS. (1) The tritiated water vapour penetrated up to 50 mm into the concrete from the surface in six months' exposure. The penetration rate of tritium in the concrete was thus appreciably first, the isotope exchange capacity of the cement paste plays an important role in tritium trapping and penetration into concrete materials when concrete is exposed to tritiated water vapour. It is required to evaluate the effect of coating on the penetration rate quantitatively from the actual tritium tests. (2) SF6 gas decreased the detritiation factor of ADS. Since the effect of SF6 depends closely on its concentration, the amount of SF6 released into the tritium handling area in an accident should be reduced by some ideas of arrangement of components in the buildings. (3) It was expected that the electrolysis cell of the ITER-WDS could endure 3 years' operation under the ITER design conditions. Measuring the concentration of the fluorine ions could be a promising technique for monitoring the damage to the electrolysis cell.

Alleviation of Facility/Engine Interactions in an Open-Jet Scramjet Test Facility

NASA Technical Reports Server (NTRS)

Albertson, Cindy W.; Emami, Saied

2001-01-01

Results of a series of shakedown tests to eliminate facility/engine interactions in an open-jet scramjet test facility are presented. The tests were conducted with the NASA DFX (Dual-Fuel eXperimental scramjet) engine in the NASA Langley Combustion Heated Scramjet Test Facility (CHSTF) in support of the Hyper-X program, The majority of the tests were conducted at a total enthalpy and pressure corresponding to Mach 5 flight at a dynamic pressure of 734 psf. The DFX is the largest engine ever tested in the CHSTF. Blockage, in terms of the projected engine area relative to the nozzle exit area, is 81% with the engine forebody leading edge aligned with the upper edge of the facility nozzle such that it ingests the nozzle boundary layer. The blockage increases to 95% with the engine forebody leading edge positioned 2 in. down in the core flow. Previous engines successfully tested in the CHSTF have had blockages of no more than 51%. Oil flow studies along with facility and engine pressure measurements were used to define flow behavior. These results guided modifications to existing aeroappliances and the design of new aeroappliances. These changes allowed fueled tests to be conducted without facility interaction effects in the data with the engine forebody leading edge positioned to ingest the facility nozzle boundary layer. Interaction effects were also reduced for tests with the engine forebody leading edge positioned 2 in. into the core flow, however some interaction effects were still evident in the engine data. A new shroud and diffuser have been designed with the goal of allowing fueled tests to be conducted with the engine forebody leading edge positioned in the core without facility interaction effects in the data. Evaluation tests of the new shroud and diffuser will be conducted once ongoing fueled engine tests have been completed.

Plasma kinetic effects on atomistic mix in one dimension and at structured interfaces (II)

NASA Astrophysics Data System (ADS)

Albright, Brian; Yin, Lin; Cooley, James; Haack, Jeffrey; Douglas, Melissa

2017-10-01

The Marble campaign seeks to develop a platform for studying mix evolution in turbulent, inhomogeneous, high-energy-density plasmas at the NIF. Marble capsules contain engineered CD foams, the pores of which are filled with hydrogen and tritium. During implosion, hydrodynamic stirring and plasma diffusivity mix tritium fuel into the surrounding CD plasma, leading to both DD and DT fusion neutron production. In this presentation, building upon prior work, kinetic particle-in-cell simulations using the VPIC code are used to examine kinetic effects on thermonuclear burn in Marble-like settings. Departures from Maxwellian distributions are observed near the interface and TN burn rates and inferred temperatures from synthetic neutron time of flight diagnostics are compared with those from treating the background species as Maxwellian. Work performed under the auspices of the U.S. DOE by the Los Alamos National Security, LLC Los Alamos National Laboratory and supported by the ASC and Science programs.

Investigation of the Possibility of Using Nuclear Magnetic Spin Alignment

NASA Technical Reports Server (NTRS)

Dent, William V., Jr.

1998-01-01

The goal of the program to investigate a "Gasdynamic fusion propulsion system for space exploration" is to develop a fusion propulsion system for a manned mission to the planet mars. A study using Deuterium and Tritium atoms are currently in progress. When these atoms under-go fusion, the resulting neutrons and alpha particles are emitted in random directions (isotropically). The probable direction of emission is equal for all directions, thus resulting in wasted energy, massive shielding and cooling requirements, and serious problems with the physics of achieving fusion. If the nuclear magnetic spin moments of the deuterium and tritium nuclei could be precisely aligned at the moment of fusion, the stream of emitted neutrons could be directed out the rear of the spacecraft for thrust and the alpha particles directed forward into an electromagnet ot produce electricity to continue operating the fusion engine. The following supporting topics are discussed: nuclear magnetic moments and spin precession in magnetic field, nuclear spin quantum mechanics, kinematics of nuclear reactions, and angular distribution of particles.

Tritium release from SS316 under vacuum condition

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

Torikai, Y.; Penzhorn, R.D.

The plasma facing surface of the ITER vacuum vessel, partly made of low carbon austenitic stainless steel type 316L, will incorporate tritium during machine operation. In this paper the kinetics of tritium release from stainless steel type 316 into vacuum and into a noble gas stream are compared and modelled. Type 316 stainless steel specimens loaded with tritium either by exposure to 1.2 kPa HT at 573 K or submersion into liquid HTO at 298 K showed characteristic thin surface layers trapping tritium in concentrations far higher than those determined in the bulk. The evolution of the tritium depth profilemore » in the bulk during heating under vacuum was non-discernible from that of tritium liberated into a stream of argon. Only the relative amount of the two released tritium-species, i.e. HT or HTO, was different. Temperature-dependent depth profiles could be predicted with a one-dimensional diffusion model. Diffusion coefficients derived from fitting of the tritium release into an evacuated vessel or a stream of argon were found to be (1.4 ± 1.0)*10{sup -7} and (1.3 ± 0.9)*10{sup -9} cm{sup 2}/s at 573 and 423 K, respectively. Polished surfaces on type SS316 stainless steel inhibit considerably the thermal release rate of tritium.« less

Comparison of the Current Center of Site Annual Neshap Dose Modeling at the Savannah River Site with Other Assessment Methods.

PubMed

Minter, Kelsey M; Jannik, G Timothy; Stagich, Brooke H; Dixon, Kenneth L; Newton, Joseph R

2018-04-01

The U.S. Environmental Protection Agency (EPA) requires the use of the model CAP88 to estimate the total effective dose (TED) to an offsite maximally exposed individual (MEI) for demonstrating compliance with 40 CFR 61, Subpart H: The National Emission Standards for Hazardous Air Pollutants (NESHAP) regulations. For NESHAP compliance at the Savannah River Site (SRS), the EPA, the U.S. Department of Energy (DOE), South Carolina's Department of Health and Environmental Control, and SRS approved a dose assessment method in 1991 that models all radiological emissions as if originating from a generalized center of site (COS) location at two allowable stack heights (0 m and 61 m). However, due to changes in SRS missions, radiological emissions are no longer evenly distributed about the COS. An area-specific simulation of the 2015 SRS radiological airborne emissions was conducted to compare to the current COS method. The results produced a slightly higher dose estimate (2.97 × 10 mSv vs. 2.22 × 10 mSv), marginally changed the overall MEI location, and noted that H-Area tritium emissions dominated the dose. Thus, an H-Area dose model was executed as a potential simplification of the area-specific simulation by adopting the COS methodology and modeling all site emissions from a single location in H-Area using six stack heights that reference stacks specific to the tritium production facilities within H-Area. This "H-Area Tritium Stacks" method produced a small increase in TED estimates (3.03 × 10 mSv vs. 2.97 × 10 mSv) when compared to the area-specific simulation. This suggests that the current COS method is still appropriate for demonstrating compliance with NESHAP regulations but that changing to the H-Area Tritium Stacks assessment method may now be a more appropriate representation of operations at SRS.

Tritium Decay Helium-3 Effects in Tungsten

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

Shimada, M.; Merrill, B. J.

2016-06-01

A critical challenge for long-term operation of ITER and beyond to a Demonstration reactor (DEMO) and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to steady-state/transient heat fluxes and intense neutral/ion particle fluxes under the extreme fusion nuclear environment, while at the same time minimizing in-vessel tritium inventories and permeation fluxes into the PFC’s coolant. Tritium will diffuse in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [1,2]. Tritium decay into helium-3 may also play a major role in microstructuralmore » evolution (e.g. helium embrittlement) in tungsten due to relatively low helium-4 production (e.g. He/dpa ratio of 0.4-0.7 appm [3]) in tungsten. Tritium-decay helium-3 effect on tungsten is hardly understood, and its database is very limited. Two tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) were exposed to high flux (ion flux of 1.0x1022 m-2s-1 and ion fluence of 1.0x1026 m-2) 0.5%T2/D2 plasma at two different temperatures (200, and 500°C) in Tritium Plasma Experiment (TPE) at Idaho National Laboratory. Tritium implanted samples were stored at ambient temperature in air for more than 3 years to investigate tritium decay helium-3 effect in tungsten. The tritium distributions on plasma-exposed was monitored by a tritium imaging plate technique during storage period [4]. Thermal desorption spectroscopy was performed with a ramp rate of 10°C/min up to 900°C to outgas residual deuterium and tritium but keep helium-3 in tungsten. These helium-3 implanted samples were exposed to deuterium plasma in TPE to investigate helium-3 effect on deuterium behavior in tungsten. The results show that tritium surface concentration in 200°C sample decreased to 30 %, but tritium surface concentration in 500°C sample did not alter over the 3 years storage period, indicating possible tritium retention in helium-3 bubble. This paper reports the initial experimental observation of tritium-decay helium-3 in tungsten exposed to deuterium/tritium plasma along with electron microscope analysis and also discusses a Tritium Migration Analysis Program (TMAP) analysis of tritium-decay helium-3 effects on tritium retention in tungsten for DEMO and future fusion reactor. [1] Y. Hatano, et.al., Nucl. Fusion 53 (2013) 073006 [2] M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008 [3] M. Sawan, Fus. Sci. Technol. 66 (2014) 272 [4] T. Otsuka, Fus. Sci. Technol. 60 (2011) 1539 This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.« less

Tritium in Exit Signs | RadTown USA | US EPA

EPA Pesticide Factsheets

2018-05-01

Many exit signs contain tritium to light the sign without batteries or electricity, which allows it to remain lit if the power goes out. Tritium is most dangerous when it is inhaled or swallowed. Never tamper with a tritium exit sign.

Improvement of tritium accountancy technology for ITER fuel cycle safety enhancement

NASA Astrophysics Data System (ADS)

O'hira, S.; Hayashi, T.; Nakamura, H.; Kobayashi, K.; Tadokoro, T.; Nakamura, H.; Itoh, T.; Yamanishi, T.; Kawamura, Y.; Iwai, Y.; Arita, T.; Maruyama, T.; Kakuta, T.; Konishi, S.; Enoeda, M.; Yamada, M.; Suzuki, T.; Nishi, M.; Nagashima, T.; Ohta, M.

2000-03-01

In order to improve the safe handling and control of tritium for the ITER fuel cycle, effective in situ tritium accounting methods have been developed at the Tritium Process Laboratory in the Japan Atomic Energy Research Institute under one of the ITER-EDA R&D tasks. The remote and multilocation analysis of process gases by an application of laser Raman spectroscopy developed and tested could provide a measurement of hydrogen isotope gases with a detection limit of 0.3 kPa analytical periods of 120 s. An in situ tritium inventory measurement by application of a `self-assaying' storage bed with 25 g tritium capacity could provide a measurement with the required detection limit of less than 1% and a design proof of a bed with 100 g tritium capacity.

Process for making solid-state radiation-emitting composition

DOEpatents

Ashley, Carol S.; Brinker, C. Jeffrey; Reed, Scott; Walko, Robert J.

1993-01-01

The invention provides a process for loading an aerogel substrate with tritium and the resultant compositions. According to the process, an aerogel substrate is hydrolyzed so that surface OH groups are formed. The hydrolyzed aerogel is then subjected to tritium exchange employing, for example, a tritium-containing gas, whereby tritium atoms replace H atoms of surface OH groups. OH and/or CH groups of residual alcohol present in the aerogel may also undergo tritium exchange.

Process for making solid-state radiation-emitting composition

DOEpatents

Ashley, C.S.; Brinker, C.J.; Reed, S.; Walko, R.J.

1993-08-31

The invention provides a process for loading an aerogel substrate with tritium and the resultant compositions. According to the process, an aerogel substrate is hydrolyzed so that surface OH groups are formed. The hydrolyzed aerogel is then subjected to tritium exchange employing, for example, a tritium-containing gas, whereby tritium atoms replace H atoms of surface OH groups. OH and/or CH groups of residual alcohol present in the aerogel may also undergo tritium exchange.

Fermilab | Tritium at Fermilab | Frequently asked questions

Science.gov Websites

computing Quantum initiatives Research and development Key discoveries Benefits of particle physics Particle Accelerators Leading accelerator technology Accelerator complex Illinois Accelerator Research Center Fermilab questions about tritium Tritium in surface water Indian Creek Kress Creek Ferry Creek Tritium in sanitary

Effect of tritium and decay helium on the fracture toughness properties of stainless steel weldments

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

Morgan, M. J.; West, S.; Tosten, M. H.

2008-07-15

J-Integral fracture toughness tests were conducted on tritium-exposed-and- aged Types 304L and 21-6-9 stainless steel weldments in order to measure the combined effects of tritium and its decay product, helium-3 on the fracture toughness properties. Initially, weldments have fracture toughness values about three times higher than base-metal values. Delta-ferrite phase in the weld microstructure improved toughness provided no tritium was present in the microstructure. After a tritium-exposure-and-aging treatment that resulted in {approx}1400 atomic parts per million (appm) dissolved tritium, both weldments and base metals had their fracture toughness values reduced to about the same level. The tritium effect was greatermore » in weldments (67 % reduction vs. 37% reduction) largely because the ductile discontinuous delta-ferrite phase was embrittled by tritium and decay helium. For both base metals and weldments, fracture toughness values decreased with increasing decay helium content in the range tested (50-800 appm). (authors)« less

TRITIUM AND DECAY HELIUM EFFECTS ON THE FRACTURE TOUGHNESS PROPERTIES OF STAINLESS STEEL WELDMENTS

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

Morgan, M; Scott West, S; Michael Tosten, M

2007-08-31

J-Integral fracture toughness tests were conducted on tritium-exposed-and-aged Types 304L and 21-6-9 stainless steel weldments in order to measure the combined effects of tritium and its decay product, helium-3 on the fracture toughness properties. Initially, weldments have fracture toughness values about three times higher than base-metal values. Delta-ferrite phase in the weld microstructure improved toughness provided no tritium was present in the microstructure. After a tritium-exposure-and-aging treatment that resulted in {approx}1400 atomic parts per million (appm) dissolved tritium, both weldments and base metals had their fracture toughness values reduced to about the same level. The tritium effect was greater inmore » weldments (67 % reduction vs. 37% reduction) largely because the ductile discontinuous delta-ferrite interfaces were embrittled by tritium and decay helium. Fracture toughness values decreased for both base metals and weldments with increasing decay helium content in the range tested (50-200 appm).« less

Studying of tritium content in snowpack of Degelen mountain range.

PubMed

Turchenko, D V; Lukashenko, S N; Aidarkhanov, A O; Lyakhova, O N

2014-06-01

The paper presents the results of investigation of tritium content in the layers of snow located in the streambeds of the "Degelen" massif contaminated with tritium. The objects of investigation were selected watercourses Karabulak, Uzynbulak, Aktybai located beyond the "Degelen" site. We studied the spatial distribution of tritium relative to the streambed of watercourses and defined the borders of the snow cover contamination. In the centre of the creek watercourses the snow contamination in the surface layer is as high as 40 000 Bq/L. The values of the background levels of tritium in areas not related to the streambed, which range from 40 to 50 Bq/L. The results of snow cover measurements in different seasonal periods were compared. The main mechanisms causing tritium transfer in snow were examined and identified. The most important mechanism of tritium transfer in the streams is tritium emanation from ice or soil surface. Copyright © 2014 Elsevier Ltd. All rights reserved.

2014 Accomplishments-Tritium aging studies on stainless steel: Fracture toughness properties of forged stainless steels-Effect of hydrogen, forging strain rate, and forging temperature

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

Morgan, Michael J.

Forged stainless steels are used as the materials of construction for tritium reservoirs. During service, tritium diffuses into the reservoir walls and radioactively decays to helium-3. Tritium and decay helium cause a higher propensity for cracking which could lead to a tritium leak or delayed failure of a tritium reservoir. The factors that affect the tendency for crack formation and propagation include: Environment; steel type and microstructure; and, vessel configuration (geometry, pressure, residual stress). Fracture toughness properties are needed for evaluating the long-term effects of tritium on their structural properties. Until now, these effects have been characterized by measuring themore » effects of tritium on the tensile and fracture toughness properties of specimens fabricated from experimental forgings in the form of forward-extruded cylinders. A key result of those studies is that the long-term cracking resistance of stainless steels in tritium service depends greatly on the interaction between decay helium and the steels’ forged microstructure. New experimental research programs are underway and are designed to measure tritium and decay helium effects on the cracking properties of stainless steels using actual tritium reservoir forgings instead of the experimental forgings of past programs. The properties measured should be more representative of actual reservoir properties because the microstructure of the specimens tested will be more like that of the tritium reservoirs. The programs are designed to measure the effects of key forging variables on tritium compatibility and include three stainless steels, multiple yield strengths, and four different forging processes. The effects on fracture toughness of hydrogen and crack orientation were measured for type 316L forgings. In addition, hydrogen effects on toughness were measured for Type 304L block forgings having two different yield strengths. Finally, fracture toughness properties of type 304L stainless steel were measured for four different forging strain rates which and two forging temperatures. Tritium exposures have been and are being conducted on companion specimens for property measurements in the upcoming years.« less

Tritium migration from a low-level radioactive-waste disposal site near Chicago, Illinois

USGS Publications Warehouse

Nicholas, J.R.; Healy, R.W.

1988-01-01

This paper describes the results of a study to determine the geologic and hydrologic factors that control migration of tritium from a closed, low-level radioactive-waste disposal site. The disposal site, which operated from 1943 to mid1949, contains waste generated by research activities at the world's first nuclear reactors. Tritium has migrated horizontally at least 1,300 feet northward in glacial drift and more than 650 feet in the underlying dolomite. Thin, gently sloping sand layers in an otherwise clayey glacial drift are major conduits for ground-water flow and tritium migration in a perched zone beneath the disposal site. Tritium concentrations in the drift beneath the disposal site exceed 100,000 nanocuries per liter. Regional horizontal joints in the dolomite are enlarged by solution and are the major conduits for ground-water flow and tritium migration in the dolomite. A weathered zone at the top of the dolomite also is a pathway for tritium migration. The maximum measured tritium concentration in the dolomite is 29.4 nanocuries per liter. Fluctuations of tritium concentration in the dolomite are the result of dilution by seasonal recharge from the drift.

Tritium migration to the surfaces of Type 316 stainless steel; aluminum 6061; and oxygen-free, high-conductivity copper

DOE PAGES

Sharpe, M.; Shmayda, W. T.; Schroder, W. U.

2016-05-25

The migration of tritium to the surfaces of aluminum 6061, oxygen-free, high-conductivity copper (OFHC), and stainless-steel 316 from the bulk metal was studied using low-pressure Tonks–Langmuir argon plasma. The plasma is shown to be effective at removing tritium from metal surfaces in a controlled manner. Tritium is removed in decreasing quantities with successive plasma exposures, which suggests a depletion of the surface and near-surface tritium inventories. A diffusion model was developed to predict tritium migration from the bulk and its accumulation in the water layers present on the metal surface. The model reproduces the rate of tritium re-growth on themore » surface for all three metals and can be used to calculate the triton solubility in the water layers present on metal surfaces. The ratio of surface-to-bulk solubilities at the water-layer/bulk-metal interface uniquely determines the concentration ratio between these two media. Removing the tritium-rich water layers induces tritium to migrate from the bulk to the surface. Furthermore, this process is driven by a concentration gradient that develops in the bulk because of the perturbation on the surface.« less

Fabrication and tritium release property of Li2TiO3-Li4SiO4 biphasic ceramics

NASA Astrophysics Data System (ADS)

Yang, Mao; Ran, Guangming; Wang, Hailiang; Dang, Chen; Huang, Zhangyi; Chen, Xiaojun; Lu, Tiecheng; Xiao, Chengjian

2018-05-01

Li2TiO3-Li4SiO4 biphasic ceramic pebbles have been developed as an advanced tritium breeder due to the potential to combine the advantages of both Li2TiO3 and Li4SiO4. Wet method was developed for the pebble fabrication and Li2TiO3-Li4SiO4 biphasic ceramic pebbles were successfully prepared by wet method using the powders synthesized by hydrothermal method. The tritium release properties of the Li2TiO3-Li4SiO4 biphasic ceramic pebbles were evaluated. The biphasic pebbles exhibited good tritium release property at low temperatures and the tritium release temperature was around 470 °C. Because of the isotope exchange reaction between H2 and tritium, the addition of 0.1%H2 to purge gas He could significantly enhance the tritium gas release and the fraction of molecular form of tritium increased from 28% to 55%. The results indicate that the Li2TiO3-Li4SiO4 biphasic ceramic pebbles fabricated by wet method exhibit good tritium release property and hold promising potential as advanced breeder pebbles.

Advancement Of Tritium Powered Betavoltaic Battery Systems FY16 EOY Report

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

Staack, G.; Gaillard, J.; Hitchcock, D.

2016-10-12

The goal of this work is to increase the power output of tritium-powered betavoltaic batteries and investigate the change in power output and film resistance in real-time during tritium loading of adsorbent films. To this end, several tritium-compatible test vessels with the capability of measuring both the resistivity of a tritium trapping film and the power output of a betavoltaic device in-situ have been designed and fabricated using four electrically insulated feedthroughs in tritium-compatible load cells. Energy conversion devices were received from Widetronix, a betavoltaic manufacturing firm based in Ithaca, NY. Thin films were deposited on the devices and cappedmore » with palladium to facilitate hydrogen loading. Gold contacts were then deposited on top of the films to allow resistivity measurements of the film during hydrogen loading. Finally, the chips were wire bonded and installed in the test cells. The cells were then baked-out under vacuum and leak checked at temperature to reduce the chances of tritium leaks during loading. Following the bake-out, IV curves were measured to verify no internal wires were compromised, and the cells were delivered to Tritium for loading. Tritium loading is anticipated in October, 2017.« less

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

NASA Astrophysics Data System (ADS)

Menard, J. E.; Brown, T.; El-Guebaly, L.; Boyer, M.; Canik, J.; Colling, B.; Raman, R.; Wang, Z.; Zhai, Y.; Buxton, P.; Covele, B.; D'Angelo, C.; Davis, A.; Gerhardt, S.; Gryaznevich, M.; Harb, M.; Hender, T. C.; Kaye, S.; Kingham, D.; Kotschenreuther, M.; Mahajan, S.; Maingi, R.; Marriott, E.; Meier, E. T.; Mynsberge, L.; Neumeyer, C.; Ono, M.; Park, J.-K.; Sabbagh, S. A.; Soukhanovskii, V.; Valanju, P.; Woolley, R.

2016-10-01

A fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR  ≈  1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m-2.2 m are described. In particular, it is found the threshold major radius for TBR  =  1 is {{R}0}≥slant 1.7 m, and a smaller R 0  =  1 m ST device has TBR  ≈  0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A  =  2, R 0  =  3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

DOE PAGES

Menard, J. E.; Brown, T.; El-Guebaly, L.; ...

2016-08-16

Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 ismore » $${{R}_{0}}\\geqslant 1.7$$ m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.« less

Tritium containing polymers having a polymer backbone substantially void of tritium

DOEpatents

Jensen, G.A.; Nelson, D.A.; Molton, P.M.

1992-03-31

A radioluminescent light source comprises a solid mixture of a phosphorescent substance and a tritiated polymer. The solid mixture forms a solid mass having length, width, and thickness dimensions, and is capable of self-support. In one aspect of the invention, the phosphorescent substance comprises solid phosphor particles supported or surrounded within a solid matrix by a tritium containing polymer. The tritium containing polymer comprises a polymer backbone which is essentially void of tritium. 2 figs.

Tritium containing polymers having a polymer backbone substantially void of tritium

DOEpatents

Jensen, George A.; Nelson, David A.; Molton, Peter M.

1992-01-01

A radioluminescent light source comprises a solid mixture of a phosphorescent substance and a tritiated polymer. The solid mixture forms a solid mass having length, width, and thickness dimensions, and is capable of self-support. In one aspect of the invention, the phosphorescent substance comprises solid phosphor particles supported or surrounded within a solid matrix by a tritium containing polymer. The tritium containing polymer comprises a polymer backbone which is essentially void of tritium.

TRITIUM BARRIER MATERIALS AND SEPARATION SYSTEMS FOR THE NGNP

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

Sherman, S; Thad Adams, T

2008-07-17

Contamination of downstream hydrogen production plants or other users of high-temperature heat is a concern of the Next Generation Nuclear Plant (NGNP) Project. Due to the high operating temperatures of the NGNP (850-900 C outlet temperature), tritium produced in the nuclear reactor can permeate through heat exchangers to reach the hydrogen production plant, where it can become incorporated into process chemicals or the hydrogen product. The concentration limit for tritium in the hydrogen product has not been established, but it is expected that any future limit on tritium concentration will be no higher than the air and water effluent limitsmore » established by the NRC and the EPA. A literature survey of tritium permeation barriers, capture systems, and mitigation measures is presented and technologies are identified that may reduce the movement of tritium to the downstream plant. Among tritium permeation barriers, oxide layers produced in-situ may provide the most suitable barriers, though it may be possible to use aluminized surfaces also. For tritium capture systems, the use of getters is recommended, and high-temperature hydride forming materials such as Ti, Zr, and Y are suggested. Tritium may also be converted to HTO in order to capture it on molecular sieves or getter materials. Counter-flow of hydrogen may reduce the flux of tritium through heat exchangers. Recommendations for research and development work are provided.« less

EFFECTS OF TRITIUM GAS EXPOSURE ON EPDM ELASTOMER

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

Clark, E.

2009-12-11

Samples of four formulations of ethylene-propylene diene monomer (EPDM) elastomer were exposed to initially pure tritium gas at one atmosphere and ambient temperature for various times up to about 420 days in closed containers. Two formulations were carbon-black-filled commercial formulations, and two were the equivalent formulations without filler synthesized for this work. Tritium effects on the samples were characterized by measuring the sample volume, mass, flexibility, and dynamic mechanical properties and by noting changes in appearance. The glass transition temperature was determined by analysis of the dynamic mechanical properties. The glass transition temperature increased significantly with tritium exposure, and themore » unfilled formulations ceased to behave as elastomers after the longest tritium exposure. The filled formulations were more resistant to tritium exposure. Tritium exposure made all samples significantly stiffer and therefore much less able to form a reliable seal when employed as O-rings. No consistent change of volume or density was observed; there was a systematic lowering of sample mass with tritium exposure. In addition, the significant radiolytic production of gas, mainly protium (H{sub 2}) and HT, by the samples when exposed to tritium was characterized by measuring total pressure in the container at the end of each exposure and by mass spectroscopy of a gas sample at the end of each exposure. The total pressure in the containers more than doubled after {approx}420 days tritium exposure.« less

Tritium environmental transport studies at TFTR

NASA Astrophysics Data System (ADS)

Ritter, P. D.; Dolan, T. J.; Longhurst, G. R.

1993-06-01

Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a week after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER).

TRITIUM EFFECTS ON WELDMENT FRACTURE TOUGHNESS

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

Morgan, M; Michael Tosten, M; Scott West, S

2006-07-17

The effects of tritium on the fracture toughness properties of Type 304L stainless steel and its weldments were measured. Fracture toughness data are needed for assessing tritium reservoir structural integrity. This report provides data from J-Integral fracture toughness tests on unexposed and tritium-exposed weldments. The effect of tritium on weldment toughness has not been measured until now. The data include tests on tritium-exposed weldments after aging for up to three years to measure the effect of increasing decay helium concentration on toughness. The results indicate that Type 304L stainless steel weldments have high fracture toughness and are resistant to tritiummore » aging effects on toughness. For unexposed alloys, weldment fracture toughness was higher than base metal toughness. Tritium-exposed-and-aged base metals and weldments had lower toughness values than unexposed ones but still retained good toughness properties. In both base metals and weldments there was an initial reduction in fracture toughness after tritium exposure but little change in fracture toughness values with increasing helium content in the range tested. Fracture modes occurred by the dimpled rupture process in unexposed and tritium-exposed steels and welds. This corroborates further the resistance of Type 304L steel to tritium embrittlement. This report fulfills the requirements for the FY06 Level 3 milestone, TSR15.3 ''Issue summary report for tritium reservoir material aging studies'' for the Enhanced Surveillance Campaign (ESC). The milestone was in support of ESC L2-1866 Milestone-''Complete an annual Enhanced Surveillance stockpile aging assessment report to support the annual assessment process''.« less

An engine awaits processing in the new engine shop at KSC

NASA Technical Reports Server (NTRS)

1998-01-01

A new Block 2A engine awaits processing in the low bay of the Space Shuttle Main Engine Processing Facility (SSMEPF). Officially opened on July 6, the new facility replaces the Shuttle Main Engine Shop. The SSMEPF is an addition to the existing Orbiter Processing Facility Bay 3. The engine is scheduled to fly on the Space Shuttle Endeavour during the STS-88 mission in December 1998.

Tritium release during nuclear power operation in China.

PubMed

Yang, D J; Chen, X Q; Li, B

2012-06-01

Overviews were evaluated of tritium releases and related doses to the public from airborne and liquid effluents from nuclear power plants on the mainland of China before 2009. The differences between tritium releases from various nuclear power plants were also evaluated. The tritium releases are mainly from liquid pathways for pressurised water reactors, but tritium releases between airborne and liquid effluents are comparable for heavy water reactors. The airborne release from a heavy water reactor is obviously higher than that from a pressurised water reactor.

Method for nondestructive fuel assay of laser fusion targets

DOEpatents

Farnum, Eugene H.; Fries, R. Jay

1976-01-01

A method for nondestructively determining the deuterium and tritium content of laser fusion targets by counting the x rays produced by the interaction of tritium beta particles with the walls of the microballoons used to contain the deuterium and tritium gas mixture under high pressure. The x rays provide a direct measure of the tritium content and a means for calculating the deuterium content using the initial known D-T ratio and the known deuterium and tritium diffusion rates.

Drum bubbler tritium processing system

DOEpatents

Rule, Keith; Gettelfinger, Geoff; Kivler, Paul

1999-01-01

A method of separating tritium oxide from a gas stream containing tritium oxide. The gas stream containing tritium oxide is fed into a container of water having a head space above the water. Bubbling the gas stream containing tritium oxide through the container of water and removing gas from the container head space above the water. Thereafter, the gas from the head space is dried to remove water vapor from the gas, and the water vapor is recycled to the container of water.

Small engine components test facility compressor testing cell at NASA Lewis Research Center

NASA Technical Reports Server (NTRS)

Brokopp, Richard A.; Gronski, Robert S.

1992-01-01

LeRC has designed and constructed a new test facility. This facility, called the Small Engine Components Facility (SECTF) is used to test gas turbines and compressors at conditions similar to actual engine conditions. The SECTF is comprised of a compressor testing cell and a turbine testing cell. Only the compressor testing cell is described. The capability of the facility, the overall facility design, the instrumentation used in the facility, and the data acquisition system are discussed in detail.

Comprehensive sets of 124Xe(n ,γ )125Xe and 124Xe(n ,2 n )123Xe cross-section data for assessment of inertial-confinement deuterium-tritium fusion plasma

NASA Astrophysics Data System (ADS)

Bhike, Megha; Fallin, B.; Gooden, M. E.; Ludin, N.; Tornow, W.

2015-01-01

Measurements of the neutron radiative-capture cross section of 124Xe have been performed for the first time for neutron energies above 100 keV. In addition, data for the 124Xe(n ,2 n )123Xe reaction cross section have been obtained from threshold to 14.8 MeV to cover the entire energy range of interest, while previous data existed only at around 14 MeV. The results of these measurements provide the basis for an alternative and sensitive diagnostic tool for investigating properties of the inertial confinement fusion plasma in deuterium-tritium (DT) capsules at the National Ignition Facility located at Lawrence Livermore National Laboratory. Here, areal density ρ R (density × radius) of the fuel, burn asymmetry, and fuel-ablator mix are of special interest. The 124Xe(n ,γ )125Xe reaction probes the down-scattered neutrons, while the 124Xe(n ,2 n )123Xe reaction provides a measure of the 14 MeV direct neutrons.

First Measurements of Deuterium-Tritium and Deuterium-Deuterium Fusion Reaction Yields in Ignition-Scalable Direct-Drive Implosions.

PubMed

Forrest, C J; Radha, P B; Knauer, J P; Glebov, V Yu; Goncharov, V N; Regan, S P; Rosenberg, M J; Sangster, T C; Shmayda, W T; Stoeckl, C; Gatu Johnson, M

2017-03-03

The deuterium-tritium (D-T) and deuterium-deuterium neutron yield ratio in cryogenic inertial confinement fusion (ICF) experiments is used to examine multifluid effects, traditionally not included in ICF modeling. This ratio has been measured for ignition-scalable direct-drive cryogenic DT implosions at the Omega Laser Facility [T. R. Boehly et al., Opt. Commun. 133, 495 (1997)OPCOB80030-401810.1016/S0030-4018(96)00325-2] using a high-dynamic-range neutron time-of-flight spectrometer. The experimentally inferred yield ratio is consistent with both the calculated values of the nuclear reaction rates and the measured preshot target-fuel composition. These observations indicate that the physical mechanisms that have been proposed to alter the fuel composition, such as species separation of the hydrogen isotopes [D. T. Casey et al., Phys. Rev. Lett. 108, 075002 (2012)PRLTAO0031-900710.1103/PhysRevLett.108.075002], are not significant during the period of peak neutron production in ignition-scalable cryogenic direct-drive DT implosions.

Hydrogeology of a low-level radioactive-waste disposal site near Sheffield, Illinois

USGS Publications Warehouse

Foster, J.B.; Erickson, J.R.; Healy, R.W.

1984-01-01

The Sheffield low-level radioactive-waste facility is located on 20 acres of rolling terrain 3 miles southwest of Sheffield, Illinois. The shallow hydrogeologic system is composed of glacial sediments. Pennsylvania shale and mudstone bedrock isolate the regional aquifers below from the hydrogeologic system in the overlying glacial deposits. Pebbly sand underlies 67 percent of the site. Two ground-water flow paths were identified. The primary path conveys ground water from the site to the east through the pebbly-sand unit; a secondary path conveys ground water to the south and east through less permeable material. The pebbly-sand unit provides an underdrain that eliminates the risk of water rising into the trenches. Digital computer model results indicate that the pebbly-sand unit controls ground-water movement. Tritium found migrating in ground water in the southeast corner of the site travels approximately 25 feet per year. A group of water samples from wells which contained the highest tritium concentrations had specific conductivities, alkalinities, hardness, and chloride, sulfate, calcium, and magnesium contents higher than normal for local shallow ground water. (USGS)

Results from the MARBLE Campaign on the National Ignition Facility: Implosion of Foam-Filled Capsules for Studying Thermonuclear Burn in the Presence of Heterogeneous Mix

NASA Astrophysics Data System (ADS)

Murphy, T. J.; Douglas, M. R.; Cardenas, T.; Devolder, B. G.; Fincke, J. R.; Gunderson, M. A.; Haines, B. M.; Hamilton, C. E.; Kim, Y. H.; Lee, M. N.; Oertel, J. A.; Olson, R. E.; Randolph, R. B.; Shah, R. C.; Smidt, J. M.

2016-10-01

The MARBLE campaign on NIF investigates the effect of heterogeneous mix on thermonuclear burn for comparison to a probability distribution function (PDF) burn model. MARBLE utilizes plastic capsules filled with deuterated plastic foam and tritium gas. The ratio of DT to DD neutron yield is indicative of the degree to which the foam and the gas atomically mix. Platform development experiments have been performed to understand the behavior of the foam and of the gas separately using two types of capsule. The first uses partially deuterated foam and hydrogen gas fill to understand the burn in the foam. The second uses undeuterated foam and deuterium gas fill to understand the dynamics of the gas. Experiments using deuterated foam and tritium gas are planned. Results of these experiments, and the implications for our understanding of thermonuclear burn in heterogeneously mixed separated reactant experiments will be discussed. This work is supported by US DOE/NNSA, performed at LANL, operated by LANS LLC under contract DE-AC52-06NA25396.

Oxidative Tritium Decontamination System

DOEpatents

Gentile, Charles A. , Guttadora, Gregory L. , Parker, John J.

2006-02-07

The Oxidative Tritium Decontamination System, OTDS, provides a method and apparatus for reduction of tritium surface contamination on various items. The OTDS employs ozone gas as oxidizing agent to convert elemental tritium to tritium oxide. Tritium oxide vapor and excess ozone gas is purged from the OTDS, for discharge to atmosphere or transport to further process. An effluent stream is subjected to a catalytic process for the decomposition of excess ozone to diatomic oxygen. One of two configurations of the OTDS is employed: dynamic apparatus equipped with agitation mechanism and large volumetric capacity for decontamination of light items, or static apparatus equipped with pressurization and evacuation capability for decontamination of heavier, delicate, and/or valuable items.

Effect of Tritium on Cracking Threshold in 7075 Aluminum

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

Duncan, A.; Morgan, M.

The effect of long-term exposure to tritium gas on the cracking threshold (K TH) of 7075 Aluminum Alloy was investigated. The alloy is the material of construction for a cell used to contain tritium in an accelerator at Jefferson Laboratory designed for inelastic scattering experiments on nucleons. The primary safety concerns for the Jefferson Laboratory tritium cell is a tritium leak due to mechanical failure of windows from hydrogen isotope embrittlement, radiation damage, or loss of target integrity from accidental excessive beam heating due to failure of the raster or grossly mis-steered beam. Experiments were conducted to investigate the potentialmore » for embrittlement of the 7075 Aluminum alloy from tritium gas.« less

Silicon Carbide as a tritium permeation barrier in tungsten plasma-facing components

NASA Astrophysics Data System (ADS)

Wright, G. M.; Durrett, M. G.; Hoover, K. W.; Kesler, L. A.; Whyte, D. G.

2015-03-01

The control of tritium inventory is of great importance in future fusion reactors, not only from a safety standpoint but also to maximize a reactor's efficiency. Due to the high mobility of hydrogenic species in tungsten (W) one concern is the loss of tritium from the system via permeation through the tungsten plasma-facing components (PFC). This can lead to loss of tritium through the cooling channels of the wall thereby mandating tritium monitoring and recovery methods for the cooling system of the first wall. The permeated tritium is then out of the fuel cycle and cannot contribute to energy production until it is recovered and recycled into the system.

In-pile test of Li 2TiO 3 pebble bed with neutron pulse operation

NASA Astrophysics Data System (ADS)

Tsuchiya, K.; Nakamichi, M.; Kikukawa, A.; Nagao, Y.; Enoeda, M.; Osaki, T.; Ioki, K.; Kawamura, H.

2002-12-01

Lithium titanate (Li 2TiO 3) is one of the candidate materials as tritium breeder in the breeding blanket of fusion reactors, and it is necessary to show the tritium release behavior of Li 2TiO 3 pebble beds. Therefore, a blanket in-pile mockup was developed and in situ tritium release experiments with the Li 2TiO 3 pebble bed were carried out in the Japan Materials Testing Reactor. In this study, the relationship between tritium release behavior from Li 2TiO 3 pebble beds and effects of various parameters were evaluated. The ( R/ G) ratio of tritium release ( R) and tritium generation ( G) was saturated when the temperature at the outside edge of the Li 2TiO 3 pebble bed became 300 °C. The tritium release amount increased cycle by cycle and saturated after about 20 pulse operations.

The Stark Effect on the Wave Function of Tritium in Relativistic Condition

NASA Astrophysics Data System (ADS)

Supriadi, B.; Prastowo, S. H. B.; Bahri, S.; Ridlo, Z. R.; Prihandono, T.

2018-03-01

Tritium Atom is one of the isotopes of Hydrogen that has two Neutrons in the nucleus and an electron that surrounds the nucleus. The Stark Effect is an effect of a shift or polarization of the atomic spectrum caused by the external electrostatic field. The interaction between the electrons and the external electric field can be reviewed using an approximation method of perturbation theory. The perturbation theory used is a time Independent non-degenerate perturbation and reviewed to second order to obtain correction of Tritium Atomic wave function. The condition that used in the system is a relativistic condition by reviewing the movement of electrons within the Atom. The effects of relativity also affect the correction of the wave function of Atom Tritium in the ground state. Tritium is radioactive material that is still relatively safe, and one of the applications of Tritium Atom is on the battery of betavoltaics (Nano Tritium Battery).

Analysis of air mass trajectories to explain observed variability of tritium in precipitation at the Southern Sierra Critical Zone Observatory, California, USA

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

Visser, Ate; Thaw, Melissa; Esser, Brad

Understanding the behavior of tritium, a radioactive isotope of hydrogen, in the environment is important to evaluate the exposure risk of anthropogenic releases, and for its application as a tracer in hydrology and oceanography. To understand and predict the variability of tritium in precipitation, HYSPLIT air mass trajectories were analyzed for 16 aggregate precipitation samples collected over a 2 year period at irregular intervals at a research site located at 2000 m elevation in the southern Sierra Nevada (California, USA). Attributing the variation in tritium to specific source areas confirms the hypothesis that higher latitude or inland sources bring highermore » tritium levels in precipitation than precipitation originating in the lower latitude Pacific Ocean. In this case, the source of precipitation accounts for 79% of the variation observed in tritium concentrations. In conclusion, air mass trajectory analysis is a promising tool to improve the predictions of tritium in precipitation at unmonitored locations and thoroughly understand the processes controlling transport of tritium in the environment.« less

Dependence of Tritium Release from Stainless Steel on Temperature and Water Vapor

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

Shmayda, W. T.; Sharpe, M.; Boyce, A. M.

The impact of water vapor and temperature on the release of tritium from stainless steel was studied. Degreased stainless steel samples loaded with tritium at room temperature following a 24-h degassing in vacuum at room temperature were subjected to increasing temperatures or humidity. In general, increasing either the sample temperature or the humidity causes an increased quantity of tritium to be removed. Increasing the temperature to 300°C in a dry gas stream results in a significant release of tritium and is therefore an effective means for reducing the tritium inventory in steel. For humid purges at 30°C, a sixfold increasemore » in humidity results in a tenfold increase in the peak outgassing rate. Increasing the humidity from 4 parts per million (ppm) to 1000 ppm when the sample temperature is 100°C causes a significant increase in the tritium outgassing rate. Finally, a simple calculation shows that only 15% of the activity present in the sample was removed in these experiments, suggesting that the surface layer of adsorbed water participates in regulating tritium desorption from the surface.« less

Dependence of Tritium Release from Stainless Steel on Temperature and Water Vapor

DOE PAGES

Shmayda, W. T.; Sharpe, M.; Boyce, A. M.; ...

2015-09-15

The impact of water vapor and temperature on the release of tritium from stainless steel was studied. Degreased stainless steel samples loaded with tritium at room temperature following a 24-h degassing in vacuum at room temperature were subjected to increasing temperatures or humidity. In general, increasing either the sample temperature or the humidity causes an increased quantity of tritium to be removed. Increasing the temperature to 300°C in a dry gas stream results in a significant release of tritium and is therefore an effective means for reducing the tritium inventory in steel. For humid purges at 30°C, a sixfold increasemore » in humidity results in a tenfold increase in the peak outgassing rate. Increasing the humidity from 4 parts per million (ppm) to 1000 ppm when the sample temperature is 100°C causes a significant increase in the tritium outgassing rate. Finally, a simple calculation shows that only 15% of the activity present in the sample was removed in these experiments, suggesting that the surface layer of adsorbed water participates in regulating tritium desorption from the surface.« less

Analysis of air mass trajectories to explain observed variability of tritium in precipitation at the Southern Sierra Critical Zone Observatory, California, USA

DOE PAGES

Visser, Ate; Thaw, Melissa; Esser, Brad

2017-11-20

Understanding the behavior of tritium, a radioactive isotope of hydrogen, in the environment is important to evaluate the exposure risk of anthropogenic releases, and for its application as a tracer in hydrology and oceanography. To understand and predict the variability of tritium in precipitation, HYSPLIT air mass trajectories were analyzed for 16 aggregate precipitation samples collected over a 2 year period at irregular intervals at a research site located at 2000 m elevation in the southern Sierra Nevada (California, USA). Attributing the variation in tritium to specific source areas confirms the hypothesis that higher latitude or inland sources bring highermore » tritium levels in precipitation than precipitation originating in the lower latitude Pacific Ocean. In this case, the source of precipitation accounts for 79% of the variation observed in tritium concentrations. In conclusion, air mass trajectory analysis is a promising tool to improve the predictions of tritium in precipitation at unmonitored locations and thoroughly understand the processes controlling transport of tritium in the environment.« less

Advances in shock timing experiments on the National Ignition Facility

NASA Astrophysics Data System (ADS)

Robey, H. F.; Celliers, P. M.; Moody, J. D.; Sater, J.; Parham, T.; Kozioziemski, B.; Dylla-Spears, R.; Ross, J. S.; LePape, S.; Ralph, J. E.; Hohenberger, M.; Dewald, E. L.; Berzak Hopkins, L.; Kroll, J. J.; Yoxall, B. E.; Hamza, A. V.; Boehly, T. R.; Nikroo, A.; Landen, O. L.; Edwards, M. J.

2016-03-01

Recent advances in shock timing experiments and analysis techniques now enable shock measurements to be performed in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility (NIF). Previous measurements of shock timing in inertial confinement fusion (ICF) implosions were performed in surrogate targets, where the solid DT ice shell and central DT gas were replaced with a continuous liquid deuterium (D2) fill. These previous experiments pose two surrogacy issues: a material surrogacy due to the difference of species (D2 vs. DT) and densities of the materials used and a geometric surrogacy due to presence of an additional interface (ice/gas) previously absent in the liquid-filled targets. This report presents experimental data and a new analysis method for validating the assumptions underlying this surrogate technique.

Time-of-Flight Measurements of Neutron Yields from Implosions at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Caggaino, Joseph

2014-10-01

Three 20-m time-of-flight detectors measure neutron spectra from implosions of deuterium-tritium targets at the National Ignition Facility. Two prominent peaks appear in the spectra from the T(d, n) and D(d, n) reactions. The ratio of yields extracted from the peaks depend on the DT and DD reaction rates and attenuation from the compressed DT fuel, which makes the ratio a diagnostic of the hotspot thermodynamics and fuel areal density. The measured peak widths provide additional constraints on reactant temperature. Recent measurements from a high-yield campaign will be presented and compared to radiation-hydrodynamic simulations of similar implosions. This research is supported by the Department of Energy National Nuclear Security Administration under Contract DE-NA0001944.

Spatial variations of tritium concentrations in groundwater collected in the southern coastal region of Fukushima, Japan, after the nuclear accident.

PubMed

Kashiwaya, Koki; Muto, Yuta; Kubo, Taiki; Ikawa, Reo; Nakaya, Shinji; Koike, Katsuaki; Marui, Atsunao

2017-10-03

Spatial variations in tritium concentrations in groundwater were identified in the southern part of the coastal region in Fukushima Prefecture, Japan. Higher tritium concentrations were measured at wells near the Fukushima Daiichi Nuclear Power Station (F1NPS). Mean tritium concentrations in precipitation in the 5 weeks after the F1NPS accident were estimated to be 433 and 139 TU at a distance of 25 and 50 km, respectively, from the F1NPS. The elevations of tritium concentrations in groundwater were calculated using a simple mixing model of the precipitation and groundwater. By assuming that these precipitation was mixed into groundwater with a background tritium concentration in a hypothetical well, concentrations of 13 and 7 TU at distances of 25 and 50 km from the F1NPS, respectively, were obtained. The calculated concentrations are consistent with those measured at the studied wells. Therefore, the spatial variation in tritium concentrations in groundwater was probably caused by precipitation with high tritium concentrations as a result of the F1NPS accident. However, the highest estimated tritium concentrations in precipitation for the study site were much lower than the WHO limits for drinking water, and the concentrations decreased to almost background level at the wells by mixing with groundwater.

Tritium levels in milk in the vicinity of chronic tritium releases.

PubMed

Le Goff, P; Guétat, Ph; Vichot, L; Leconte, N; Badot, P M; Gaucheron, F; Fromm, M

2016-01-01

Tritium is the radioactive isotope of hydrogen. It can be integrated into most biological molecules. Even though its radiotoxicity is weak, the effects of tritium can be increased following concentration in critical compartments of living organisms. For a better understanding of tritium circulation in the environment and to highlight transfer constants between compartments, we studied the tritiation of different agricultural matrices chronically exposed to tritium. Milk is one of the most frequently monitored foodstuffs in the vicinity of points known for chronic release of radionuclides firstly because dairy products find their way into most homes but also because it integrates deposition over large areas at a local scale. It is a food which contains all the main nutrients, especially proteins, carbohydrates and lipids. We thus studied the tritium levels of milk in chronic exposure conditions by comparing the tritiation of the main hydrogenated components of milk, first, component by component, then, sample by sample. Significant correlations were found between the specific activities of drinking water and free water of milk as well as between the tritium levels of cattle feed dry matter and of the main organic components of milk. Our findings stress the importance of the metabolism on the distribution of tritium in the different compartments. Overall, dilution of hydrogen in the environmental compartments was found to play an important role dimming possible isotopic effects even in a food chain chronically exposed to tritium. Copyright © 2015 Elsevier Ltd. All rights reserved.

Estimating cancer risk in relation to tritium exposure from routine operation of a nuclear-generating station in Pickering, Ontario.

PubMed

Wanigaratne, S; Holowaty, E; Jiang, H; Norwood, T A; Pietrusiak, M A; Brown, P

2013-09-01

Evidence suggests that current levels of tritium emissions from CANDU reactors in Canada are not related to adverse health effects. However, these studies lack tritium-specific dose data and have small numbers of cases. The purpose of our study was to determine whether tritium emitted from a nuclear-generating station during routine operation is associated with risk of cancer in Pickering, Ontario. A retrospective cohort was formed through linkage of Pickering and north Oshawa residents (1985) to incident cancer cases (1985-2005). We examined all sites combined, leukemia, lung, thyroid and childhood cancers (6-19 years) for males and females as well as female breast cancer. Tritium estimates were based on an atmospheric dispersion model, incorporating characteristics of annual tritium emissions and meteorology. Tritium concentration estimates were assigned to each cohort member based on exact location of residence. Person-years analysis was used to determine whether observed cancer cases were higher than expected. Cox proportional hazards regression was used to determine whether tritium was associated with radiation-sensitive cancers in Pickering. Person-years analysis showed female childhood cancer cases to be significantly higher than expected (standardized incidence ratio [SIR] = 1.99, 95% confidence interval [CI]: 1.08-3.38). The issue of multiple comparisons is the most likely explanation for this finding. Cox models revealed that female lung cancer was significantly higher in Pickering versus north Oshawa (HR = 2.34, 95% CI: 1.23-4.46) and that tritium was not associated with increased risk. The improved methodology used in this study adds to our understanding of cancer risks associated with low-dose tritium exposure. Tritium estimates were not associated with increased risk of radiationsensitive cancers in Pickering.

Institute of Electrical and Electronics Engineers, Nuclear Science Symposium, 18th, and Nuclear Power Systems Symposium, 3rd, San Francisco, Calif., November 3-5, 1971, Proceedings.

NASA Technical Reports Server (NTRS)

1972-01-01

Potential advantages of fusion power reactors are discussed together with the protection of the public from radioactivity produced in nuclear power reactors, and the significance of tritium releases to the environment. Other subjects considered are biomedical instrumentation, radiation damage problems, low level environmental radionuclide analysis systems, nuclear techniques in environmental research, nuclear instrumentation, and space and plasma instrumentation. Individual items are abstracted in this issue.

VAPOR PRESSURE ISOTOPE EFFECTS IN THE MEASUREMENT OF ENVIRONMENTAL TRITIUM SAMPLES.

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

Kuhne, W.

2012-12-03

Standard procedures for the measurement of tritium in water samples often require distillation of an appropriate sample aliquot. This distillation process may result in a fractionation of tritiated water and regular light water due to the vapor pressure isotope effect, introducing either a bias or an additional contribution to the total tritium measurement uncertainty. The magnitude of the vapor pressure isotope effect is characterized as functions of the amount of water distilled from the sample aliquot and the heat settings for the distillation process. The tritium concentration in the distillate is higher than the tritium concentration in the sample earlymore » in the distillation process, it then sharply decreases due to the vapor pressure isotope effect and becomes lower than the tritium concentration in the sample, until the high tritium concentration retained in the boiling flask is evaporated at the end of the process. At that time, the tritium concentration in the distillate again overestimates the sample tritium concentration. The vapor pressure isotope effect is more pronounced the slower the evaporation and distillation process is conducted; a lower heat setting during the evaporation of the sample results in a larger bias in the tritium measurement. The experimental setup used and the fact that the current study allowed for an investigation of the relative change in vapor pressure isotope effect in the course of the distillation process distinguish it from and extend previously published measurements. The separation factor as a quantitative measure of the vapor pressure isotope effect is found to assume values of 1.034 {+-} 0.033, 1.052 {+-} 0.025, and 1.066 {+-} 0.037, depending on the vigor of the boiling process during distillation of the sample. A lower heat setting in the experimental setup, and therefore a less vigorous boiling process, results in a larger value for the separation factor. For a tritium measurement in water samples, this implies that the tritium concentration could be underestimated by 3 - 6%.« less

Application of tritium in precipitation and baseflow in Japan: a case study of groundwater transit times and storage in Hokkaido watersheds

NASA Astrophysics Data System (ADS)

Gusyev, Maksym A.; Morgenstern, Uwe; Stewart, Michael K.; Yamazaki, Yusuke; Kashiwaya, Kazuhisa; Nishihara, Terumasa; Kuribayashi, Daisuke; Sawano, Hisaya; Iwami, Yoichi

2016-07-01

In this study, we demonstrate the application of tritium in precipitation and baseflow to estimate groundwater transit times and storage volumes in Hokkaido, Japan. To establish the long-term history of tritium concentration in Japanese precipitation, we used tritium data from the global network of isotopes in precipitation and from local studies in Japan. The record developed for Tokyo area precipitation was scaled for Hokkaido using tritium values for precipitation based on wine grown at Hokkaido. Then, tritium concentrations measured with high accuracy in river water from Hokkaido, Japan, were compared to this scaled precipitation record and used to estimate groundwater mean transit times (MTTs). A total of 16 river water samples in Hokkaido were collected in June, July, and October 2014 at 12 locations with altitudes between 22 and 831 m above sea level and catchment areas between 14 and 377 km2. Measured tritium concentrations were between 4.07 (± 0.07) TU and 5.29 (± 0.09) TU in June, 5.06 (± 0.09) TU in July, and between 3.75 (± 0.07) TU and 4.85 (± 0.07) TU in October. We utilised TracerLPM (Jurgens et al., 2012) for MTT estimation and introduced a Visual Basic module to automatically simulate tritium concentrations and relative errors for selected ranges of MTTs, exponential-piston ratios, and scaling factors of tritium input. Using the exponential (70 %) piston flow (30 %) model (E70 %PM), we simulated unique MTTs for seven river samples collected in six Hokkaido headwater catchments because their low tritium concentrations were no longer ambiguous. These river catchments are clustered in similar hydrogeological settings of Quaternary lava as well as Tertiary propylite formations near Sapporo city. However, nine river samples from six other catchments produced up to three possible MTT values with E70 % PM due to the interference by the tritium from the atmospheric hydrogen bomb testing 5-6 decades ago. For these catchments, we show that tritium in Japanese groundwater will reach natural levels in a decade, when one tritium measurement will be sufficient to estimate a unique MTT. Using a series of tritium measurements over the next few years with 3-year intervals will enable us to estimate the correct MTT without ambiguity in this period. These unique MTTs will allow estimation of groundwater storage volumes for water resources management during droughts and improvement of numerical model simulations. For example, the groundwater storage ranges between 0.013 and 5.07 km3 with saturated water thickness from 0.2 and 24 m. In summary, we emphasise three important points from our findings: (1) one tritium measurement is already sufficient to estimate MTTs for some Japanese catchments, (2) the hydrogeological settings control the tritium transit times of subsurface groundwater storage during baseflow, and (3) in the future, one tritium measurement will be sufficient to estimate MTTs in most Japanese watersheds.

Estimation of recharge rates to the sand and gravel aquifer using environmental tritium, Nantucket Island, Massachusetts

USGS Publications Warehouse

Knott, Jayne Fifield; Olimpio, Julio C.

1986-01-01

Estimation of the average annual rate of ground-water recharge to sand and gravel aquifers using elevated tritium concentrations in ground water is an alternative to traditional steady-state and water-balance recharge-rate methods. The concept of the tritium tracer method is that the average annual rate of ground-water recharge over a period of time can be calculated from the depth of the peak tritium concentration in the aquifer. Assuming that ground-water flow is vertically downward and that aquifer properties are reasonably homogeneous, and knowing the date of maximum tritium concentration in precipitation and the current depth to the tritium peak from the water table, the average recharge rate can be calculated. The method, which is a direct-measurement technique, was applied at two sites on Nantucket Island, Massachusetts. At site 1, the average annual recharge rate between 1964 and 1983 was 26.1 inches per year, or 68 percent of the average annual precipitation, and the estimated uncertainty is ?15 percent. At site 2, the multilevel water samplers were not constructed deep enough to determine the peak concentration of tritium in ground water. The tritium profile at site 2 resembles the upper part of the tritium profile at site 1 and indicates that the average recharge rate was at least 16 .7 inches per year, or at least 44 percent of the average annual precipitation. The Nantucket tritium recharge rates clearly are higher than rates determined elsewhere in southeastern Massachusetts using the tritium, water-table-fluctuation, and water-balance (Thornthwaite) methods, regardless of the method or the area. Because the recharge potential on Nantucket is so high (runoff is only 2 percent of the total water balance), the tritium recharge rates probably represent the effective upper limit for ground-water recharge in this region. The recharge-rate values used by Guswa and LeBlanc (1985) and LeBlanc (1984) in their ground-water-flow computer models of Cape Cod are 20 to 30 percent lower than this upper limit. The accuracy of the tritium method is dependent on two key factors: the accuracy of the effective-porosity data, and the sampling interval used at the site. For some sites, the need for recharge-rate data may require a determination as statistically accurate as that which can be provided by the tritium method. However, the tritium method is more costly and more time consuming than the other methods because numerous wells must be drilled and installed and because many water samples must be analyzed for tritium, to a very small level of analytical detection. For many sites, a less accurate, less expensive, and faster method of recharge-rate determination might be more satisfactory . The factor that most seriously limits the usefulness of the tritium tracer method is the current depth of the tritium peak. Water with peak concentrations of tritium entered the ground more than 20 years ago, and, according to the Nantucket data, that water now is more than 100 feet below the land surface. This suggests that the tracer method will work only in sand and gravel aquifers that are exceedingly thick by New England standards. Conversely, the results suggest that the method may work in areas where saturated thicknesses are less than 100 feet and the rate of vertical ground-water movement is relatively slow, such as in till and in silt- and clay-rich sand and gravel deposits.

The use of tritium rich capsules with 25-35% deuterium to achieve ignition at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Wilson, D. C.; Spears, B. K.; Hatchett, S. P., Ii; Cerjan, C. J.; Springer, P. T.; Clark, D. S.; Edwards, M. J.; Salmonson, J. D.; Weber, S. V.; Hammel, B. A.; Grim, G. P.; Herrmann, H. W.; Wilke, M. D.

2010-08-01

Diagnostics such as neutron yield, ion temperature, image size and shape, and bang time in capsules with >~25 % deuterium fuel show changes due to burn product heating. The comparison of performance between a THD(2%) and THD(35%) can help predict ignition in a TD(50%) capsule. Surrogacy of THD capsules to TD(50%) is incomplete due to variations in fuel molecular vapour pressures. TD(25-35%) capsules might be preferred to study hot spot heating, but at the risk of increased fuel/ablator mixing.

Hard x-ray (>100 keV) imager to measure hot electron preheat for indirectly driven capsule implosions on the NIF.

PubMed

Döppner, T; Dewald, E L; Divol, L; Thomas, C A; Burns, S; Celliers, P M; Izumi, N; Kline, J L; LaCaille, G; McNaney, J M; Prasad, R R; Robey, H F; Glenzer, S H; Landen, O L

2012-10-01

We have fielded a hard x-ray (>100 keV) imager with high aspect ratio pinholes to measure the spatially resolved bremsstrahlung emission from energetic electrons slowing in a plastic ablator shell during indirectly driven implosions at the National Ignition Facility. These electrons are generated in laser plasma interactions and are a source of preheat to the deuterium-tritium fuel. First measurements show that hot electron preheat does not limit obtaining the fuel areal densities required for ignition and burn.

Evaluation of Ruthenium Capture Methods for Tritium Pretreatment Off-Gas Streams

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

Spencer, Barry B.; Jubin, Robert Thomas; Bruffey, Stephanie H.

2017-07-01

In the reprocessing of used nuclear fuel, radioactive elements are released into various plant off-gas streams. While much research and development has focused on the abatement of the volatile nuclides 3H, 14C, 85Kr, and 129I, the potential release of semivolatile isotopes that could also report to the off-gas streams in a reprocessing facility has been examined. Ruthenium (as 106Ru) has been identified as one of the semivolatile nuclides requiring the greatest degree of abatement prior to discharging the plant off-gas to the environment.

A model function of the global bomb tritium distribution in precipitation, 1960-1986

NASA Astrophysics Data System (ADS)

Doney, Scott C.; Glover, David M.; Jenkins, William J.

1992-04-01

The paper presents a model function for predicting the annual mean concentration of the decay-corrected bomb tritium in precipitation over the time period 1960-1986. The model was developed using the World Meteorological Organization/International Atomic Energy Agency data for tritium precipitation. The resulting tritium function is global in scope and includes both marine and continental data. Estimates were obtained of the seasonal cycle of tritium in precipitation, which may be useful for studying atmospheric transport and oceanic processes, such as convection and subduction that occur on seasonal timescales.

An analysis of workers' tritium concentration in urine samples as a function of time after intake at Korean pressurised heavy water reactors.

PubMed

Kim, Hee Geun; Kong, Tae Young

2012-12-01

In general, internal exposure from tritium at pressurised heavy water reactors (PHWRs) accounts for ∼20-40 % of the total radiation dose. Tritium usually reaches the equilibrium concentration after a few hours inside the body and is then excreted from the body with an effective half-life in the order of 10 d. In this study, tritium metabolism was reviewed using its excretion rate in urine samples of workers at Korean PHWRs. The tritium concentration in workers' urine samples was also measured as a function of time after intake. On the basis of the monitoring results, changes in the tritium concentration inside the body were then analysed.

This photographic copy of an engineering drawing shows floor plans, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

This photographic copy of an engineering drawing shows floor plans, sections and elevations of Building E-86, with details typical of the steel frame and "Transite" building construction at JPL Edwards Facility. California Institute of Technology, Jet Propulsion Laboratory, Facilities Engineering and Construction Office: "Casting & Curing, Building E-86, Floor Plan, Elevations & Section," drawing no. E86/6, 25 February 1977. California Institute of Technology, Jet Propulsion Laboratory, Plant Engineering: engineering drawings of structures at JPL Edwards Facility. Drawings on file at JPL Plant Engineering, Pasadena, California - Jet Propulsion Laboratory Edwards Facility, Casting & Curing Building, Edwards Air Force Base, Boron, Kern County, CA

Modeling tritium transport through a deep unsaturated zone in an arid environment

USGS Publications Warehouse

Mayers, C.J.; Andraski, Brian J.; Cooper, C.A.; Wheatcraft, S.W.; Stonestrom, David A.; Michel, R.L.

2005-01-01

Understanding transport of tritium (3H) in unsaturated zones is critical to evaluating options for waste isolation. Tritium typically is a large component of low-level radioactive waste (LLRW). Studies at the U.S. Geological Survey's Amargosa Desert Research Site (ADRS) in Nevada investigate 3H transport from a closed LLRW facility. Two boreholes are 100 and 160 m from the nearest waste trench and extend to the water table at 110 m. Soil-water vapor samples from the deep boreholes show elevated levels of 3H at all depths. The objectives of this study were to (i) test source thermal and gas-advection mechanisms driving 3H transport and (ii) evaluate model sensitivity to these mechanisms and to selected physical and hydraulic properties including porosity, tortuosity, and anisotropy. A two-dimensional numerical model incorporated a non-isothermal, heterogeneous domain of the unsaturated zone and instantaneous isotopic equilibrium. The TOUGH2 code was used; however, it required modification to account for temperature dependence of both the Henry's law equilibrium constant and isotopic fractionation with respect to tritiated water. Increases in source temperature, pressure, and porosity enhanced 3H migration, but failed to match measured 3H distributions. All anisotropic simulations with a source pressure component resembled, in shape, the upper portion of the 3H distribution of the nearest borehole. Isotopic equilibrium limited migration of 3H, while effects of radioactive decay were negligible. A 500 Pa pressure increase above ambient pressure in conjunction with a high degree of anisotropy (1:100) was necessary for simulated 3H transport to reach the nearest borehole.

Neutron imaging with bubble chambers for inertial confinement fusion

NASA Astrophysics Data System (ADS)

Ghilea, Marian C.

One of the main methods to obtain energy from controlled thermonuclear fusion is inertial confinement fusion (ICF), a process where nuclear fusion reactions are initiated by heating and compressing a fuel target, typically in the form of a pellet that contains deuterium and tritium, relying on the inertia of the fuel mass to provide confinement. In inertial confinement fusion experiments, it is important to distinguish failure mechanisms of the imploding capsule and unambiguously diagnose compression and hot spot formation in the fuel. Neutron imaging provides such a technique and bubble chambers are capable of generating higher resolution images than other types of neutron detectors. This thesis explores the use of a liquid bubble chamber to record high yield 14.1 MeV neutrons resulting from deuterium-tritium fusion reactions on ICF experiments. A design tool to deconvolve and reconstruct penumbral and pinhole neutron images was created, using an original ray tracing concept to simulate the neutron images. The design tool proved that misalignment and aperture fabrication errors can significantly decrease the resolution of the reconstructed neutron image. A theoretical model to describe the mechanism of bubble formation was developed. A bubble chamber for neutron imaging with Freon 115 as active medium was designed and implemented for the OMEGA laser system. High neutron yields resulting from deuterium-tritium capsule implosions were recorded. The bubble density was too low for neutron imaging on OMEGA but agreed with the model of bubble formation. The research done in here shows that bubble detectors are a promising technology for the higher neutron yields expected at National Ignition Facility (NIF).

7. Historic aerial photo of rocket engine test facility complex, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

7. Historic aerial photo of rocket engine test facility complex, June 1962. On file at NASA Plumbrook Research Center, Sandusky, Ohio. NASA GRC photo number C-60674. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

Tritium pellet injector for the tokamak fusion test reactor

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Baylor, L. R.; Combs, S. K.; Fisher, P. W.; Foust, C. R.; Milora, S. L.

The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the FY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability.

Tritium behavior pattern in some soil-plant systems in a tropical environment

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

Soma, S.D.; Iyengar, T.S.; Sadarangani, S.H.

1975-01-01

A study of the distribution pattern of tritium in the soil/plant environment gives a valuable ecological information on the natural water balance. The results of such a study for the conditions obtaining in India are given in this paper. Field studies are carried out by injection of tritium into some soil/ plant systems and following the transfer pathways. The method of extraction for tissue-free-water-tritium (TFWT) is based on the vacuum freeze-drying technique while the tissue-bound-tritium (TBT) is estimated by a modified version of the Shoniger method. The determination of residence time of tritium in aqueous and organic phase in amore » number of tropical trees has been carried out both for stem- injection as well as intake from the soil. From the results of this study the tree biomass and transpiration rates have been determined. The tritium profile over time, for an acute exposure in certain trees such as Morinda Tinetoria, Achras Sapota etc. shows significantly different patterns compared to the normal pattern shown by Mangifera Indica, Terminalia Catappa, Ficus Glomerata etc. The period of investigation in each case varied from 400 to 1000 h. In most of the cases, the TBT fractions were very low compared to TFWT fractions in the initial stages. The tritium behaviour in the tree reflects significant characteristics of the tritium behaviour in the soil system. The authors have found that the leaf sampling can be used as an indicator of total environmental tritium behaviour. (auth)« less

Tritium recapture behavior at a nuclear power reactor due to airborne releases.

PubMed

Harris, Jason T; Miller, David W; Foster, Doug W

2008-08-01

This paper describes the initiatives taken by Cook Nuclear Plant to study the on-site behavior of recaptured tritium released in its airborne effluents. Recapture is the process where a released radioactive effluent, in this case tritium, is brought back on-site through some mechanism. Precipitation, shifts in wind direction, or anthropogenic structures that restrict or alter effluent movement can all lead to recapture. The investigation was started after tritium was detected in the north storm drain outfall. Recent inadvertent tritium releases by several other nuclear power plants, many of which entered the groundwater, have led to increased surveillance and scrutiny by regulatory authorities and the general public. To determine the source of tritium in the outfall, an on-site surface water, well water, rainwater and air-conditioning condensate monitoring program was begun. Washout coefficients were also determined to compare with results reported by other nuclear power plants. Program monitoring revealed detectable tritium concentrations in several precipitation sample locations downwind of the two monitored containment building release vents. Tritium was found in higher concentrations in air-conditioning condensate, with a mean value of 528 Bq L(-1) (14,300 pCi L(-1)). The condensate, and to a lesser extent rainwater, were contributing to the tritium found in the north storm drain outfall. Maximum concentration values for each sample type were used to estimate the most conservative dose. A maximum dose of 1.1 x 10(-10) mSv (1.1 x 10(-8) mrem) total body was calculated to determine the health impact of the tritium detected.

Harbors.

DTIC Science & Technology

1981-07-01

CONTRACT OR GRANT NUMBER(e) Naval Facilities Engineering Command 200 Stovall Street r Alexandria, VA 22332 (Code 0453) s. PERFORMING ORGANIZATION NAME...AND ADDRESS 10. PROGRAM ELEMENT. PROJECT. TASK • Naval Facilities Engineering Command AREA & WORK UNIT NUMBERS < 200 Stovall Street Engineering and...Design Alexandria, VA 22332 It. CONTROLLING OFFICE NAME AND ADDRESS 12. REPORT DATE ~ Naval Facilities Engineering Command (Code10432) July 1981 200

Organically bound tritium (OBT) for various plants in the vicinity of a continuous atmospheric tritium release.

PubMed

Vichot, L; Boyer, C; Boissieux, T; Losset, Y; Pierrat, D

2008-10-01

In order to quantify tritium impact on the environmental, we studied vegetation continuously exposed to a tritiated atmosphere. We chose lichens as bio-indicators, trees for determination of past tritium releases of the Valduc Centre, and lettuce as edible vegetables for dose calculation regarding neighbourhood. The Pasquill and Doury models from the literature were tested to estimate tritium concentration in the air around vegetable for distance from the release point less than 500 m. The results in tree rings show that organically bound tritium (OBT) concentration was strongly correlated with tritium releases. Using the GASCON model, the modelled variation of OBT concentration with distance was correlated with the measurements. Although lichens are recognized as bio-indicators, our experiments show that they were not convenient for environmental surveys because their age is not definitive. Thus, tritium integration time cannot be precisely determined. Furthermore, their biological metabolism is not well known and tritium concentration appears to be largely dependent on species. An average conversion rate of HTO to OBT was determined for lettuce of about 0.20-0.24% h(-1). Nevertheless, even if it is equivalent to values already published in the literature for other vegetation, we have shown that this conversion rate, established by weekly samples, varies by a factor of 10 during the different stages of lettuce development, and that its variation is linked to the biomass derivative.

Multiscale integral analysis of a HT leakage in a fusion nuclear power plant

NASA Astrophysics Data System (ADS)

Velarde, M.; Fradera, J.; Perlado, J. M.; Zamora, I.; Martínez-Saban, E.; Colomer, C.; Briani, P.

2016-05-01

The present work presents an example of the application of an integral methodology based on a multiscale analysis that covers the whole tritium cycle within a nuclear fusion power plant, from a micro scale, analyzing key components where tritium is leaked through permeation, to a macro scale, considering its atmospheric transport. A leakage from the Nuclear Power Plants, (NPP) primary to the secondary side of a heat exchanger (HEX) is considered for the present example. Both primary and secondary loop coolants are assumed to be He. Leakage is placed inside the HEX, leaking tritium in elementary tritium (HT) form to the secondary loop where it permeates through the piping structural material to the exterior. The Heating Ventilation and Air Conditioning (HVAC) system removes the leaked tritium towards the NPP exhaust. The HEX is modelled with system codes and coupled to Computational Fluid Dynamic (CFD) to account for tritium dispersion inside the nuclear power plants buildings and in site environment. Finally, tritium dispersion is calculated with an atmospheric transport code and a dosimetry analysis is carried out. Results show how the implemented methodology is capable of assessing the impact of tritium from the microscale to the atmospheric scale including the dosimetric aspect.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

NASA Astrophysics Data System (ADS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Safe, Cost Effective Management of Inactive Facilities at the Savannah River Site

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

Austin, W. E.; Yannitell, D. M.; Freeman, D. W.

The Savannah River Site is part of the U.S. Department of Energy complex. It was constructed during the early 1950s to produce basic materials (such as plutonium-239 and tritium) used in the production of nuclear weapons. The 310-square-mile site is located in South Carolina, about 12 miles south of Aiken, South Carolina, and about 15 miles southeast of Augusta, Georgia. Savannah River Site (SRS) has approximately 200 facilities identified as inactive. These facilities range in size and complexity from large nuclear reactors to small storage buildings. These facilities are located throughout the site including three reactor areas, the heavy watermore » plant area, the manufacturing area, and other research and support areas. Unlike DOE Closure Sites such as Hanford and Rocky Flats, SRS is a Project Completion Site with continuing missions. As facilities complete their defined mission, they are shutdown and transferred from operations to the facility disposition program. At the SRS, Facilities Decontamination and Decommissioning (FDD) personnel manage the disposition phase of a inactive facility's life cycle in a manner that minimizes life cycle cost without compromising (1) the health or safety of workers and the public or (2) the quality of the environment. The disposition phase begins upon completion of operations shutdown and extends through establishing the final end-state. FDD has developed innovative programs to manage their responsibilities within a constrained budget.« less

The stark effect on the spectrum energy of tritium in first excited state with relativistic condition

NASA Astrophysics Data System (ADS)

Prastowo, S. H. B.; Supriadi, B.; Bahri, S.; Ridlo, Z. R.

2018-04-01

This research discussed about the correction of Stark Effect on Tritium atoms in the first excited state with relativistic conditions. The approach used to solve this Stark Effect correction was the perturbation theory which was from time independent degenerate perturbation theory to second-order correction. The Stark Effect on the excited state made the spectrum energy polarization of Tritium which was included in the isotope of hydrogen with an electron moving around the nucleus with high velocity. Hence, the relativistic correction affected the spectrum energy shift. Tritium was a radioactive material having half-time 12,3 years and relatively safe. The Tritium application was a material for the manufacture of nuclear battery. The most effective external electric field that should give to Tritium was 108 V/mith the total correction energy that was 0,97398557 × 10-21 Joule. Therefore, its effect reduced the binding energy between electron and nucleus, and increased the power of Tritium Betavoltaics Battery.

Tritium processing for the European test blanket systems: current status of the design and development strategy

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

Ricapito, I.; Calderoni, P.; Poitevin, Y.

2015-03-15

Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for themore » design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)« less

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

Wertsching, Alan Kevin; Trantor, Troy Joseph; Ebner, Matthias Anthony

A method and device for producing secure, high-density tritium bonded with carbon. A substrate comprising carbon is provided. A precursor is intercalated between carbon in the substrate. The precursor intercalated in the substrate is irradiated until at least a portion of the precursor, preferably a majority of the precursor, is transmutated into tritium and bonds with carbon of the substrate forming bonded tritium. The resulting bonded tritium, tritium bonded with carbon, produces electrons via beta decay. The substrate is preferably a substrate from the list of substrates consisting of highly-ordered pyrolytic graphite, carbon fibers, carbon nanotunes, buckministerfullerenes, and combinations thereof.more » The precursor is preferably boron-10, more preferably lithium-6. Preferably, thermal neutrons are used to irradiate the precursor. The resulting bonded tritium is preferably used to generate electricity either directly or indirectly.« less

9. Historic aerial photo of rocket engine test facility complex, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

9. Historic aerial photo of rocket engine test facility complex, June 11, 1965. On file at NASA Plumbrook Research Center, Sandusky, Ohio. NASA GRC photo number C-65-1270. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

10. Historic photo of rendering of rocket engine test facility ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Historic photo of rendering of rocket engine test facility complex, April 28, 1964. On file at NASA Plumbrook Research Center, Sandusky, Ohio. NASA GRC photo number C-69472. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

8. Historic aerial photo of rocket engine test facility complex, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

8. Historic aerial photo of rocket engine test facility complex, June 11, 1965. On file at NASA Plumbrook Research Center, Sandusky, Ohio. NASA GRC photo number C-65-1271. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

Tritium safety study using Caisson Assembly (CATS) at TPL/JAEA

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

Hayashi, T.; Kobayashi, K.; Iwai, Y.

Tritium confinement is required as the most important safety Junction for a fusion reactor. In order to demonstrate the confinement performance experimentally, an unique equipment, called CATS: Caisson Assembly for Tritium Safety study, was installed in Tritium Process Laboratory of Japan Atomic Energy Agency and operated for about 10 years. Tritium confinement and migration data in CATS have been accumulated and dynamic simulation code was accumulated using these data. Contamination and decontamination behavior on various materials and new safety equipment functions have been investigated under collaborations with a lot of laboratories and universities. (authors)

The Reactions of Recoil Tritium with Anilides (in Japanese)

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

OKAMOTO, Jiro; TSUCHIHASHI, Gen-ichi

1961-01-01

The distribution of tritium in some tritiated anilides (acetanilide, propionanilide, n-butylanilide, iso-butylanilide) which were produced by irradiation of mixed powder of anilides and lithium carbonate, were investigated. The tritium contents of the ortho-, meta-, and para-positions in the anilides were obtained by the activity measurement of some derivatives. The reactivities of ortho-position for tritium decreases in the order acetanilide, propionanilide, nbutylanilide, iso-butylanilide, perhaps because of steric interference of alkyl groups. The contents of incorporated tritium in alkyl groups were 13.2%, 31.7%, 31.1%, and 45.4%, for acetanilide, propionanilide, n-butylanilide, iso- butylanilide, respectively.

Facilities | Argonne National Laboratory

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Skip to main content Argonne National Laboratory Toggle Navigation Toggle Search Research Facilities Advanced Powertrain Research Facility Center for Transportation Research Distributed Energy Research Center Engine Research Facility Heat Transfer Laboratory Materials Engineering Research Facility

Next generation gamma-ray Cherenkov detectors for the National Ignition Facility.

PubMed

Herrmann, H W; Kim, Y H; McEvoy, A M; Zylstra, A B; Young, C S; Lopez, F E; Griego, J R; Fatherley, V E; Oertel, J A; Stoeffl, W; Khater, H; Hernandez, J E; Carpenter, A; Rubery, M S; Horsfield, C J; Gales, S; Leatherland, A; Hilsabeck, T; Kilkenny, J D; Malone, R M; Hares, J D; Milnes, J; Shmayda, W T; Stoeckl, C; Batha, S H

2016-11-01

The newest generation of Gas Cherenkov Detector (GCD-3) employed in Inertial Confinement Fusion experiments at the Omega Laser Facility has provided improved performance over previous generations. Comparison of reaction histories measured using two different deuterium-tritium fusion products, namely gamma rays using GCD and neutrons using Neutron Temporal Diagnostic (NTD), have provided added credibility to both techniques. GCD-3 is now being brought to the National Ignition Facility (NIF) to supplement the existing Gamma Reaction History (GRH-6m) located 6 m from target chamber center (TCC). Initially it will be located in a reentrant well located 3.9 m from TCC. Data from GCD-3 will inform the design of a heavily-shielded "Super" GCD to be located as close as 20 cm from TCC. It will also provide a test-bed for faster optical detectors, potentially lowering the temporal resolution from the current ∼100 ps state-of-the-art photomultiplier tubes (PMT) to ∼10 ps Pulse Dilation PMT technology currently under development.

Hydrologic conditions at the Idaho National Engineering Laboratory, Idaho, emphasis; 1974-1978

USGS Publications Warehouse

Barraclough, Jack T.; Lewis, Barney D.; Jensen, Rodger G.

1981-01-01

Aqueous chemical and radioactive wastes have been discharged to shallow ponds and to shallow or deep wells on the Idaho National Engineering Laboratory (INEL) since 1952 and has affected the quality of the ground water in the underlying Snake River Plain aquifer. Ongoing studies conducted from 1974 through 1978 have shown the perpetuation of a perched ground-water zone in the basalt underlying the waste disposal ponds at the INEL 's Test Reactor Area and of several waste plumes in the regional aquifer created by deep well disposal at the Idaho Chemical Processing Plant (ICPP). The perched zone contains tritium, chromium-51, cobalt-60, strontium-90, and several nonradioactive chemicals. Tritium has formed the largest waste plume south of the ICPP, and accounts for 95 percent of the total radioacticity disposed of through the ICPP disposal well. Waste plumes with similar configurations and flowpaths contain sodium, chloride, and nitrate. Strontium-90, iodine-129, and cesium-137 are also discharged through the well but they are sorbed from solution as they move through the aquifer or are discharged in very small quantities. Strontium-90 and iodine-129 have formed small waste plumes and cesium-137 is not detectable in ground-water samples. Radionuclide plume size and concentrations therein are controlled by aquifer flow conditions, the quantity discharged, radioactive decay, sorption, dilution by dispersion, and perhaps other chemical reactions. Chemical wastes are subject to the same processes except for radioactive decay. (USGS)

Effectiveness of passivation techniques on hydrogen desorption in a tritium environment

NASA Astrophysics Data System (ADS)

Woodall, Steven Michael

2009-11-01

Tritium is a radioactive isotope of hydrogen. It is used as a fuel in fusion reactors, a booster material in nuclear weapons and as a light source in commercial applications. When tritium is used in fusion reactors, and especially when used in the manufacture of nuclear weapons, purity is critical. For U.S. Department of Energy use, tritium is recycled by Savannah River Site in South Carolina and is processed to a minimum purity of 99.5%. For use elsewhere in the country, it must be shipped and stored, while maintaining the highest purity possible. As an isotope of hydrogen it exchanges easily with the most common isotope of hydrogen, protium. Stainless steel bottles are used to transport and store tritium. Protium, present in air, becomes associated in and on the surface of stainless steel during and after the manufacture of the steel. When filled, the tritium within the bottle exchanges with the protium in and on the surface of the stainless steel, slowly contaminating the pure tritium with protium. The stainless steel is therefore passivated to minimize the protium outgrowth of the bottles into the pure tritium. This research is to determine how effective different passivation techniques are in minimizing the contamination of tritium with protium. Additionally, this research will attempt to determine a relationship between surface chemistry of passivated steels and protium contamination of tritium. The conclusions of this research found that passivated bottles by two companies which routinely provide passivated materials to the US Department of Energy provide low levels of protium outgrowth into pure tritium. A bottle passivated with a material to prevent excessive corrosion in a highly corrosive environment, and a clean and polished bottle provided outgrowth rates roughly twice those of the passivated bottles above. Beyond generally high levels of chromium, oxygen, iron and nickel in the passivated bottles, there did not appear to be a strong correlation between surface chemistry in the surface of the bottles and protium outgrowth rates.

2017 Accomplishments – Tritium Aging Studies on Stainless Steel Weldments and Heat-Affected Zones

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

Morgan, Michael J.; Hitchcock, Dale; Krentz, Tim

In this study, the combined effects tritium and decay helium in forged and welded Types 304L and 21-6-9 stainless steels were studied. To measure these effects, fracture mechanic specimens were thermally precharged with tritium and aged for approximately 17 years to build in decay helium from tritium decay prior to testing. The results are compared to earlier measurements on the same alloys and weldments (4-5, 8-9). In support of Enhanced Surveillance, “Tritium Effects on Materials”, the fracture toughness properties of long-aged tritium-charged stainless-steel base metals and weldments were measured and compared to earlier measurements. The fracture-toughness data were measured bymore » thermally precharging as-forged and as-welded specimens with tritium gas at 34.5 MPa and 350°C and aging for approximately 17 years to build-in decay helium prior to testing. These data result from the longest aged specimens ever tested in the history of the tritium effects programs at Savannah River and the fracture toughness values measured were the lowest ever recorded for tritium-exposed stainless steel. For Type 21-6-9 stainless steel, fracture toughness values were reduced to less than 2-4% of the as-forged values to 41 lbs / in specimens that contained more than 1300 appm helium from tritium decay. The fracture toughness properties of long-aged weldments were also measured. The fracture toughness reductions were not as severe because the specimens did not retain as much tritium from the charging and aging as did the base metals. For Type 304L weldments, the specimens in this study contained approximately 600 appm helium and their fracture toughness values averaged 750 lbs / in. The results for other steels and weldments are reported and additional tests will be conducted during FY18.« less

Tritium in the western Mediterranean Sea during 1981 Phycemed cruise

NASA Astrophysics Data System (ADS)

Andrie, Chantal; Merlivat, Liliane

1988-02-01

We report on simultaneous hydrological and tritium data taken in the western Mediterranean Sea during April 1981 and which implement our knowledge of the spatial and temporal variability of the convection process occurring in the Northern Basin (Gulf of Lion, Ligurian Sea). The renewal time of the deep waters in the Medoc area is calculated to be 11 ± 2 years using a box-model assymption. An important local phenomenon of "cascading" off the Ebro River near the Spanish coast is, noticeable by the use of tritium data. In the Sardinia Straits area tritium data indicate very active mixing between 100 and 500 m depth. The tritium subsurface maxima in Sardinia Straits suggests the influence of not only the Levantine Intermediate Water (LIW) but also an important shallower component. In waters deeper than 500m, an active mixing occurs between the deep water and the LIW via an intermediate water mass from the Tyrrhenian Sea by "salt-fingering". Assuming a two end-member mixing. We determine the deep tritium content in the Sardinia Channel to be 1.8 TU. For comparison, the deep tritium content of the Northern Basin is equal to 1.3 TU. Tritium data relative to the Alboran Sea show that a layer of high tritium content persists all along its path from Sardifia to Gibraltar on a density surface shallower than the intermediate water. The homogeneity of the deep tritium concentrations between 1200 m depth and the bottom corroborate the upward "pumping" and westward circulation of deep waters along the continental slope of the North African Shelf. From the data measured in the Sardinia Straits and in the Alboran Sea, and upper limit of the deep advection rate of the order of 0.5 cm s-1 is estimated.

Lithographic Printing Via Two-Photon Polymerization of Engineered Foams

DOE PAGES

Herman, Matthew J.; Peterson, Dominic; Henderson, Kevin; ...

2017-11-29

Understanding deuterium-tritium mix in capsules is critical to achieving fusion within inertial confined fusion experiments. One method of understanding how the mix of hydrogen fuels can be controlled is by creating various structured deuterated foams and filling the capsule with liquid tritium. Historically, these materials have been a stochastically structured gas-blown foam. Later, to improve the uniformity of this material, pore formers have been used which are then chemically removed, leaving behind a foam of monodisperse voids. However, this technique is still imperfect in that fragments of the pore templating particles may not be completely removed and the void distributionmore » may not be uniform over the size scale of the capsule. Recently, advances in three-dimensional printing suggest that it can be used to create microlattices and capsule walls in one single print. Demonstrated in this paper are proof-of-concept microlattices produced using two-photon polymerization with submicrometer resolution of various structures as well as a microlattice-containing capsule. Finally, with this technology, complete control of the mixing structure is possible, amenable to modeling and easily modified for tailored target design.« less

Tritium release from neutron-irradiated Li 2O sintered pellets: porosity dependence

NASA Astrophysics Data System (ADS)

Tanifuji, Takaaki; Yamaki, Daiju; Takahashi, Tadashi; Iwamoto, Akira

2000-12-01

The tritium release behaviour from sintered Li 2O pellets of various densities (71-98.5% theoretical density, T.D.) has been investigated by heating tests at a constant rate. It is shown that the tritium release rate depends on porosity at densities above 87% T.D., while no dependence was observed at densities below 86% T.D. The tritium release process is thought to consist of three stages described as follows: (1) the liberation of tritium trapped at point defects due to their recovery (peak at around 570 K); (2) the advection through interconnected pores via adsorption and desorption on their inner walls and diffusion in the gas phase of interconnected pores (peak at around 620 K); (3) the dissolution and release of tritium trapped in closed pores (peaks at around 700, 830 and 1000 K).

Neutral Beam Injection in the JET Trace Tritium Experiment

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

Surrey, E.; Ciric, D.; Cox, S. J.

Operation of the JET Neutral Beam Injectors with tritium is described. Supplying the tritium feed via the special electrically grounded gas feed compromised the performance of the up-graded high current triode Positive Ion Neutral Injectors (PINI) due to gas starvation of the source and the methods adopted to ameliorate this effect are described. A total of 362 PINI beam pulses were requested, circulating a total of 4.73g tritium, of which 9.3mg was injected into the torus. Safety considerations required a continuous, cumulative total to be maintained of the mass of tritium adsorbed onto the cryo-pumping panel; a daily limit ofmore » 0.5g was adopted for the Trace Tritium Experiment (TTE). A subsequent clean up phase using 115keV deuterium beams completed the isotopic exchange of components in the beamline.« less

Comparison of balance of tritium activity in waste water from nuclear power plants and at selected monitoring sites in the Vltava River, Elbe River and Jihlava (Dyje) River catchments in the Czech Republic.

PubMed

Hanslík, Eduard; Marešová, Diana; Juranová, Eva; Sedlářová, Barbora

2017-12-01

During the routine operation, nuclear power plants discharge waste water containing a certain amount of radioactivity, whose main component is the artificial radionuclide tritium. The amounts of tritium released into the environment are kept within the legal requirements, which minimize the noxious effects of radioactivity, but the activity concentration is well measurable in surface water of the recipient. This study compares amount of tritium activity in waste water from nuclear power plants and the tritium activity detected at selected relevant sites of surface water quality monitoring. The situation is assessed in the catchment of the Vltava and Elbe Rivers, affected by the Temelín Nuclear Power Plant as well as in the Jihlava River catchment (the Danube River catchment respectively), where the waste water of the Dukovany Nuclear Power Plant is discharged. The results show a good agreement of the amount of released tritium stated by the power plant operator and the tritium amount detected in the surface water and highlighted the importance of a robust independent monitoring of tritium discharged from a nuclear power plant which could be carried out by water management authorities. The outputs of independent monitoring allow validating the values reported by a polluter and expand opportunities of using tritium as e.g. tracer. Copyright © 2017 Elsevier Ltd. All rights reserved.

6. Historic photo of rocket engine test facility Building 202 ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

6. Historic photo of rocket engine test facility Building 202 complex in operation at night, September 12, 1957. On file at NASA Plumbrook Research Center, Sandusky, Ohio. NASA GRC photo number C-45924. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

13. Historic drawing of rocket engine test facility layout, including ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

13. Historic drawing of rocket engine test facility layout, including Buildings 202, 205, 206, and 206A, February 3, 1984. NASA GRC drawing number CF-101539. On file at NASA Glenn Research Center. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

Recovery of tritium from tritiated molecules

DOEpatents

Swansiger, W.A.

1984-10-17

This invention relates to the recovery of tritium from various tritiated molecules by reaction with uranium. More particularly, the invention relates to the recovery of tritium from tritiated molecules by reaction with uranium wherein the reaction is conducted in a reactor which permits the reaction to occur as a moving front reaction from the point where the tritium enters the reactor charged with uranium down the reactor until the uranium is exhausted.

18 years of continuous observation of tritium and atmospheric precipitations in Ramnicu Valcea (Romania): A time series analysis.

PubMed

Duliu, Octavian G; Varlam, Carmen; Shnawaw, Muataz Dheyaa

2018-05-16

To get more information on the origin of tritium and to evidence any possible presence of anthropogenic sources, between January 1999 and December 2016, the precipitation level and tritium concentration were monthly recorded and investigated by the Cryogenic Institute of Ramnicu Valcea, Romania. Compared with similar data covering a radius of about 1200 km westward, the measurements gave similar results concerning the time evolution of tritium content and precipitation level for the entire time interval excepting the period between 2009 and 2011 when the tritium concentrations showed a slight increase, most probable due to the activity of neighboring experimental pilot plant for tritium and deuterium separation. Regardless this fact, all data pointed towards a steady tendency of tritium concentrations to decrease with an annual rate of about 1.4 ± 0.05%. The experimental data on precipitation levels and tritium concentrations form two complete time series whose time series analysis showed, at p < 0.01, the presence of a single one-year periodicity whose coincident maximums which correspond to late spring - early summer months suggest the existence of the Spring Leak mechanism with a possible contribution of the soil moisture remobilization during the warm period. Copyright © 2018 Elsevier Ltd. All rights reserved.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

DOE PAGES

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

2015-03-19

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

In situ measurement of tritium permeation through stainless steel

NASA Astrophysics Data System (ADS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2013-06-01

The TMIST-2 irradiation experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory to evaluate tritium permeation through Type 316 stainless steel (316 SS). The interior of a 316 SS seamless tube specimen was exposed to a 4He carrier gas mixed with a specified quantity of tritium (T2) to yield partial pressures of 0.1, 5, and 50 Pa at 292 °C and 330 °C. In situ tritium permeation measurements were made by passing a He-Ne sweep gas over the outer surface of the specimen to carry the permeated tritium to a bubbler column for liquid scintillation counting. Results from in situ permeation measurements were compared with predictions based on an ex-reactor permeation correlation in the literature. In situ permeation data were also used to derive an in-reactor permeation correlation as a function of temperature and pressure over the ranges considered in this study. In addition, the triton recoil contribution to tritium permeation, which results from the transmutation of 3He to T, was also evaluated by introducing a 4He carrier gas mixed with 3He at a partial pressure of 1013 Pa at 330 °C. Less than 3% of the tritium resulting from 3He transmutation contributed to tritium permeation.

Home page of Arnold Air Force Base

Science.gov Websites

time to reflect on the men and women who have gi... Facebook Logo Free-jet engine test at AEDC facility record for free-jet mode engines by achieving transonic speeds! @AEDCnews https://t.co/6lD4T5bnte Free-jet engine test at AEDC facility sets record Free-jet engine test at AEDC facility sets record

Test Stand at the Rocket Engine Test Facility

NASA Image and Video Library

1973-02-21

The thrust stand in the Rocket Engine Test Facility at the National Aeronautics and Space Administration (NASA) Lewis Research Center in Cleveland, Ohio. The Rocket Engine Test Facility was constructed in the mid-1950s to expand upon the smaller test cells built a decade before at the Rocket Laboratory. The $2.5-million Rocket Engine Test Facility could test larger hydrogen-fluorine and hydrogen-oxygen rocket thrust chambers with thrust levels up to 20,000 pounds. Test Stand A, seen in this photograph, was designed to fire vertically mounted rocket engines downward. The exhaust passed through an exhaust gas scrubber and muffler before being vented into the atmosphere. Lewis researchers in the early 1970s used the Rocket Engine Test Facility to perform basic research that could be utilized by designers of the Space Shuttle Main Engines. A new electronic ignition system and timer were installed at the facility for these tests. Lewis researchers demonstrated the benefits of ceramic thermal coatings for the engine’s thrust chamber and determined the optimal composite material for the coatings. They compared the thermal-coated thrust chamber to traditional unlined high-temperature thrust chambers. There were more than 17,000 different configurations tested on this stand between 1973 and 1976. The Rocket Engine Test Facility was later designated a National Historic Landmark for its role in the development of liquid hydrogen as a propellant.

Power-scaling performance of a three-dimensional tritium betavoltaic diode

NASA Astrophysics Data System (ADS)

Liu, Baojun; Chen, Kevin P.; Kherani, Nazir P.; Zukotynski, Stefan

2009-12-01

Three-dimensional diodes fabricated by electrochemical etching are exposed to tritium gas at pressures from 0.05 to 33 atm at room temperature to examine its power scaling performance. It is shown that the three-dimensional microporous structure overcomes the self-absorption limited saturation of beta flux at high tritium pressures. These results are contrasted against the three-dimensional device powered in one instance by tritium absorbed in the near surface region of the three-dimensional microporous network, and in another by a planar scandium tritide foil. These findings suggest that direct tritium occlusion in the near surface of three-dimensional diode can improve the specific power production.

[Value of the tritium test for determining the fat content in the body of rats].

PubMed

Pisarchuk, K L

1990-01-01

An indirect method for estimation of the fat percentage in the animal organism, a tritium test, was studied on laboratory male rats aged 4 and 12 months. Results obtained from the tritium test and direct chemical analysis were compared. With age a mean absolute error of the tritium test increased (from 1 to 8%) as against actual values of the water and fat percentage in the organism obtained by a direct chemical analysis. The data obtained testify to the relative insolvency of the tritium test, as well as the necessity to carry additional investigations in order to obtain adequate data.

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

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

Phillips, H.; Dean, J.; Privas, E.

2015-03-15

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

[Study on Tritium Content in Soil at Sites of Nuclear Explosions on the Territory of Semipalatinsk Test Site].

PubMed

Timonova, L V; Lyakhova, O N; Lukashenko, S N; Aidarkhanov, A O

2015-01-01

As a result of investigations carried out on the territory of Semipalatinsk Test Site, tritium was found in different environmental objects--surface and ground waters, vegetation, air environment, and snow cover. The analysis of the data obtained has shown that contamination of environmental objects at the Semipalatinsk Test Site with tritium is associated with the places where underground nuclear tests were performed. Since tritium can originate from an activation reaction and be trapped by pock particles during a test, it was decided to examine the soil in the sites where surface and excavation tests took place. It was found that the concentration of tritium in soil correlates with the concentration of europium. Probably, the concentration of tritium in the soil depends on the character and yield of the tests performed. Findings of the study have revealed that tritium can be found in soil in significant amounts not only in sites where underground nuclear tests took place but also in sites where surface and excavation nuclear tests were carried out.

Effect of the self-pumped limiter concept on the tritium fuel cycle

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

Finn, P.A.; Sze, D.K.; Hassanein, A.

1988-09-01

The self-pumped limiter concept was the impurity control system for the reactor in the Tokamak Power Systems Study (TPSS). The use of a self-pumped limiter had a major impact on the design of the tritium systems of the TPSS fusion reactor. The self-pumped limiter functions by depositing the helium ash under a layer of metal (vanadium). The majority of the hydrogen species are recycled at the plasma edge; a small fraction permeates to the blanket/coolant which was lithium in TPSS. Use of the self-pumped limiter results in the elimination of the plasma processing system. Thus, the blanket tritium processing systemmore » becomes the major tritium system. The main advantages achieved for the tritium systems with a self-pumped limiter are a reduction in the capital cost of tritium processing equipment as well as a reduction in building space, a reduced tritium inventory for processing and for reserve storage, an increase in the inherent safety of the fusion plant and an improvement in economics for a fusion world economy.« less

Recovery of Retained Tritium from Graphite Tile of JT-60U

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

Takeishi, Toshiharu; Katayama, Kazunari; Nishikawa, Masabumi

Tritium thermal release and full combustion with oxygen were performed on isotropic graphite tiles used for plasma facing material of JT-60U. Approximately 50-80 % of tritium was released by dry argon gas purge and 20-50 % of tritium was released by humid argon gas purge up to 800-1200 deg. C within one day, respectively. Further several percent of tritium was released by full combustion with oxygen. It was experimentally confirmed that all retained tritium is not released by thermal dry gas purge and by use of isotope exchange reaction at high temperature in such a short period. In the fullmore » combustion operation, isotropic graphite begins to combust at higher temperature than 650 deg. C, but effective combustion temperature was higher than 700 deg. C. Since it is very difficult to heat the graphite tile attached on the wall of vacuum vessel at higher than 700 deg. C, it is considered to be not easy to recover all the tritium retained in the graphite while in the vacuum vessel.« less

EFFECTS OF TRITIUM GAS EXPOSURE ON THE GLASS TRANSITION TEMPERATURE OF EPDM ELASTOMER AND ON THE CONDUCTIVITY OF POLYANILINE

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

Clark, E; Marie Kane, M

2008-12-12

Four formulations of EPDM (ethylene-propylene diene monomer) elastomer were exposed to tritium gas initially at one atmosphere and ambient temperature for between three and four months in closed containers. Material properties that were characterized include density, volume, mass, appearance, flexibility, and dynamic mechanical properties. The glass transition temperature was determined by analysis of the dynamic mechanical property data per ASTM standards. EPDM samples released significant amounts of gas when exposed to tritium, and the glass transition temperature increased by about 3 C. during the exposure. Effects of ultraviolet and gamma irradiation on the surface electrical conductivity of two types ofmore » polyaniline films are also documented as complementary results to planned tritium exposures. Future work will determine the effects of tritium gas exposure on the electrical conductivity of polyaniline films, to demonstrate whether such films can be used as a sensor to detect tritium. Surface conductivity was significantly reduced by irradiation with both gamma rays and ultraviolet light. The results of the gamma and UV experiments will be correlated with the tritium exposure results.« less

Light water detritiation

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

Fedorchenko, O.A.; Aleksee, I.A.; Bondarenko, S.D.

2015-03-15

Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWHmore » process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m{sup 3}/day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m{sup 3}), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the hot tower. The introduction of some quantity of deuterium into the gaseous flow before the hot tower lowers the tritium separation factor in that column. One possible variant of deuterium introduction to the hot tower of the first stage was modelled. The decontamination capacity increases by a 2.5 factor.« less

Tritium in [15O]water, its identification and removal.

PubMed

Sasaki, T; Ishii, S; Tomiyoshi, K; Ido, T; Miyauchi, J; Senda, M

2000-02-01

The present investigation was undertaken to identify the long-lived radionuclide and its chemical forms existing in [15O]water which was synthesized from 15O produced by the nuclear reaction 14N(d,n)15O, and to develop a method for its removal to facilitate radioactive waste disposal. The long-lived nuclide was identified as tritium based on a comparison of its physical half-life and the energy spectrum of beta-rays with those of tritium. The major chemical form of tritium in the target gas was estimated to be molecular hydrogen. The tritium radioactivity was completely removed without a serious loss occurring to the yield of [15O]water by passing the irradiated target gas over a heated palladium catalyst followed by a calcium chloride column before the final synthesis of the [15O]water. This provided a practical way of removing tritium from the [15O]water.

In situ measurements of tritium evapotranspiration (³H-ET) flux over grass and soil using the gradient and eddy covariance experimental methods and the FAO-56 model.

PubMed

Connan, O; Maro, D; Hébert, D; Solier, L; Caldeira Ideas, P; Laguionie, P; St-Amant, N

2015-10-01

The behaviour of tritium in the environment is linked to the water cycle. We compare three methods of calculating the tritium evapotranspiration flux from grassland cover. The gradient and eddy covariance methods, together with a method based on the theoretical Penmann-Monteith model were tested in a study carried out in 2013 in an environment characterised by high levels of tritium activity. The results show that each of the three methods gave similar results. The various constraints applying to each method are discussed. The results show a tritium evapotranspiration flux of around 15 mBq m(-2) s(-1) in this environment. These results will be used to improve the entry parameters for the general models of tritium transfers in the environment. Copyright © 2015 Elsevier Ltd. All rights reserved.

Study on Tritium Removal Performance by Gas Separation Membrane with Reflux Flow for Tritium Removal System of Fusion Reactor

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

Iwai, Yasunori; Yamanishi, Toshihiko; Hayashi, Takumi

2005-07-15

Addition of gas separation membrane process into the usual tritium removal process from an indoor atmosphere is attractive for a fusion plant, where a large amount of atmosphere should be processed. As a manner to improve the partial pressure difference between feed and permeated side, intended reflux of vapor and the hydrogen concentrated at permeated side is conceived to enlarge the partial pressure difference. Membrane separation with reflux flow has been proposed as an attractive process to enhance the recovery ratio of tritium component. Effect of reflux on the recovery ratio of tritium component was evaluated by numerical analysis. Themore » effect of reflux on separation performance becomes striking as the target species have higher permeability coefficients. Hence, the gas separation by membrane with reflux flow is favorable for tritium recovery.« less

Optimization of simultaneous tritium–radiocarbon internal gas proportional counting

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

Bonicalzi, R. M.; Aalseth, C. E.; Day, A. R.

Specific environmental applications can benefit from dual tritium and radiocarbon measurements in a single compound. Assuming typical environmental levels, it is often the low tritium activity relative to the higher radiocarbon activity that limits the dual measurement. In this paper, we explore the parameter space for a combined tritium and radiocarbon measurement using a methane sample mixed with an argon fill gas in low-background proportional counters of a specific design. We present an optimized methane percentage, detector fill pressure, and analysis energy windows to maximize measurement sensitivity while minimizing count time. The final optimized method uses a 9-atm fill ofmore » P35 (35% methane, 65% argon), and a tritium analysis window from 1.5 to 10.3 keV, which stops short of the tritium beta decay endpoint energy of 18.6 keV. This method optimizes tritium counting efficiency while minimizing radiocarbon beta decay interference.« less

Design and tritium permeation analysis of China HCCB TBM port cell

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

Jiangfeng, S.; Guoqiang, H.; Zhiyong, H.

2015-03-15

China is planning to develop a helium-cooled ceramic breeder (HCCB) test blanket module (TBM) on ITER to test key blanket technologies. In this paper, the design and tritium permeation analysis of China HCCB TBM port cell are introduced. A theoretical model has been developed to estimate tritium permeation rates and leak rates from the components and pipes which China has scheduled to house in the port cell. It is shown that on normal working conditions, the permeation and leak rate of the systems in the port cell will be no higher than 1.58 Ci/d without the use of tritium permeationmore » barriers, and 0.10 Ci/d with the use of tritium permeation barriers. It also appears that tritium permeation barriers are necessary for high temperature components such as the reduction bed and the heater.« less

Development of a Tritium Extruder for ITER Pellet Injection

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

M.J. Gouge; P.W. Fisher

As part of the International Thermonuclear Experimental Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. Hydrogenic pellets will be used in ITER to sustain the fusion power in the plasma core and may be crucial in reducing first-wall tritium inventories by a process of "isotopic fueling" in which tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge. This repeating single-stage pneumatic pellet injector, called the Tritium-Proof-of-Principle Phase II (TPOP-II) Pellet Injector, has a piston-driven mechanical extruder andmore » is designed to extrude and accelerate hydrogenic pellets sized for the ITER device. The TPOP-II program has the following development goals: evaluate the feasibility of extruding tritium and deuterium-tritium (D-T) mixtures for use in future pellet injection systems; determine the mechanical and thermal properties of tritium and D-T extrusions; integrate, test, and evaluate the extruder in a repeating, single-stage light gas gun that is sized for the ITER application (pellet diameter -7 to 8 mm); evaluate options for recycling propellant and extruder exhaust gas; and evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory that requires secondary and room containment systems. In tests with deuterium feed at ORNL, up to 13 pellets per extrusion have been extruded at rates up to 1 Hz and accelerated to speeds of 1.0 to 1.1 km/s, using hydrogen propellant gas at a supply pressure of 65 bar. Initially, deuterium pellets 7.5 mm in diameter and 11 mm in length were produced-the largest cryogenic pellets produced by the fusion program to date. These pellets represent about a 10% density perturbation to ITER. Subsequently, the extruder nozzle was modified to produce pellets that are almost 7.5-mm right circular cylinders. Tritium and D-T pellets have been produced in experiments at the Los Alamos National Laboratory Tritium Systems Test Assembly. About 38 g of tritium have been utilized in the experiment. The tritium was received in eight batches, six from product containers and two from the Isotope Separation System. Two types of runs were made: those in which the material was only extruded and those in which pellets were produced and fired with deuterium propellant. A total of 36 TZ runs and 28 D-T runs have been made. A total of 36 pure tritium runs and 28 D-T mixture runs were made. Extrusion experiments indicate that both T2 and D-T will require higher extrusion forces than D2 by about a factor of two.« less

KSC-98pc786

NASA Image and Video Library

1998-07-06

James W. Tibble (pointing at engine), an Engine Systems/Ground Support Equipment team manager for Rocketdyne, discusses the operation of a Space Shuttle Main Engine with Robert B. Sieck, director of Shuttle Processing; U.S. Congressman Dave Weldon; and KSC Center Director Roy D. Bridges Jr. Following the ribbon cutting ceremony for KSC's new 34,600-square-foot Space Shuttle Main Engine Processing Facility (SSMEPF), KSC employees and media explored the facility. A major addition to the existing Orbiter Processing Facility Bay 3, the SSMEPF replaces the Shuttle Main Engine Shop located in the Vehicle Assembly Building (VAB). The decision to move the shop out of the VAB was prompted by safety considerations and recent engine processing improvements. The first three main engines to be processed in the new facility will fly on Shuttle Endeavour's STS-88 mission in December 1998

The SSMEPF opens with a ribbon-cutting ceremony

NASA Technical Reports Server (NTRS)

1998-01-01

James W. Tibble (pointing at engine), an Engine Systems/Ground Support Equipment team manager for Rocketdyne, discusses the operation of a Space Shuttle Main Engine with Robert B. Sieck, director of Shuttle Processing; U.S. Congressman Dave Weldon; and KSC Center Director Roy D. Bridges Jr. Following the ribbon cutting ceremony for KSC's new 34,600-square-foot Space Shuttle Main Engine Processing Facility (SSMEPF), KSC employees and media explored the facility. A major addition to the existing Orbiter Processing Facility Bay 3, the SSMEPF replaces the Shuttle Main Engine Shop located in the Vehicle Assembly Building (VAB). The decision to move the shop out of the VAB was prompted by safety considerations and recent engine processing improvements. The first three main engines to be processed in the new facility will fly on Shuttle Endeavour's STS-88 mission in December 1998.

Estimating subsurface water volumes and transit times in Hokkaido river catchments, Japan, using high-accuracy tritium analysis

NASA Astrophysics Data System (ADS)

Gusyev, Maksym; Yamazaki, Yusuke; Morgenstern, Uwe; Stewart, Mike; Kashiwaya, Kazuhisa; Hirai, Yasuyuki; Kuribayashi, Daisuke; Sawano, Hisaya

2015-04-01

The goal of this study is to estimate subsurface water transit times and volumes in headwater catchments of Hokkaido, Japan, using the New Zealand high-accuracy tritium analysis technique. Transit time provides insights into the subsurface water storage and therefore provides a robust and quick approach to quantifying the subsurface groundwater volume. Our method is based on tritium measurements in river water. Tritium is a component of meteoric water, decays with a half-life of 12.32 years, and is inert in the subsurface after the water enters the groundwater system. Therefore, tritium is ideally suited for characterization of the catchment's responses and can provide information on mean water transit times up to 200 years. Only in recent years has it become possible to use tritium for dating of stream and river water, due to the fading impact of the bomb-tritium from thermo-nuclear weapons testing, and due to improved measurement accuracy for the extremely low natural tritium concentrations. Transit time of the water discharge is one of the most crucial parameters for understanding the response of catchments and estimating subsurface water volume. While many tritium transit time studies have been conducted in New Zealand, only a limited number of tritium studies have been conducted in Japan. In addition, the meteorological, orographic and geological conditions of Hokkaido Island are similar to those in parts of New Zealand, allowing for comparison between these regions. In 2014, three field trips were conducted in Hokkaido in June, July and October to sample river water at river gauging stations operated by the Ministry of Land, Infrastructure, Transport and Tourism (MLIT). These stations have altitudes between 36 m and 860 m MSL and drainage areas between 45 and 377 km2. Each sampled point is located upstream of MLIT dams, with hourly measurements of precipitation and river water levels enabling us to distinguish between the snow melt and baseflow contributions to the river discharge. For the June sampling, the tritium and stable isotope results indicate below normal river discharges with a strong contribution of snow melt at some sampling points, and relatively short groundwater transit times. The tritium concentration results are used to interpret mean transit times (MTTs) for each sampling point using a tritium input curve constructed from historical International Atomic Energy Agency and available Japanese data, and subsurface volumes are estimated from the MTTs and measured river discharges.

Radionuclide Basics: Tritium

EPA Pesticide Factsheets

Tritium is a hydrogen atom that has two neutrons in the nucleus and one proton. It is radioactive and behaves like other forms of hydrogen in the environment. Tritium is produced naturally in the upper atmosphere and as a byproduct of nuclear fission.

2016 Accomplishments. Tritium aging studies on stainless steel. Forging process effects on the fracture toughness properties of tritium-precharged stainless steel

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

Morgan, Michael J.

Forged austenitic stainless steels are used as the materials of construction for pressure vessels designed to contain tritium at high pressure. These steels are highly resistant to tritium-assisted fracture but their resistance can depend on the details of the forging microstructure. During FY16, the effects of forging strain rate and deformation temperature on the fracture toughness properties of tritium-exposed-and-aged Type 304L stainless steel were studied. Forgings were produced from a single heat of steel using four types of production forging equipment – hydraulic press, mechanical press, screw press, and high-energy-rate forging (HERF). Each machine imparted a different nominal strain ratemore » during the deformation. The objective of the study was to characterize the J-Integral fracture toughness properties as a function of the industrial strain rate and temperature. The second objective was to measure the effects of tritium and decay helium on toughness. Tritium and decay helium effects were measured by thermally precharging the as-forged specimens with tritium gas at 34.5 MPa and 350°C and aging for up to five years at -80°C to build-in decay helium prior to testing. The results of this study show that the fracture toughness properties of the as-forged steels vary with forging strain rate and forging temperature. The effect is largely due to yield strength as the higher-strength forgings had the lower toughness values. For non-charged specimens, fracture toughness properties were improved by forging at 871°C versus 816°C and Screw-Press forgings tended to have lower fracture toughness values than the other forgings. Tritium exposures reduced the fracture toughness values remarkably to fracture toughness values averaging 10-20% of as-forged values. However, forging strain rate and temperature had little or no effect on the fracture toughness after tritium precharging and aging. The result was confirmed by fractography which indicated that fracture modes in the tritium-exposed specimens were similar for all forgings. Another FY16 objective was to prepare fracture toughness specimens from Types 304L and 21-6-9 stainless steel weldments and heat-affected zones (HAZ) for tritium charging.« less

The global geochemistry of bomb-produced tritium - General circulation model compared to available observations and traditional interpretations

NASA Technical Reports Server (NTRS)

Koster, Randal D.; Broecker, Wallace S.; Jouzel, Jean; Suozzo, Robert J.; Russell, Gary L.; Rind, David

1989-01-01

Observational evidence suggests that of the tritium produced during nuclear bomb tests that has already reached the ocean, more than twice as much arrived through vapor impact as through precipitation. In the present study, the Goddard Institute for Space Studies 8 x 10 deg atmospheric general circulation model is used to simulate tritium transport from the upper atmosphere to the ocean. The simulation indicates that tritium delivery to the ocean via vapor impact is about equal to that via precipitation. The model result is relatively insensitive to several imposed changes in tritium source location, in model parameterizations, and in model resolution. Possible reasons for the discrepancy are explored.

Cryogenci DT and D2 Targets for Inertial Confinement Fusion

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

Sangster, T.C.; Betti, R.; Craxton, R.S.

Ignition target designs for inertial confinement fusion on the National Ignition Facility (NIF) are based on a spherical ablator containing a solid, cryogenic-fuel layer of deuterium and tritium. The need for solid-fuel layers was recognized more than 30 years ago and considerable effort has resulted in the production of cryogenic targets that meet most of the critical fabrication tolerances for ignition on the NIf. Significant progress with the formation and characterization of cryogenic targets for both direct and x-ray drive will be described. Results from recent cryogenic implosions will also be presented.

Early Career: The search for weakly interacting dark matter with liquid xenon

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

Hall, Carter

2017-02-08

We report results from a search for weakly interacting dark matter particles obtained with the LUX experiment. LUX was located at a depth of 4850 feet at the Sanford Underground Research Facility in Lead, South Dakota from 2013 through 2016. It found no evidence for dark matter particle interactions and set new constraints on the properties of such particles for masses between 6 GeV and 100 TeV. The work reported here also characterized the performance of such experiments by developing a new calibration technique based upon a tritium beta decay source.

Facilities Engineering in NASA

NASA Technical Reports Server (NTRS)

Pagluiso, M. A.

1970-01-01

An overview of NASA facilities is given outlining some of the more interesting and unique aspects of engineering and facilities associated with the space program. Outlined are some of the policies under which the Office of Facilities conducts its business. Included are environmental quality control measures.

THERMOGRAVIMETRIC CHARACTERIZATION OF GLOVEBOX GLOVES

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

Korinko, P.

An experimental project was initiated to characterize mass loss when heating different polymer glovebox glove material samples to three elevated temperatures, 90, 120, and 150 C. Samples from ten different polymeric gloves that are being considered for use in the tritium gloveboxes were tested. The intent of the study was to determine the amount of material lost. These data will be used in a subsequent study to characterize the composition of the material lost. One goal of the study was to determine which glove composition would least affect the glovebox atmosphere stripper system. Samples lost most of the mass inmore » the initial 60 minutes of thermal exposure and as expected increasing the temperature increased the mass loss and shortened the time to achieve a steady state loss. The most mass loss was experienced by Jung butyl-Hypalon{reg_sign} at 146 C with 12.9% mass loss followed by Piercan Hypalon{reg_sign} at 144 C with 11.4 % mass loss and Jung butyl-Viton{reg_sign} at 140 C with 5.2% mass loss. The least mass loss was experienced by the Jung Viton{reg_sign} and the Piercan polyurethane. Unlike the permeation testing (1) the vendor and fabrication route influences the amount of gaseous species that is evolved. Additional testing to characterize these products is recommended. Savannah River Site (SRS) has many gloveboxes deployed in the Tritium Facility. These gloveboxes are used to protect the workers and to ensure a suitable environment in which to handle tritium gas products. The gas atmosphere in the gloveboxes is purified using a stripper system. The process gas strippers collect molecules that may have hydrogen or its isotopes attached, e.g., waters of hydration, acids, etc. Recently, sulfur containing compounds were detected in the stripper system and the presence of these compounds accelerates the stripper system's aging process. This accelerated aging requires the strippers to be replaced more often which can impact the facility's schedule and operational cost. It was posited that sulfur bearing and other volatile compounds were derived from glove off-gassing. Due to the large number of gloves in the facility, small mass loss from each glove could result in a significant total mass of undesirable material entering the glovebox atmosphere and subsequently the stripper system. A thermogravimetric analysis (TGA) study was conducted to determine the amount of low temperature volatiles that may be expected to offgas from the gloves. The data were taken on relatively small samples but are normalized with respect to the sample's surface area. Additional testing is needed to determine the composition of the off-gassing species. The TGA study was conducted to ascertain the magnitude of the issue and to determine if further experimentation is warranted or necessary.« less

Short papers of the Fourth International Conference, Geochronology, Cosmochronology, Isotope Geology, 1978

USGS Publications Warehouse

Zartman, Robert E.

1978-01-01

Tritium content of both hot and cold waters in Yellowstone National Park was used to infer something of the ground-water system feeding hot springs and geysers. Curves in three figures show: (1) Tritium content of water leaving piston flow and well mixed ground-water systems in Yellowstone Park; (2) tritium in precipitation, mixed reservoirs, and cold waters of Yellowstone Park, and (3) tritium in mixed reservoirs and hot waters of Yellowstone Park. (Woodard-USGS)

Region-specific tritium enrichment, and not differential beta-absorption, is the major cause of 'quenching' in film autoradiography.

PubMed

McEachron, D L; Nissanov, J; Tretiak, O J

1997-06-01

Tritium quenching refers to the situation in which estimates of tritium content generated by film autoradiography depend on the chemical composition of the tissue as well as on the concentration of the radioisotope. When analysing thin brain sections, for example, regions rich in lipid content generate reduced optical densities on x-ray film compared with lipid-poor regions even when the total tissue concentration of tritium in those regions is identical. We hypothesize that the dried thickness of regions within sections depends upon the relative concentrations and types of lipid within the regions. Areas low in white matter dry thinner than areas high in white matter, leading to a relative enrichment of tritium in the thinner regions. To test this model, a series of brain pastes were made with different concentrations of grey and white matter and impregnated with equal amounts of tritium. The thickness of dried sections was compared with percentage of white matter and apparent radioactive content as determined by autoradiogram analysis. The results demonstrated that thickness increased, and apparent radioactivity decreased, with higher percentages of white matter. In the second experiment, thickness measurements from dried sections were successfully used to correct the apparent radioisotope content of autoradiograms created from tritium containing white- and grey-matter tissue slices. We conclude that within-section thickness variation is the major physical cause for 'tritium quenching'.

Vertical profile of tritium concentration in air during a chronic atmospheric HT release.

PubMed

Noguchi, Hiroshi; Yokoyama, Sumi

2003-03-01

The vertical profiles of tritium gas and tritiated water concentrations in air, which would have an influence on the assessment of tritium doses as well as on the environmental monitoring of tritium, were measured in a chronic tritium gas release experiment performed in Canada in 1994. While both of the profiles were rather uniform during the day because of atmospheric mixing, large gradients of the profiles were observed at night. The gradient coefficients of the profiles were derived from the measurements. Correlations were analyzed between the gradient coefficients and meteorological conditions: solar radiation, wind speed, and turbulent diffusivity. It was found that the solar radiation was highly correlated with the gradient coefficients of tritium gas and tritiated water profiles and that the wind speed and turbulent diffusivity showed weaker correlations with those of tritiated water profiles. A one-dimensional tritium transport model was developed to analyze the vertical diffusion of tritiated water re-emitted from the ground into the atmosphere. The model consists of processes of tritium gas deposition to soil including oxidation into tritiated water, reemission of tritiated water, dilution of tritiated water in soil by rain, and vertical diffusion of tritiated water in the atmosphere. The model accurately represents the accumulation of tritiated water in soil water and the time variations and vertical profiles of tritiated water concentrations in air.

The effects of dual-domain mass transfer on the tritium-helium-3 dating method.

PubMed

Neumann, Rebecca B; Labolle, Eric M; Harvey, Charles F

2008-07-01

Diffusion of tritiated water (referred to as tritium) and helium-3 between mobile and immobile regions in aquifers (mass transfer) can affect tritium and helium-3 concentrations and hence tritium-helium-3 (3H/3He) ages that are used to estimate aquifer recharge and groundwater residence times. Tritium and helium-3 chromatographically separate during transport because their molecular diffusion coefficients differ. Simulations of tritium and helium-3 transport and diffusive mass transfer along stream tubes show that mass transfer can shift the 3H/3He age of the tritium and helium-3 concentration ([3H + 3He]) peak to dates much younger than the 1963 peak in atmospheric tritium. Furthermore, diffusive mass-transfer can cause the 3H/3He age to become younger downstream along a stream tube, even as the mean water-age must increase. Simulated patterns of [3H + 3He] versus 3H/3He age using a mass transfer model appear consistent with a variety of field data. These results suggest that diffusive mass transfer should be considered, especially when the [3H + 3He] peak is not well defined or appears younger than the atmospheric peak. 3H/3He data provide information about upstream mass-transfer processes that could be used to constrain mass-transfer models; however, uncritical acceptance of 3H/3He dates from aquifers with immobile regions could be misleading.

Isotopic fractionation of tritium in biological systems.

PubMed

Le Goff, Pierre; Fromm, Michel; Vichot, Laurent; Badot, Pierre-Marie; Guétat, Philippe

2014-04-01

Isotopic fractionation of tritium is a highly relevant issue in radiation protection and requires certain radioecological considerations. Sound evaluation of this factor is indeed necessary to determine whether environmental compartments are enriched/depleted in tritium or if tritium is, on the contrary, isotopically well-distributed in a given system. The ubiquity of tritium and the standard analytical methods used to assay it may induce biases in both the measurement and the signification that is accorded to the so-called fractionation: based on an exhaustive review of the literature, we show how, sometimes large deviations may appear. It is shown that when comparing the non-exchangeable fraction of organically bound tritium (neOBT) to another fraction of tritium (e.g. tritiated water) the preparation of samples and the measurement of neOBT reported frequently led to underestimation of the ratio of tritium to hydrogen (T/H) in the non-exchangeable compartment by a factor of 5% to 50%. In the present study, corrections are proposed for most of the biological matrices studied so far. Nevertheless, the values of isotopic fractionation reported in the literature remain difficult to compare with each other, especially since the physical quantities and units often vary between authors. Some improvements are proposed to better define what should encompass the concepts of exchangeable and non-exchangeable fractions. Copyright © 2014 Elsevier Ltd. All rights reserved.

Field-Reversed Configuration Power Plant Critical-Issue Scoping Study

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

Santarius, J. F.; Mogahed, E. A.; Emmert, G. A.

A team from the Universities of Wisconsin, Washington, and Illinois performed an engineering scoping study of critical issues for field-reversed configuration (FRC) power plants. The key tasks for this research were (1) systems analysis for deuterium-tritium (D-T) FRC fusion power plants, and (2) conceptual design of the blanket and shield module for an FRC fusion core. For the engineering conceptual design of the fusion core, the project team focused on intermediate-term technology. For example, one decision was to use steele structure. The FRC systems analysis led to a fusion power plant with attractive features including modest size, cylindrical symmetry, goodmore » thermal efficiency (52%), relatively easy maintenance, and a high ratio of electric power to fusion core mass, indicating that it would have favorable economics.« less

Developing the Pulsed Fission-Fusion (PuFF) Engine

NASA Technical Reports Server (NTRS)

Adams, Robert B.; Cassibry, Jason; Bradley, David; Fabisinski, Leo; Statham, Geoffrey

2014-01-01

In September 2013 the NASA Innovative Advanced Concept (NIAC) organization awarded a phase I contract to the PuFF team. Our phase 1 proposal researched a pulsed fission-fusion propulsion system that compressed a target of deuterium (D) and tritium (T) as a mixture in a column, surrounded concentrically by Uranium. The target is surrounded by liquid lithium. A high power current would flow down the liquid lithium and the resulting Lorentz force would compress the column by roughly a factor of 10. The compressed column would reach criticality and a combination of fission and fusion reactions would occur. Our Phase I results, summarized herein, review our estimates of engine and vehicle performance, our work to date to model the fission-fusion reaction, and our initial efforts in experimental analysis.

7. This photographic copy of an engineering drawing displays the ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

7. This photographic copy of an engineering drawing displays the building's floor plan in its 1995 arrangement, with rooms designated. California Institute of Technology, Jet Propulsion Laboratory, Facilities Engineering and Construction Office, "Addition to Weigh & Control Bldg. E-35, Demolition, Floor and Roof Plans," drawing no. E35/3-0, October 5, 1983. California Institute of Technology, Jet Propulsion Laboratory, Plant Engineering: engineering drawings of structures at JPL Edwards Facility. Drawings on file at JPL Plant Engineering, Pasadena, California. - Jet Propulsion Laboratory Edwards Facility, Weigh & Control Building, Edwards Air Force Base, Boron, Kern County, CA

4. This photographic copy of an engineering drawing shows the ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

4. This photographic copy of an engineering drawing shows the plan and details for Test Stand "G" and the placement of the vibrator. California Institute of Technology, Jet Propulsion Laboratory, Plant Engineering: "Vibration Test Facility-Bldg E-72, Floor & Roof Plans, Sections, Details & Door Schedule," drawing no. E72/2-5, 21 May 1964. California Institute of Technology, Jet Propulsion Laboratory, Plant Engineering: engineering drawings of structures at JPL Edwards Facility. Drawings on file at JPL Plant Engineering, Pasadena, California. - Jet Propulsion Laboratory Edwards Facility, Test Stand G, Edwards Air Force Base, Boron, Kern County, CA

Modeling and analysis of tritium dynamics in a DT fusion fuel cycle

NASA Astrophysics Data System (ADS)

Kuan, William

1998-11-01

A number of crucial design issues have a profound effect on the dynamics of the tritium fuel cycle in a DT fusion reactor, where the development of appropriate solutions to these issues is of particular importance to the introduction of fusion as a commercial system. Such tritium-related issues can be classified according to their operational, safety, and economic impact to the operation of the reactor during its lifetime. Given such key design issues inherent in next generation fusion devices using the DT fuel cycle development of appropriate models can then lead to optimized designs of the fusion fuel cycle for different types of DT fusion reactors. In this work, two different types of modeling approaches are developed and their application to solving key tritium issues presented. For the first approach, time-dependent inventories, concentrations, and flow rates characterizing the main subsystems of the fuel cycle are simulated with a new dynamic modular model of a fusion reactor's fuel cycle, named X-TRUFFLES (X-Windows TRitiUm Fusion Fuel cycLE dynamic Simulation). The complex dynamic behavior of the recycled fuel within each of the modeled subsystems is investigated using this new integrated model for different reactor scenarios and design approaches. Results for a proposed fuel cycle design taking into account current technologies are presented, including sensitivity studies. Ways to minimize the tritium inventory are also assessed by examining various design options that could be used to minimize local and global tritium inventories. The second modeling approach involves an analytical model to be used for the calculation of the required tritium breeding ratio, i.e., a primary design issue which relates directly to the feasibility and economics of DT fusion systems. A time-integrated global tritium balance scheme is developed and appropriate analytical expressions are derived for tritium self-sufficiency relevant parameters. The easy exploration of the large parameter space of the fusion fuel cycle can thus be conducted as opposed to previous modeling approaches. Future guidance for R&D (research and development) in fusion nuclear technology is discussed in view of possible routes to take in reducing the tritium breeding requirements of DT fusion reactors.

Advice on the setting up of a workshop for treating tritium gas light sources at 527 ecw at Dongen

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

Not Available

1978-05-01

Tritium occurs in light sources mainly in the form of hydrogen gas, but also a certain amount in the form of tritiated water vapor. From a radiation-hygienic standpoint, the latter form determines the safety regulations to be taken, because this radioactive water vapor is absorbed to a considerable amount by the human body via inhalation and via the skin. The work space must satisfy various demands. The distances over which the apparatus and accessories are transported must be as short as possible. The floors must be seamless, the walls must be decontaminated. There must be storage in the work roommore » for radioactive materials and this facility must be fireproof. The apparatus must work on a seamless and well decontaminated working surface. The air velocity in the opening on the front side must amount to approximately 40 cm/sec with normal use. A ventilator can be placed in the ceiling with a water-tight design. The air supply in the space must be regulated in such a way that the whole space is provided with fresh air.« less

Pb17Li and lithium: A thermodynamic rationalisation of their radically different chemistry

NASA Astrophysics Data System (ADS)

Hubberstey, Peter

1997-08-01

The contrasting chemistry of Pb17Li and lithium is attributed to their lithium activities. PbLi alloys exhibit marked negative deviations from ideality owing to 'chemical short range order', giving γ Li = 7.26 × 10 -4, aLi = 1.23 × 10 -4 and overlineGLi = -57.8 kJ mol -1 in Pb-17Li at 773 K. This overlineGLi value is sufficiently negative to prevent the reaction of Pb17Li with gaseous hydrogen and nitrogen to form LiH and Li 3N but not with oxygen containing gases to form Li 2O. Similarly, nitride and carbide ceramics are compatible with Pb-17Li but oxide ceramics are liable to degradation. In contrast, unit activity liquid lithium reacts with all the gases and, depending on their free energy of formation, some of the ceramics. Wherea, dissolved oxygen is corrosive in Pb-17Li, giving LiCrO 2, dissolved nitrogen adopts the corrosive role in lithium giving Li 9CrN 5. The instability of LiH in Pb-17Li renders tritium extraction facile; this contrasts with lithium for which tritium extraction is difficult owing to LiH formation.

Shock timing measurements and analysis in deuterium-tritium-ice layered capsule implosions on NIF

NASA Astrophysics Data System (ADS)

Robey, H. F.; Celliers, P. M.; Moody, J. D.; Sater, J.; Parham, T.; Kozioziemski, B.; Dylla-Spears, R.; Ross, J. S.; LePape, S.; Ralph, J. E.; Hohenberger, M.; Dewald, E. L.; Berzak Hopkins, L.; Kroll, J. J.; Yoxall, B. E.; Hamza, A. V.; Boehly, T. R.; Nikroo, A.; Landen, O. L.; Edwards, M. J.

2014-02-01

Recent advances in shock timing experiments and analysis techniques now enable shock measurements to be performed in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility (NIF). Previous measurements of shock timing in inertial confinement fusion implosions [Boehly et al., Phys. Rev. Lett. 106, 195005 (2011); Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] were performed in surrogate targets, where the solid DT ice shell and central DT gas were replaced with a continuous liquid deuterium (D2) fill. These previous experiments pose two surrogacy issues: a material surrogacy due to the difference of species (D2 vs. DT) and densities of the materials used and a geometric surrogacy due to presence of an additional interface (ice/gas) previously absent in the liquid-filled targets. This report presents experimental data and a new analysis method for validating the assumptions underlying this surrogate technique. Comparison of the data with simulation shows good agreement for the timing of the first three shocks, but reveals a considerable discrepancy in the timing of the 4th shock in DT ice layered implosions. Electron preheat is examined as a potential cause of the observed discrepancy in the 4th shock timing.

Experimental investigation on charcoal adsorption for cryogenic pump application

NASA Astrophysics Data System (ADS)

Scannapiego, Matthieu; Day, Christian

2017-12-01

Fusion reactors are generating energy by nuclear fusion between deuterium and tritium. In order to evacuate the high gas throughputs from the plasma exhaust, large pumping speed systems are required. Within the European Fusion Programme, the Karlsruhe Institute of Technology (KIT) has taken the lead to design a three-stage cryogenic pump that can provide a separation function of hydrogen isotopes from the remaining gases; hence limiting the tritium inventory in the machine. A primary input parameter for the detailed design of a cryopump is the sticking coefficient between the gas and the pumping surface. For this purpose, the so-called TIMO open panel pump experiment was conducted in the TIMO-2 test facility at KIT in order to measure pumping speeds on an activated carbon surface cooled at temperatures between 6 K and 22 K, for various pure gases and gas mixtures, under fusion relevant gas flow conditions, and for two different geometrical pump configurations. The influences of the panel temperature, the gas throughput and the intake gas temperature on the pumping speed have been characterized, providing valuable qualitative results for the design of the three-stage cryopump. In a future work, supporting Monte Carlo simulations should allow for derivation of the sticking coefficients.

Neutron temporal diagnostic for high-yield deuterium–tritium cryogenic implosions on OMEGA

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

Stoeckl, C.; Boni, R.; Ehrne, F.

A next-generation neutron temporal diagnostic (NTD) capable of recording high-quality data for the highest anticipated yield cryogenic deuterium–tritium (DT) implosion experiments was recently installed at the Omega Laser Facility. A high-quality measurement of the neutron production width is required to determine the hot-spot pressure achieved in inertial confinement fusion experiments—a key metric in assessing the quality of these implosions. The design of this NTD is based on a fast-rise-time plastic scintillator, which converts the neutron kinetic energy to 350- to 450-nm-wavelength light. The light from the scintillator inside the nose-cone assembly is relayed ∼16 m to a streak camera inmore » a well-shielded location. An ∼200× reduction in neutron background was observed during the first high-yield DT cryogenic implosions compared to the current NTD installation on OMEGA. An impulse response of ∼40 ± 10 ps was measured in a dedicated experiment using hard x-rays from a planar target irradiated with a 10-ps short pulse from the OMEGA EP laser. The measured instrument response includes contributions from the scintillator rise time, optical relay, and streak camera.« less

Understanding fuel magnetization and mix using secondary nuclear reactions in magneto-inertial fusion.

PubMed

Schmit, P F; Knapp, P F; Hansen, S B; Gomez, M R; Hahn, K D; Sinars, D B; Peterson, K J; Slutz, S A; Sefkow, A B; Awe, T J; Harding, E; Jennings, C A; Chandler, G A; Cooper, G W; Cuneo, M E; Geissel, M; Harvey-Thompson, A J; Herrmann, M C; Hess, M H; Johns, O; Lamppa, D C; Martin, M R; McBride, R D; Porter, J L; Robertson, G K; Rochau, G A; Rovang, D C; Ruiz, C L; Savage, M E; Smith, I C; Stygar, W A; Vesey, R A

2014-10-10

Magnetizing the fuel in inertial confinement fusion relaxes ignition requirements by reducing thermal conductivity and changing the physics of burn product confinement. Diagnosing the level of fuel magnetization during burn is critical to understanding target performance in magneto-inertial fusion (MIF) implosions. In pure deuterium fusion plasma, 1.01 MeV tritons are emitted during deuterium-deuterium fusion and can undergo secondary deuterium-tritium reactions before exiting the fuel. Increasing the fuel magnetization elongates the path lengths through the fuel of some of the tritons, enhancing their probability of reaction. Based on this feature, a method to diagnose fuel magnetization using the ratio of overall deuterium-tritium to deuterium-deuterium neutron yields is developed. Analysis of anisotropies in the secondary neutron energy spectra further constrain the measurement. Secondary reactions also are shown to provide an upper bound for the volumetric fuel-pusher mix in MIF. The analysis is applied to recent MIF experiments [M. R. Gomez et al., Phys. Rev. Lett. 113, 155003 (2014)] on the Z Pulsed Power Facility, indicating that significant magnetic confinement of charged burn products was achieved and suggesting a relatively low-mix environment. Both of these are essential features of future ignition-scale MIF designs.

Comparison of the recently proposed super-Marx generator approach to thermonuclear ignition with the deuterium-tritium laser fusion-fission hybrid concept by the Lawrence Livermore National Laboratory

DOE PAGES

Winterberg, F.

2009-01-01

The recently proposed super-Marx generator pure deuterium microdetonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser deuterium-tritium fusion-fission hybrid concept (LIFE). In a super-Marx generator, a large number of ordinary Marx generators charge up a much larger second stage ultrahigh voltage Marx generator from which for the ignition of a pure deuterium microexplosion an intense GeV ion beam can be extracted. Typical examples of the LIFE concept are a fusion gain of 30 and a fission gain of 10, making up a total gain of 300, with about ten times more energy released into fissionmore » as compared to fusion. This means the substantial release of fission products, as in fissionless pure fission reactors. In the super-Marx approach for the ignition of pure deuterium microdetonation, a gain of the same magnitude can, in theory, be reached. If feasible, the super-Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of thermonuclear microexplosions.« less

Methods for tritium labeling

DOEpatents

Andres, Hendrik; Morimoto, Hiromi; Williams, Philip G.

1993-01-01

Reagents and processes for reductively introducing deuterium or tritium into organic molecules are described. The reagents are deuterium or tritium analogs of trialkyl boranes, borane or alkali metal aluminum hydrides. The process involves forming these reagents in situ from alkali metal tritides or deuterides.

Process for exchanging hydrogen isotopes between gaseous hydrogen and water

DOEpatents

Hindin, Saul G.; Roberts, George W.

1980-08-12

A process for exchanging isotopes of hydrogen, particularly tritium, between gaseous hydrogen and water is provided whereby gaseous hydrogen depeleted in tritium and liquid or gaseous water containing tritium are reacted in the presence of a metallic catalyst.

A free-piston Stirling engine/linear alternator controls and load interaction test facility

NASA Technical Reports Server (NTRS)

Rauch, Jeffrey S.; Kankam, M. David; Santiago, Walter; Madi, Frank J.

1992-01-01

A test facility at LeRC was assembled for evaluating free-piston Stirling engine/linear alternator control options, and interaction with various electrical loads. This facility is based on a 'SPIKE' engine/alternator. The engine/alternator, a multi-purpose load system, a digital computer based load and facility control, and a data acquisition system with both steady-periodic and transient capability are described. Preliminary steady-periodic results are included for several operating modes of a digital AC parasitic load control. Preliminary results on the transient response to switching a resistive AC user load are discussed.

Structural dynamics verification facility study

NASA Technical Reports Server (NTRS)

Kiraly, L. J.; Hirchbein, M. S.; Mcaleese, J. M.; Fleming, D. P.

1981-01-01

The need for a structural dynamics verification facility to support structures programs was studied. Most of the industry operated facilities are used for highly focused research, component development, and problem solving, and are not used for the generic understanding of the coupled dynamic response of major engine subsystems. Capabilities for the proposed facility include: the ability to both excite and measure coupled structural dynamic response of elastic blades on elastic shafting, the mechanical simulation of various dynamical loadings representative of those seen in operating engines, and the measurement of engine dynamic deflections and interface forces caused by alternative engine mounting configurations and compliances.

Experiments with high-voltage insulators in the presence of tritium

NASA Astrophysics Data System (ADS)

Grisham, L. R.; Falter, H.; Causey, R.; Chrisman, W.; Stevenson, T.; Wright, K.

1991-02-01

During the final deuterium-tritium phases of the TFTR and JET tokamaks half of the neutral injectors will be used to produce tritium neutral beams to maintain an equal mix of deuterium and tritium in the core plasma, and such requirements may also occur in future devices. This will require that the voltage hold off capabilities of the high voltage insulators in the accelerators be unimpaired by any charge buildups associated with the beta decay of adsorbed layers. We report tests in which we measured the drain currents under high dc voltage of TFTR and JET accelerator insulators while they were successively exposed to vacuum, deuterium and tritium. There did not appear to be any substantial reduction in hold-off capability with tritium, although at some voltages there was a small increase in the leakage current. We also compared the breakdown properties of a plastic tubing filled with deuterium and then tritium at varying pressures, since such tubing has been considered as a high-voltage break in the gas feed system for TFTR, and the presence of large numbers of electron-ion pairs might lead to enhanced Paschen breakdown. We found no significant differences in the behavior for the geometry used.

Age Distribution of Groundwater

NASA Astrophysics Data System (ADS)

Morgenstern, U.; Daughney, C. J.

2012-04-01

Groundwater at the discharge point comprises a mixture of water from different flow lines with different travel time and therefore has no discrete age but an age distribution. The age distribution can be assessed by measuring how a pulse shaped tracer moves through the groundwater system. Detection of the time delay and the dispersion of the peak in the groundwater compared to the tracer input reveals the mean residence time and the mixing parameter. Tritium from nuclear weapons testing in the early 1960s resulted in a peak-shaped tritium input to the whole hydrologic system on earth. Tritium is the ideal tracer for groundwater because it is an isotope of hydrogen and therefore is part of the water molecule. Tritium time series data that encompass the passage of the bomb tritium pulse through the groundwater system in all common hydrogeologic situations in New Zealand demonstrate a semi-systematic pattern between age distribution parameters and hydrologic situation. The data in general indicate high fraction of mixing, but in some cases also indicate high piston flow. We will show that still, 45 years after the peak of the bomb tritium, it is possible to assess accurately the parameters of age distributions by measuring the tail of the bomb tritium.

Tritium as an indicator of ground-water age in Central Wisconsin

USGS Publications Warehouse

Bradbury, Kenneth R.

1991-01-01

In regions where ground water is generally younger than about 30 years, developing the tritium input history of an area for comparison with the current tritium content of ground water allows quantitative estimates of minimum ground-water age. The tritium input history for central Wisconsin has been constructed using precipitation tritium measured at Madison, Wisconsin and elsewhere. Weighted tritium inputs to ground water reached a peak of over 2,000 TU in 1964, and have declined since that time to about 20-30 TU at present. In the Buena Vista basin in central Wisconsin, most ground-water samples contained elevated levels of tritium, and estimated minimum ground-water ages in the basin ranged from less than one year to over 33 years. Ground water in mapped recharge areas was generally younger than ground water in discharge areas, and estimated ground-water ages were consistent with flow system interpretations based on other data. Estimated minimum ground-water ages increased with depth in areas of downward ground-water movement. However, water recharging through thick moraine sediments was older than water in other recharge areas, reflecting slower infiltration through the sandy till of the moraine.

Continuous production of tritium in an isotope-production reactor with a separate circulation system

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

Tritium volume activity in the Baltic Sea in 1987-1989

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

Styro, D.B.; Korotkov, V.P.

Tritium volume activities measured in the Baltic Sea are summarized in this paper. Activity levels were determined by the liquid scintillation method with a LS-1000 counter. The field investigations showed that the tritium volume activity in the Baltic Sea can change substantially in absolute magnitude. Therefore, average volume activity is used as an indicator of natural content. Correlations between calculated (averaged) tritium activity levels and the Chernobyl accident are very briefly discussed. 7 refs., 2 figs., 1 tab.

Tritium distribution in ground water around large underground fusion explosions

USGS Publications Warehouse

Stead, F.W.

1963-01-01

Tritium will be released in significant amounts from large underground nuclear fusion explosions in the Plowshare Program. The tritium could become highly concentrated in nearby ground waters, and could be of equal or more importance as a possible contaminant than other long-lived fission-product and induced radionuclides. Behavior of tritiated water in particular hydrologic and geologic environments, as illustrated by hypothetical explosions in dolomite and tuff, must be carefully evaluated to predict under what conditions high groundwater concentrations of tritium might occur.

Rayleigh Scattering for Measuring Flow in a Nozzle Testing Facility

NASA Technical Reports Server (NTRS)

Gomez, Carlos R.; Panda, Jayanta

2006-01-01

A molecular Rayleigh-scattering-based air-density measurement system was built in a large nozzle-and-engine-component test facility for surveying supersonic plumes from jet-engine exhaust. A molecular Rayleigh-scattering-based air-density measurement system was built in a large nozzle-and-enginecomponent test facility for surveying supersonic plumes from jet-engine exhaust

Final Environmental Assessment: Base-Wide Building Demolition Arnold Air Force Base, Tennessee

DTIC Science & Technology

2006-02-01

Building • Engine Test Facility ( ETF )-B Exhauster • ETF -A Airside • ETF -A Exhauster • ETF -A Reefer • CE Facility • Rocket Storage • Von Karman Gas...Executive Order ESA Endangered Species Act ETF Engine Test Facility FamCamp Family Camping Area P:\\ARNOLDAFB\\333402DO42COMPLIANCE\\DEMOLITION...Fabrication Shop • Natural Resources Building • Salt Storage Building • Administration Building • Engine Test Facility ( ETF )-B Exhauster • ETF -A

Use of isotopic data to estimate water residence times of the Finger Lakes, New York

USGS Publications Warehouse

Michel, Robert L.; Kraemer, Thomas F.

1995-01-01

Water retention times in the Finger Lakes, a group of 11 lakes in central New York with similar hydrologic and climatic characteristics, were estimated by use of a tritium-balance model. During July 1991, samples were collected from the 11 lakes and selected tributary streams and were analyzed for tritium, deuterium, and oxygen-18. Additional samples from some of the sites were collected in 1990, 1992 and 1993. Tritium concentration in lake water ranged from 24.6 Tritium Units (TU) (Otisco Lake) to 43.2 TU (Seneca Lake).The parameters in the model used to obtain water retention time (WRT) included relative humidity, evaporation rate, tritium concentrations of inflowing water and lake water, and WRT of the lake. A historical record of tritium concentrations in precipitation and runoff was obtained from rainfall data at Ottawa, Canada, analyses of local wines produced during 1977–1991, and streamflow samples collected in 1990–1991. The model was simulated in yearly steps for 1953–1991, and the WRT was varied to reproduce tritium concentrations measured in each lake in 1991. Water retention times obtained from model simulations ranged from 1 year for Otisco Lake to 12 years for Seneca Lake, and with the exception of Seneca Lake and Skaneateles Lake, were in agreement with earlier estimates obtained from runoff estimates and chloride balances. The sensitivity of the model to parameter changes was tested to determine possible reasons for the differences calculated for WRT's for Seneca Lake and Skaneateles Lake. The shorter WRT obtained from tritium data for Lake Seneca (12 years as compared to 18 years) can be explained by a yearly addition of less than 3% by lake volume of ground water to the lake, the exact percentage depending on tritium concentration in the ground water.

Tritium retention in S-65 beryllium after 100 eV plasma exposure

NASA Astrophysics Data System (ADS)

Causey, Rion A.; Longhurst, Glen R.; Harbin, Wally

1997-02-01

The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 × 10 17 ( D + T)/ cm2s up to about 3 × 10 18 ( D + T)/ cm2s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H 2 gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 × 10 17 ( D + T)/ cm2 at 373 K to a low of 1 × 10 16 ( D + T)/ cm2 at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C = 0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters.

Modeling of HT and HTO release from irradiated lithium metazirconate

NASA Astrophysics Data System (ADS)

Beloglazov, S.; Nishikawa, M.; Glugla, M.; Kinjyo, T.

2004-08-01

Modeling studies of tritium release from irradiated Li 2ZrO 3 (MAPI) pebbles have been carried out in order to evaluate the effect of purge gas composition on tritium release behavior. The release characteristics were obtained by temperature programmed desorption (TPD) technique in the series of post-irradiation experiments in JRR-4 research reactor of JAERI. Nitrogen with hydrogen at various partial pressures (100 and 1000 Pa) was used as a purge gas. Two sets of ionization chambers and its dedicated electrometers allowed the tritium concentration to be monitored in the chemical form of HT and overall tritium concentration in the mixture HT and HTO simultaneously during desorption runs. The tritium release curves were numerically fitted in order to evaluate the mass transfer coefficients.

Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

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

Shaver, Mark W.; Lanning, Donald D.

2010-02-01

The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum ofmore » the individual components equaling the measured values.« less

This photocopy of an engineering drawing shows the floor plan ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

This photocopy of an engineering drawing shows the floor plan of the Liner Lab, including room functions. Austin, Field & Fry, Architects Engineers, 22311 West Third Street, Los Angeles 57, California: Edwards Test Station Complex Phase II, Jet Propulsion Laboratory, California Institute of Technology, Edwards Air Force Base, Edwards, California: "Liner Laboratory, Floor Plan and Schedules," drawing no. E33/4-2, 26 June 1962. California Institute of Technology, Jet Propulsion Laboratory, Plant Engineering: engineering drawings of structures at JPL Edwards Facility. Drawings on file at JPL Plant Engineering, Pasadena, California. California Institute of Technology, Jet Propulsion Laboratory, Plant Engineering: engineering drawings of structures at JPL Edwards Facility. Drawings on file at JPL Plant Engineering, Pasadena, California - Jet Propulsion Laboratory Edwards Facility, Liner Laboratory, Edwards Air Force Base, Boron, Kern County, CA

NASA Lewis Wind Tunnel Model Systems Criteria

NASA Technical Reports Server (NTRS)

Soeder, Ronald H.; Haller, Henry C.

1994-01-01

This report describes criteria for the design, analysis, quality assurance, and documentation of models or test articles that are to be tested in the aeropropulsion facilities at the NASA Lewis Research Center. The report presents three methods for computing model allowable stresses on the basis of the yield stress or ultimate stress, and it gives quality assurance criteria for models tested in Lewis' aeropropulsion facilities. Both customer-furnished model systems and in-house model systems are discussed. The functions of the facility manager, project engineer, operations engineer, research engineer, and facility electrical engineer are defined. The format for pretest meetings, prerun safety meetings, and the model criteria review are outlined Then, the format for the model systems report (a requirement for each model that is to be tested at NASA Lewis) is described, the engineers that are responsible for developing the model systems report are listed, and the time table for its delivery to the facility manager is given.

Export Control Requirements for Tritium Processing Design and R&D

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

Hollis, William Kirk; Maynard, Sarah-Jane Wadsworth

This document will address requirements of export control associated with tritium plant design and processes. Los Alamos National Laboratory has been working in the area of tritium plant system design and research and development (R&D) since the early 1970’s at the Tritium Systems Test Assembly (TSTA). This work has continued to the current date with projects associated with the ITER project and other Office of Science Fusion Energy Science (OS-FES) funded programs. ITER is currently the highest funding area for the DOE OS-FES. Although export control issues have been integrated into these projects in the past a general guidance documentmore » has not been available for reference in this area. To address concerns with currently funded tritium plant programs and assist future projects for FES, this document will identify the key reference documents and specific sections within related to tritium research. Guidance as to the application of these sections will be discussed with specific detail to publications and work with foreign nationals.« less

Export Control Requirements for Tritium Processing Design and R&D

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

Hollis, William Kirk; Maynard, Sarah-Jane Wadsworth

2015-10-30

This document will address requirements of export control associated with tritium plant design and processes. Los Alamos National Laboratory has been working in the area of tritium plant system design and research and development (R&D) since the early 1970’s at the Tritium Systems Test Assembly (TSTA). This work has continued to the current date with projects associated with the ITER project and other Office of Science Fusion Energy Science (OS-FES) funded programs. ITER is currently the highest funding area for the DOE OS-FES. Although export control issues have been integrated into these projects in the past a general guidance documentmore » has not been available for reference in this area. To address concerns with currently funded tritium plant programs and assist future projects for FES, this document will identify the key reference documents and specific sections within related to tritium research. Guidance as to the application of these sections will be discussed with specific detail to publications and work with foreign nationals.« less

Modeling of tritium transport in a fusion reactor pin-type solid breeder blanket using the diffuse code

NASA Astrophysics Data System (ADS)

Martin, Rodger; Ghoniem, Nasr M.

1986-11-01

A pin-type fusion reactor blanket is designed using γ-LiAlO 2 solid tritium breeder. Tritium transport and diffusive inventory are modeled using the DIFFUSE code. Two approaches are used to obtain characteristic LiAlO 2 grain temperatures. DIFFUSE provides intragranular diffusive inventories which scale up to blanket size. These results compare well with a numerical analysis, giving a steady-state blanket tritium inventory of 13 g. Start-up transient inventories are modeled using DIFFUSE for both full and restricted coolant flow. Full flow gives rapid inventory buildup while restricted flow prevents this buildup. Inventories after shutdown are modeled: reduced cooling is found to have little effect on removing tritium, but preheating rapidly purges inventory. DIFFUSE provides parametric modeling of solid breeder density, radiation, and surface effects. 100% dense pins are found to give massive inventory and marginal tritium release. Only large trapping energies and concentrations significantly increase inventory. Diatomic surface recombination is only significant at high temperatures.

Implementation of two-phase tritium models for helium bubbles in HCLL breeding blanket modules

NASA Astrophysics Data System (ADS)

Fradera, J.; Sedano, L.; Mas de les Valls, E.; Batet, L.

2011-10-01

Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the breeding blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phenomena in liquid Pb17.5Li and EUROFER is proposed. In a first approximation, it has been considered that He bubbles can be represented as a passive scalar. The nucleation model is based on the classical theory and includes a simplified bubble growth model. The model captures the interaction of tritium with bubbles and tritium diffusion through walls. Results show the influence of helium cavitation on tritium inventory and the importance of simulating the system walls instead of imposing fixed boundary conditions.

Radiological Impact of Tritium from Gaseous Effluent Releases at Cook Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Young, Joshua Allan

The purpose of this study was to investigate the washout of tritiated water by snow and rain from gaseous effluent releases at Donald C. Cook Nuclear Power Plant. Primary concepts studied were determination of washout coefficients for rainfall and snowfall; correlations between rainfall and snow fall tritium concentrations with tritium concentrations in the spent fuel pool, reactor cooling systems, and tritium release rates; and calculations of received doses from the process of recapture. The dose calculations are under the assumption of a maximally exposed individual to get the most conservative estimate of the effect that washout of tritiated water has on individuals around the plant site. This study is in addition to previous work that has been conducted at Cook Nuclear Power Plant for several years. The calculated washout coefficients were typically within the range of 1x10-7s -1 to 1x10-5s-1. A strong correlation between tritium concentration within the spent fuel pool and the tritium release rates was determined.

Solid state tritium detector for biomedical applications

NASA Astrophysics Data System (ADS)

Gordon, J. S.; Farrell, R.; Daley, K.; Oakes, C. E.

1994-08-01

Radioactive labeling of proteins is a very important technique used in biomedical research to identify, isolate, and investigate the expression and properties of proteins in biological systems. In such procedures, the preferred radiolabel is often tritium. Presently, binding assays involving tritium are carried out using inconvenient and expensive techniques which rely on the use of scintillation fluid counting systems. This traditional method involves both time-consuming laboratory protocols and the generation of substantial quantities of radioactive and chemical waste. We have developed a novel technology to measure the tritium content of biological specimens that does not rely on scintillation fluids. The tritiated samples can be positioned directly under a large area, monolithic array of specially prepared avalanche photodiodes (APDs) which record the tritium activity distribution at each point within the field of view of the array. The 1 mm(sup 2) sensing elements exhibit an intrinsic tritium beta detection efficiency of 27% with high gain uniformity and very low cross talk.

32 CFR 766.8 - Procedure for review, approval, execution and distribution of aviation facility licenses.

Code of Federal Regulations, 2010 CFR

2010-07-01

... CIVIL AIRCRAFT § 766.8 Procedure for review, approval, execution and distribution of aviation facility... license and Certificate of Insurance to the Commander, Naval Facilities Engineering Command or his... Facilities Engineering Command or his designated representative. (1) Upon receipt, the Commander, Naval...

Cytogenetic damage analysis in mice chronically exposed to low-dose internal tritium beta-particle radiation.

PubMed

Roch-Lefèvre, Sandrine; Grégoire, Eric; Martin-Bodiot, Cécile; Flegal, Matthew; Fréneau, Amélie; Blimkie, Melinda; Bannister, Laura; Wyatt, Heather; Barquinero, Joan-Francesc; Roy, Laurence; Benadjaoud, Mohamed; Priest, Nick; Jourdain, Jean-René; Klokov, Dmitry

2018-06-08

The aim of this study was to carry out a comprehensive examination of potential genotoxic effects of low doses of tritium delivered chronically to mice and to compare these effects to the ones resulting from equivalent doses of gamma-irradiation. Mice were chronically exposed for one or eight months to either tritiated water (HTO) or organically bound tritium (OBT) in drinking water at concentrations of 10 kBq/L, 1 MBq/L or 20 MBq/L. Dose rates of internal β-particle resulting from such tritium treatments were calculated and matching external gamma-exposures were carried out. We measured cytogenetic damage in bone marrow and in peripheral blood lymphocytes (PBLs) and the cumulative tritium doses (0.009 - 181 mGy) were used to evaluate the dose-response of OBT in PBLs, as well as its relative biological effectiveness (RBE). Neither tritium, nor gamma exposures produced genotoxic effects in bone marrow. However, significant increases in chromosome damage rates in PBLs were found as a result of chronic OBT exposures at 1 and 20 M Bq/L, but not at 10 kBq/L. When compared to an external acute gamma-exposure ex vivo , the RBE of OBT for chromosome aberrations induction was evaluated to be significantly higher than 1 at cumulative tritium doses below 10 mGy. Although found non-existent at 10 kBq/L (the WHO limit), the genotoxic potential of low doses of tritium (>10 kBq/L), mainly OBT, may be higher than currently assumed.

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

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

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

DOE PAGES

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

2018-03-20

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Tritium, deuterium, and helium permeation through EPDM O-rings

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

Swansiger, W.A.

1992-03-01

This paper discusses tritium permeabilities determined at room temperature, 1.0 MPa (150 psia) tritium for three 23.4 cm diameter EPDM (ethylene-propylene-diene monomer) O-rings using a full-scale mock-up of the Al-SX shipping container seal geometry. The AL-SX container is being developed by Sandia National Laboratories for shipping tritium reservoirs. To determine the tritium permeation rate as a function of temperature, a 50.8 mm diameter EPDM O-ring was tested from room temperature to 150{degrees}C at a pressure of 1.0 MPa. Additional permeation measurements were made under the following test conditions: deuterium and helium-4 at room temperature and a pressure of 1.0 MPamore » using the full-scale AL-SX fixture, tritium from 0.1 MPa to 1.0 MPa at 142{degrees}C using the 50.8 mm fixture, and deuterium form room temperature to 150{degrees}C at a pressure of 1.0 MPa using the three full-scale O-rings showed the average room temperature, 1.0 MPa steady state tritium permeation rate to be about 1 {times} 10{sup {minus}2} Pa-liter/sec (7.6 {times} 10{sup {minus}5} torr-liter/sec or 1 {times} 10{sup {minus}4} std cc/sec), well within the allowable limit of 7.1 {times} 10{sup {minus}2} Pa-liter/sec for tritium release form the AL-SX container.« less

Monte Carlo calculations of the incineration of plutonium and minor actinides of laser fusion inertial confinement fusion fission energy (LIFE) engine

NASA Astrophysics Data System (ADS)

Adem, ACIR; Eşref, BAYSAL

2018-07-01

In this paper, neutronic analysis in a laser fusion inertial confinement fusion fission energy (LIFE) engine fuelled plutonium and minor actinides using a MCNP codes was investigated. LIFE engine fuel zone contained 10 vol% TRISO particles and 90 vol% natural lithium coolant mixture. TRISO fuel compositions have Mod①: reactor grade plutonium (RG-Pu), Mod②: weapon grade plutonium (WG-Pu) and Mod③: minor actinides (MAs). Tritium breeding ratios (TBR) were computed as 1.52, 1.62 and 1.46 for Mod①, Mod② and Mod③, respectively. The operation period was computed as ∼21 years when the reference TBR > 1.05 for a self-sustained reactor for all investigated cases. Blanket energy multiplication values (M) were calculated as 4.18, 4.95 and 3.75 for Mod①, Mod② and Mod③, respectively. The burnup (BU) values were obtained as ∼1230, ∼1550 and ∼1060 GWd tM–1, respectively. As a result, the higher BU were provided with using TRISO particles for all cases in LIFE engine.

Measurements of the 169Tm(n,2n)168Tm cross section between 9.0 and 17.5 MeV

NASA Astrophysics Data System (ADS)

Soter, J.; Bhike, Megha; Krishichayan, Fnu; Finch, S. W.; Tornow, W.

2016-09-01

Measurements of the 169Tm(n,2n)168Tm cross section have been performed in 0.5 MeV intervals for neutron energies ranging from 9.0 MeV to 17.5 MeV in order to resolve discrepancies in the current literature data. The neutron activation technique was used with 90Zr and 197Au as monitor foils. After irradiation, de-excitation gamma rays were recorded off-line with High-Purity Germanium (HPGE) detectors in TUNL's Low-Background Counting Facility. In addition, data for the 169Tm(n,3n)167Tm reaction have also been obtained from 15.5 MeV to 17.5 MeV. The results of these measurements provide the basis for investigating properties of the interial confinement fusion plasma in deuterium-tritium (DT) capsules at the National Ignition Facility located at Lawrence Livermore National Laboratory.

Measurements of an ablator-gas atomic mix in indirectly driven implosions at the National Ignition Facility.

PubMed

Smalyuk, V A; Tipton, R E; Pino, J E; Casey, D T; Grim, G P; Remington, B A; Rowley, D P; Weber, S V; Barrios, M; Benedetti, L R; Bleuel, D L; Bradley, D K; Caggiano, J A; Callahan, D A; Cerjan, C J; Clark, D S; Edgell, D H; Edwards, M J; Frenje, J A; Gatu-Johnson, M; Glebov, V Y; Glenn, S; Haan, S W; Hamza, A; Hatarik, R; Hsing, W W; Izumi, N; Khan, S; Kilkenny, J D; Kline, J; Knauer, J; Landen, O L; Ma, T; McNaney, J M; Mintz, M; Moore, A; Nikroo, A; Pak, A; Parham, T; Petrasso, R; Sayre, D B; Schneider, M B; Tommasini, R; Town, R P; Widmann, K; Wilson, D C; Yeamans, C B

2014-01-17

We present the first results from an experimental campaign to measure the atomic ablator-gas mix in the deceleration phase of gas-filled capsule implosions on the National Ignition Facility. Plastic capsules containing CD layers were filled with tritium gas; as the reactants are initially separated, DT fusion yield provides a direct measure of the atomic mix of ablator into the hot spot gas. Capsules were imploded with x rays generated in hohlraums with peak radiation temperatures of ∼294  eV. While the TT fusion reaction probes conditions in the central part (core) of the implosion hot spot, the DT reaction probes a mixed region on the outer part of the hot spot near the ablator-hot-spot interface. Experimental data were used to develop and validate the atomic-mix model used in two-dimensional simulations.

F-Area Hazardous Waste Management Facility groundwater monitoring report, Third and fourth quarters 1995: Volume 1

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

NONE

1996-03-01

Groundwater at the F-Area Hazardous Waste Management Facility (HWMF) is monitored in compliance with applicable regulations. Monitoring results are compared to the South Carolina Department of Health and Environmental Control (SCDHEC) Groundwater Protection Standard (GWPS). Historically and currently, gross alpha, nitrates, nonvolatile beta, and tritium are among the primary constituents to exceed standards. Numerous other radionuclides and hazardous constituents also exceed the GWPS in the groundwater during the second half of 1995, notably cadmium, lead, radium-226, radium-228, strontium-90, and total alpha-emitting radium. The elevated constituents were found primarily in the water table (aquifer zone IIB{sub 2}), however, several other aquifermore » unit monitoring wells contained elevated levels of constituents. Water-level maps indicate that the groundwater flow rates and directions at the F-Area HWMF have remained relatively constant since the basins ceased to be active in 1988.« less

The magnetic recoil spectrometer for measurements of the absolute neutron spectrum at OMEGA and the NIF

DOE PAGES

Casey, D. T.; Frenje, J. A.; Gatu Johnson, M.; ...

2013-04-18

The neutron spectrum produced by deuterium-tritium (DT) inertial confinement fusion implosions contains a wealth of information about implosion performance including the DT yield, iontemperature, and areal-density. The Magnetic Recoil Spectrometer (MRS) has been used at both the OMEGA laser facility and the National Ignition Facility (NIF) to measure the absolute neutron spectrum from 3 to 30 MeV at OMEGA and 3 to 36 MeV at the NIF. These measurements have been used to diagnose the performance of cryogenic target implosions to unprecedented accuracy. Interpretation of MRS data requires a detailed understanding of the MRS response and background. This paper describesmore » ab initio characterization of the system involving Monte Carlo simulations of the MRS response in addition to the commission experiments for in situ calibration of the systems on OMEGA and the NIF.« less

Cryogenic target system for hydrogen layering

DOE PAGES

Parham, T.; Kozioziemski, B.; Atkinson, D.; ...

2015-11-24

Here, a cryogenic target positioning system was designed and installed on the National Ignition Facility (NIF) target chamber. This instrument incorporates the ability to fill, form, and characterize the NIF targets with hydrogen isotopes needed for ignition experiments inside the NIF target bay then transport and position them in the target chamber. This effort brought to fruition years of research in growing and metrologizing high-quality hydrogen fuel layers and landed it in an especially demanding operations environment in the NIF facility. D-T (deuterium-tritium) layers for NIF ignition experiments have extremely tight specifications and must be grown in a very highlymore » constrained environment: a NIF ignition target inside a cryogenic target positioner inside the NIF target bay. Exquisite control of temperature, pressure, contaminant level, and thermal uniformity are necessary throughout seed formation and layer growth to create an essentially-groove-free single crystal layer.« less

The magnetic recoil spectrometer for measurements of the absolute neutron spectrum at OMEGA and the NIF.

PubMed

Casey, D T; Frenje, J A; Johnson, M Gatu; Séguin, F H; Li, C K; Petrasso, R D; Glebov, V Yu; Katz, J; Magoon, J; Meyerhofer, D D; Sangster, T C; Shoup, M; Ulreich, J; Ashabranner, R C; Bionta, R M; Carpenter, A C; Felker, B; Khater, H Y; LePape, S; MacKinnon, A; McKernan, M A; Moran, M; Rygg, J R; Yeoman, M F; Zacharias, R; Leeper, R J; Fletcher, K; Farrell, M; Jasion, D; Kilkenny, J; Paguio, R

2013-04-01

The neutron spectrum produced by deuterium-tritium (DT) inertial confinement fusion implosions contains a wealth of information about implosion performance including the DT yield, ion-temperature, and areal-density. The Magnetic Recoil Spectrometer (MRS) has been used at both the OMEGA laser facility and the National Ignition Facility (NIF) to measure the absolute neutron spectrum from 3 to 30 MeV at OMEGA and 3 to 36 MeV at the NIF. These measurements have been used to diagnose the performance of cryogenic target implosions to unprecedented accuracy. Interpretation of MRS data requires a detailed understanding of the MRS response and background. This paper describes ab initio characterization of the system involving Monte Carlo simulations of the MRS response in addition to the commission experiments for in situ calibration of the systems on OMEGA and the NIF.

First downscattered neutron images from Inertial Confinement Fusion experiments at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Guler, Nevzat; Aragonez, Robert J.; Archuleta, Thomas N.; Batha, Steven H.; Clark, David D.; Clark, Deborah J.; Danly, Chris R.; Day, Robert D.; Fatherley, Valerie E.; Finch, Joshua P.; Gallegos, Robert A.; Garcia, Felix P.; Grim, Gary; Hsu, Albert H.; Jaramillo, Steven A.; Loomis, Eric N.; Mares, Danielle; Martinson, Drew D.; Merrill, Frank E.; Morgan, George L.; Munson, Carter; Murphy, Thomas J.; Oertel, John A.; Polk, Paul J.; Schmidt, Derek W.; Tregillis, Ian L.; Valdez, Adelaida C.; Volegov, Petr L.; Wang, Tai-Sen F.; Wilde, Carl H.; Wilke, Mark D.; Wilson, Douglas C.; Atkinson, Dennis P.; Bower, Dan E.; Drury, Owen B.; Dzenitis, John M.; Felker, Brian; Fittinghoff, David N.; Frank, Matthias; Liddick, Sean N.; Moran, Michael J.; Roberson, George P.; Weiss, Paul; Buckles, Robert A.; Cradick, Jerry R.; Kaufman, Morris I.; Lutz, Steve S.; Malone, Robert M.; Traille, Albert

2013-11-01

Inertial Confinement Fusion experiments at the National Ignition Facility (NIF) are designed to understand and test the basic principles of self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT) filled cryogenic plastic (CH) capsules. The experimental campaign is ongoing to tune the implosions and characterize the burning plasma conditions. Nuclear diagnostics play an important role in measuring the characteristics of these burning plasmas, providing feedback to improve the implosion dynamics. The Neutron Imaging (NI) diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by collecting images at two different energy bands for primary (13-15 MeV) and downscattered (10-12 MeV) neutrons. From these distributions, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. The first downscattered neutron images from imploding ICF capsules are shown in this paper.

Inertial Confinement Fusion and the National Ignition Facility (NIF)

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

Ross, P.

2012-08-29

Inertial confinement fusion (ICF) seeks to provide sustainable fusion energy by compressing frozen deuterium and tritium fuel to extremely high densities. The advantages of fusion vs. fission are discussed, including total energy per reaction and energy per nucleon. The Lawson Criterion, defining the requirements for ignition, is derived and explained. Different confinement methods and their implications are discussed. The feasibility of creating a power plant using ICF is analyzed using realistic and feasible numbers. The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory is shown as a significant step forward toward making a fusion power plant based on ICF.more » NIF is the world’s largest laser, delivering 1.8 MJ of energy, with a peak power greater than 500 TW. NIF is actively striving toward the goal of fusion energy. Other uses for NIF are discussed.« less

Predicting water quality changes from artificial recharge sources to nearby wellfields

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

Moran, J.E.

1998-01-23

Isotope tracer technologies have proven to be powerful tools for addressing questions related to surface water-ground water interactions. The Alameda County Water District artificially recharges tens of thousands of acre-ft of water annually, delivered from Alameda Creek in order to augment dwindling ground water supplies, and to maintain a barrier to seawater intrusion. The authors are using a suite of isotope tracers to track water movement, source characteristics and accompanying water quality changes from ACWD recharge facilities to nearby wells. The data gathered during the three year project will allow quantification of dilution by ambient basin ground water, subsurface travelmore » times, and several key water quality parameters, including degree of degradation of organic compounds, the fate of trace metals during recharge and subsurface transport, and sources and transport of major ions (salts). Reconnaissance work was carried out on naturally occurring isotopes in order to better understand the hydrogeology of the ground water basin. The basin is dissected by the Hayward Fault, and geologic conditions vary greatly on either side of the fault. Stable isotopes of oxygen, carbon, helium and other noble gases, along with radiocarbon and tritium were measured on water samples from production and monitoring wells. The goal of the reconnaissance work was to age date the water at various depths and distances from the recharge ponds, to examine the chemical evolution of the water with age, and to examine the water for source-related variations in isotope composition. Ground water ages were calculated by the tritium-helium method for three production wells in the Peralta-Tyson wellfield (in the Above Hayward Fault sub-basin), and for a monitoring well positioned between the recharge facilities and production wells, screened at three discreet intervals.« less

The VATO project: An original methodology to study the transfer of tritium as HT and HTO in grassland ecosystem.

PubMed

Maro, D; Vermorel, F; Rozet, M; Aulagnier, C; Hébert, D; Le Dizès, S; Voiseux, C; Solier, L; Cossonnet, C; Godinot, C; Fiévet, B; Laguionie, P; Connan, O; Cazimajou, O; Morillon, M; Lamotte, M

2017-02-01

Tritium ( 3 H) is mainly released into the environment by nuclear power plants, military nuclear facilities and nuclear reprocessing plants. The construction of new nuclear facilities in the world as well as the evolution of nuclear fuel management might lead to an increase of 3 H discharges from the nuclear industry. The VATO project was set up by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and EDF (Electricité de France) to reduce the uncertainties in the knowledge about transfers of 3 H from an atmospheric source (currently releasing HT and HTO) to a grassland ecosystem. A fully instrumented technical platform with specifically designed materials was set up downwind of the AREVA NC La Hague reprocessing plant (Northwest of the France). This study, started in 2013, was conducted in four main steps to provide an hourly data set of 3 H concentrations in the environment, adequate to develop and/or validate transfer models. It consisted first in characterizing the physico-chemical forms of 3 H present in the air around the plant. Then, 3 H transfer kinetics to grass were quantified regarding contributions from various compartments of the environment. For this purpose, an original experimental procedure was provided to take account for biases due to rehydration of freeze-dried samples for the determination of OBT activity concentrations in biological samples. In a third step, the 3 H concentrations measured in the air and in rainwater were reconstructed at hourly intervals. Finally, a data processing technique was used to determine the biological half-lives of OBT in grass. Copyright © 2016 Elsevier Ltd. All rights reserved.

8. Photocopy of engineering drawing. AETR DIGS FACILITY THEODOLITE AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

8. Photocopy of engineering drawing. AETR DIGS FACILITY THEODOLITE AND PRISM SHELTER: SECTIONS AND DETAILS, 1971. - Cape Canaveral Air Station, Launch Complex 17, Facility 28413, East end of Lighthouse Road, Cape Canaveral, Brevard County, FL

7 CFR 1942.20 - Community Facility Guides.

Code of Federal Regulations, 2011 CFR

2011-01-01

.... (7) Guide 7—Preliminary Engineering Report Water Facility. (8) Guide 8—Preliminary Engineering Report Sewerage Systems. (9) Guide 9—Preliminary Engineering Report Solid Waste Disposal Systems. (10) Guide 10—Preliminary Engineering Report Storm Waste-Water Disposal. (11) Guide 11—Daily Inspection Report. (12) Guide...

Tritium trick

NASA Technical Reports Server (NTRS)

Green, W. V.; Zukas, E. G.; Eash, D. T.

1971-01-01

Large controlled amounts of helium in uniform concentration in thick samples can be obtained through the radioactive decay of dissolved tritium gas to He3. The term, tritium trick, applies to the case when helium, added by this method, is used to simulate (n,alpha) production of helium in simulated hard flux radiation damage studies.

Portable Intelligent Tritium in Air Monitor

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

Purghel, L.; Calin, M.R.; Bartos, D.

2005-07-15

The tritium detection method used for this monitor is original, patented in Romania. The detection unit consists of a single ionization chamber, a special fast preamplifier and a dedicated software associated to the detection unit, for signals processing. Some results concerning the tritium in relative strong gamma-ray fields are presented.

Lithium aluminate/zirconium material useful in the production of tritium

DOEpatents

Cawley, W.E.; Trapp, T.J.

A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

Lithium aluminate/zirconium material useful in the production of tritium

DOEpatents

Cawley, W.E.; Trapp, T.J.

1984-10-09

A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

Lithium aluminate/zirconium material useful in the production of tritium

DOEpatents

Cawley, William E.; Trapp, Turner J.

1984-10-09

A composition is described useful in the production of tritium in a nuclear eactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

Preparatory steps for a robust dynamic model for organically bound tritium dynamics in agricultural crops

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

Melintescu, A.; Galeriu, D.; Diabate, S.

2015-03-15

The processes involved in tritium transfer in crops are complex and regulated by many feedback mechanisms. A full mechanistic model is difficult to develop due to the complexity of the processes involved in tritium transfer and environmental conditions. First, a review of existing models (ORYZA2000, CROPTRIT and WOFOST) presenting their features and limits, is made. Secondly, the preparatory steps for a robust model are discussed, considering the role of dry matter and photosynthesis contribution to the OBT (Organically Bound Tritium) dynamics in crops.

Modeling and experiments on tritium permeation in fusion reactor blankets

NASA Astrophysics Data System (ADS)

Holland, D. F.; Longhurst, G. R.

The determination of tritium loss from helium-cooled fusion breeding blankets are discussed. The issues are: (1) applicability of present models to permeation at low tritium pressures; (2) effectiveness of oxide layers in reducing permeation; (3) effectiveness of hydrogen addition as a means to lower tritium permeation; and (4) effectiveness of conversion to tritiated water and subsequent trapping to reduce permeation. Theoretical models applicable to these issues are discussed, and results of experiments in two areas are presented; permeation of mixtures of hydrogen isotopes and conversion to tritiated water.

Hydrogeology, ground-water flow, and tritium movement at low-level radioactive-waste disposal site near Sheffield, Illinois

USGS Publications Warehouse

Garklavs, George; Healy, R.W.

1986-01-01

Groundwater flow and tritium movement are described at and near a low-level radioactive waste disposal site near Sheffield, Illinois. Flow in the shallow aquifer is confined to three basins that ultimately drain into a stripmine lake. Most of the flow from the site is through a buried, pebbly sandfilled channel. Remaining flow is toward alluvium of an existing stream. Conceptual flow models for the two largest basins are used to improve definition of flow velocity and direction. Flow velocities range from about 25 to 2,500 ft/yr. Tritium was found in all three basins. The most extensive migration of tritium is coincident with buried channel. Tritium concentrations ranged from detection level to more than 300 nanocuries/L. (USGS)

Tritium waste package

DOEpatents

Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

1995-01-01

A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

Tritium waste package

DOEpatents

Rossmassler, R.; Ciebiera, L.; Tulipano, F.J.; Vinson, S.; Walters, R.T.

1995-11-07

A containment and waste package system for processing and shipping tritium oxide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within the outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen and oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB. 1 fig.

The Windowless Gaseous Tritium Source (WGTS) of the KATRIN experiment

NASA Astrophysics Data System (ADS)

Heizmann, Florian; Seitz-Moskaliuk, Hendrik; KATRIN Collaboration

2017-09-01

The Karlsruhe Tritium Neutrino Experiment (KATRIN) will perform a direct, kinematics-based measurement of the neutrino mass with a sensitivity of 200 meV (90 % C. L.), which will be reached after 3 years of measurement time. The neutrino mass is obtained by investigating the shape of the energy spectrum of tritium β-decay electrons close to the endpoint at 18.6 keV with a spectrometer of MAC-E filter type. This contribution reviews the current status of the tritium source cryostat and magnet system which is currently in its first cool-down phase. Furthermore, the next steps of the comprehensive pre-tritium measurement programme to characterise the apparatus and investigate important systematics are outlined. This work is supported by BMBF (05A14VK2) and the Helmholtz Association.

Assessment of tritium in the Savannah River Site environment

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

Carlton, W.H.; Murphy, C.E. Jr.; Bauer, L.R.

1993-10-01

This report is the first revision to a series of reports on radionuclides inn the SRS environment. Tritium was chosen as the first radionuclide in the series because the calculations used to assess the dose to the offsite population from SRS releases indicate that the dose due to tritium, through of small consequence, is one of the most important the radionuclides. This was recognized early in the site operation, and extensive measurements of tritium in the atmosphere, surface water, and ground water exist due to the effort of the Environmental Monitoring Section. In addition, research into the transport and fatemore » of tritium in the environment has been supported at the SRS by both the local Department of Energy (DOE) Office and DOE`s Office of Health and Environmental Research.« less

Project 8: Towards cyclotron radiation emission spectroscopy on tritium

NASA Astrophysics Data System (ADS)

Fertl, Martin; Project 8 Collaboration

2017-01-01

Project 8 aims to determine the neutrino mass by making a precise measurement of the beta decay of molecular tritium (Q = 18.6 keV) using the recently demonstrated the technique of cyclotron radiation emission spectroscopy (CRES). We report on results for calibration measurements performed with Kr-83m in a gas cell that fulfills the stringent requirements for a measurement using tritium: cryogenic operation, safe tritium handling, a non-magnetic design, and a good microwave guide performance. The phased program that allows Project 8 to probe the neutrino mass range accessible using molecular tritium is described. Major financial support by the U.S. Department of Energy, Office of Science, Office of Nuclear Physics to the University of Washington under Award Number DE-FG02-97ER41020 is acknowledged

REVIEWS OF TOPICAL PROBLEMS: Helium-isotope mass-spectrometric method for studying tritium beta decay (idea, experiment, nuclear and molecular physics applications)

NASA Astrophysics Data System (ADS)

Akulov, Yuii A.; Mamyrin, Boris A.

2003-11-01

Experimental data on the variation of tritium nucleus beta decay constant caused by the interaction of the resulting beta-electron with orbital electrons and shell vacancies are reviewed for free atomic tritium and molecular tritium and used to obtain the half-life of atomic tritium (T1/2)a=(12.264±0.018) y, the half-life of the free triton (T1/2)t=(12.238±0.020) y, the axial-vector-to-vector weak-interaction coupling constant ratio (GA/GV)t=-1.2646 ± 0.0035 for beta decay of the triton, and an independent estimate of the free neutron lifetime τn= (890.3 ± 3.9stat ± 1.4syst) s.

The effects of numerical-model complexity and observation type on estimated porosity values

USGS Publications Warehouse

Starn, Jeffrey; Bagtzoglou, Amvrossios C.; Green, Christopher T.

2015-01-01

The relative merits of model complexity and types of observations employed in model calibration are compared. An existing groundwater flow model coupled with an advective transport simulation of the Salt Lake Valley, Utah (USA), is adapted for advective transport, and effective porosity is adjusted until simulated tritium concentrations match concentrations in samples from wells. Two calibration approaches are used: a “complex” highly parameterized porosity field and a “simple” parsimonious model of porosity distribution. The use of an atmospheric tracer (tritium in this case) and apparent ages (from tritium/helium) in model calibration also are discussed. Of the models tested, the complex model (with tritium concentrations and tritium/helium apparent ages) performs best. Although tritium breakthrough curves simulated by complex and simple models are very generally similar, and there is value in the simple model, the complex model is supported by a more realistic porosity distribution and a greater number of estimable parameters. Culling the best quality data did not lead to better calibration, possibly because of processes and aquifer characteristics that are not simulated. Despite many factors that contribute to shortcomings of both the models and the data, useful information is obtained from all the models evaluated. Although any particular prediction of tritium breakthrough may have large errors, overall, the models mimic observed trends.

Biological effects of tritium on fish cells in the concentration range of international drinking water standards.

PubMed

Stuart, Marilyne; Festarini, Amy; Schleicher, Krista; Tan, Elizabeth; Kim, Sang Bog; Wen, Kendall; Gawlik, Jilian; Ulsh, Brant

2016-10-01

To evaluate whether the current Canadian tritium drinking water limit is protective of aquatic biota, an in vitro study was designed to assess the biological effects of low concentrations of tritium, similar to what would typically be found near a Canadian nuclear power station, and higher concentrations spanning the range of international tritium drinking water standards. Channel catfish peripheral blood B-lymphoblast and fathead minnow testis cells were exposed to 10-100,000 Bq l(-1) of tritium, after which eight molecular and cellular endpoints were assessed. Increased numbers of DNA strand breaks were observed and ATP levels were increased. There were no increases in γH2AX-mediated DNA repair. No differences in cell growth were noted. Exposure to the lowest concentrations of tritium were associated with a modest increase in the viability of fathead minnow testicular cells. Using the micronucleus assay, an adaptive response was observed in catfish B-lymphoblasts. Using molecular endpoints, biological responses to tritium in the range of Canadian and international drinking water standards were observed. At the cellular level, no detrimental effects were noted on growth or cycling, and protective effects were observed as an increase in cell viability and an induced resistance to a large challenge dose.

Midterm Summary of Japan-US Fusion Cooperation Program TITAN

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

Muroga, Takeo; Sze, Dai-Kai; Sokolov, Mikhail

2011-01-01

Japan-US cooperation program TITAN (Tritium, Irradiation and Thermofluid for America and Nippon) started in April 2007 as 6-year project. This is the summary report at the midterm of the project. Historical overview of the Japan-US cooperation programs and direction of the TITAN project in its second half are presented in addition to the technical highlights. Blankets are component systems whose principal functions are extraction of heat and tritium. Thus it is crucial to clarify the potentiality for controlling heat and tritium flow throughout the first wall, blanket and out-of-vessel recovery systems. The TITAN project continues the JUPITER-II activity but extendsmore » its scope including the first wall and the recovery systems with the title of 'Tritium and thermofluid control for magnetic and inertial confinement systems'. The objective of the program is to clarify the mechanisms of tritium and heat transfer throughout the first-wall, the blanket and the heat/tritium recovery systems under specific conditions to fusion such as irradiation, high heat flux, circulation and high magnetic fields. Based on integrated models, the breeding, transfer, inventory of tritium and heat extraction properties will be evaluated for some representative liquid breeder blankets and the necessary database will be obtained for focused research in the future.« less

New Safety and Technical Challenges and Operational Experience on the JET First Trace Tritium Experiment

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

Pearce, R.J.H.; Bell, A.C.; Brennan, D.

'Trace Tritium Experiments' (TTE) were successfully performed on JET in 2003. The Campaign marked the first use of tritium in JET plasmas since the Deuterium-Tritium Experiment (DTE1) Campaign in 1997, and was the first use of tritium in experiments under the EFDA organisation with the UKAEA as JET Operator. The safety and regulatory preparations for the experiment were extensive. Since JET has been operated by the UKAEA the operations have followed the model of a licensed nuclear site. The safe operation of the JET torus is demonstrated in a safety case. Key Safety Management Requirement (KSMR) and Key Safety Relatedmore » Equipment (KSRE) are identified in the Safety Case for DT operation. The safe operation of the torus is within the bounds of, and under the control of, an Authority to Operate (ATO). New technical challenges were presented by the need to inject and account for small quantities of tritium in very short pulses ({approx}80ms), with an accurate time stamp. The safety and operational management of the campaign are described. Valuable lessons were learned which would help in running future experiments. It is concluded that JET is in a strong position to run future trace tritium and full DT discharges.« less

Tritium hydrology of the Mississippi River basin

USGS Publications Warehouse

Michel, R.L.

2004-01-01

In the early 1960s, the US Geological Survey began routinely analysing river water samples for tritium concentrations at locations within the Mississippi River basin. The sites included the main stem of the Mississippi River (at Luling Ferry, Louisiana), and three of its major tributaries, the Ohio River (at Markland Dam, Kentucky), the upper Missouri River (at Nebraska City, Nebraska) and the Arkansas River (near Van Buren, Arkansas). The measurements cover the period during the peak of the bomb-produced tritium transient when tritium concentrations in precipitation rose above natural levels by two to three orders of magnitude. Using measurements of tritium concentrations in precipitation, a tritium input function was established for the river basins above the Ohio River, Missouri River and Arkansas River sampling locations. Owing to the extent of the basin above the Luling Ferry site, no input function was developed for that location. The input functions for the Ohio and Missouri Rivers were then used in a two-component mixing model to estimate residence times of water within these two basins. (The Arkansas River was not modelled because of extremely large yearly variations in flow during the peak of the tritium transient.) The two components used were: (i) recent precipitation (prompt outflow) and (ii) waters derived from the long-term groundwater reservoir of the basin. The tritium concentration of the second component is a function of the atmospheric input and the residence times of the groundwaters within the basin. Using yearly time periods, the parameters of the model were varied until a best fit was obtained between modelled and measured tritium data. The results from the model indicate that about 40% of the flow in the Ohio River was from prompt outflow, as compared with 10% for the Missouri River. Mean residence times of 10 years were calculated for the groundwater component of the Ohio River versus 4 years for the Missouri River. The mass flux of tritium through the Mississippi Basin and its tributaries was calculated during the years that tritium measurements were made. The cumulative fluxes, calculated in grams of 3II were: (i) 160 g for the Ohio (1961-1986), (ii) 98 g for the upper Missouri (1963-1997), (iii) 30 g for the Arkansas (1961-1997) and (iv) 780 g for the Mississippi (1961-1997). Published in 2004 by John Wiley and Sons, Ltd.

Tritium, deuterium, and oxygen-18 in water collected from unsaturated sediments near a low-level radioactive-waste burial site south of Beatty, Nevada

USGS Publications Warehouse

Prudic, David E.; Stonestrom, David A.; Striegl, Robert G.

1997-01-01

Pore water was extracted in March 1996 from cores collected from test holes UZB-1 and UZB-2 drilled November 1992 and September 1993, respectively, in the Amargosa Desert south of Beatty, Nevada. The test holes are part of a study to determine factors affecting water and gas movement through unsaturated sediments. The holes are about 100 meters south of the southwest corner of the fence enclosing a commercial burial area for low-level radioactive waste. Water vapor collected from test hole UZB-2 in April 1994 and July 1995 had tritium concentrations greater than would be expected from atmospheric deposition. An apparatus was built in which pore water was extracted by cryodistillation from the previously obtained core samples. The extracted core water was analyzed for the radioactive isotope tritium and for the stable isotopes deuterium (D) and oxygen-18 (18O). The isotopic composition of core water was compared with that of water vapor previously collected from air ports in test hole UZB-2 and to additional samples collected during May 1996. Core water becomes increasingly depleted in D and 18O from the land surface to a depth of 30 meters, indicating that net evaporation of water is occurring near the land surface. Below a depth of 30 meters the stable-isotopic composition of core water becomes nearly constant and roughly equal to that of ground water. The stable isotopes plot on an evaporation trend. The source of the partly evaporated water could be either ground water or past precipitation having the same average isotopic composition as ground water but not modern precipitation, based on 18 months of record. Profiles of D and 18O in water vapor roughly parallel those in core water. The stable isotopes of core water appear to be in isotopic equilibrium with water vapor from UZB-2 when temperature-dependent fractionation is considered. The data are consistent with the hypothesis of evaporative discharge of ground water at the land surface. The concentration of tritium in core water from depths less than 50 meters was higher than that of present-day atmospheric air, indicating that elevated tritium concentrations preceded the drilling. The concentrations of tritium in core water from the deepest sample (85 meters) and in UZB-2 groundwater (110 meters) were below detection. Thus, tritium in the unsaturated zone is not being introduced through ground water. The shape of the tritium profile for core water was similar to the shape of the tritium profile for water vapor collected April 1994, except that concentrations were consistently lower in core water than in water vapor. Tritium concentrations in water vapor increased from April 1994 to May 1996. Similar to the stable isotopes, the highest tritium concentrations were measured at shallow depths. Concentrations of tritium in water vapor during core collection were estimated assuming isotopic equilibrium with core water. The computed concentrations for November 1992 and September 1993 form consistent temporal trends with subsequent tritium concentrations in water vapor collected April 1994, July 1995, and May 1996. Observations of a bimodal distribution of tritium, in which the highest concentrations are in a gravel layer at a depth of 1-2 meters, indicate lateral migration of tritium through the vicinity of UZB-2.

Control Room at the NACA’s Rocket Engine Test Facility

NASA Image and Video Library

1957-05-21

Test engineers monitor an engine firing from the control room of the Rocket Engine Test Facility at the National Advisory Committee for Aeronautics (NACA) Lewis Flight Propulsion Laboratory. The Rocket Engine Test Facility, built in the early 1950s, had a rocket stand designed to evaluate high-energy propellants and rocket engine designs. The facility was used to study numerous different types of rocket engines including the Pratt and Whitney RL-10 engine for the Centaur rocket and Rocketdyne’s F-1 and J-2 engines for the Saturn rockets. The Rocket Engine Test Facility was built in a ravine at the far end of the laboratory because of its use of the dangerous propellants such as liquid hydrogen and liquid fluorine. The control room was located in a building 1,600 feet north of the test stand to protect the engineers running the tests. The main control and instrument consoles were centrally located in the control room and surrounded by boards controlling and monitoring the major valves, pumps, motors, and actuators. A camera system at the test stand allowed the operators to view the tests, but the researchers were reliant on data recording equipment, sensors, and other devices to provide test data. The facility’s control room was upgraded several times over the years. Programmable logic controllers replaced the electro-mechanical control devices. The new controllers were programed to operate the valves and actuators controlling the fuel, oxidant, and ignition sequence according to a predetermined time schedule.

Test results and facility description for a 40-kilowatt stirling engine

NASA Technical Reports Server (NTRS)

Kelm, G. G.; Cairelli, J. E.; Walter, R. J.

1981-01-01

A 40 kilowatt Stirling engine, its test support facilities, and the experimental procedures used for these tests are described. Operating experience with the engine is discussed, and some initial test results are presented

Providing security for automated process control systems at hydropower engineering facilities

NASA Astrophysics Data System (ADS)

Vasiliev, Y. S.; Zegzhda, P. D.; Zegzhda, D. P.

2016-12-01

This article suggests the concept of a cyberphysical system to manage computer security of automated process control systems at hydropower engineering facilities. According to the authors, this system consists of a set of information processing tools and computer-controlled physical devices. Examples of cyber attacks on power engineering facilities are provided, and a strategy of improving cybersecurity of hydropower engineering systems is suggested. The architecture of the multilevel protection of the automated process control system (APCS) of power engineering facilities is given, including security systems, control systems, access control, encryption, secure virtual private network of subsystems for monitoring and analysis of security events. The distinctive aspect of the approach is consideration of interrelations and cyber threats, arising when SCADA is integrated with the unified enterprise information system.

Steve Nixon | NREL

Science.gov Websites

NREL. Steve has an extensive background in facilities engineering, facilities management, and Energy Manager, and a Project Management Professional. Prior to joining NREL, Steve was the Facilities manufacturing engineering, business application programming, and business process management positions

7. Photocopy of engineering drawing. AETR DIGS FACILITY THEODOLITE AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

7. Photocopy of engineering drawing. AETR DIGS FACILITY THEODOLITE AND PRISM SHELTER: ELEVATIONS, FLOOR AND FOUNDATION PLANS, 1971. - Cape Canaveral Air Station, Launch Complex 17, Facility 28413, East end of Lighthouse Road, Cape Canaveral, Brevard County, FL

Next Generation Gamma-Ray Cherenkov Detectors for the National Ignition Facility

DOE PAGES

Herrmann, Hans W.; Kim, Yong Ho; McEvoy, Aaron Matthew; ...

2016-10-19

The newest generation of Gas Cherenkov Detector (GCD-3) employed in Inertial Confinement Fusion experiments at the Omega Laser Facility has provided improved performance over previous generations. Comparison of reaction histories measured using two different deuterium-tritium fusion products, namely gamma rays using GCD and neutrons using Neutron Temporal Diagnostic (NTD), have provided added credibility to both techniques. GCD-3 is now being brought to the National Ignition Facility (NIF) to supplement the existing Gamma Reaction History (GRH-6m) located 6 m from target chamber center (TCC). Initially it will be located in a reentrant well located 3.9 m from TCC. Data from GCD-3more » will inform the design of a heavily-shielded “Super” GCD to be located as close as 20 cm from TCC. In conclusion, it will also provide a test-bed for faster optical detectors, potentially lowering the temporal resolution from the current ~100 ps state-of-the-art photomultiplier tubes (PMT) to ~10 ps Pulse Dilation PMT technology currently under development.« less

Tritium and radioactive carbon (14C) analyses of gas collected from unsaturated sediments next to a low-level radioactive-waste burial site south of Beatty, Nevada, April 1994 and July 1995

USGS Publications Warehouse

Prudic, David E.; Striegl, Robert G.

1995-01-01

Tritium activities in water vapor and radioactive carbon (14C) activities in carbon dioxide were determined in gas samples pumped from small-diameter air ports installed in a test hole within the unsaturated sediments next to a commercial burial site for low-level radioactive waste south of Beatty, Nevada. In April 1994, gas samples were collected from test hole UZB-2, which was drilled about 350 feet south of the southwest corner of the fence enclosing the burial site. The test hole is part of a study to determine the depth to which atmospheric air circulates through the unsaturated sediments at the desert site. Laboratory results completed in May 1995 show activities of tritium and 14C were greater than expected, with measured tritium in the water vapor as high as 762 tritium units at a depth of 79 feet and measured 14C in carbon dioxide as high as 1,700 percent modern carbon at a depth of 18 feet.In July 1995, the uppermost five air ports in test hole UZB-2 were resampled. In addition, water vapor was collected for tritium analyses at a distant test hole, and water vapor for tritium analyses and carbon dioxide for 14C analyses were collected from three depths at the research shaft about 200 feet north of test hole UZB-2, and at two shallow probes (depth of 5.5 feet) next to the fence enclosing the burial site. Analyses of samples collected in the upper 112 feet from test hole UZB-2 in July 1995 show the same distribution of tritium and 14C as analyses of samples collected in April 1994, except that activities were somewhat greater in July. The greatest activities of tritium and 14C were measured from a shallow probe next to the fence with activities of 29,400 tritium units and 517,000 percent modern carbon, respectively.

EXTRACTION OF FRACTURE-MECHANICS AND TRANSMISSION-ELECTRON-MICROSCOPY SAMPLES FROM TRITIUM-EXPOSED RESERVOIRS USING ELECTRIC-DISCHARGE MACHINING

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

Morgan, M; Ken Imrich, K; Michael Tosten, M

2006-08-31

The Enhanced Surveillance Campaign is funding a program to investigate tritium aging effects on the structural properties of tritium reservoir steels. The program is designed to investigate how the structural properties of reservoir steels change during tritium service and to examine the role of microstructure and reservoir manufacturing on tritium compatibility. New surveillance tests are also being developed that can better gauge the long-term effects of tritium and its radioactive decay product, helium-3, on the properties of reservoir steels. In order to conduct these investigations, three types of samples are needed from returned reservoirs: tensile, fracture mechanics, and transmission-electron microscopymore » (TEM). An earlier report demonstrated how the electric-discharge machining (EDM) technique can be used for cutting tensile samples from serial sections of a 3T reservoir and how yield strength, ultimate strength and elongation could be measured from those samples. In this report, EDM was used successfully to section sub-sized fracture-mechanics samples from the inner and outer walls of a 3T reservoir and TEM samples from serial sections of a 1M reservoir. This report fulfills the requirements for the FY06 Level 3 milestone, TSR 15.1 ''Cut Fracture-Mechanics Samples from Tritium-Exposed Reservoir'' and TSR 15.2 ''Cut Transmission-electron-microscopy foils from Tritium-Exposed Reservoir'' for the Enhance Surveillance Campaign (ESC). This was in support of ESC L2-1870 Milestone-''Provide aging and lifetime assessments of selected components and materials for multiple enduring stockpile systems''.« less

Convert Ten Foot Environmental Test Chamber into an Ion Engine Test Chamber

NASA Technical Reports Server (NTRS)

VanVelzer, Paul

2006-01-01

The 10 Foot Space Simulator at the Jet Propulsion Laboratory has been used for the last 40 years to test numerous spacecraft, including the Ranger series, several Mariner class, among many others and finally, the Spirit and Opportunity Mars Rovers. The request was made to convert this facility to an Ion Engine test facility, with a possible long term life test. The Ion engine was to propel the Prometheus spacecraft to Jupiter's moons. This paper discusses the challenges that were met, both from a procedural and physical standpoint. The converted facility must operate unattended, support a 30 Kw Ion Engine, operate economically, and be easily converted back to former operation as a spacecraft test facility.

Site Characterization Report (Building 202). Volume 2. Appendicies A-H.

DTIC Science & Technology

1996-04-01

Bionetics,Groundwater and Wells, Environmental Science and Engineering, Inc., Installation Assessment of ERADCOM Activities, Environmental Science and...Engineering, Inc., Plan for the Assessment of Contamination at Woodbridge Research Facility, Environmental Science and Engineering, Inc., Remedial...Action Plan for the Woodbridge Research Facility PCB Disposal Site, Environmental Science and Engineering, Inc., Remedial Investigation and

6. Photocopy of engineering drawing. AETR DIGS FACILITY THEODOLITE AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

6. Photocopy of engineering drawing. AETR DIGS FACILITY THEODOLITE AND PRISM SHELTER: MONUMENT LOCATION AND LINE-OF-SIGHT PLAN, 1972. - Cape Canaveral Air Station, Launch Complex 17, Facility 28413, East end of Lighthouse Road, Cape Canaveral, Brevard County, FL

Evaluating All-Metal Valves for Use in a Tritium Environment

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

Houk, L.; Payton, A.

In the tritium gas processing system, it is desired to minimize polymer components due to their degradation from tritium exposure (beta decay). One source of polymers in the tritium process is valve components. A vendor has been identified that manufactures a valve that is marketed as being made from all-metal construction. This manufacturer, Ham-Let Group, manufactures a diaphragm valve (3LE series) that claims to be made entirely of metal. SRNL procured twelve (12) Ham-Let diaphragm valves for characterization and evaluation. The characterization tests include identification of the maximum pressure of these valves by performing pressure and burst tests. Leak testsmore » were performed to ensure the valves do not exceed the acceptable leak rate for tritium service. These valves were then cycled in a nitrogen gas and/or vacuum environment to ensure they would be durable in a process environment. They were subsequently leak tested per ASTM protocol to ensure that the valves maintained their leak tight integrity. A detailed material analysis was also conducted to determine hydrogen and tritium compatibility.« less

Tritium Effects on Fracture Toughness of Stainless Steel Weldments

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

MORGAN, MICHAEL; CHAPMAN, G. K.; TOSTEN, M. H.

2005-05-12

The effects of tritium on the fracture toughness properties of Type 304L and Type 21-6-9 stainless steel weldments were measured. Weldments were tritium-charged-and-aged and then tested in order to measure the effect of the increasing decay helium content on toughness. The results were compared to uncharged and hydrogen-charged samples. For unexposed weldments having 8-12 volume percent retained delta ferrite, fracture toughness was higher than base metal toughness. At higher levels of weld ferrite, the fracture toughness decreased to values below that of the base metal. Hydrogen-charged and tritium-charged weldments had lower toughness values than similarly charged base metals and toughnessmore » decreased further with increasing weld ferrite content. The effect of decay helium content was inconclusive because of tritium off-gassing losses during handling, storage and testing. Fracture modes were dominated by the dimpled rupture process in unexposed weldments. In hydrogen and tritium-exposed weldments, the fracture modes depended on the weld ferrite content. At high ferrite contents, hydrogen-induced transgranular fracture of the weld ferrite phase was observed.« less

Ground test facility for SEI nuclear rocket engines

NASA Astrophysics Data System (ADS)

Harmon, Charles D.; Ottinger, Cathy A.; Sanchez, Lawrence C.; Shipers, Larry R.

1992-07-01

Nuclear (fission) thermal propulsion has been identified as a critical technology for a manned mission to Mars by the year 2019. Facilities are required that will support ground tests to qualify the nuclear rocket engine design, which must support a realistic thermal and neutronic environment in which the fuel elements will operate at a fraction of the power for a flight weight reactor/engine. This paper describes the design of a fuel element ground test facility, with a strong emphasis on safety and economy. The details of major structures and support systems of the facility are discussed, and a design diagram of the test facility structures is presented.

Langley Mach 4 scramjet test facility

NASA Technical Reports Server (NTRS)

Andrews, E. H., Jr.; Torrence, M. G.; Anderson, G. Y.; Northam, G. B.; Mackley, E. A.

1985-01-01

An engine test facility was constructed at the NASA Langley Research Center in support of a supersonic combustion ramjet (scramjet) technology development program. Hydrogen combustion in air with oxygen replenishment provides simulated air at Mach 4 flight velocity, pressure, and true total temperature for an altitude range from 57,000 to 86,000 feet. A facility nozzle with a 13 in square exit produces a Mach 3.5 free jet flow for engine propulsion tests. The facility is described and calibration results are presented which demonstrate the suitability of the test flow for conducting scramjet engine research.

Tritium in water vapor in the shallow unsaturated zone at the Amargosa Desert Research Site

USGS Publications Warehouse

Healy, Richard W.; Striegl, Robert G.; Michel, Robert L.; Prudic, David E.; Andraski, Brian J.; Morganwalp, David W.; Buxton, Herbert T.

1999-01-01

Samples of water vapor in soil gas were obtained at the U.S. Geological Survey's Amargosa Desert Research Site in 1997 and 1998 from a depth of 1.5 m (meters) within a 300 m by 300 m grid that lies immediately to the south and west of a low-level radioactive-waste disposal site. The gas samples were analyzed for tritium. Fifty-eight samples were collected in May 1997; 61 samples were collected in June 1998. Measured tritium concentrations ranged from 16 ± 9 TU (tritium units) to 36,900 ± 300 TU in 1997, and from 6 ± 6 TU to 37,360 ± 450 TU in 1998. Concentrations decreased from northeast to southwest across the grid. In general, there was very little difference in tritium concentrations between the two sampling periods.

A laboratory information management system for the analysis of tritium (3H) in environmental waters.

PubMed

Belachew, Dagnachew Legesse; Terzer-Wassmuth, Stefan; Wassenaar, Leonard I; Klaus, Philipp M; Copia, Lorenzo; Araguás, Luis J Araguás; Aggarwal, Pradeep

2018-07-01

Accurate and precise measurements of low levels of tritium ( 3 H) in environmental waters are difficult to attain due to complex steps of sample preparation, electrolytic enrichment, liquid scintillation decay counting, and extensive data processing. We present a Microsoft Access™ relational database application, TRIMS (Tritium Information Management System) to assist with sample and data processing of tritium analysis by managing the processes from sample registration and analysis to reporting and archiving. A complete uncertainty propagation algorithm ensures tritium results are reported with robust uncertainty metrics. TRIMS will help to increase laboratory productivity and improve the accuracy and precision of 3 H assays. The software supports several enrichment protocols and LSC counter types. TRIMS is available for download at no cost from the IAEA at www.iaea.org/water. Copyright © 2018 Elsevier Ltd. All rights reserved.

Direct LiT Electrolysis in a Metallic Fusion Blanket

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

Olson, Luke

2016-09-30

A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium formore » the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.« less

Studying the spatial organization of membrane proteins by means of tritium stratigraphy: bacteriorhodopsin in purple membrane.

PubMed

Shishkov, A V; Ksenofontov, A L; Bogacheva, E N; Kordyukova, L V; Badun, G A; Alekseevsky, A V; Tsetlin, V I; Baratova, L A

2002-05-15

The topography of bacteriorhodopsin (bR) in situ was earlier studied by using the tritium bombardment approach [Eur. J. Biochem. 178 (1988) 123]. Now, having the X-ray crystallography data of bR at atom resolution [Proc. Natl. Acad. Sci. 95 (1998) 11673], we estimated the influence of membrane environment (lipid and protein) on tritium incorporation into amino acid residues forming transmembrane helices. We have determined the tritium flux attenuation coefficients for residues 10-29 of helix A. They turned out to be low (0.04+/-0.02 A(-1)) for residues adjacent to the lipid matrix, and almost fourfold higher (0.15+/-0.05 A(-1)) for those oriented to the neighboring transmembrane helices. We believe that tritium incorporation data could help modeling transmembrane segment arrangement in the membrane.

The Concentration Of Tritium In Urine And Internal Radiation Dose Estimation Of PTNBR Radiation Workers

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

Tjahaja, Poppy Intan; Sukmabuana, Putu; Aisyah, Neneng Nur

2010-12-23

The operation of Triga 2000 reactor in Nuclear Technology Center for Materials and Radiometry (PTNBR BATAN) normally produce tritium radionuclide which is the activation product of deuterium atom in reactor primary cooling water. According to previous monitoring, tritium was detected with the concentration of 8.236{+-}0.677 kBq/L and 1.704{+-}0.046 Bq/L in the primary cooling water and in reactor hall air, respectively. The tritium in reactor hall air chronically can be inhaled by the workers. In this research, tritium content in radiation workers' urine was determined to estimate the internal radiation doses received by the workers. About 50-100 mL of urine samplesmore » were collected from 48 PTNBR workers that is classified as 24 radiation workers and 24 administration staffs as a control. Urine samples of 25 mL were then prepared by active charcoal and KMnO{sub 4} addition and followed with complete distillation. The 2 mL of distillate was added with 13 mL scintillator, shaked vigorously and remained in cool and dark condition for about 24 hours. The tritium in the samples was then measured using liquid scintillation counter (LSC) for 1 hour. From the measurement results it was obtained that the tritium concentration in the urine of radiation workers were in the range of not detected and 5.191 Bq/mL, whereas in the administration staffs the concentration were between not detected and 4.607 Bq/mL. Internally radiation doses were calculated using the tritium concentration data, and it was found the averages about 0.602 {mu}Sv/year and 0.532 {mu}Sv/year for radiation workers and administration staffs, respectively. The doses received by the workers were lower than that of the permissible doses from tritium, i.e. 40 {mu}Sv/year.« less

Coal-Oil Mixtures Problems and Opportunities,

DTIC Science & Technology

1982-01-15

Ernest C. Friedrich Ashland Oil, Inc. New Richmond, Ohio Cleveland, Ohio Florida Power Corporation American Refining Co., Inc. 3201 34th St. South...Room 1A 518, The Pentagon USAF Institute of Technology Washington, DC 20310 AFIT/DED Wright Patterson AFB, OH 45433 Commander-in-Chief USA, Europe...Engineer Facilities Engineer Fort A P Hill Lone Star Army Ammunition Plant Bowling Green, VA 22427 Texarkana , TX 75501 Facilities Engineer Facilities

Scintillation imaging of tritium radioactivity distribution during tritiated thymidine uptake by PC12 cells using a melt-on scintillator.

PubMed

Irikura, Namiko; Miyoshi, Hirokazu; Shinohara, Yasuo

2017-02-01

A scintillation image of tritium fixed in a melt-on scintillator was obtained using a charged-coupled device (CCD) imager, and a linear relationship was observed between the intensity of the scintillation image and the radioactivity of tritium. In a [ 3 H]thymidine uptake experiment, a linear correlation between the intensity of the CCD image and the dilution ratio of cells was confirmed. Scintillation imaging has the potential for use in direct observation of tritium radioactivity distribution. Copyright © 2016 Elsevier Ltd. All rights reserved.

Radioactive emission data from Canadian nuclear generating stations, 1988 to 1997. Report number INFO-0210/Rev.8

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

NONE

1998-12-31

This edition incorporates histograms for each nuclear generating station (NGS) displaying the annual gaseous emissions containing tritium, in the form of tritium oxide, noble gases, iodine-131, and radioactive particulates, as well as the annual liquid emissions containing tritium, in the form of tritiated water, and gross beta-gamma activity. For Pickering NGS A and Gentilly 2, annual emissions of carbon-14 are depicted; and for Darlington NGS A, airborne emissions of elemental tritium since 1988 are shown. In each case, the emission data are compared to the derived emission limits.

Synthesis of labeled compounds using recovered tritium from expired beta light sources

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

Matei, L.; Postolache, C.; Bubueanu, G.

2008-07-15

In this paper, the technological procedures for extracting tritium from beta light source are highlighted. The recovered tritium was used in the synthesis of organically labeled compounds and in the preparation of tritiated water (HTO) with high specific activity. Technological procedures for treatment of beta light sources consist of: envelope breaking into evacuated enclosure, the radioactive gaseous mixture pumping and its storage on metallic sodium. The mixtures of T{sub 2} and {sup 3}He were used in the synthesis of tritium labeled steroid hormones, nucleosides analogues and for the preparation of HTO with high radioactivity concentrations. (authors)

Analysis of fusion neutron spectral widths in high-foot implosions at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Grim, Gary; Caggiano, Joseph; Callahan, Debra; Casey, Daniel; Cerjan, Charles; Clark, Daniel; Tilo, Doeppner; Eckart, Mark; Field, John; Frenje, Lars; Gatu-Johnson, Maria; Hartouni, Edward; Hatarik, Robert; Hurricane, Omar; Kilkenny, Joseph; Knauer, James; Ma, Tammy; Mannion, Owen; Munro, David; Park, Hye-Sook; Sayre, Daniel; Spears, Brian; Yeamans, Charles

2015-11-01

We present the latest results of thermal temperature analyses of cryogenically layered deuterium-tritium implosions at the NIF using data from the ``High Foot'' campaign. Data from new analysis methods and interpreted in the context of new theoretical developments will be reported. These data will include DD and DT apparent ion temperatures, their uniformity with direction, inferred plasma thermal temperature, as well as the magnitude of non-thermal contributions to the spectral widths. Work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344.

Simulation and assessment of ion kinetic effects in a direct-drive capsule implosion experiment

DOE PAGES

Le, Ari Yitzchak; Kwan, Thomas J. T.; Schmitt, Mark J.; ...

2016-10-24

The first simulations employing a kinetic treatment of both fuel and shell ions to model inertial confinement fusion experiments are presented, including results showing the importance of kinetic physics processes in altering fusion burn. A pair of direct drive capsule implosions performed at the OMEGA facility with two different gas fills of deuterium, tritium, and helium-3 are analyzed. During implosion shock convergence, highly non-Maxwellian ion velocity distributions and separations in the density and temperature amongst the ion species are observed. Finally, diffusion of fuel into the capsule shell is identified as a principal process that degrades fusion burn performance.

Current Trends of Blanket Research and Deveopment in Japan 4.Blanket Technology Development Using ITER for Demonstration and Commercial Fusion Power Plant

NASA Astrophysics Data System (ADS)

Akiba, Masato; Jitsukawa, Shiroh; Muroga, Takeo

This paper describes the status of blanket technology and material development for fusion power demonstration plants and commercial fusion plants. In particular, the ITER Test Blanket Module, IFMIF, JAERI/DOE HFIR and JUPITER-II projects are highlighted, which have the important role to develop these technology. The ITER Test Blanket Module project has been conducted to demonstrate tritium breeding and power generation using test blanket modules, which will be installed into the ITER facility. For structural material development, the present research status is overviewed on reduced activation ferritic steel, vanadium alloys, and SiC/SiC composites.

Response of LaBr3(Ce) scintillators to 2.5 MeV fusion neutrons.

PubMed

Cazzaniga, C; Nocente, M; Tardocchi, M; Croci, G; Giacomelli, L; Angelone, M; Pillon, M; Villari, S; Weller, A; Petrizzi, L; Gorini, G

2013-12-01

Measurements of the response of LaBr3(Ce) to 2.5 MeV neutrons have been carried out at the Frascati Neutron Generator and at tokamak facilities with deuterium plasmas. The observed spectrum has been interpreted by means of a Monte Carlo model. It is found that the main contributor to the measured response is neutron inelastic scattering on (79)Br, (81)Br, and (139)La. An extrapolation of the count rate response to 14 MeV neutrons from deuterium-tritium plasmas is also presented. The results are of relevance for the design of γ-ray diagnostics of fusion burning plasmas.

Tritium Records to Trace Stratospheric Moisture Inputs in Antarctica

NASA Astrophysics Data System (ADS)

Fourré, E.; Landais, A.; Cauquoin, A.; Jean-Baptiste, P.; Lipenkov, V.; Petit, J.-R.

2018-03-01

Better assessing the dynamic of stratosphere-troposphere exchange is a key point to improve our understanding of the climate dynamic in the East Antarctica Plateau, a region where stratospheric inputs are expected to be important. Although tritium (3H or T), a nuclide naturally produced mainly in the stratosphere and rapidly entering the water cycle as HTO, seems a first-rate tracer to study these processes, tritium data are very sparse in this region. We present the first high-resolution measurements of tritium concentration over the last 50 years in three snow pits drilled at the Vostok station. Natural variability of the tritium records reveals two prominent frequencies, one at about 10 years (to be related to the solar Schwabe cycles) and the other one at a shorter periodicity: despite dating uncertainty at this short scale, a good correlation is observed between 3H and Na+ and an anticorrelation between 3H and δ18O measured on an individual pit. The outputs from the LMDZ Atmospheric General Circulation Model including stable water isotopes and tritium show the same 3H-δ18O anticorrelation and allow further investigation on the associated mechanism. At the interannual scale, the modeled 3H variability matches well with the Southern Annular Mode index. At the seasonal scale, we show that modeled stratospheric tritium inputs in the troposphere are favored in winter cold and dry conditions.

Observation of tritium in gas/plasma loaded titanium samples

NASA Astrophysics Data System (ADS)

Srinivasan, M.; Shyam, A.; Kaushik, T. C.; Rout, R. K.; Kulkarni, L. V.; Krishnan, M. S.; Malhotra, S. K.; Nagvenkar, V. G.; Iyengar, P. K.

1991-05-01

The observation of significant neutron yield from gas loaded titanium samples at Frascati in April 1989 opened up an alternate pathway to the investigation of anomalous nuclear phenomena in deuterium/solid systems, complimenting the electrolytic approach. Since then at least six different groups have successfully measured burst neutron emission from deuterated titanium shavings following the Frascati methodology, the special feature of which was the use of liquid nitrogen to create repeated thermal cycles resulting in the production of non-equilibrium conditions in the deuterated samples. At Trombay several variations of the gas loading procedure have been investigated including induction heating of single machined titanium targets in a glass chamber as well as use of a plasma focus device for deuteriding its central titanium electrode. Stemming from earlier observations both at BARC and elsewhere that tritium yield is ≂108 times higher than neutron output in cold fusion experiments, we have channelised our efforts to the search for tritium rather than neutrons. The presence of tritium in a variety gas/plasma loaded titanium samples has been established successfully through a direct measurement of the radiations emitted as a result of tritium decay, in contradistinction to other groups who have looked for tritium in the extracted gases. In some samples we have thus observed tritium levels of over 10 MBq with a corresponding (t/d) ratio of ≳10-5.

Chromosome aberrations in workers occupationally exposed to tritium.

PubMed

Tawn, E Janet; Curwen, Gillian B; Riddell, Anthony E

2018-06-01

This paper reports the findings of an historical chromosome analysis for unstable aberrations, undertaken on 34 nuclear workers with monitored exposure to tritium. The mean recorded β-particle dose from tritium was 9.33 mGy (range 0.25-79.71 mGy) and the mean occupational dose from external, mainly γ-ray, irradiation was 1.94 mGy (range 0.00-7.71 mGy). The dicentric frequency of 1.91 ± 0.53 × 10 -3 per cell was significantly raised, in comparison with that of 0.61 ± 0.30 × 10 -3 per cell for a group of 66 comparable worker controls unexposed to occupational radiation. The frequency of total aberrations was also significantly higher in the tritium workers. Comparisons with in vitro studies indicate that at these dose levels an increase in aberration frequency is not expected. However, the available historical tritium dose records were produced for the purposes of radiological protection and based on a methodology that has since been updated, so tritium doses are subject to considerable uncertainty. It is therefore recommended that, if possible, tritium doses are reassessed using information on historical recording practices in combination with current dosimetry methodology, and that further chromosome studies are undertaken using modern FISH techniques to establish stable aberration frequencies, as these will provide information on a cumulative biological effect.

Vanadium hydride deuterium-tritium generator

DOEpatents

Christensen, Leslie D.

1982-01-01

A pressure controlled vanadium hydride gas generator to provide deuterium-tritium gas in a series of pressure increments. A high pressure chamber filled with vanadium-deuterium-tritium hydride is surrounded by a heater which controls the hydride temperature. The heater is actuated by a power controller which responds to the difference signal between the actual pressure signal and a programmed pressure signal.

Tritium labeling of organic compounds deposited on porous structures

DOEpatents

Ehrenkaufer, Richard L. E.; Wolf, Alfred P.; Hembree, Wylie C.

1979-01-01

An improved process for labeling organic compounds with tritium is carried out by depositing the selected compound on the extensive surface of a porous structure such as a membrane filter and exposing the membrane containing the compound to tritium gas activated by the microwave discharge technique. The labeled compound is then recovered from the porous structure.

12. Historic plot plan and drawings index for rocket engine ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

12. Historic plot plan and drawings index for rocket engine test facility, June 28, 1956. NASA GRC drawing number CE-101810. On file at NASA Glenn Research Center. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

5. Historic photo of scale model of rocket engine test ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

5. Historic photo of scale model of rocket engine test facility, June 18, 1957. On file at NASA Plumbrook Research Center, Sandusky, Ohio. NASA GRC photo number C-45264. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 4: Supplementary engineering data

NASA Astrophysics Data System (ADS)

1981-09-01

The reference conceptual design of the Magnetohydrodynamic Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD is summarized. Main elements of the design are identified and explained, and the rationale behind them is reviewed. Major systems and plant facilities are listed and discussed. Construction cost and schedule estimates, and identification of engineering issues that should be reexamined are also given. The latest (1980-1981) information from the MHD technology program are integrated with the elements of a conventional steam power electric generating plant. Supplementary Engineering Data (Issues, Background, Performance Assurance Plan, Design Details, System Design Descriptions and Related Drawings) is presented.

Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 4: Supplementary engineering data

NASA Technical Reports Server (NTRS)

1981-01-01

The reference conceptual design of the Magnetohydrodynamic Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD is summarized. Main elements of the design are identified and explained, and the rationale behind them is reviewed. Major systems and plant facilities are listed and discussed. Construction cost and schedule estimates, and identification of engineering issues that should be reexamined are also given. The latest (1980-1981) information from the MHD technology program are integrated with the elements of a conventional steam power electric generating plant. Supplementary Engineering Data (Issues, Background, Performance Assurance Plan, Design Details, System Design Descriptions and Related Drawings) is presented.

The Advanced Monitoring Systems Initiative--Performance Monitoring for DOE Environmental Remediation and Contaminant Containment

NASA Astrophysics Data System (ADS)

Haas, W. J.; Venedam, R. J.; Lohrstorfer, C. F.; Weeks, S. J.

2005-05-01

The Advanced Monitoring System Initiative (AMSI) is a new approach to accelerate the development and application of advanced sensors and monitoring systems in support of Department of Energy needs in monitoring the performance of environmental remediation and contaminant containment activities. The Nevada Site Office of the National Nuclear Security Administration (NNSA) and Bechtel Nevada manage AMSI, with funding provided by the DOE Office of Environmental Management (DOE EM). AMSI has easy access to unique facilities and capabilities available at the Nevada Test Site (NTS), including the Hazardous Materials (HazMat) Spill Center, a one-of-a-kind facility built and permitted for releases of hazardous materials for training purposes, field-test detection, plume dispersion experimentation, and equipment and materials testing under controlled conditions. AMSI also has easy access to the facilities and considerable capabilities of the DOE and NNSA National Laboratories, the Special Technologies Laboratory, Remote Sensing Laboratory, Desert Research Institute, and Nevada Universities. AMSI provides rapid prototyping, systems integration, and field-testing, including assistance during initial site deployment. The emphasis is on application. Important features of the AMSI approach are: (1) customer investment, involvement and commitment to use - including definition of needs, desired mode of operation, and performance requirements; and (2) employment of a complete systems engineering approach, which allows the developer to focus maximum attention on the essential new sensing element or elements while AMSI assumes principal responsibility for infrastructure support elements such as power, packaging, and general data acquisition, control, communication, visualization and analysis software for support of decisions. This presentation describes: (1) the needs for sensors and performance monitoring for environmental systems as seen by the DOE Long Term Stewardship Science and Technology Roadmap and the Long Term Monitoring Sensors and Analytical Methods Workshop, and (2) AMSI operating characteristics and progress in addressing those needs. Topics addressed will include: vadose zone and groundwater tritium monitoring, a wireless moisture monitoring system, Cr(VI) and CCl4 monitoring using a commercially available "universal sensor platform", strontium-90 and technetium-99 monitoring, and area chemical monitoring using an array of multi-chemical sensors.

Residence times in river basins as determined by analysis of long-term tritium records

USGS Publications Warehouse

Michel, R.L.

1992-01-01

The US Geological Survey has maintained a network of stations to collect samples for the measurement of tritium concentrations in precipitation and streamflow since the early 1960s. Tritium data from outflow waters of river basins draining 4500-75000 km2 are used to determine average residence times of water within the basins. The basins studied are the Colorado River above Cisco, Utah; the Kissimmee River above Lake Okeechobee, Florida; the Mississippi River above Anoka, Minnesota; the Neuse River above Streets Ferry Bridge near Vanceboro, North Carolina; the Potomac River above Point of Rocks, Maryland; the Sacramento River above Sacramento, California; the Susquehanna River above Harrisburg, Pennsylvania. The basins are modeled with the assumption that the outflow in the river comes from two sources-prompt (within-year) runoff from precipitation, and flow from the long-term reservoirs of the basin. Tritium concentration in the outflow water of the basin is dependent on three factors: (1) tritium concentration in runoff from the long-term reservoir, which depends on the residence time for the reservoir and historical tritium concentrations in precipitation; (2) tritium concentrations in precipitation (the within-year runoff component); (3) relative contributions of flow from the long-term and within-year components. Predicted tritium concentrations for the outflow water in the river basins were calculated for different residence times and for different relative contributions from the two reservoirs. A box model was used to calculate tritium concentrations in the long-term reservoir. Calculated values of outflow tritium concentrations for the basin were regressed against the measured data to obtain a slope as close as possible to 1. These regressions assumed an intercept of zero and were carried out for different values of residence time and reservoir contribution to maximize the fit of modeled versus actual data for all the above rivers. The final slopes of the fitted regression lines ranged from 0.95 to 1.01 (correlation coefficient > 0.96) for the basins studied. Values for the residence time of waters within the basins and average relative contributions of the within-year and long-term reservoirs to outflow were obtained. Values for river basin residence times ranged from 2 years for the Kissimmee River basin to 20 years for the Potomac River basin. The residence times indicate the time scale in which the basin responds to anthropogenic inputs. The modeled tritium concentrations for the basins also furnish input data for urban and agricultural settings where these river waters are used. ?? 1992.

33 CFR 143.120 - Floating OCS facilities.

Code of Federal Regulations, 2010 CFR

2010-07-01

...) OUTER CONTINENTAL SHELF ACTIVITIES DESIGN AND EQUIPMENT OCS Facilities § 143.120 Floating OCS facilities... (Marine Engineering) and J (Electrical Engineering) of 46 CFR chapter I and 46 CFR part 108 (Design and Equipment). Where unusual design or equipment needs make compliance impracticable, alternative proposals...

Tritium as tracer of groundwater pollution extension: case study of Andralanitra landfill site, Antananarivo-Madagascar

NASA Astrophysics Data System (ADS)

Ramaroson, Voahirana; Rakotomalala, Christian Ulrich; Rajaobelison, Joel; Fareze, Lahimamy Paul; Razafitsalama, Falintsoa A.; Rasolofonirina, Mamiseheno

2018-05-01

This study aims to understand the extension of groundwater pollution downstream of a landfill, Andralanitra-Antananarivo-Madagascar. Twenty-one samples, composed of dug well waters, spring waters, river, and lake, were measured in stable isotopes ( δ 2H, δ 18O) and tritium. Results showed that only two dug well waters, collected at the immediate vicinity of the landfill, have high tritium activities (22.82 TU and 10.43 TU), probably of artificial origin. Both upstream and further downstream of the landfill, tritium activities represent natural source, with values varying from 0.17 TU to 1.46 TU upstream and from 0.88 TU to 1.88 TU further downstream. Stable isotope data suggest that recharge occurs through infiltration of slightly evaporated rainfall. Using the radioactive decay equation, the calculated tracer ages related to two recent ground water samples collected down gradient of the landfill lay between [8-15] years and [4-7] years, taking into account the uncertainty of tritium measurements. For the calculation, a value of 2.36 TU was taken as A o. The latter was estimated based on similarity between stable isotope compositions of nearby spring and dug well waters as well as tritium activities of the local precipitation. Calculation of the tritium activities from the contaminated water point having 22.82 TU to further downstream using the calculated tracer ages showed values of one order of magnitude higher than the measured values. The absence of hydrological connection from the contaminated water point to further downstream the landfill would explain the lower tritium activities measured. Groundwater pollution seems to be limited to the closest proximity of the landfill.

Tritium power source for long-lived sensors

NASA Astrophysics Data System (ADS)

Litz, M. S.; Katsis, D. C.; Russo, J. A.; Carroll, J. J.

2014-06-01

A tritium-based indirect converting photovoltaic (PV) power source has been designed and prototyped as a long-lived (~15 years) power source for sensor networks. Tritium is a biologically benign beta emitter and low-cost isotope acquired from commercial vendors for this purpose. The power source combines tritium encapsulated with a radioluminescent phosphor coupled to a commercial PV cell. The tritium, phosphor, and PV components are packaged inside a BA5590-style military-model enclosure. The package has been approved by the nuclear regulatory commission (NRC) for use by DOD. The power source is designed to produce 100μW electrical power for an unattended radiation sensor (scintillator and avalanche photodiode) that can detect a 20 μCi source of 137Cs at three meters. This beta emitting indirect photon conversion design is presented as step towards the development of practical, logistically acceptable, lowcost long-lived compact power sources for unattended sensor applications in battlefield awareness and environmental detection.

Tritium target manufacturing for use in accelerators

NASA Astrophysics Data System (ADS)

Bach, P.; Monnin, C.; Van Rompay, M.; Ballanger, A.

2001-07-01

As a neutron tube manufacturer, SODERN is now in charge of manufacturing tritium targets for accelerators, in cooperation with CEA/DAM/DTMN in Valduc. Specific deuterium and tritium targets are manufactured on request, according to the requirements of the users, starting from titanium target on copper substrate, and going to more sophisticated devices. A wide range of possible uses is covered, including thin targets for neutron calibration, thick targets with controlled loading of deuterium and tritium, rotating targets for higher lifetimes, or large size rotating targets for accelerators used in boron neutron therapy. Activity of targets lies in the 1 to 1000 Curie, diameter of targets being up to 30 cm. Special targets are also considered, including surface layer targets for lowering tritium desorption under irradiation, or those made from different kinds of occluders such as titanium, zirconium, erbium, scandium, with different substrates. It is then possible to optimize either neutron output, or lifetime and stability, or thermal behavior.

Tritium saturation in plasma-facing materials surfaces1

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Anderl, Robert A.; Causey, Rion A.; Federici, Gianfranco; Haasz, Anthony A.; Pawelko, Robert J.

1998-10-01

Plasma-facing components in the International Thermonuclear Experimental Reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10 20-10 23 particles/m 2s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments.

Measurement of helium isotopes in soil gas as an indicator of tritium groundwater contamination.

PubMed

Olsen, Khris B; Dresel, P Evan; Evans, John C; McMahon, William J; Poreda, Robert

2006-05-01

The focus of this study was to define the shape and extent of tritium groundwater contamination emanating from a legacy burial ground and to identify vadose zone sources of tritium using helium isotopes (3He and 4He) in soil gas. Helium isotopes were measured in soil-gas samples collected from 70 sampling points around the perimeter and downgradient of a burial ground that contains buried radioactive solid waste. The soil-gas samples were analyzed for helium isotopes using rare gas mass spectrometry. 3He/4He ratios, reported as normalized to the air ratio (RA), were used to locate the tritium groundwater plume emanating from the burial ground. The 3He (excess) suggested that the general location of the tritium source is within the burial ground. This study clearly demonstrated the efficacy of the 3He method for application to similar sites elsewhere within the DOE weapons complex.

Conceptual design of the MHD Engineering Test Facility

NASA Technical Reports Server (NTRS)

Bents, D. J.; Bercaw, R. W.; Burkhart, J. A.; Mroz, T. S.; Rigo, H. S.; Pearson, C. V.; Warinner, D. K.; Hatch, A. M.; Borden, M.; Giza, D. A.

1981-01-01

The reference conceptual design of the MHD engineering test facility, a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commerical feasibility of open cycle MHD is summarized. Main elements of the design are identified and explained, and the rationale behind them is reviewed. Major systems and plant facilities are listed and discussed. Construction cost and schedule estimates are included and the engineering issues that should be reexamined are identified.

Neutronics and activation analysis of lithium-based ternary alloys in IFE blankets

DOE PAGES

Jolodosky, Alejandra; Kramer, Kevin; Meier, Wayne; ...

2016-04-09

Here we report that an attractive feature of using liquid lithium as the breeder and coolant in fusion blankets is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. The Lawrence Livermore National Laboratory is carrying an effort to develop a lithium-based alloy that maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) and at the same time reduces overall flammability concerns. This study evaluates the neutronics performance of lithium-based alloys inmore » the blanket of an inertial fusion energy chamber in order to inform such development. 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and the fusion energy multiplication factor (EMF). It was found that elements that exhibit low absorption cross sections and higher q-values such as lead, tin, and strontium, perform well with those that have high neutron multiplication such as lead and bismuth. These elements meet TBR constrains ranging from 1.02 to 1.1. However, most alloys do not reach EMFs greater than 1.15. Additionally, it was found that enriching lithium significantly increases the TBR and decreases the minimum lithium concentration by more than 60%. The amount of enrichment depends on how much total lithium is in the alloy to begin with. Alloys that performed well in the TBR and EMF calculations were considered for activation analysis. Activation simulations were executed with 50 years of irradiation and 300 years of cooling. It was discovered that bismuth is a poor choice due to achieving the highest decay heat, contact dose rates, and accident doses. In addition, it does not meet the waste disposal ratings (WDR). Some of the activation results for alloys with tin, zinc, and gallium were in the higher end and should be considered secondary to elements such as strontium and barium that had overall better results. The results of this study along with other considerations such as thermodynamics, and chemical reactivity will help down select a preferred lithium ternary alloy.« less

Area G Perimeter Surface-Soil and Single-Stage Water Sampling: Environmental Surveillance for Fiscal Years 1996 and 1997, Group ESH-19

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

Marquis Childs; Ron Conrad

1998-10-01

Area Gin Technical Area 54, has been the principal facility at Los Alamos National Laboratory for the storage and disposal of low-level, solid mixed, and transuranic radioactive waste since 1957. Soil samples were analyzed for tritium, isotopic plutonium, americium-241, and cesium-137. Thirteen metals-silver, arsenic, barium, beryllium, cadmium, chromium, mercury, nickel, lead, antimony, selenium, thallium and zinc-were analyzed on filtered-sediment fractions of the single-stage samples using standard analytical chemistry techniques. During the two years of sampling discussed in this report elevated levels of tritium (as high as 716,000 pCi/L) in soil were found for sampling sites adjacent to the tritium burialmore » shafts located on the south- central perimeter of Area G. Additionally, tritium concentrations in soil as high as 38,300 pCi/L were detected adjacent to the TRU pads in the northeast comer of Area G. Plutonium-238 activities in FY96 soils ranged from 0.001-2.866 pCi/g, with an average concentration of 0.336& 0.734 pCdg. Pu-238 activities in FY97 soils ranged from 0.002-4.890 pCi/g, with an average concentration of 0.437 & 0.928 pCdg. Pu-239 activities in FY96 soils ranged from 0.009 to 1.62 pCdg, with an average of 0.177- 0.297 pCdg. Pu-239 activities in FY97 soils ranged from 0.005 to 1.71 pCi/g, with an average of 0.290- 0.415 pCi/g. The locations of elevated plutonium readings were consistent with the history of plutonium disposal at Area G. The two areas of elevated Am-241 activity reflected the elevated activities found for plutonium, the average values for Am-241 on soils were 0.6-2.07 pCi/g, and 0.10-0.14 pCi/g respectively for samples collected in FY96 and FY97. CS-137 activities in soils had average values of 0.33 pCi/g, and 0.28 pCi/g respectively for samples collected in FY96 and 97. There was no perimeter area where soil concentrations of CS-137 were significantly elevated.« less

POLLUTION PREVENTION OPPORTUNITY ASSESSMENT OF THE U.S. ARMY CORPS OF ENGINEERS CIVIL WORKS FACILITIES

EPA Science Inventory

The Pollution Prevention Opportunity Assessments (PPOA) summarized here were conducted at the following representative Army Corps of Engineers (USAGE) Civil Works facilities: Pittsburgh Engineering Warehouse and Repair Station (PEWARS) and Emsworth Locks and Dams in Pittsburgh, P...

11. Historic photo of cutaway rendering of rocket engine test ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

11. Historic photo of cutaway rendering of rocket engine test facility complex, June 11, 1965. On file at NASA Plumbrook Research Center, Sandusky, Ohio. NASA GRC photo number C-74433. - Rocket Engine Testing Facility, NASA Glenn Research Center, Cleveland, Cuyahoga County, OH

Investigation of Workplace-like Calibration Fields via a Deuterium-Tritium (D-T) Neutron Generator.

PubMed

Mozhayev, Andrey V; Piper, Roman K; Rathbone, Bruce A; McDonald, Joseph C

2017-04-01

Radiation survey meters and personal dosimeters are typically calibrated in reference neutron fields based on conventional radionuclide sources, such as americium-beryllium (Am-Be) or californium-252 (Cf), either unmodified or heavy-water moderated. However, these calibration neutron fields differ significantly from the workplace fields in which most of these survey meters and dosimeters are being used. Although some detectors are designed to yield an approximately dose-equivalent response over a particular neutron energy range, the response of other detectors is highly dependent upon neutron energy. This, in turn, can result in significant over- or underestimation of the intensity of neutron radiation and/or personal dose equivalent determined in the work environment. The use of simulated workplace neutron calibration fields that more closely match those present at the workplace could improve the accuracy of worker, and workplace, neutron dose assessment. This work provides an overview of the neutron fields found around nuclear power reactors and interim spent fuel storage installations based on available data. The feasibility of producing workplace-like calibration fields in an existing calibration facility has been investigated via Monte Carlo simulations. Several moderating assembly configurations, paired with a neutron generator using the deuterium tritium (D-T) fusion reaction, were explored.

Shock timing measurements and analysis in deuterium-tritium-ice layered capsule implosions on NIF

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

Robey, H. F.; Celliers, P. M.; Moody, J. D.

2014-02-15

Recent advances in shock timing experiments and analysis techniques now enable shock measurements to be performed in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility (NIF). Previous measurements of shock timing in inertial confinement fusion implosions [Boehly et al., Phys. Rev. Lett. 106, 195005 (2011); Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] were performed in surrogate targets, where the solid DT ice shell and central DT gas were replaced with a continuous liquid deuterium (D2) fill. These previous experiments pose two surrogacy issues: a material surrogacy due to the difference of species (D2 vs.more » DT) and densities of the materials used and a geometric surrogacy due to presence of an additional interface (ice/gas) previously absent in the liquid-filled targets. This report presents experimental data and a new analysis method for validating the assumptions underlying this surrogate technique. Comparison of the data with simulation shows good agreement for the timing of the first three shocks, but reveals a considerable discrepancy in the timing of the 4th shock in DT ice layered implosions. Electron preheat is examined as a potential cause of the observed discrepancy in the 4th shock timing.« less

Tritium Packages and 17th RH Canister Categories of Transuranic Waste Stored Below Ground within Area G

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

Hargis, Kenneth Marshall

A large wildfire called the Las Conchas Fire burned large areas near Los Alamos National Laboratory (LANL) in 2011 and heightened public concern and news media attention over transuranic (TRU) waste stored at LANL’s Technical Area 54 (TA-54) Area G waste management facility. The removal of TRU waste from Area G had been placed at a lower priority in budget decisions for environmental cleanup at LANL because TRU waste removal is not included in the March 2005 Compliance Order on Consent (Reference 1) that is the primary regulatory driver for environmental cleanup at LANL. The Consent Order is a settlementmore » agreement between LANL and the New Mexico Environment Department (NMED) that contains specific requirements and schedules for cleaning up historical contamination at the LANL site. After the Las Conchas Fire, discussions were held by the U.S. Department of Energy (DOE) with the NMED on accelerating TRU waste removal from LANL and disposing it at the Waste Isolation Pilot Plant (WIPP). This report summarizes available information on the origin, configuration, and composition of the waste containers within the Tritium Packages and 17th RH Canister categories; their physical and radiological characteristics; the results of the radioassays; and potential issues in retrieval and processing of the waste containers.« less

Code JEF Facilities Engineering Home Page for the Internet

NASA Technical Reports Server (NTRS)

Mahaffey, Valerie A.; Harrison, Marla J. (Technical Monitor)

1995-01-01

There are always many activities going on in JEF. We work on and manage the Construction of Facilities (C of F) projects at NASA-Ames. We are constantly designing or analyzing a new facility or project, or a modification to an existing facility. Every day we answer numerous questions about engineering policy, codes and standards, we attend design reviews, we count dollars and we make sure that everything at the Center is designed and built according to good engineering judgment. In addition, we study literature and attend conferences to make sure that we keep current on new legislation and standards.

Hydrogen isotope separation from water

DOEpatents

Jensen, R.J.

1975-09-01

A process for separating tritium from tritium-containing water or deuterium enrichment from water is described. The process involves selective, laser-induced two-photon excitation and photodissociation of those water molecules containing deuterium or tritium followed by immediate reaction of the photodissociation products with a scavenger gas which does not substantially absorb the laser light. The reaction products are then separated from the undissociated water. (auth)