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Sample records for tritium shipping package

  1. Tritium waste package

    Science.gov (United States)

    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.

  2. DYNAMIC ANALYSIS OF THE BULK TRITIUM SHIPPING PACKAGE SUBJECTED TO CLOSURE TORQUES AND SEQUENTIAL IMPACTS

    International Nuclear Information System (INIS)

    Wu, T; Paul Blanton, P; Kurt Eberl, K

    2007-01-01

    This paper presents a finite-element technique to simulate the structural responses and to evaluate the cumulative damage of a radioactive material packaging requiring bolt closure-tightening torque and subjected to the scenarios of the Hypothetical Accident Conditions (HAC) defined in the Code of Federal Regulations Title 10 part 71 (10CFR71). Existing finite-element methods for modeling closure stresses from bolt pre-load are not readily adaptable to dynamic analyses. The HAC events are required to occur sequentially per 10CFR71 and thus the evaluation of the cumulative damage is desirable. Generally, each HAC event is analyzed separately and the cumulative damage is partially addressed by superposition. This results in relying on additional physical testing to comply with 10CFR71 requirements for assessment of cumulative damage. The proposed technique utilizes the combination of kinematic constraints, rigid-body motions and structural deformations to overcome some of the difficulties encountered in modeling the effect of cumulative damage. This methodology provides improved numerical solutions in compliance with the 10CFR71 requirements for sequential HAC tests. Analyses were performed for the Bulk Tritium Shipping Package (BTSP) designed by Savannah River National Laboratory to demonstrate the applications of the technique. The methodology proposed simulates the closure bolt torque preload followed by the sequential HAC events, the 30-foot drop and the 30-foot dynamic crush. The analytical results will be compared to the package test data

  3. Safety analysis report on Model UC-609 shipping package

    International Nuclear Information System (INIS)

    Sandberg, R.R.

    1977-08-01

    This Safety Analysis Report for Packaging demonstrates that model UC-609 shipping package can safely transport tritium in any of its forms. The package and its contents are described. The package when subjected to the transport conditions specified in the Code of Federal Regulations, Title 10, Part 71 is evaluated. Finally, compliance with these regulations is discussed

  4. Safety analysis report: packages. LP-12 tritium package (packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Gates, A.A.; McCarthy, P.G.; Edl, J.W.

    1975-05-01

    Elemental tritium is shipped at low pressure in a stainless steel container (LP-12) within an aluminum vessel and surrounded by 3.9 in.-thick Celotex insulation in a steel drum. Information is presented on the packaging design, evaluation of the structural, thermal, containment, shielding, and criticality characteristics of the package, procedures for loading, unloading, transporting, and testing the LP-12, and quality assurance requirements. (U.S.)

  5. Safety analysis report: packages. LP-50 tritium package (packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Gates, A.A.; McCarthy, P.G.; Edl, J.W.

    1975-04-01

    Elemental tritium is shipped at low pressure in a stainless steel container (LP-50) sealed within an aluminum vessel and surrounded by a minimum of 4-in. thick Celotex insulation in a steel drum. The structural, thermal, containment, shielding, and criticality safety aspects of this package are evaluated. Procedures for loading and unloading, empty cask transport, acceptance testing and maintenance, and quality assurance requirements for the LP-50 package are described in detail. (U.S.)

  6. Safety analysis report; packages LP-50 tritium package. (Packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Gates, A.A.; McCarthy, P.G.; Edl, J.W.; Chalfant, G.G.

    1975-05-01

    Elemental tritium is shipped at low pressure in a stainless steel container (LP-50) surrounded by an aluminum vessel and Celotex insulation at least 4 in. thick in a steel drum. The total weight of the package is 260 lbs maximum. The various components that constitute the package are described and are shown in 7 figures. The safety analysis includes: structural evaluations; thermal evaluations; containment; operating procedures; acceptance tests and maintenance program; and design review

  7. Recovery and packaging of tritium from Canadian heavy water reactors

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Goyette, V.; Harrison, T.E.; Miller, J.M.

    1985-01-01

    The Tritium Extraction Plant being built at Chalk River Nuclear Laboratories (CRNL) will be the first industrial scale demonstration of the Liquid Phase Catalytic Exchange (LPCE) process for transfer of tritium from heavy water to deuterium. The plant will also demonstrate new technology in the areas of electrolytic cells for D 2 generation, water cooled recombiners, metal hydride packaging and magnetically coupled blowers for tritium service. It will be used to detritiate the heavy water in Atomic Energy of Canada Limited's (AECL) reactors

  8. MODEL 9975 SHIPPING PACKAGE FABRICATION PROBLEMS AND SOLUTIONS

    Energy Technology Data Exchange (ETDEWEB)

    May, C; Allen Smith, A

    2008-05-07

    The Model 9975 Shipping Package is the latest in a series (9965, 9968, etc.) of radioactive material shipping packages that have been the mainstay for shipping radioactive materials for several years. The double containment vessels are relatively simple designs using pipe and pipe cap in conjunction with the Chalfont closure to provide a leak-tight vessel. The fabrication appears simple in nature, but the history of fabrication tells us there are pitfalls in the different fabrication methods and sequences. This paper will review the problems that have arisen during fabrication and precautions that should be taken to meet specifications and tolerances. The problems and precautions can also be applied to the Models 9977 and 9978 Shipping Packages.

  9. Inspection, testing, and operating requiremens for the packaging and shipping of uranium trioxide in 55-gallon Department of Transportation (DOT) Specification 6M shipping packagings

    International Nuclear Information System (INIS)

    Toomer, D.V.

    1991-06-01

    This document identifies the inspection, testing and operating requirements for the packaging, loading, and shipping of uranium trioxide (UO 3 ) in 55-gallon DOT Specification 6M shipping packagings from the Idaho Chemical Processing Plant (ICPP). Compliance with this document assures established controls for the purchasing, packaging, loading, and shipping of DOT Specification 6M shipping packagings are maintained in strict accordance with applicable Code of Federal Regulations (CFRs) and Department of Energy (DOE) Orders. 7 refs., 3 figs., 1 tab

  10. Certification testing for the ES-2 shipping package

    International Nuclear Information System (INIS)

    Feldman, M.R.; Byington, G.A.; Handy, K.D.; Shappert, L.B.; Handy, K.D.; Oaks, R.E. Jr.; Stumpfl, E.

    1998-02-01

    The ES-2 is a multiconfiguration, Type B fissile material shipping package, designed by the Y-12 Nuclear Packaging Systems. It is unique in that a castable refractory material performs primary impact absorption and thermal insulation duties. This material, unlike the insulation often used in fissile material packages, such as Celotex and various foams, is fireproof at temperatures associated with Type B package testing (800 C). The ES-2 is designed to permit the use of three different containment vessels which can result in as many as six different configurations. Eight prototype units were manufactured and successfully tested to US Federal Regulatory Requirements

  11. Tritium

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The role played the large amount supply of tritium and its effects are broadly reviewed. This report is divided into four parts. The introductory part includes the history of tritium research. The second part deals with the physicochemical properties of tritium and the compounds containing tritium such as tritium water and labeled compounds, and with the isotope effects and self radiation effects of tritium. The third part deals with the tritium production by artificial reaction. Attention is directed to the future productivity of tritium from B, Be, N, C, O, etc. by using the beams of high energy protons or neutrons. The problems of the accepting market and the accuracy of estimating manufacturing cost are discussed. The expansion of production may bring upon the reduction of cost but also a large possibility of social impact. The irradiation problem and handling problem in view of environmental preservation are discussed. The fourth part deals with the use of tritium as a target, as a source of radiation or light, and its utilization for geochemistry. The future development of the solid tritium target capable of elongating the life of neutron sources is expected. The rust thickness of the surface of iron can be measured with the X-ray of Ti-T or Zr-T. The tritium can substitute self-light emission paint or lamp. The tritium is suitable for tracing the movement of sea water and land surface water because of its long half life. (Iwakiri, K.)

  12. Examination of shipping package 9975-04985

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-01

    Package 9975-04985 was examined following the identification of several unexpected conditions during surveillance activities. A heavy layer of corrosion product on the shield and the shield outer diameter being larger that allowed by drawing tolerances contributed to a very tight fit between the upper fiberboard assembly and shield. The average corrosion rate for the shield is estimated to be 0.0018 inch/year or less, which falls within the bounding rate of 0.002 inch/year that has been previously recommended for these packages. Several apparent foreign objects were noted within the package. One object observed on the air shield was identified as tape. The other objects were comprised of mostly fine fibers from the cane fiberboard. It is postulated that the upper and lower fiberboard assemblies were able to rub against each other due to the upper fiberboard assembly being held tight to the shield, and a few stray cane chips became frayed under vibratory motions.

  13. Quality assurance program description for shipping packages of radioactive material

    International Nuclear Information System (INIS)

    1978-01-01

    This quality assurance plan describes the quality assurance program at the Pacific Northwest Laboratory (PNL), for shipping packages of radioactive material. The purpose of this report is to describe how PNL will comply with the Code of Federal Regulations, Title 10, Part 71, Appendix E. In compliance with the instructions from the Nuclear Regulatory Commission (NRC), the 18 criteria from Appendix E are covered

  14. 9975 Shipping package component long-term degradation rates

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    Special nuclear materials are being stored in the K-Area Complex using 3013 containers that are held within Model 9975 shipping packages. The service life for these packages in storage was recently increased from 15 to 20 years, since some of these packages have been stored for nearly 15 years. A strategy is also being developed whereby such storage might be extended beyond 20 years. This strategy is based on recent calculations that support acceptable 9975 package performance for 20 years with internal heat loads up to 19 watts, and identifies a lower heat load limit for which the package components should degrade at half the bounding rate or less, thus doubling the effective storage life for these lower wattage packages. The components of the 9975 package that are sensitive to aging under storage conditions are the fiberboard overpack and the O-ring seals, although some degradation of the lead shield and outer drum are also possible. This report summarizes degradation rates applicable to lower heat load storage conditions. In particular, the O-ring seals should provide leak-tight performance for more than 40 years in packages for which their maximum temperature is ≤135 °F. Similarly, the fiberboard should remain acceptable in performance of its required safety functions for up to 40 years in packages with a maximum fiberboard temperature ≤125 °F.

  15. Testing and evaluation of used UF6 shipping packages

    International Nuclear Information System (INIS)

    Barlow, C.R.; Ziehlke, K.T.; Pryor, W.A.; Housholder, W.R.

    1986-01-01

    Damage to shipping packages and their components, whether due to normal environmental exposure or because of unforeseen accidents, requires occasional reevaluation and requalification to assure the suitability of the packages for continued service. Pressure tests have been conducted on used 30B cylinders that were involved in a warehouse fire to assess the possible damaging effects of the exposure. Deteriorated and mechanically damaged overpacks which had absorbed large quantities of water in service were subjected to drying tests, foam properties were evaluated, and a rehabilitation procedure was developed to allow return of such packages to service. Protective shipping packages show deterioration with extended service, principally structural damage from rough handling and rust damage from exposure to weather and from storage practices which may promote absorption of water by the insulating foam. The structural and thermal properties of the phenolic foam insulation from used shipping packages were found to be not adversely affected by absorbed water, and they were not degraded by the drying process. In order to slow or avoid continuing rust damage in renovated packages, however, a process was developed for drying the foam in commercial facilities prior to restoration work on used or damaged containers. Two 2-1/2-ton steel transport cylinders, type 30B, were involved in a warehouse fire where portions of the cylinders were estimated to have reached a temperature of 1600 0 F. The cylinders were empty at the time of the fire and thus were not in the protective packages in which full product cylinders are handled while in transit. Hydrostatic burst tests showed that the integrity of the cylinders was not degraded by the fire exposure. They withstood test pressures in excess of 10 times the design pressure, and showed a volume expansion of 30% above the original capacity before rupturing in a completely ductile fashion

  16. Tritium

    International Nuclear Information System (INIS)

    Fiege, A.

    1992-07-01

    This report contains information on chemical and physical properties, occurence, production, use, technology, release, radioecology, radiobiology, dose estimates, radioprotection and legal aspects of tritium. The objective of this report is to provide a reliable data base for the public discussion on tritium, especially with regard to its use in future nuclear fusion plants and its radiological assessment. (orig.) [de

  17. 9975 SHIPPING PACKAGE LIFE EXTENSION SURVEILLANCE PROGRAM RESULTS SUMMARY

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.; Dunn, K.; Hackney, B.; Hoffman, E.; Skidmore, E.

    2011-01-06

    Results from the 9975 Surveillance Program at the Savannah River Site (SRS) are summarized for justification to extend the life of the 9975 packages currently stored in the K-Area Materials Storage (KAMS) facility from 10 years to 15 years. This justification is established with the stipulation that surveillance activities will continue throughout this extended time to ensure the continued integrity of the 9975 materials of construction and to further understand the currently identified degradation mechanisms. The current 10 year storage life was developed prior to storage. A subsequent report was later used to extend the qualification of the 9975 shipping packages for 2 years for shipping plus 10 years for storage. However the qualification for the storage period was provided by the monitoring requirements of the Storage and Surveillance Program. This report summarizes efforts to determine a new safe storage limit for the 9975 shipping package based on the surveillance data collected since 2005 when the surveillance program began. KAMS is a zero-release facility that depends upon containment by the 9975 to meet design basis storage requirements. Therefore, to confirm the continued integrity of the 9975 packages while stored in KAMS, a 9975 Storage and Surveillance Program was implemented alongside the DOE required Integrated Surveillance Program (ISP) for 3013 plutonium-bearing containers. The 9975 Storage and Surveillance Program performs field surveillance as well as accelerated aging tests to ensure any degradation due to aging, to the extent that could affect packaging performance, is detected in advance of such degradation occurring in the field. The Program has demonstrated that the 9975 package has a robust design that can perform under a variety of conditions. As such the primary emphasis of the on-going 9975 Surveillance Program is an aging study of the 9975 Viton(reg.sign) GLT containment vessel O-rings and the Celotex(reg.sign) fiberboard thermal

  18. Tritium

    Science.gov (United States)

    2011-11-01

    QUESTIONS 5 4 CONSTRAINTS OF BASIC PHYSICS 7 4.1 Neutronics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.2 Tritium Burnup Fraction...Requirements for tritium-breeding should be one of the key tradeoffs 5 in reactor design. For example, maximizing the burnup fraction fb (a plasma-physics...account for the heterogeneous geometry and structural materials of the reactor, it should be possible to calculate values of the TBR that will be

  19. Destructive examination of shipping package 9975-02101

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-05-01

    Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02101 as part of the comprehensive Model 9975 package surveillance program. This package is one of ten high-wattage packages that were selected for field surveillance in FY15, and was identified to contain several non-conforming conditions. Most of these conditions (mold, stains, drum corrosion, calculated fiberboard dimensions and fiberboard damage) relate to the accumulation of water in the outer and lower portions of the cane fiberboard assembly. In the short term, this causes local but reversible changes in the fiberboard properties. Long-term effects can include the permanent loss of fiberboard properties (thus far observed only in the bottom fiberboard layers) and reduced drum integrity due to corrosion. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Three of the other nine FY15 high-wattage packages examined in the K-Area Complex showed similar behavior. Corrosion of the overpack drum has been seen primarily in those packages with relatively severe fiberboard degradation. Visual examination of the drums in storage for external corrosion should be considered as a screening tool to identify additional packages with potential fiberboard degradation. Where overpack drum corrosion has been observed, it is typically heaviest adjacent to the stitch welds along the bottom edge. It is possible that changes to the stitch weld design would reduce the degree of corrosion in this area, but would not eliminate it. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement with the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.

  20. Aging Behavior of the EPDM O-Rings in the H1616 Shipping Package

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.; Stefek, T.; Skidmore, E.

    2015-06-09

    The H1616 shipping package is used within the DOE complex for shipping tritium reservoirs. The annual recertification frequency can create logistical difficulties with other constraints on the timing of shipments; thus, a longer re-certification period is desirable. The ethylene propylene diene monomer (EPDM) O-rings used in the H1616 shipping package are being aged and tested at the Savannah River National Laboratory (SRNL) to provide a technical basis for extending the annual maintenance of the H1616 shipping package. H1616 EPDM O-rings are being aged at elevated temperature, and tested for degradation in mechanical properties, compression stress relaxation (CSR) behavior, and leak performance. Mechanical properties of aged O-rings show significant degradation can occur, but an inert atmosphere (argon backfill) greatly reduces the rate of degradation. The CSR behavior of O-rings was evaluated in air at 79 to 177 °C. These collective data were used to develop a predictive model for extrapolation of CSR behavior to relevant service temperatures (<67 °C). O-rings were also aged in H1616 Containment Vessels (CV) in an inert atmosphere at 71 to 149 °C. The vessels are helium leak tested periodically to determine if they continue to remain leak-tight. The vessel tests provide a solid demonstration that the H1616 O-rings will remain leak-tight at temperatures up to 113 °C for up to approximately 2.3 years. Significantly longer periods of leak-tight service are expected at the lower temperatures actually experienced in service. The predictive model developed from the CSR data conservatively indicates a service life of ~5 years at 67 °C. Although the relationship between CSR behavior and leak-tight performance has not been established for this design, the CSR predictions for this O-ring are conservative relative to leak-tight performance. Based on the collective data developed to date, SRNL has recommended that the maintenance interval for the H1616 package be

  1. Safety analysis report for packaging Lawrence Livermore Laboratories shipping containers

    International Nuclear Information System (INIS)

    Evans, J.H.

    1975-12-01

    The Lawrence Livermore Laboratories shipping containers were designed at Oak Ridge National Laboratory for use in transporting weapons and nuclear components. The design for the containers was evaluated to show compliance with applicable regulations governing packages in which radioactive and fissile materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the containers relative to the standards for the normal conditions of transport. A full-scale container test model was destructively tested to verify compliance with the standards for the accident conditions. The results of the analytical evaluations and the tests demonstrate that the design for the Lawrence Livermore Laboratories shipping containers is in compliance with the applicable regulations

  2. Safety analysis report: packages. DOT specification 7A - Type A container Mark 15 sludge shipping package

    International Nuclear Information System (INIS)

    Zeh, C.W.

    1985-03-01

    Sludge or filter cake containing 1.1 wt % 235 U enriched uranium from Mark 15 fabrication will be packaged in 55-gallon containers and shipped from SRP to NLO, Fernald, Ohio for recovery of product. About 7 Metric Tons (MT) of filter cake will be produced from fabricating Mark 15 slugs each reactor charge. Packaged shipments of this material, consigned as exclusive use, will be shipped to NLO in Department of Transportation (DOT) Specification 7A - Type A packages under a DOE Certificate of Compliance for Fissile Class III shipments. ''Type A packaging'' is designed to retain containment and shielding integrity under normal conditions of transport. This report documents compliance of the package design, construction methods, material and test performance with DOT Specification 7A. This DOT 7A Type A package contains a carbon steel outer container which is a 0.060-in.-thick 55-gal, galvanized drum equipped with a gasketed closure. The outer container encloses a 0.090-in.-thick, open head, polyethylene liner with lid

  3. Destructive examination of shipping package 9975-02644

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-23

    Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02644 as part of a comprehensive SRS surveillance program for plutonium material stored in the K-Area Complex (KAC). During the field surveillance inspection of this package in KAC, three non-conforming conditions were noted: the axial gap of 1.389 inch exceeded the 1 inch maximum criterion, the exposed height of the lead shield was greater than the 4.65 inch maximum criterion, and the difference between the upper assembly inside height and the exposed height of the lead shield was less than the 0.425 inch minimum criterion. All three of these observations relate to axial shrinkage of the lower fiberboard assembly. In addition, liquid water (condensation) was observed on the interior of the drum lid, the thermal blanket and the air shield.

  4. Destructive Examination of Shipping Package 9975-02019

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-06-13

    Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02019 as part of a comprehensive SRS surveillance program for plutonium material stored in the K-Area Complex (KAC). During the field surveillance inspection of this package in KAC, two non-conforming conditions were noted: the axial gap of 1.577 inch exceeded the 1 inch maximum criterion, and two areas of dried glue residue were noted on the upper fiberboard subassembly. This package was subsequently transferred to SRNL for more detailed inspection and destructive examination. In addition to the conditions noted in KAC, the following conditions were noted: - Numerous small spots of corrosion were observed along the bottom edge of the drum. - In addition to the smeared glue residue on the upper fiberboard subassembly, there was also a small dark stain. - Mold was present on the side and bottom of the lower fiberboard subassembly. Dark stains from elevated moisture content were also present in these areas. - A dark spot with possible light corrosion was observed on the primary containment vessel flange, and corresponding rub marks were observed on the secondary containment vessel ID. - The fiberboard thermal conductivity in the radial orientation was above the specified range. When the test was repeated with slightly lower moisture content, the result was acceptable. The moisture content for both tests was within a range typical of other packages in storage. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement within the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.

  5. Packaging Certification Program Methodology For Determining Dose Rates For Small Gram Quantities In Shipping Packagings

    International Nuclear Information System (INIS)

    Nathan, S.

    2011-01-01

    The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials, are significantly less hazardous than large amounts of the same materials. This study describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package compliant with 10 CFR Part 71 external radiation level limits regulations. These regulations require packaging for the shipment of radioactive materials perform, under both normal and accident conditions, the essential functions of material containment, subcriticality, and maintain external radiation levels within regulatory limits. 10 CFR 71.33(b)(1)(2) and (3) state radioactive and fissile materials must be identified and their maximum quantity, chemical and physical forms be included in an application. Furthermore, the U.S. Federal Regulations require application contain an evaluation demonstrating the package (i.e., the packaging and its contents) satisfies the external radiation standards for all packages (10 CFR 71.31(2), 71.35(a), and 71.47). By placing the contents in a He leak-tight containment vessel, and limiting the mass to ensure subcriticality, the first two essential functions are readily met. Some isotopes emit sufficiently strong photon radiation that small amounts of material can yield a large external dose rate. Quantifying of the dose rate for a proposed content is a challenging issue for the SGQ approach. It is essential to quantify external radiation levels from several common gamma and neutron sources that can be safely placed in a specific packaging, to ensure compliance with federal regulations. The Packaging Certification Program (PCP) Methodology for Determining Dose Rate for Small Gram Quantities in Shipping Packagings described in this report provides bounding mass limits for a set of proposed SGQ isotopes. Methodology calculations were performed to estimate external radiation

  6. Safety analysis report: packages 238Pu oxide shipping cask (packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Evans, J.E.; Gates, A.A.

    1975-06-01

    Plutonium-238 (as PuO 2 powder) is shipped in triple-container stainless steel shipping casks in compliance with ERDA Manual Chapter 0529 (ERDAM 0529), Safety Standards for the Packaging of Fissile and Other Radioactive Materials. (U.S.)

  7. LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, K. [Savannah River National Laboratory; Bellamy, S. [Savannah River National Laboratory; Daugherty, W. [Savannah River National Laboratory; Sindelar, R. [Savannah River National Laboratory; Skidmore, E. [Savannah River National Laboratory

    2013-08-18

    Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintain integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.

  8. Dynamic analysis to establish normal shock and vibration of radioactive material shipping packages

    International Nuclear Information System (INIS)

    Fields, S.R.

    1980-01-01

    A computer model, CARDS (Cask-Railcar Dynamic Simulator) was developed to provide input data for a broad range of radioactive material package-tiedown structural assessments. CARDS simulates the dynamic behavior of shipping packages and their transporters during normal transport conditions. The model will be used to identify parameters which significantly affect the normal shock and vibration environments which, in turn, provide the basis for determining the forces transmitted to the packages

  9. Humidity Data for 9975 Shipping Packages with Softwood Fiberboard

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-12

    The 9975 surveillance program is developing a technical basis to support extending the storage period of 9975 packages in K-Area Complex beyond the currently approved 15 years. A key element of this effort is developing a better understanding of degradation of the fiberboard assembly under storage conditions. This degradation is influenced greatly by the moisture content of the fiberboard, which is not well characterized on an individual package basis. Direct measurements of humidity and fiberboard moisture content have been made on two test packages with softwood fiberboard and varying internal heat levels from 0 up to 19W. Comparable measurements with cane fiberboard have been reported previously. With an internal heat load, a temperature gradient in the fiberboard assembly leads to varying relative humidity in the air around the fiberboard. However, the absolute humidity tends to remain approximately constant throughout the package, especially at lower heat loads.

  10. DOE handbook: Tritium handling and safe storage

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    The DOE Handbook was developed as an educational supplement and reference for operations and maintenance personnel. Most of the tritium publications are written from a radiological protection perspective. This handbook provides more extensive guidance and advice on the null range of tritium operations. This handbook can be used by personnel involved in the full range of tritium handling from receipt to ultimate disposal. Compliance issues are addressed at each stage of handling. This handbook can also be used as a reference for those individuals involved in real time determination of bounding doses resulting from inadvertent tritium releases. This handbook provides useful information for establishing processes and procedures for the receipt, storage, assay, handling, packaging, and shipping of tritium and tritiated wastes. It includes discussions and advice on compliance-based issues and adds insight to those areas that currently possess unclear DOE guidance.

  11. DOE handbook: Tritium handling and safe storage

    International Nuclear Information System (INIS)

    1999-03-01

    The DOE Handbook was developed as an educational supplement and reference for operations and maintenance personnel. Most of the tritium publications are written from a radiological protection perspective. This handbook provides more extensive guidance and advice on the null range of tritium operations. This handbook can be used by personnel involved in the full range of tritium handling from receipt to ultimate disposal. Compliance issues are addressed at each stage of handling. This handbook can also be used as a reference for those individuals involved in real time determination of bounding doses resulting from inadvertent tritium releases. This handbook provides useful information for establishing processes and procedures for the receipt, storage, assay, handling, packaging, and shipping of tritium and tritiated wastes. It includes discussions and advice on compliance-based issues and adds insight to those areas that currently possess unclear DOE guidance

  12. Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

    International Nuclear Information System (INIS)

    1991-01-01

    The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173

  13. Acoustic performance design and optimal allocation of sound package in ship cabin noise reduction

    Directory of Open Access Journals (Sweden)

    YANG Deqing

    2017-08-01

    Full Text Available The sound package in noise reduction design of ship cabins has become the main approach for the future. The sound package is a specially designed acoustic component consisting of damping materials, absorption materials, sound isolation materials and base structural materials which can achieve the prescribed performance of noise reduction. Based on the Statistical Energy Analysis(SEAmethod, quick evaluation and design methods, and the optimal allocation theory of sound packages are investigated. The standard numerical acoustic performance evaluation model, sound package optimization design model and sound package optimal allocation model are presented. A genetic algorithm is applied to solve the presented optimization problems. Design examples demonstrate the validity and efficiency of the proposed models and solutions. The presented theory and methods benefit the standardization and programming of sound package design, and decrease noise reduction costs.

  14. Design of shipping packages to transport varying radioisotopic source materials for future space and terrestrial missions

    International Nuclear Information System (INIS)

    Barklay, C.D.

    1995-01-01

    The exploration of space will begin with manned missions to the moon and to Mars, first for scientific discoveries, then for mining and manufacturing. Because of the great financial costs of this type of exploration, it can only be accomplished through an international team effort. This unified effort must include the design, planning and, execution phases of future space missions, extending down to such activities as isotope processing, and shipping package design, fabrication, and certification. All aspects of this effort potentially involve the use of radioisotopes in some capacity, and the transportation of these radioisotopes will be impossible without a shipping package that is certified by the Nuclear Regulatory Commission or the U.S. Department of Energy for domestic shipments, and the U.S. Department of Transportation or the International Atomic Energy Agency for international shipments. To remain without the international regulatory constraints, and still support the needs of new and challenging space missions conducted within ever-shrinking budgets, shipping package concepts must be innovative. A shipping package must also be versatile enough to be reconfigured to transport the varying radioisotopic source materials that may be required to support future space and terrestrial missions. One such package is the Mound USA/9516/B(U)F. Taking into consideration the potential need to transport specific types of radioisotopes, approximations of dose rates at specific distances were determined taking into account the attenuation of dose rate with distance for varying radioisotopic source materials. As a result, it has been determined that the shipping package requirements that will be demanded by future space (and terrestrial) missions can be met by making minor modifications to the USA/9516/B(U)F. copyright 1995 American Institute of Physics

  15. Safety Analysis Report--packages: cobalt-60 shipping cask (packaging of radioactive and fissile materials). Final report

    International Nuclear Information System (INIS)

    Cherney, J.R.; Gates, A.A.

    1976-08-01

    Safety Analysis Report DPSPU 73-124-1 is revised to permit shipment of 8 slugs (max), each slug containing an average of 43,750 curies of 60 Co and contained in a stainless steel can with a double-plug closure in cobalt-60 shipping casks in compliance with 10 CFR 71, Packaging of Radioactive Materials for Transport. This is in addition to the forms already discussed in DPSPU 73-124-1

  16. Preparation of the Shippingport reactor pressure vessel shipping package

    International Nuclear Information System (INIS)

    Yannitell, D.M.

    1988-01-01

    Shippingport Station Decommissioning Project is the removal and shipment the Reactor Pressure Vessel (PRV) and its associated Neutron Shield Tank (NST) to the government owned Hanford Reservation in Richland, Washington. Engineering studies considered the alternatives for removal and shipment of the RPV/NST. These included segmentation for subsequent truck shipments, and one-piece removal with barge or rail shipment. Although the analysis indicated that current technology could be utilized to accomplish either alternative, one-piece removal of the RPV was selected as the safest, most cost effective method. When compared to segmentation, it was estimated that one-piece removal would reduce the duration of the Project by 1 year, reduce cost by $4 M, and result in a savings of radiation exposure of 150 man-Rem. Rail transportation of an integral RPV/NST package is not feasible due to the physical size of the package. 5 refs., 1 fig

  17. Safety analysis report for packaging (SARP) of the Oak Ridge National Laboratory. TRU curium shipping container

    International Nuclear Information System (INIS)

    Box, W.D.; Klima, B.B.; Seagren, R.D.; Shappert, L.B.; Aramayo, G.A.

    1980-06-01

    An analytical evaluation of the Oak Ridge National Laboratory Transuranium (TRU) Curium Shipping Container was made to demonstrate its compliance with the regulations governing offsite shipment of packages containing radioactive material. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the container complies with the applicable regulations

  18. Safety analysis report for packaging (SARP) of the Oak Ridge National Laboratory. TRU curium shipping container

    Energy Technology Data Exchange (ETDEWEB)

    Box, W.D.; Klima, B.B.; Seagren, R.D.; Shappert, L.B.; Aramayo, G.A.

    1980-06-01

    An analytical evaluation of the Oak Ridge National Laboratory Transuranium (TRU) Curium Shipping Container was made to demonstrate its compliance with the regulations governing offsite shipment of packages containing radioactive material. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the container complies with the applicable regulations.

  19. REVIEW OF CLEANING SOLUTIONS FOR USE ON COMPONENTS OF THE 9975 SHIPPING PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.

    2013-09-30

    Several candidate cleaning products have been reviewed for use as a disinfectant on 9975 shipping package components which contain or have contacted mold. Following review of the compatibility of these products with each component, ammonia (ammonium hydroxide diluted to 1.5 wt% concentration) appears compatible with all package components that it might contact. Each of the other candidate products is incompatible with one or more package components. Accordingly, ammonia is recommended for this purpose. It is further recommended that all components which are disinfected be subsequently rinsed with di-ionized or distilled water.

  20. Guidelines for safe design of shipping packages against brittle fracture

    International Nuclear Information System (INIS)

    1993-08-01

    In 1992, the ninth meeting of the Standing Advisory Group on the Safe Transport of Radioactive Materials recommended the publication of this TECDOC in an effort to promote the widest debate on the criteria for the brittle fracture safe design of transport packages. The published IAEA advice on the influence of brittle fracture on material integrity is contained in Appendix IX of the Advisory Material for the IAEA Regulations for the Safety Transport of Radioactive Material (1985 Edition, as amended 1990), Safety Series No. 37. This guidance is limited in scope, dealing only with ferritic steels in general terms. It is becoming more common for designers to specify materials other than austenitic stainless steel for packaging components. The data on ferritic steels cannot be assumed to apply to other metals, hence the need for further guidance on the development of relationships describing material properties at low temperatures. The methods described in this TECDOC will be considered by the Revision Panel for inclusion in the 1996 Edition of the IAEA Regulations for the Safe Transport of Radioactive Material and the supporting documents. If accepted by the Revision Panel, this advice will be a candidate for upgrading to a Safety Practice. In the interim period, this TECDOC offers provisional advice on brittle fracture evaluation. It is acknowledged that, at this stage, the views expressed do not necessarily reflect those of the governments of Member States or organizations under whose auspices this manuscript was produced. Refs and figs

  1. DEMONSTRATION OF EQUIVALENCY OF CANE AND SOFTWOOD BASED CELOTEX FOR MODEL 9975 SHIPPING PACKAGES

    International Nuclear Information System (INIS)

    Watkins, R; Jason Varble, J

    2008-01-01

    Cane-based Celotex(trademark) has been used extensively in various Department of Energy (DOE) packages as a thermal insulator and impact absorber. Cane-based Celotex(trademark) fiberboard was only manufactured by Knight-Celotex Fiberboard at their Marrero Plant in Louisiana. However, Knight-Celotex Fiberboard shut down their Marrero Plant in early 2007 due to impacts from hurricane Katrina and other economic factors. Therefore, cane-based Celotex(trademark) fiberboard is no longer available for use in the manufacture of new shipping packages requiring the material as a component. Current consolidation plans for the DOE Complex require the procurement of several thousand new Model 9975 shipping packages requiring cane-based Celotex(trademark) fiberboard. Therefore, an alternative to cane-based Celotex(trademark) fiberboard is needed. Knight-Celotex currently manufactures Celotex(trademark) fiberboard from other cellulosic materials, such as hardwood and softwood. A review of the relevant literature has shown that softwood-based Celotex(trademark) meets all parameters important to the Model 9975 shipping package

  2. Examination Of Shipping Packages 9975-01818, 9975-01903 AND 9975-02287

    International Nuclear Information System (INIS)

    Daugherty, W.

    2009-01-01

    Three 9975 shipping packages were examined to investigate the non-conforming condition of an axial air gap greater than 1 inch. This condition typically indicates the presence of excess moisture in the fiberboard overpack, and may be accompanied by degradation in the fiberboard properties. The package with the largest axial air gap (9975-01818, with an air gap of 1.437 inches) was found to contain significant excess moisture, and the lower fiberboard assembly was covered with mold and was significantly degraded in strength. This condition is very similar to that observed previously in package 9975-01819. Both packages (-1818 and -1819) appear to contain a similar amount of excess moisture, which was previously estimated for 9975-01819 as ∼2.5 liters. The condition of 9975-01818 was also evidenced by several rust spots along the bottom chime of the drum, although no significant rust was noted on the closure bolts. Packages 9975-01903 and 9975-02287 were also examined. The axial air gap in these two packages was less than in 9975-01818, but still exceeded 1 inch. These two packages contained elevated moisture levels, although not significantly higher than seen in other 'typical' packages. The fiberboard in these two packages was of sound integrity, and appeared generally consistent with undegraded material. A few small patches of mold on and near the bottom of the fiberboard in 9975-01903 appeared dormant. No mold was observed on package 9975-02287. The SPA will provide recommendations on possible follow-up activities with these three packages. This might include a demonstration in SRNL of whether removal of the caplugs from similar packages would facilitate removal of excess moisture. Future efforts should also include an assessment of using the 1 inch axial gap criterion as a valid indicator of fiberboard degradation.

  3. EXAMINATION OF SHIPPING PACKAGES 9975-01818, 9975-01903 AND 9975-02287

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.

    2009-11-18

    Three 9975 shipping packages were examined to investigate the non-conforming condition of an axial air gap greater than 1 inch. This condition typically indicates the presence of excess moisture in the fiberboard overpack, and may be accompanied by degradation in the fiberboard properties. The package with the largest axial air gap (9975-01818, with an air gap of 1.437 inches) was found to contain significant excess moisture, and the lower fiberboard assembly was covered with mold and was significantly degraded in strength. This condition is very similar to that observed previously in package 9975-01819. Both packages (-1818 and -1819) appear to contain a similar amount of excess moisture, which was previously estimated for 9975-01819 as {approx}2.5 liters. The condition of 9975-01818 was also evidenced by several rust spots along the bottom chime of the drum, although no significant rust was noted on the closure bolts. Packages 9975-01903 and 9975-02287 were also examined. The axial air gap in these two packages was less than in 9975-01818, but still exceeded 1 inch. These two packages contained elevated moisture levels, although not significantly higher than seen in other 'typical' packages. The fiberboard in these two packages was of sound integrity, and appeared generally consistent with undegraded material. A few small patches of mold on and near the bottom of the fiberboard in 9975-01903 appeared dormant. No mold was observed on package 9975-02287. The SPA will provide recommendations on possible follow-up activities with these three packages. This might include a demonstration in SRNL of whether removal of the caplugs from similar packages would facilitate removal of excess moisture. Future efforts should also include an assessment of using the 1 inch axial gap criterion as a valid indicator of fiberboard degradation.

  4. Examination of Shipping Packages 9975-02694 and 9975-02729

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-11-09

    SRNL has assisted in the examination of two 9975 shipping packages (9975-02694 and 9975-02729) following their use for storage of nuclear material in K-Area Complex (KAC). Both of these were targeted for examination because they were exposed to water from a leaking expansion joint during storage. When first opened in KAC, the axial gap was found to be in excess of the 1 inch maximum criterion, signaling the potential for degradation of the fiberboard overpack and drum. Within each package the highest fiberboard moisture levels were observed in the bottom layers, but no mold was observed in either package. The fiberboard in each package appeared to retain good integrity consistent with non-degraded material. Minor corrosion was also observed on these drums along the lower stitch welds and on several closure bolts.

  5. Design of an experiment to measure fire exposure of packages aboard container cargo ships

    International Nuclear Information System (INIS)

    Koski, J.A.

    1998-01-01

    The test described in this paper is intended to measure the typical accident environment for a radioactive materials package in a fire abroad a container cargo ship. A stack of nice used standard cargo containers will be variously loaded with empty packages, simulated packages and combustible cargo and placed over a large hydrocarbon pool fire of one hour duration. Fire environments, both inside and outside the containers, typical of on-deck stowage will be measured as well as the potential for container-to-container fire spread. With the use of the inverse heat conduction calculations, the local heat transfer to the simulated packages can be estimated from thermocouple data. Data recorded will also provide information on fire durations in each container, fire intensity and container-to-container fire spread characteristics. (authors)

  6. Evaluating the loss of a LWR spent fuel or plutonium shipping package into the sea

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Baker, D.A.

    1976-06-01

    As the nations of the world turn to nuclear power for an energy source, commerce in nuclear fuel cycle materials will increase. Some of this commerce will be transported by sea. Such shipments give rise to the possibility of loss of these materials into the sea. This paper discusses the postulated accidental loss of two materials, light water reactor (LWR) spent fuel and plutonium, at sea. The losses considered are that of a single shipping package which is either undamaged or damaged by fire prior to the loss. The containment failure of the package in the sea,

  7. 9975 Shipping Package Performance Of Alternate Materials For Long-Term Storage Application

    International Nuclear Information System (INIS)

    Skidmore, E.; Hoffman, E.; Daugherty, W.

    2010-01-01

    The Model 9975 shipping package specifies the materials of construction for its various components. With the loss of availability of material for two components (cane fiberboard overpack and Viton(reg s ign) GLT O-rings), alternate materials of construction were identified and approved for use for transport (softwood fiberboard and Viton(reg s ign) GLT-S O-rings). As these shipping packages are part of a long-term storage configuration at the Savannah River Site, additional testing is in progress to verify satisfactory long-term performance of the alternate materials under storage conditions. The test results to date can be compared to comparable results on the original materials of construction to draw preliminary conclusions on the performance of the replacement materials.

  8. Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory TRU Californium Shipping Container

    International Nuclear Information System (INIS)

    Box, W.D.; Shappert, L.B.; Seagren, R.D.; Klima, B.B.; Jurgensen, M.C.; Hammond, C.R.; Watson, C.D.

    1980-01-01

    An analytical evaluation of the Oak Ridge National Laboratory TRU Californium Shipping Container was made in order to demonstrate its compliance with the regulations governing off-site shipment of packages that contain radioactive material. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of this evaluation demonstrate that the container complies with the applicable regulations

  9. PROPERTIES OF FIBERBOARD OVERPACK MATERIAL IN THE 9975 SHIPPING PACKAGE FOLLOWING THERMAL AGING

    International Nuclear Information System (INIS)

    Daugherty, W

    2007-01-01

    Many radioactive material shipping packages incorporate cane fiberboard overpacks for thermal insulation and impact resistance. Mechanical, thermal and physical properties have been measured on cane fiberboard following thermal aging in several temperature/humidity environments. Several of the measured properties change significantly over time in the more severe environments, while other properties are relatively constant. These properties continue to be tracked, with the goal of developing a model for predicting a service life under long-term storage conditions

  10. AGING BEHAVIOR OF VITON O-RING SEALS IN THE 9975 SHIPPING PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Skidmore, E.; Daugherty, W.; Hoffman, E.; Dunn, K.; Bellamy, S.

    2012-01-13

    The Savannah River Site (SRS) is storing plutonium (Pu) materials in the K-Area Materials Storage (KAMS) facility. The Pu materials were packaged according to the DOE-STD-3013 standard and shipped to the SRS in Type B 9975 packages. The robust 9975 shipping package was not designed for long-term product storage, but it is a specified part of the storage configuration and the KAMS facility safety basis credits the 9975 design with containment. Within the 9975 package, nested stainless steel containment vessels are closed with dual O-ring seals based on Viton{reg_sign} GLT or GLT-S fluoroelastomer. The aging behavior of the O-ring compounds is being studied to provide the facility with advanced notice of nonconformance and to develop life prediction models. A combination of field surveillance, leak testing of surrogate fixtures aged at bounding service temperatures, and accelerated-aging methodologies based on compression stress-relaxation and oxygen consumption analysis is being used to evaluate seal performance. A summary of the surveillance program relative to seal aging behavior is presented.

  11. An analysis of the qualification criteria for small radioactive material shipping packages

    Energy Technology Data Exchange (ETDEWEB)

    McClure, J.D.

    1983-05-01

    The RAM package design certification process has two important elements, testing and acceptance. These terms sound very similar but they have specific meanings. Qualification testing in the context of this study is the imposition of simulated accident test conditions upon the candidate package design. (Normal transportation environments may also be included.) Following qualification testing, the acceptance criteria provide the performance levels which, if demonstrated, indicate the ability of the RAM package to sustain the severity of the qualification testing sequence and yet maintain specified levels of package integrity. This study has used Severities of Transportation Accidents as a data base to examine the regulatory test criteria which are required to be met by small packages containing Type B quantities of radioactive material (RAM). The basic findings indicate that the present regulatory test standards provide significantly higher levels of protection for the surface transportation modes (truck, rail) than for RAM packages shipped by aircraft. It should also be noted that various risk assessment studies have shown that the risk to the public due to severe transport accidents by surface and air transport modes is very low. A key element in this study was the quantification of the severity of the transportation accident environment and the severity of the present qualification test standards (called qualification test standards in this document) so that a direct comparison could be made between them to assess the effectiveness of the existing qualification test standards. The manner in which this was accomplished is described.

  12. An analysis of the qualification criteria for small radioactive material shipping packages

    International Nuclear Information System (INIS)

    McClure, J.D.

    1983-05-01

    The RAM package design certification process has two important elements, testing and acceptance. These terms sound very similar but they have specific meanings. Qualification testing in the context of this study is the imposition of simulated accident test conditions upon the candidate package design. (Normal transportation environments may also be included.) Following qualification testing, the acceptance criteria provide the performance levels which, if demonstrated, indicate the ability of the RAM package to sustain the severity of the qualification testing sequence and yet maintain specified levels of package integrity. This study has used Severities of Transportation Accidents as a data base to examine the regulatory test criteria which are required to be met by small packages containing Type B quantities of radioactive material (RAM). The basic findings indicate that the present regulatory test standards provide significantly higher levels of protection for the surface transportation modes (truck, rail) than for RAM packages shipped by aircraft. It should also be noted that various risk assessment studies have shown that the risk to the public due to severe transport accidents by surface and air transport modes is very low. A key element in this study was the quantification of the severity of the transportation accident environment and the severity of the present qualification test standards (called qualification test standards in this document) so that a direct comparison could be made between them to assess the effectiveness of the existing qualification test standards. The manner in which this was accomplished is described

  13. Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory Shipping Cask D-38. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Box, W.D.; Shappert, L.B.; Seagren, R.D.; Watson, C.D.; Hammond, C.R.; Klima, B.B.

    1979-09-01

    An analytical evaluation of the Oak Ridge National Laboratory Shipping Cask D-38 (solids shipments) was made to demonstrate its compliance with the regulations governing off-site radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the cask complies with the applicable regulations.

  14. Aging Behavior of the Viton® Fluoroelastomer O-Rings in the 9975 Shipping Package

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.; Mcwilliams, A.; Skidmore, E.

    2015-06-09

    The 9975 Type B shipping package is used within the DOE complex for shipping special nuclear materials. This package is re-certified annually in accordance with Safety Analysis Report requirements. The package is also used at the Savannah River Site as part of the long-term storage configuration of special nuclear materials. As such, the packages do not undergo annual recertification during storage, with uncertainty as to how long some of the package components will meet their functional requirements in the storage environment. The packages are currently approved for up to 15 years storage, and work continues to provide a technical basis to extend that period. This paper describes efforts by the Savannah River National Laboratory (SRNL) to extend the service life estimate of Viton® GLT and GLT-S fluoroelastomer O-rings used in the 9975 shipping package. O-rings of both compositions are undergoing accelerated aging at elevated temperature, and are periodically tested for compression stress relaxation (CSR) behavior and leak performance. The CSR behavior of O-rings was evaluated at temperatures from 79 °C to 177 °C. These collective data were used to develop predictive models for extrapolation of CSR behavior to relevant service temperatures (< 75 °C). O-rings were also aged in Primary Containment Vessel (PCV) fixtures at temperatures ranging from 79 °C to 232 °C. The fixtures are helium leak tested periodically to determine if they remain leak-tight. The PCV fixture tests demonstrate that the 9975 O-rings will remain leak-tight at temperatures up to 149 °C for 3 years or more, and no leak failures have been observed with up to 8 years aging at 93 °C. Significantly longer periods of leak-tight service are expected at the lower temperatures actually experienced in the storage environment. The predictive model developed from the CSR data conservatively indicates a service life of more than 20 years at the bounding temperature of 75 °C. Although the

  15. Underground Test Area Subproject Phase I Data Analysis Task. Volume VII - Tritium Transport Model Documentation Package

    Energy Technology Data Exchange (ETDEWEB)

    None

    1996-12-01

    Volume VII of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the tritium transport model documentation. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  16. Physical test report to drop test of a 9975 radioactive material shipping packaging

    International Nuclear Information System (INIS)

    Blanton, P.S.

    1997-01-01

    This report presents the drop test results for the 9975 radioactive material shipping package being dropped 30 feet onto a unyielding surface followed by a 40-inch puncture pin drop. The purpose of these drops was to show that the package lid would remain attached to the drum. The 30-foot drop was designed to weaken the lid closure lug while still maintaining maximum extension of the lugs from the drum surface. This was accomplished by angling the drum approximately 30 degrees from horizontal in an inverted position. In this position, the drum was rotated slightly so as not to embed the closure lugs into the drum as a result of the 30-foot drop. It was determined that this orientation would maximize deformation to the closure ring around the closure lug while still maintaining the extension of the lugs from the package surface. The second drop was from 40 inches above a 40-inch tall 6-inch diameter puncture pin. The package was angled 10 degrees from vertical and aligned over the puncture pin to solidly hit the drum lug(s) in an attempt to disengage the lid when dropped.Tests were performed in response to DOE EM-76 review Q5 inquires that questioned the capability of the 9975 drum lid to remain in place under this test sequence. Two packages were dropped utilizing this sequence, a 9974 and 9975. Test results for the 9974 package are reported in WSRC-RP-97-00945. A series of 40-inch puncture pin tests were also performed on undamaged 9975 and 9974 packages

  17. Beneficial uses shipping system (BUSS) cask, safety analysis report for packaging: Volumes 1 and 2

    International Nuclear Information System (INIS)

    Ferrell, P.C.

    1997-01-01

    The Beneficial Uses Shipping System (BUSS) cask Safety Analysis Report for Packaging (SARP) was originally prepared by Sandia National Laboratory (SNL). After the certification process was completed, the ownership of the BUSS cask and associated SARP was transferred from SNL to the DOE Hanford site in Richland, Washington. During timely renewal of the BUSS cask certificate of compliance, the SARP was revised to (1) respond to the timely renewal questions, (2) consolidate the previous revision made by SNL, and (3) bring the SARP into compliance with the 1996 version of 10 CFR 71. Since the BUSS cask is now the responsibility of RL, the SARP was reissued as a Hanford document

  18. Estimated Radiation Dose from a MOX Fuel Shipping Package that is Hypothetically Submerged in the Sea

    International Nuclear Information System (INIS)

    Tsumune, D.; Suzuki, H.; Saegusa, T.; Maruyama, K.; Ito, C.; Watabe, N.

    2000-01-01

    The sea transport of fresh MOX fuel from Europe to Japan is under consideration. For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of the IMO. For transport of radioactive materials, there is a safety standard stipulated in the Regulations for the Safe Transport of Radioactive Materials issued by the IAEA. Under these, code and standard, fresh MOX fuel will be transported safely by sea. However, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk from some unexpected cause. In both cases, for a package sunk in the coastal region and for one sunk in the open sea, the evaluated results of the dose equivalent by radiation exposure of the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv.y -1 ). (author)

  19. Estimated Radiation Dose from a MOX Fuel Shipping Package that is Hypothetically Submerged in the Sea

    Energy Technology Data Exchange (ETDEWEB)

    Tsumune, D.; Suzuki, H.; Saegusa, T.; Maruyama, K.; Ito, C.; Watabe, N

    2000-07-01

    The sea transport of fresh MOX fuel from Europe to Japan is under consideration. For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of the IMO. For transport of radioactive materials, there is a safety standard stipulated in the Regulations for the Safe Transport of Radioactive Materials issued by the IAEA. Under these, code and standard, fresh MOX fuel will be transported safely by sea. However, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk from some unexpected cause. In both cases, for a package sunk in the coastal region and for one sunk in the open sea, the evaluated results of the dose equivalent by radiation exposure of the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv.y{sup -1}). (author)

  20. Dose Assessment for Public due to Packages Shipping Radioactive Materials Hypothetically Sunk on a Continental Shelf

    Energy Technology Data Exchange (ETDEWEB)

    Tsumune, D.; Saegusa, T.; Suzuki, H.; Watabe, N.; Asano, H.; Maruyama, K.; Kinehara, Y

    2000-07-01

    Radioactive materials such as spent fuel (SF), PuO{sub 2} powder, high level wastes (HLW) and fresh mixed oxide (MOX) fuel have been transported by sea between Europe and Japan for many years. Dose assessments for the public have been performed in the past for situations assuming packages shipping radioactive materials are hypothetically sunk on a continental shelf. These studies employed various conditions and methods in their assessments and the results were not always the same. In this study, the dose assessment for all types of package was performed under the same conditions and by the same methods. The effective dose to the members of the public for all materials is lower than previous evaluations due to more realistic assumptions used in this study. These evaluated effective doses are far less than the ICRP recommended limit (1 mSv.year{sup -1} averaged over 5 years). (author)

  1. Functional assessment of hospital laboratory packaging and shipping preparedness in New York State.

    Science.gov (United States)

    Pennell, Paula A; Antenucci, Alan J; Brennan, Lynn E; Burhans, Robert L; Ostrowski, Stephanie E

    2008-01-01

    The 2006-2007 New York State (NYS) Hospital Laboratory Drill Series was implemented in order to test notification, referral and packaging and shipping (P&S) procedures at acute care hospital facilities (statewide, excluding New York City) that submit suspect bioterrorism (BT), chemical terrorism (CT), and/or pandemic influenza (Pan Flu) clinical specimens to the NYS Department of Health (DOH) Wadsworth Center for confirmatory testing. Results showed that 97% and 84% of hospital facilities had the ability to directly access the notification network and retrieve drill guidance, respectively. Most hospital laboratories (92%) demonstrated the ability to refer specimens to the Wadsworth Center laboratory. Evaluation of specimen submissions found that 68% of BT packages, 27% of Pan Flu packages, and 20% of CT packages arrived to the laboratory with no P&S deficiencies. It can be concluded that acute care hospital facilities in NYS are more prepared to refer and submit clinical specimens during a BT public health emergency than during a Pan Flu or CT emergency event.

  2. REVIEW OF AGING DATA ON EPDM O-RINGS IN THE H1616 SHIPPING PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Skidmore, E.

    2012-03-27

    Currently, all H1616 shipping package containers undergo annual re-verification testing, including containment vessel leak testing to verify leak-tightness (<1 x 10{sup -7} ref cc/sec air) as per ANSI N14.5. The purpose of this literature review is to supplement aging studies currently being performed by SRNL on the EPDM O-rings to provide the technical basis for extending annual re-verification testing for the H1616 shipping package and to predict the life of the seals at bounding service conditions. The available data suggest that the EPDM O-rings can retain significant mechanical properties and sealing force at or below bounding service temperatures (169 F or 76 C) beyond the 1 year maintenance period. Interpretation of available data suggests that a service life of at least 2 years and potentially 4-6 years may be possible at bounding temperatures. Seal lifetimes at lower, more realistic temperatures will likely be longer. Being a hydrocarbon elastomer, EPDM O-rings may exhibit an inhibition period due to the presence of antioxidants. Once antioxidants are consumed, mechanical properties and seal performance could decline at a faster rate. Testing is being performed to validate the assumptions outlined in this report and to assess the long-term performance of O-ring seals under actual service conditions.

  3. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G{sub eff}, is defined by: G{sub eff} - {Sigma}{sub M} (F{sub M} x G{sub M}) F{sub M}-fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials.

  4. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    International Nuclear Information System (INIS)

    1994-10-01

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G eff , is defined by: G eff - Σ M (F M x G M ) F M -fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials

  5. Consequences of postulated losses of LWR spent fuel and plutonium shipping packages at sea

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Baker, D.A.; Beyer, C.E.; Friley, J.R.; Mandel, S.; Peterson, P.L.; Sominen, F.A.

    1977-10-01

    The potential consequences of the loss of a large spent fuel cask and of a single 6M plutonium shipping package into the sea for two specific accident cases are estimated. The radiation dose to man through the marine food chain following the loss of undamaged and fire-damaged packages to the continental shelf and in the deep ocean are conservatively estimated. Two failure mechanisms that could lead to release of radioactive material after loss of packages into the ocean have been considered: corrosion and hydrostatic pressure. A third possible mechanism is thermal overpressurization following burial in marine sediments. It was determined that the seals or pressure relief devices on an undamaged spent fuel cask might fail from hydrostatic forces for losses on the continental shelf although some cask designs would retain their integrity at this depth. The population dose to man through the marine food chain following these scenarios has been estimated. The dose estimates are made relating the radioactive material released and the seafood productivity in the region of the release. Doses are based on a one-year consumption of contaminated seafood. The loss of a single plutonium package on the continental shelf is estimated to produce a population dose commitment of less than 250 man-rem for recycle plutonium. The dose commitment to the average individual is less than one millirem. Doses for losses of undamaged casks to the continental shelf and deep ocean and for loss of a fire-damaged cask to the deep ocean were determined to be several orders of magnitude smaller. 22 tables, 10 figures

  6. Tritium Packages and 17th RH Canister Categories of Transuranic Waste Stored Below Ground within Area G

    Energy Technology Data Exchange (ETDEWEB)

    Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-01

    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 settlement 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.

  7. Safety analysis report for packaging for the Idaho National Engineering Laboratory TRA Type 1 Shipping Container and TRA Type 2 Shipping Capsule

    International Nuclear Information System (INIS)

    Havlovick, B.J.

    1992-01-01

    The TRA Type I Shipping Container and TRA Type II Shipping Capsule were designed and fabricated at the Idaho National Engineering Laboratory as special form containers for the transport of non-fissile radioisotopes and fissile radioisotopes in exempt quantities. The Type I container measures 0.75 in. outside diameter and 3.000 in long. The Type II capsule is 0.495 in. outside diameter 2.000 in. long. The container and capsule were tested and evaluated to determine their compliance with Title 49 Code of Federal Regulations 173, which governs packages for special form radioactive material. This report is based upon those tests and evaluations. The results of those tests and evaluations demonstrate the container and capsule are in full compliance with the special form shipping container regulations of 49 CFR 173

  8. Source term development for tritium at the Sheffield disposal site

    International Nuclear Information System (INIS)

    MacKenzie, D.R.; Barletta, R.E.; Smalley, J.F.; Kempf, C.R.; Davis, R.E.

    1984-01-01

    The Sheffield low-level radioactive waste disposal site, which ceased operation in 1978, has been the focus of modeling efforts by the NRC for the purpose of predicting long-term site behavior. To provide the NRC with the information required for its modeling effort, a study to define the source term for tritium in eight trenches at the Sheffield site has been undertaken. Tritium is of special interest since significant concentrations of the isotope have been found in groundwater samples taken at the site and at locations outside the site boundary. Previous estimates of tritium site inventory at Sheffield are in wide disagreement. In this study, the tritium inventory in the eight trenches was estimated by reviewing the radioactive shipping records (RSRs) for waste buried in these trenches. It has been found that the tritium shipped for burial at the site was probably higher than previously estimated. In the eight trenches surveyed, which amount to roughly one half the total volume and activity buried at Sheffield, approximately 2350 Ci of tritium from non-fuel cycle sources were identified. The review of RSRs also formed the basis for obtaining waste package descriptions and for contacting large waste generators to obtain more detailed information regarding these waste packages. As a result of this review and the selected generator contacts, the non-fuel cycle tritium waste was categorized. The tritium releases from each of these waste categories were modeled. The results of this modeling effort are presented for each of the eight trenches selected. 3 references, 2 figures

  9. CSR behavior and aging model for the Viton© Fluorelastomer O-rings in the 9975 shipping package

    International Nuclear Information System (INIS)

    Mcwilliams, A. J.; Daugherty, W. L.; Skidmore, T. E.

    2015-01-01

    The 9975 Type B shipping package is used within the DOE complex for shipping special nuclear materials. This package is re-certified annually in accordance with Safety Analysis Report for Packaging (SARP) requirements. The package is also used at the Savannah River Site as part of the long-term storage configuration of special nuclear materials. As such, the packages do not undergo annual recertification during storage, with uncertainty as to how long some of the package components will meet their functional requirements in the storage environment. The packages are currently approved for up to 15 years storage, and work continues to provide a technical basis to extend that period. This report describes efforts by the Savannah River National Laboratory (SRNL) to extend the service life estimate of Viton® GLT and GLT-S fluoroelastomer O-rings used in the 9975 shipping package. O-rings of both GLT and GLT-S compositions are undergoing accelerated aging at elevated temperature, and are periodically tested for compression stress relaxation (CSR) behavior. The CSR behavior of O-rings was evaluated at temperatures from 175 to 400 °F. These collective data were used to develop predictive models for extrapolation of CSR behavior to relevant service temperatures (< 156 °F). The predictive model developed from the CSR data conservatively indicates a service life of approximately 37 years for Viton GLT O-rings at the maximum effective service temperature of 156 °F. The estimated service life for Viton GLT-S O-rings is significantly longer.

  10. Examination of shipping packages 9975-01641, 9975-01692, 9975-03373, 9975-02101 AND 9975-02713

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-01-01

    SRNL has assisted in the examination of five 9975 shipping packages following storage of nuclear material in K-Area Complex (KAC). Two packages (9975-01641 and -01692) with water intrusion resulting from a roof leak were selected for detailed examination after internal fiberboard degradation (mold) was observed. 9975-01692 contained regions of saturated fiberboard and significant mold, while the second package was less degraded. A third package (9975-03373) was removed from storage for routine surveillance activities, and set aside for further examination after a musty odor was noted inside. No additional degradation was noted in 9975-03373, but the lower assembly could not be removed from the drum for detailed examination. Two additional packages (9975-02101 and -02713) identified for further examination were among a larger group selected for surveillance as part of a specific focus on high-wattage packages. These two packages displayed several non-conforming conditions, including the following: (1) the axial gap criterion was exceeded, (2) a significant concentration of moisture was found in the bottom fiberboard layers, with active mold in this area, (3) condensation and/or water stains were observed on internal components (drum, lid, air shield), and (4) both drums contained localized corrosion along the bottom lip. It is recommended that a new screening check be implemented for packages that are removed from storage, as well as high wattage packages remaining in storage. An initial survey for corrosion along the drum bottom lip of high wattage packages could identify potential degraded packages for future surveillance focus. In addition, after packages have been removed from storage (and unloaded), the drum bottom lip and underside should be inspected for corrosion. The presence of corrosion could signal the need to remove the lower fiberboard assembly for further inspection of the fiberboard and drum prior to recertification of the package.

  11. Dose assessment for public by packages shipping radioactive materials hypothetically sunk on the continental shelf. Annex 3

    International Nuclear Information System (INIS)

    Tsumune, Daisuke; Suzuki, Hiroshi; Saegusa, Toshiari; Watabe, Naohito; Asano, Hiroyuki; Maruyama, Koki; Kinehara, Yoshiki

    2001-01-01

    Radioactive materials such as spent fuel (SF), PuO 2 powder, high level wastes (HLW) and fresh mixed oxide (MOX) fuel have been transported by sea between Europe and Japan. Dose assessments for public have been performed in the past when the packages shipping radioactive materials were hypothetically sunk on the continental shelf. These studies employed various conditions and methods in their assessments and the results were not always the same. In this study, the dose assessment for these packages was performed under the same conditions and by the same methods. The effective dose equivalents of radiation exposure to the public for all materials become smaller than the previous evaluations due to more realistic assumption in this study. These evaluated results are far less than the effective dose equivalent limit (1 mSv year -1 ) by the ICRP recommendation. (author)

  12. Materials Assessment of Insulating Foam in the 9977 Shipping Package for Long-Term Storage - Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    McWilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-01

    The 9977 shipping package is being evaluated for long-term storage applications in the K-Area Complex (KAC) with specific focus on the packaging foam material. A rigid closed cell polyurethane foam, LAST-A-FOAM® FR-3716, produced by General Plastics Manufacturing Company is sprayed and expands to fill the void between the inner container and the outer shell of the package. The foam is sealed in this annular space and is not accessible. During shipping and storage, the foam experiences higher than ambient temperatures from the heat generated by nuclear material within the package creating the potential for degradation of the foam. A series of experiments is underway to determine the extent of foam degradation. Foam samples of three densities have been aging at elevated temperatures 160 °F, 160 °F + 50% relative humidity (RH), 185 °F, 215 °F, and 250 °F since 2014. Samples were periodically removed and tested. After approximately 80 weeks, samples conditioned at 160 °F, 160 °F + 50% RH, and 185 °F have retained initial property values while samples conditioned at 215 °F have reduced intumescence. Samples conditioned at 250 °F have shown the most degradation, loss of volume, mass, absorbed energy under compression, intumescence, and increased flammability. Based on the initial data, temperatures up to 185 °F have not yet shown an adverse effect on the foam properties and it is recommended that exposure of FR-3716 foam to temperatures in excess of 250 °F be avoided or minimized. Testing will continue beyond the 96 week mark. This will provide additional data to help define the long-term behavior for the lower temperature conditions. Additional testing will be pursued in an attempt to identify transition points (threshold times and temperatures) at the higher temperatures of interest, as well as possible benefits of aging within the relatively oxygen-free environment the foam experiences inside the 9977 shipping package.

  13. Environmental Tritium.

    OpenAIRE

    1984-01-01

    Environmental tritium was first observed in a helium fraction at a liquid air production facility in Germany in 1949. During the 1950s and early 1960s, huge amounts of artificial tritium were released into the atmosphere by nuclear testing. The environmental tritium level increased to more than 200 times the natural tritium level. Since the signing of a test ban treaty in 1963, the environmental tritium level has decreased, and analysis of recent Japanese rain samples has shown that the envir...

  14. Safety analysis report: packages. Argonne National Laboratory SLSF test train shipping container, P-1 shipment. Fissile material. Final report

    International Nuclear Information System (INIS)

    Meyer, C.A.

    1975-06-01

    The package is used to ship an instrumented test fuel bundle (test train) containing fissile material. The package assembly is Argonne National Laboratory (ANL) Model R1010-0032. The shipment is fissile class III. The packaging consists of an outer carbon steel container into which an inner container is placed; the inner container is separated from the outer container by urethane foam cushioning material. The test train is supported in the inner container by a series of transverse supports spaced along the length of the test train. Both the inner and outer containers are closed with bolted covers. The covers do not seal the containers in a leaktight manner. The gross weight of the shipment is about 8350 lb. The unirradiated fissile material content is less than 3 kg of UO 2 of up to 93.2 percent enrichment. This is a Type A quantity (transport group III and less than 3 curies) of radioactive material which does not require shielding, cooling or heating, or neutron absorption or moderation functions in its packaging. The maximum exterior dimensions of the container are 37 ft 11 in. long, 24 1 / 2 in. wide, and 19 3 / 4 in. high

  15. Safety Analysis Report: Packages, Pu oxide and Am oxide shipping cask: Packaging of fissile and other radioactive materials: Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1984-12-01

    The PuO 2 cask or 5320-3 cask is designed for shipment of americium or plutonium by surface transportation modes. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, dated 12/21/76, which invokes Title 10 Code of Federal Regulations, Part 71 (10 CFR 71) ''Packaging of Radioactive Materials for Transport,'' and Title 49 CFR Parts 171.179 ''Hazardous Materials Regulations.'' (US DOE Order 4580.1A, Chapter III, superseded manual chapter 0529 effective May 1981, but it retained the same 10 CFR 71 and 49 CFR 171-179 references

  16. Safety analysis report: packages. Pu oxide and Am oxide shipping cask (Packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1980-05-01

    The PuO 2 cask or SP 5320-2 and 3 cask is designed for surface shipment of americium or plutonium. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, and Chapter I, Interstate Commerce Commission. The package has been assessed for transport of up to 357 grams of plutonium (403 grams PuO 2 powder) and up to 176 grams of americium (200 grams AmO 2 powder), having a maximum decay heat of 203 watts. Criticality evaluation alone would allow the shipment as Fissile Class II but the radiation level of the cask, measured at the time of shipment, may exceed 50 mrem/h at the surface and require shipment as Fissile Class III. Sample calculations address only the more restrictive of the two materials, which in most cases is 238 PuO 2

  17. AGING AND SURVEILLANCE OF VITON GLT O-RINGS IN MODEL 9975 SHIPPING PACKAGES

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E; Skidmore,E; Daugherty, W; Dunn, A; Dunn, K

    2009-06-26

    Radioactive material packages (DOT Type B) such as the Model 9975 are used to transport Pu-bearing materials. The 9975 package provides double payload containment via nested stainless steel primary (PCV) and secondary (SCV) containment vessels. The containment vessels are closed by a conical plug sealed with dual O-rings (Figure 1) made of Parker Seals compound V0835-75, based on Viton{reg_sign} GLT fluoroelastomer. The outer O-ring is credited as being leaktight per ANSI N14.5 with a leak rate of <1E-07 ref cc/sec. The 9975 package is being used for interim storage in the K-Area Material Storage (KAMS) facility at the Savannah River Site. The aging performance of the O-rings is being studied to provide the storage facility a technical basis for service life prediction and safety analysis.

  18. An adaptive variable neighborhood search algorithm for a vehicle routing problem arising in small package shipping

    NARCIS (Netherlands)

    Stenger, Andreas; Vigo, Daniele; Enz, Steffen; Schwind, Michael

    2013-01-01

    In this paper, we investigate a routing problem arising in the last-mile delivery of small packages. The problem, called Multi-Depot Vehicle Routing Problem with Private fleet and Common carriers (MDVRPPC), is an extension of the Multi-Depot Vehicle Routing Problem (MDVRP) where customers can either

  19. STATUS REPORT FOR AGING STUDIES OF EPDM O-RING MATERIAL FOR THE H1616 SHIPPING PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Stefek, T.; Daugherty, W.; Skidmore, E.

    2012-08-31

    This is an interim status report for tasks carried out per Task Technical Plan SRNL-STI-2011-00506. A series of tasks/experiments are being performed at the Savannah River National Laboratory to monitor the aging performance of ethylene propylene diene monomer (EPDM) Orings used in the H1616 shipping package. The data will support the technical basis to extend the annual maintenance of the EPDM O-rings in the H1616 shipping package and to predict the life of the seals at bounding service conditions. Current expectations are that the O-rings will maintain a seal at bounding normal temperatures in service (152 F) for at least 12 months. The baseline aging data review suggests that the EPDM O-rings are likely to retain significant mechanical properties and sealing force at bounding service temperatures to provide a service life of at least 2 years. At lower, more realistic temperatures, longer service life is likely. Parallel compression stress relaxation and vessel leak test efforts are in progress to further validate this assessment and quantify a more realistic service life prediction. The H1616 shipping package O-rings were evaluated for baseline property data as part of this test program. This was done to provide a basis for comparison of changes in material properties and performance parameters as a function of aging. This initial characterization was limited to physical and mechanical properties, namely hardness, thickness and tensile strength. These properties appear to be consistent with O-ring specifications. Three H1616-1 Containment Vessels were placed in test conditions and are aging at temperatures ranging from 160 to 300 F. The vessels were Helium leak-tested initially and have been tested at periodic intervals after cooling to room temperature to determine if they meet the criterion of leaktightness defined in ANSI standard N14.5-97 (< 1E-07 std cc air/sec at room temperature). To date, no leak test failures have occurred. The cumulative time at

  20. Quality and physiological responses of two late-season sweet cherry cultivars 'Lapins' and 'Skeena' to modified atmosphere packaging (MAP) during simulated long distance ocean shipping

    Science.gov (United States)

    Flavor loss, skin darkening, pitting, splitting, pedicel browning, and decay are the major quality deteriorations in sweet cherries during storage/shipping. In this research, three modified atmosphere packaging (MAP) liners with varied gas permeability were evaluated for the effect on quality deteri...

  1. Refurbishment and modification of existing protective shipping packages (for 30-inch UF{sub 6} cylinders) per USDOT specification No. USA-DOT-21PF-1A

    Energy Technology Data Exchange (ETDEWEB)

    Housholder, W.R. [Nuclear Containers, Incorporated, Elizabethton, TN (United States)

    1991-12-31

    This paper addresses the refurbishment procedures for existing shipping containers for 30-inch diameter UF{sub 6} cylinders in accordance with DOT Specification 21PF-1 and the criteria used to determine rejection when such packages are unsuitable for refurbishment.

  2. Tritium conference days; Journees tritium

    Energy Technology Data Exchange (ETDEWEB)

    Garnier-Laplace, J.; Lebaron-Jacobs, L.; Sene, M.; Devin, P.; Chretien, V.; Le Guen, B.; Guetat, Ph.; Baglan, N.; Ansoborlo, E.; Boyer, C.; Masson, M.; Bailly-Du-Bois, P.; Jenkinson, St.; Wakeford, R.; Saintigny, Y.; Romeo, P.H.; Thompson, P.; Leterq, D.; Chastagner, F.; Cortes, P.; Philippe, M.; Paquet, F.; Fournier, M.

    2009-07-01

    This document gathers the slides of the available presentations given during this conference day. Twenty presentations out of 21 are assembled in the document and deal with: 1 - tritium in the environment (J. Garnier-Laplace); 2 - status of knowledge about tritium impact on health (L. Lebaron-Jacobs); 3 - tritium, discrete but present everywhere (M. Sene); 4 - management of tritium effluents from Areva NC La Hague site - related impact and monitoring (P. Devin); 5 - tritium effluents and impact in the vicinity of EDF's power plants (V. Chretien and B. Le Guen); 6 - contribution of CEA-Valduc centre monitoring to the knowledge of atmospheric tritiated water transfers to the different compartments of the environment (P. Guetat); 7 - tritium analysis in environment samples: constraints and means (N. Baglan); 8 - organically-linked tritium: the analyst view (E. Ansoborlo); 9 - study of tritium transfers to plants via OBT/HTO{sub air} and OBT/HTO{sub free} (C. Boyer); 10 - tritium in the British Channel (M. Masson and P. Bailly-Du-Bois); 11 - tritium in British coastal waters (S. Jenkinson); 12 - recent results from epidemiology (R. Wakeford); 13 - effects of tritiated thymidine on hematopoietic stem cells (P.H. Romeo); 14 - tritium management issue in Canada: the point of view from authorities (P. Thompson); 15 - experience feedback of the detritiation process of Valduc centre (D. Leterq); 16 - difficulties linked with tritiated wastes confinement (F. Chastagner); 17 - optimisation of tritium management in the ITER project (P. Cortes); 18 - elements of thought about the management of tritium generated by nuclear facilities (M. Philippe); 19 - CIPR's position about the calculation of doses and risks linked with tritium exposure (F. Paquet); 20 - tritium think tanks (M. Fournier). (J.S.)

  3. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

    International Nuclear Information System (INIS)

    1994-10-01

    The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results

  4. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results.

  5. Tritium-Powered Radiation Sensor Network

    Science.gov (United States)

    2015-09-01

    sources Tritium-filled, phosphor-lined vials are commercially available for night-vision gear applications.16 The tritium capsules purchased...support and are filled with silicon (Si) rubber adhesive producing sealed modules that meet DOT shipping regulations. The BA5590-style military...Ag) phosphor in the tritium capsules was not always a spectral match to all PV cells. We discovered that space-grade heterojunction PV cells (multi

  6. Guidelines for conducting impact tests on shipping packages for radioactive material

    International Nuclear Information System (INIS)

    Mok, G.C.; Carlson, R.W.; Lu, S.C.; Fischer, L.E.

    1995-09-01

    Federal regulation (10 CFR Part 71) specifies a number of impact conditions (free-drop, penetration, and puncture), under which a package for the transport of radioactive materials must be tested or evaluated to demonstrate compliance with the regulation. This report is a comprehensive guide to the planning and execution of these impact tests. The report identifies the required considerations for both the design, pre-, and post-test inspections of the test model and the measurement, recording, analysis, and reporting of the test data. The report also presents reasons for the requirements, identifies the major difficulties in meeting these requirements, and suggests possible methods to overcome the difficulties. Discussed in substantial detail is the use of scale models and instrumented measurements

  7. Tritium sources

    International Nuclear Information System (INIS)

    Glodic, S.; Boreli, F.

    1993-01-01

    Tritium is the only radioactive isotope of hydrogen. It directly follows the metabolism of water and it can be bound into genetic material, so it is very important to control levels of contamination. In order to define the state of contamination it is necessary to establish 'zero level', i.e. actual global inventory. The importance of tritium contamination monitoring increases with the development of fusion power installations. Different sources of tritium are analyzed and summarized in this paper. (author)

  8. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1989-01-01

    A general synthesis about tritium storage is achieved in this paper and a particular attention is given to practical application in the Fusion Technology Program. Tritium, storage under gaseous form and solid form are discussed (characteristics, advantages, disadvantages and equipments). The way of tritium storage is then discussed and a choice established as a function of a logic which takes into account the main working parameters

  9. Roadmapping - A Tool for Resolving Science and Technology Issues Related to Processing, Packaging, and Shipping Nuclear Materials and Waste

    Energy Technology Data Exchange (ETDEWEB)

    Luke, Dale Elden; Dixon, Brent Wayne; Murphy, James Anthony

    2002-06-01

    Roadmapping is an effective methodology to identify and link technology development and deployment efforts to a program's or project's needs and requirements. Roadmapping focuses on needed technical support to the baselines (and to alternatives to the baselines) where the probability of success is low (high uncertainty) and the consequences of failure are relatively high (high programmatic risk, higher cost, longer schedule, or higher ES&H risk). The roadmap identifies where emphasis is needed, i.e., areas where investments are large, the return on investment is high, or the timing is crucial. The development of a roadmap typically involves problem definition (current state versus the desired state) and major steps (functions) needed to reach the desired state. For Nuclear Materials (NM), the functions could include processing, packaging, storage, shipping, and/or final disposition of the material. Each function is examined to determine what technical development would be needed to make the function perform as desired. This requires a good understanding of the current state of technology and technology development and validation activities to ensure the viability of each step. In NM disposition projects, timing is crucial! Technology must be deployed within the project window to be of value. Roadmaps set the stage to keep the technology development and deployment focused on project milestones and ensure that the technologies are sufficiently mature when needed to mitigate project risk and meet project commitments. A recent roadmapping activity involved a 'cross-program' effort, which included NM programs, to address an area of significant concern to the Department of Energy (DOE) related to gas generation issues, particularly hydrogen. The roadmap that was developed defined major gas generation issues within the DOE complex and research that has been and is being conducted to address gas generation concerns. The roadmap also provided the basis

  10. Tritium handling experience at TFTR

    International Nuclear Information System (INIS)

    Anderson, J.L.; Gentile, C.; Hosea, J.

    1994-01-01

    In December 1993 the high power D-T experimental program on the Tokamak Fusion Test Reactor (TFTR) began. The transit the TFTR from a DOE general use facility to a low hazard category III nuclear facility has been completed successfully. The low hazard nuclear facility designation that the allowable on-site tritium inventory not exceed 50,000 Curies (1 Ci = 37 GBq). This is a TFTR Technical Safety Requirement. Tritium sealed in approved shipping containers does riot count against this inventory limit A second Technical Safety Requirement at TFTR is to have no more than 25,000 Ci at risk in a single location. From December, 1993 through mid-August, 1994 about 20 grams of tritium have been used in two gas injector assemblies and twelve neutral beam tritium injectors. The gas injected into TFTR vacuum is pumped by helium cryo-panels in the four neutral beam boxes. During non-operating periods the cryo-panels are warmed and the hydrogen am released and pumped into gas holding tanks in the tritium area. Gas in the holding tanks is oxidized in the Torus Cleanup System (TCS) and the hydrogen isotopes are collected, as water, on disposable molecular sieve beds (DMSB). These beds are then removed from the system and shipped off-site for tritium recovery or for long-term storage. Several problems in the tritium cleanup systems have occurred following a leak of sulfur hexafluoride (SF 6 ) from a neutral hewn high voltage enclosure ion source and subsequent pumping to the gas holding tanks. These problems included failure of several-moisture sensors, false readings on tritium monitors and, partial loss of catalytic activity in the TCS recombiner. Procedures for dealing with and removing this contaminant gas had to be developed and implemented. The results from this occurrence provide valuable guidance for future tritium burning fusion machines

  11. Sixth Status Report: Testing of Aged Softwood Fiberboard Material for the 9975 Shipping Package

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-03-31

    Samples have been prepared from several 9975 lower fiberboard subassemblies fabricated from softwood fiberboard. Physical, mechanical and thermal properties have been measured following varying periods of conditioning in each of several environments. These tests have been conducted in the same manner as previous testing on cane fiberboard samples. Overall, similar aging trends are observed for softwood and cane fiberboard samples, with a few differences. Some softwood fiberboard properties tend to degrade faster in some environments, while some cane fiberboard properties degrade faster in the two most aggressive environments. As a result, it is premature to assume both materials will age at the same rates, and the preliminary aging models developed for cane fiberboard might not apply to softwood fiberboard. However, it is expected that both cane and softwood fiberboard assemblies will perform satisfactorily in conforming packages stored in a typical KAC storage environment for up to 15 years. Samples from an additional 3 softwood fiberboard assemblies have begun aging during the past year to provide information on the variability of softwood fiberboard behavior. Aging and testing of softwood fiberboard will continue and additional data will be collected to support development of an aging model specific to softwood fiberboard.

  12. Tritium trick

    Science.gov (United States)

    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.

  13. TFTR tritium inventory accountability system

    International Nuclear Information System (INIS)

    Saville, C.; Ascione, G.; Elwood, S.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Stencel, J.; Voorhees, D.; Tilson, C.

    1995-01-01

    This paper discusses the program, PPPL (Princeton Plasma Physics Laboratory) Material Control and Accountability Plan, that has been implemented to track US Department of Energy's tritium and all other accountable source material. Specifically, this paper details the methods used to measure tritium in various systems at the Tokamak Fusion Test Reactor; resolve inventory differences; perform inventory by difference inside the Tokamak; process and measure plasma exhaust and other effluent gas streams; process, measure and ship scrap or waste tritium on molecular sieve beds; and detail organizational structure of the Material Control and Accountability group. In addition, this paper describes a Unix-based computerized software system developed at PPPL to account for all tritium movements throughout the facility. 5 refs., 2 figs

  14. TFTR tritium inventory accountability system

    Energy Technology Data Exchange (ETDEWEB)

    Saville, C.; Ascione, G.; Elwood, S.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Stencel, J.; Voorhees, D.; Tilson, C. [Plasma Physics Lab., Princeton, NJ (United States)

    1995-10-01

    This paper discusses the program, PPPL (Princeton Plasma Physics Laboratory) Material Control and Accountability Plan, that has been implemented to track US Department of Energy`s tritium and all other accountable source material. Specifically, this paper details the methods used to measure tritium in various systems at the Tokamak Fusion Test Reactor; resolve inventory differences; perform inventory by difference inside the Tokamak; process and measure plasma exhaust and other effluent gas streams; process, measure and ship scrap or waste tritium on molecular sieve beds; and detail organizational structure of the Material Control and Accountability group. In addition, this paper describes a Unix-based computerized software system developed at PPPL to account for all tritium movements throughout the facility. 5 refs., 2 figs.

  15. Package

    Directory of Open Access Journals (Sweden)

    Arsić Zoran

    2013-01-01

    Full Text Available It is duty of the seller to pack the goods in a manner which assures their safe arrival and enables their handling in transit and at the place of destination. The problem of packing is relevant in two main respects. First of all the buyer is in certain circumstances entitled to refuse acceptance of the goods if they are not properly packed. Second, the package is relevant to calculation of price and freight based on weight. In the case of export trade, the package should conform to the legislation in the country of destination. The impact of package on environment is regulated by environment protection regulation of Republic if Serbia.

  16. Tritium effluent removal system

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Gibbs, G.E.

    1978-01-01

    An air detritiation system has been developed and is in routine use for removing tritium and tritiated compounds from glovebox effluent streams before they are released to the atmosphere. The system is also used, in combination with temporary enclosures, to contain and decontaminate airborne releases resulting from the opening of tritium containment systems during maintenance and repair operations. This detritiation system, which services all the tritium handling areas at Mound Facility, has played an important role in reducing effluents and maintaining them at 2 percent of the level of 8 y ago. The system has a capacity of 1.7 m 3 /min and has operated around the clock for several years. A refrigerated in-line filtration system removes water, mercury, or pump oil and other organics from gaseous waste streams. The filtered waste stream is then heated and passed through two different types of oxidizing beds; the resulting tritiated water is collected on molecular sieve dryer beds. Liquids obtained from regenerating the dryers and from the refrigerated filtration system are collected and transferred to a waste solidification and packaging station. Component redundancy and by-pass capabilities ensure uninterrupted system operation during maintenance. When processing capacity is exceeded, an evacuated storage tank of 45 m 3 is automatically opened to the inlet side of the system. The gaseous effluent from the system is monitored for tritium content and recycled or released directly to the stack. The average release is less than 1 Ci/day. The tritium effluent can be reduced by isotopically swamping the tritium; this is accomplished by adding hydrogen prior to the oxidizer beds, or by adding water to the stream between the two final dryer beds

  17. Tritium research activities in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Jung, E-mail: kjjung@nfri.re.kr [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Yun, Sei-Hun, E-mail: shyun@nfri.re.kr [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Chang, Min Ho; Kang, Hyun-Goo; Chung, Dongyou; Cho, Seungyon; Lee, Hyeon Gon [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Chung, Hongsuk; Choi, Woo-Seok [Korea Atomic Energy Research Institute, Yusung-gu, Daejeon 305-353 (Korea, Republic of); Song, Kyu-Min; Moon, Chang-Bae [Korea Hydro & Nuclear Power Central Research Institute, Yusung-gu, Daejeon 305-343 (Korea, Republic of); Lee, Euy Soo [Dongguk University, Jung-gu, Seoul, 100-715 (Korea, Republic of); Cho, Jungho; Kim, Dong-Sun [Kongju National University, Cheonan, Chungnam, 330-717 (Korea, Republic of); Moon, Hung-Man [Daesung Industrial Gases Co., Ltd., Danwon-gu, Ansan-si, Gyeonggi-do, 425-090 (Korea, Republic of); Noh, Seung Jeong [Dankook University, Suji-gu, Yongin-si, Gyeonggi-do, 448-701 (Korea, Republic of); Ju, Hyunchul [Inha University, Nam-gu, Incheon, 402-751 (Korea, Republic of); Hong, Tae-Whan [Korea National University of Transportation, Chungju, Chungbuk, 380-702 (Korea, Republic of)

    2016-12-15

    Highlights: • NFRI, KAERI and KHNP CRI are major leading group for the ITER tritium SDS design; studying engineering, simulation of hydride bed, risk analysis (on safety, HAZOP), basic study, control logic & sequential operation, and others. KHNP has WTRF which gives favorable experiences for collaboration researchers. • Supplementary research partners: Five Universities (Dongguk University and POSTECH, Inha University, Dankook University, Korea National Transport University, and Kongju National University) and one industrial company (Daesung Industrial Gases Co., Ltd.); studying on basic and engineering, programming & simulation on the various topics for ITER tritium SDS, TEP, ISS, ADS, and etc. - Abstract: Major progress in tritium research in the Republic of Korea began when Korea became responsible for ITER tritium Storage and Delivery System (SDS) procurement package which is part of the ITER Fuel Cycle. To deliver the tritium SDS package, a variety of research institutes, universities and industry have respectively taken roles and responsibilities in developing technologies that have led to significant progress. This paper presents the current work and status of tritium related technological research and development (R&D) in Korea and introduces future R&D plans in the area of fuel cycle systems for fusion power generation.

  18. Environmental tritium

    International Nuclear Information System (INIS)

    Gans, I.

    1974-10-01

    The radioactive hydrogen isotope tritium can be found in all water occurrences. The concentration of natural tritium measured before 1954 amounts to 26 picocuries per liter in precipitation, 5 to 20 picocuries per liter in surface water, and 1 picocurie per liter in sea water. Since then, due to thermonuclear waepons tests in the atmosphere, considerably higher concentrations have been measured - 1963 the annual mean for precipitation went up to 10 4 picocuries per liter. Today in Middle Europe some hundred picocuries per liter are found in precipitation and surface water, less than 100 picocuries per liter in sea water, and in general less than 15 picocuries per liter in ground water. Artificial tritium today is applied in large scale in research and industry. It is of special importance as waste in the peaceful uses of nuclear energy. In the future, however, tritium emissions from nuclear power plants are less important than releases from reprocessing plants. Estimations show that the global environmental impact is small. For regions with a large density of nuclear power installations, radiation exposures of the order of magnitude of 10 mrem are predicted with pessimistic assumptions. More realistic assumptions lead to dose values of about 0.1 mrem caused by the influence of tritium. This is 80% of the dose caused by the release of radioactive material from nuclear power installations. (orig.) [de

  19. The Chalk River Tritium Extraction Plant

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Harrison, T.E.; Spagnolo, D.A.

    1990-01-01

    The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1-35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T 2 . The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: it produces pure tritium for use in fusion research and development, and it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors (author)

  20. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Cole, M.J.; Combs, S.K.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foust, C.R.; Langley, R.A.; Milora, S.L.; Qualls, A.L.; Wilgen, J.B.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

    1992-01-01

    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) phase. The existing TFTR deuterium pellet injector (DPI) has been 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 to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at 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. The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were 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 pellet. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  1. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    Energy Technology Data Exchange (ETDEWEB)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  2. What is included with your online e-cigarette order? An analysis of e-cigarette shipping, product and packaging features.

    Science.gov (United States)

    Kong, Amanda Y; Derrick, Jason C; Abrantes, Anthony S; Williams, Rebecca S

    2016-06-29

    The electronic cigarette industry is growing, with youth using e-cigarettes at higher rates than they are using cigarettes, and retail and online sales projected to reach $10 billion in 2017. Minimal regulation of the production and marketing of e-cigarettes exists to date, which has allowed companies to promote unsupported claims. We assessed the shipping, product features and packaging of a wide variety of e-cigarettes purchased online by adults and youth. The most popular internet e-cigarette vendors were identified from a larger study of internet tobacco vendors. Between August 2013 and June 2014, adults made 56 purchase attempts from online vendors, and youth made 98 attempts. Packages received were assessed for exterior and internal packaging features, including product information, health warnings and additional materials. We analysed a total of 125 orders featuring 86 unique brands of e-cigarettes. The contents were rarely indicated on package exteriors. Product information came with just 60% of orders and just 38.4% included an instruction manual. Only 44.6% of products included a health warning, and some had unsupported claims, such as lack of secondhand smoke exposure. Additionally, some products were leaking e-liquid and battery fluid on arrival. A large variety of e-cigarette products are manufactured and marketed to consumers. Many products do not include instructions for use, and unsupported claims are being presented to consumers. Effective federal regulation of the manufacturing, packaging, product information and health claims surrounding e-cigarettes is necessary to ensure consumers are presented with accurate e-cigarette use information. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

  3. Tritium handling in vacuum systems

    Energy Technology Data Exchange (ETDEWEB)

    Gill, J.T. [Monsanto Research Corp., Miamisburg, OH (United States). Mound Facility; Coffin, D.O. [Los Alamos National Lab., NM (United States)

    1986-10-01

    This report provides a course in Tritium handling in vacuum systems. Topics presented are: Properties of Tritium; Tritium compatibility of materials; Tritium-compatible vacuum equipment; and Tritium waste treatment.

  4. Magmatic tritium

    International Nuclear Information System (INIS)

    Goff, F.; Aams, A.I.; McMurtry, G.M.; Shevenell, L.; Pettit, D.R.; Stimac, J.A.; Werner, C.

    1997-01-01

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory. Detailed geochemical sampling of high-temperature fumaroles, background water, and fresh magmatic products from 14 active volcanoes reveal that they do not produce measurable amounts of tritium ( 3 H) of deep origin ( 2 O). On the other hand, all volcanoes produce mixtures of meteoric and magmatic fluids that contain measurable 3 H from the meteoric end-member. The results show that cold fusion is probably not a significant deep earth process but the samples and data have wide application to a host of other volcanological topics

  5. Tritium accountancy

    International Nuclear Information System (INIS)

    Avenhaus, R.; Spannagel, G.

    1995-01-01

    Conventional accountancy means that for a given material balance area and a given interval of time the tritium balance is established so that at the end of that interval of time the book inventory is compared with the measured inventory. In this way, an optimal effectiveness of accountancy is achieved. However, there are still further objectives of accountancy, namely the timely detection of anomalies as well as the localization of anomalies in a major system. It can be shown that each of these objectives can be optimized only at the expense of the others. Recently, Near-Real-Time Accountancy procedures have been studied; their methodological background as well as their merits will be discussed. (orig.)

  6. Magmatic tritium

    Energy Technology Data Exchange (ETDEWEB)

    Goff, F.; Aams, A.I. [Los Alamos National Lab., NM (United States); McMurtry, G.M. [Univ. of Hawaii, Honolulu, HI (United States); Shevenell, L. [Univ. of Nevada, Reno, NV (United States); Pettit, D.R. [National Aeronautics and Space Administration (United States); Stimac, J.A. [Union Geothermal Company (United States); Werner, C. [Pennsylvania State Univ., University Park, PA (United States)

    1997-07-01

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory. Detailed geochemical sampling of high-temperature fumaroles, background water, and fresh magmatic products from 14 active volcanoes reveal that they do not produce measurable amounts of tritium ({sup 3}H) of deep origin (<0.1 T.U. or <0.32 pCi/kg H{sub 2}O). On the other hand, all volcanoes produce mixtures of meteoric and magmatic fluids that contain measurable {sup 3}H from the meteoric end-member. The results show that cold fusion is probably not a significant deep earth process but the samples and data have wide application to a host of other volcanological topics.

  7. Status Report - Softwood Fiberboard Properties And Degradation Rates For Storage Of The 9975 Shipping Package In KAC

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-12-22

    Thermal, mechanical and physical properties have been measured on softwood fiberboard samples following accelerated aging for up to approximately 7 years. The aging environments have included elevated temperature < 250 ºF (the maximum allowed service temperature for fiberboard in 9975 packages) and elevated humidity. The results from this testing have been analyzed, and preliminary aging models fit to the data. Correlations relating several properties (thermal conductivity, energy absorption, weight, dimensions and density) to their rate of change in potential storage environments have been developed. Combined with acceptance criteria and an estimate of the actual conditions the fiberboard experiences in KAC, these models allow development of service life predictions.

  8. Status Report - Cane Fiberboard Properties And Degradation Rates For Storage Of The 9975 Shipping Package In KAC

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-12-22

    Thermal, mechanical and physical properties have been measured on cane fiberboard samples following accelerated aging for up to approximately 10 years. The aging environments have included elevated temperature < 250 ºF (the maximum allowed service temperature for fiberboard in 9975 packages) and elevated humidity. The results from this testing have been analyzed, and aging models fit to the data. Correlations relating several properties (thermal conductivity, energy absorption, weight, dimensions and density) to their rate of change in potential storage environments have been developed. Combined with an estimate of the actual conditions the fiberboard experiences in KAC, these models allow development of service life predictions.

  9. Analysis of a ship-to-ship collision

    International Nuclear Information System (INIS)

    Porter, V.L.; Ammerman, D.J.

    1996-01-01

    Sandia National Laboratories is involved in a safety assessment for the shipment of radioactive material by sea. One part of this study is investigation of the consequences of ship-to-ship collisions. This paper describes two sets of finite element analyses performed to assess the structural response of a small freighter and the loading imparted to radioactive material (RAM) packages during several postulated collision scenarios with another ship. The first series of analyses was performed to evaluate the amount of penetration of the freighter hull by a striking ship of various masses and initial velocities. Although these analyses included a representation of a single RAM package, the package was not impacted during the collision so forces on the package could not be computed. Therefore, a second series of analyses incorporating a representation of a row of seven packages was performed to ensure direct package impact by the striking ship. Average forces on a package were evaluated for several initial velocities and masses of the striking ship. In addition to. providing insight to ship and package response during a few postulated ship collisions scenarios, these analyses will be used to benchmark simpler ship collision models used in probabilistic risk assessment analyses

  10. Shipment and Storage Containers for Tritium Production Transportation Casks

    International Nuclear Information System (INIS)

    Massey, W.M.

    1998-04-01

    The need for a shipping and storage container for the Tritium production transportation casks is addressed in this report. It is concluded that a shipping and storage container is not required. A recommendation is made to eliminate the requirement for this container because structural support and inerting requirements can be satisfied completely by the cask with a removable basket

  11. International shipment of light weight radioisotopic heater units (LWRHU) using the USA/9516/B(U)F Mound 1 kW shipping package in support of the ''Pluto Express'' mission

    International Nuclear Information System (INIS)

    Barklay, Chadwick D.; Merten, C. William

    1997-01-01

    Radioisotopes have provided heat that has been used to maintain specific operating environments within remote satellites and spacecraft. For the 'Pluto Express' mission the 238 PuO 2 fueled light weight radioisotopic heater unit (LWRHU) will be used within the spacecraft. Since the current plan for the 'Pluto Express' mission incorporates the use of a Russian launch platform for the spacecraft, the LWRHUs must be transported in an internationally certified shipping container. An internationally certified shipping package that is versatile enough to be reconfigured to transport the LWRHUs that will be required to support the 'Pluto Express' mission is the Mound USA/9516/B(U)F

  12. Tritium activities in Canada

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1995-01-01

    Canadian tritium activites comprise three major interests: utilites, light manufacturers, and fusion. There are 21 operating CANDU reactors in Canada; 19 with Ontario Hydro and one each with Hydro Quebec and New Brunswick Power. There are two light manufacturers, two primary tritium research facilities (at AECL Chalk River and Ontario Hydro Technologies), and a number of industry and universities involved in design, construction, and general support of the other tritium activities. The largest tritum program is in support of the CANDU reactors, which generate tritium in the heavy water as a by-product of normal operation. Currently, there are about 12 kg of tritium locked up in the heavy water coolant and moderator of these reactors. The fusion work is complementary to the light manufacturing, and is concerned with tritium handling for the ITER program. This included design, development and application of technologies related to Isotope Separation, tritium handling, (tritiated) gas separation, tritium-materials interaction, and plasma fueling

  13. The Tritium White Paper

    International Nuclear Information System (INIS)

    2009-01-01

    This publication proposes a synthesis of the activities of two work-groups between May 2008 and April 2010. It reports the ASN's (the French Agency for Nuclear Safety) point of view, describes its activities and actions, and gives some recommendations. It gives a large and detailed overview of the knowledge status on tritium: tritium source inventory, tritium origin, management processes, capture techniques, reduction, tritium metrology, impact on the environment, impacts on human beings

  14. Tritium in the Channel

    International Nuclear Information System (INIS)

    Masson, M.; Fievet, B.; Bailly-Du-Bois, P.; Olivier, A.; Tenailleau, L.

    2009-01-01

    After having recalled that sea waters entering the Channel exhibit a natural concentration of tritium, the authors outline that spent nuclear fuel reprocessing plants are now the main sources of tritium for marine ecosystems as some oceanographic campaigns showed it. If data about the presence of tritium in water are numerous, data concerning the presence of tritiated water and of organically bound tritium in organisms are much less frequent. However, some surveys have been performed along the Channel French coasts

  15. Development of Tritium Permeation Analysis Code and Tritium Transport in a High Temperature Gas-Cooled Reactor Coupled with Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2010-06-01

    Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.

  16. TFTR tritium handling concepts

    International Nuclear Information System (INIS)

    Garber, H.J.

    1976-01-01

    The Tokamak Fusion Test Reactor, to be located on the Princeton Forrestal Campus, is expected to operate with 1 to 2.5 MA tritium--deuterium plasmas, with the pulses involving injection of 50 to 150 Ci (5 to 16 mg) of tritium. Attainment of fusion conditions is based on generation of an approximately 1 keV tritium plasma by ohmic heating and conversion to a moderately hot tritium--deuterium ion plasma by injection of a ''preheating'' deuterium neutral beam (40 to 80 keV), followed by injection of a ''reacting'' beam of high energy neutral deuterium (120 to 150 keV). Additionally, compressions accompany the beam injections. Environmental, safety and cost considerations led to the decision to limit the amount of tritium gas on-site to that required for an experiment, maintaining all other tritium in ''solidified'' form. The form of the tritium supply is as uranium tritide, while the spent tritium and other hydrogen isotopes are getter-trapped by zirconium--aluminum alloy. The issues treated include: (1) design concepts for the tritium generator and its purification, dispensing, replenishment, containment, and containment--cleanup systems; (2) features of the spent plasma trapping system, particularly the regenerable absorption cartridges, their integration into the vacuum system, and the handling of non-getterables; (3) tritium permeation through the equipment and the anticipated releases to the environment; (4) overview of the tritium related ventilation systems; and (5) design bases for the facility's tritium clean-up systems

  17. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

    The main objective of this publication is to provide practical guidance and recommendations on operational radiation protection aspects related to the safe handling of tritium in laboratories, industrial-scale nuclear facilities such as heavy-water reactors, tritium removal plants and fission fuel reprocessing plants, and facilities for manufacturing commercial tritium-containing devices and radiochemicals. The requirements of nuclear fusion reactors are not addressed specifically, since there is as yet no tritium handling experience with them. However, much of the material covered is expected to be relevant to them as well. Annex III briefly addresses problems in the comparatively small-scale use of tritium at universities, medical research centres and similar establishments. However, the main subject of this publication is the handling of larger quantities of tritium. Operational aspects include designing for tritium safety, safe handling practice, the selection of tritium-compatible materials and equipment, exposure assessment, monitoring, contamination control and the design and use of personal protective equipment. This publication does not address the technologies involved in tritium control and cleanup of effluents, tritium removal, or immobilization and disposal of tritium wastes, nor does it address the environmental behaviour of tritium. Refs, figs and tabs

  18. Surface tritium contamination studies

    International Nuclear Information System (INIS)

    Sienkiewicz, C.J.

    1986-01-01

    Glovebox wipe surveys were conducted to correlate surface tritium contamination with atmospheric tritium levels. Surface contamination was examined as a function of tritium concentration and limited to the HT/T 2 form. The previously predicted relationship between atmospheric HTO concentration and cleanup times was examined in order to predict a model for atmospheric detritiation of stainless steel enclosures. 2 figures, 2 tables

  19. Tritium conference days

    International Nuclear Information System (INIS)

    Garnier-Laplace, J.; Lebaron-Jacobs, L.; Sene, M.; Devin, P.; Chretien, V.; Le Guen, B.; Guetat, Ph.; Baglan, N.; Ansoborlo, E.; Boyer, C.; Masson, M.; Bailly-Du-Bois, P.; Jenkinson, St.; Wakeford, R.; Saintigny, Y.; Romeo, P.H.; Thompson, P.; Leterq, D.; Chastagner, F.; Cortes, P.; Philippe, M.; Paquet, F.; Fournier, M.

    2009-01-01

    This document gathers the slides of the available presentations given during this conference day. Twenty presentations out of 21 are assembled in the document and deal with: 1 - tritium in the environment (J. Garnier-Laplace); 2 - status of knowledge about tritium impact on health (L. Lebaron-Jacobs); 3 - tritium, discrete but present everywhere (M. Sene); 4 - management of tritium effluents from Areva NC La Hague site - related impact and monitoring (P. Devin); 5 - tritium effluents and impact in the vicinity of EDF's power plants (V. Chretien and B. Le Guen); 6 - contribution of CEA-Valduc centre monitoring to the knowledge of atmospheric tritiated water transfers to the different compartments of the environment (P. Guetat); 7 - tritium analysis in environment samples: constraints and means (N. Baglan); 8 - organically-linked tritium: the analyst view (E. Ansoborlo); 9 - study of tritium transfers to plants via OBT/HTO air and OBT/HTO free (C. Boyer); 10 - tritium in the British Channel (M. Masson and P. Bailly-Du-Bois); 11 - tritium in British coastal waters (S. Jenkinson); 12 - recent results from epidemiology (R. Wakeford); 13 - effects of tritiated thymidine on hematopoietic stem cells (P.H. Romeo); 14 - tritium management issue in Canada: the point of view from authorities (P. Thompson); 15 - experience feedback of the detritiation process of Valduc centre (D. Leterq); 16 - difficulties linked with tritiated wastes confinement (F. Chastagner); 17 - optimisation of tritium management in the ITER project (P. Cortes); 18 - elements of thought about the management of tritium generated by nuclear facilities (M. Philippe); 19 - CIPR's position about the calculation of doses and risks linked with tritium exposure (F. Paquet); 20 - tritium think tanks (M. Fournier). (J.S.)

  20. Environmental aspects of tritium

    International Nuclear Information System (INIS)

    Quisenberry, D.R.

    1979-01-01

    The potential radiological implications of environmental tritium releases must be determined in order to develop a programme for dealing with the tritium inventory predicted for the nuclear power industry which, though still in its infancy, produces tritium in megacurie quantities annually. Should the development of fusion power generation become a reality, it will create a potential source for large releases of tritium, much of it in the gaseous state. At present about 90% of the tritium produced enters the environment through gaseous and liquid effluents and is deposited in the hydrosphere as tritiated water. Tritium can be assimilated by plants and animals and organically bound, regardless of the exposure pathway. However, there appears to be no concentration factor relative to hydrogen at any level of food chains analysed to date. The body burden, for man, is dependent on the exposure pathway and tissue-bound fractions are primarily the result of organically bound tritium in food. (author)

  1. Tritium pellet injector results

    International Nuclear Information System (INIS)

    Fisher, P.W.; Bauer, M.L.; Baylor, L.R.; Deleanu, L.E.; Fehling, D.T.; Milora, S.L.; Whitson, J.C.

    1988-01-01

    Injection of solid tritium pellets is considered to be the most promising way of fueling fusion reactors. The Tritium Proof-of- Principle (TPOP) experiment has demonstrated the feasibility of forming and accelerating tritium pellets. This injector is based on the pneumatic pipe-gun concept, in which pellets are formed in situ in the barrel and accelerated with high-pressure gas. This injector is ideal for tritium service because there are no moving parts inside the gun and because no excess tritium is required in the pellet production process. Removal of 3 He from tritium to prevent blocking of the cryopumping action by the noncondensible gas has been demonstrated with a cryogenic separator. Pellet velocities of 1280 m/s have been achieved for 4-mm-diam by 4-mm-long cylindrical tritium pellets with hydrogen propellant at 6.96 MPa (1000 psi). 10 refs., 10 figs

  2. Tritium sampling and measurement

    International Nuclear Information System (INIS)

    Wood, M.J.; McElroy, R.G.; Surette, R.A.; Brown, R.M.

    1993-01-01

    Current methods for sampling and measuring tritium are described. Although the basic techniques have not changed significantly over the last 10 y, there have been several notable improvements in tritium measurement instrumentation. The design and quality of commercial ion-chamber-based and gas-flow-proportional-counter-based tritium monitors for tritium-in-air have improved, an indirect result of fusion-related research in the 1980s. For tritium-in-water analysis, commercial low-level liquid scintillation spectrometers capable of detecting tritium-in-water concentrations as low as 0.65 Bq L-1 for counting times of 500 min are available. The most sensitive method for tritium-in-water analysis is still 3He mass spectrometry. Concentrations as low as 0.35 mBq L-1 can be detected with current equipment. Passive tritium-oxide-in-air samplers are now being used for workplace monitoring and even in some environmental sampling applications. The reliability, convenience, and low cost of passive tritium-oxide-in-air samplers make them attractive options for many monitoring applications. Airflow proportional counters currently under development look promising for measuring tritium-in-air in the presence of high gamma and/or noble gas backgrounds. However, these detectors are currently limited by their poor performance in humidities over 30%. 133 refs

  3. Confinement and Tritium Stripping Systems for APT Tritium Processing

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Heung, L.K.

    1997-10-20

    This report identifies functions and requirements for the tritium process confinement and clean-up system (PCCS) and provides supporting technical information for the selection and design of tritium confinement, clean-up (stripping) and recovery technologies for new tritium processing facilities in the Accelerator for the Production of Tritium (APT). The results of a survey of tritium confinement and clean-up systems for large-scale tritium handling facilities and recommendations for the APT are also presented.

  4. Environmental tritium in trees

    International Nuclear Information System (INIS)

    Brown, R.M.

    1979-01-01

    The distribution of environmental tritium in the free water and organically bound hydrogen of trees growing in the vicinity of the Chalk River Nuclear Laboratories (CRNL) has been studied. The regional dispersal of HTO in the atmosphere has been observed by surveying the tritium content of leaf moisture. Measurement of the distribution of organically bound tritium in the wood of tree ring sequences has given information on past concentrations of HTO taken up by trees growing in the CRNL Liquid Waste Disposal Area. For samples at background environmental levels, cellulose separation and analysis was done. The pattern of bomb tritium in precipitation of 1955-68 was observed to be preserved in the organically bound tritium of a tree ring sequence. Reactor tritium was discernible in a tree growing at a distance of 10 km from CRNL. These techniques provide convenient means of monitoring dispersal of HTO from nuclear facilities. (author)

  5. NNSA TRITIUM SUPPLY CHAIN

    Energy Technology Data Exchange (ETDEWEB)

    Wyrick, Steven [Savannah River National Laboratory, Aiken, SC, USA; Cordaro, Joseph [Savannah River National Laboratory, Aiken, SC, USA; Founds, Nanette [National Nuclear Security Administration, Albuquerque, NM, USA; Chambellan, Curtis [National Nuclear Security Administration, Albuquerque, NM, USA

    2013-08-21

    Savannah River Site plays a critical role in the Tritium Production Supply Chain for the National Nuclear Security Administration (NNSA). The entire process includes: • Production of Tritium Producing Burnable Absorber Rods (TPBARs) at the Westinghouse WesDyne Nuclear Fuels Plant in Columbia, South Carolina • Production of unobligated Low Enriched Uranium (LEU) at the United States Enrichment Corporation (USEC) in Portsmouth, Ohio • Irradiation of TPBARs with the LEU at the Tennessee Valley Authority (TVA) Watts Bar Reactor • Extraction of tritium from the irradiated TPBARs at the Tritium Extraction Facility (TEF) at Savannah River Site • Processing the tritium at the Savannah River Site, which includes removal of nonhydrogen species and separation of the hydrogen isotopes of protium, deuterium and tritium.

  6. High-pressure tritium

    International Nuclear Information System (INIS)

    Coffin, D.O.

    1976-01-01

    Some solutions to problems of compressing and containing tritium gas to 200 MPa at 700 0 K are discussed. The principal emphasis is on commercial compressors and high-pressure equipment that can be easily modified by the researcher for safe use with tritium. Experience with metal bellows and diaphragm compressors has been favorable. Selection of materials, fittings, and gauges for high-pressure tritium work is also reviewed briefly

  7. Tritium in metals

    International Nuclear Information System (INIS)

    Schober, T.

    1990-01-01

    In this Chapter a review is given of some of the important features of metal tritides as opposed to hydrides and deuterides. After an introduction to the topics of tritium and tritium in metals information will be presented on a variety of metal-tritium systems. Of main interest here are the differences from the classic hydrogen behavior; the so called isotope effect. A second important topic is that of aging effects produced by the accumulation of 3 He in the samples. (orig.)

  8. Radionuclide Basics: Tritium

    Science.gov (United States)

    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.

  9. Protection against tritium radiations

    International Nuclear Information System (INIS)

    Bal, Georges

    1964-05-01

    This report presents the main characteristics of tritium, describes how it is produced as a natural or as an artificial radio-element. It outlines the hazards related to this material, presents how materials and tools are contaminated and decontaminated. It addresses the issue of permissible maximum limits: factors of assessment of the risk induced by tritium, maximum permissible activity in body water, maximum permissible concentrations in the atmosphere. It describes the measurement of tritium activity: generalities, measurement of gas activity and of tritiated water steam, tritium-induced ionisation in an ionisation chamber, measurement systems using ionisation chambers, discontinuous detection of tritium-containing water in the air, detection of surface contamination [fr

  10. Tritium fuel cycle modeling and tritium breeding analysis for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli; Pan, Lei; Lv, Zhongliang; Li, Wei; Zeng, Qin, E-mail: zengqin@ustc.edu.cn

    2016-05-15

    Highlights: • A modified tritium fuel cycle model with more detailed subsystems was developed. • The mean residence time method applied to tritium fuel cycle calculation was updated. • Tritium fuel cycle analysis for CFETR was carried out. - Abstract: Attaining tritium self-sufficiency is a critical goal for fusion reactor operated on the D–T fuel cycle. The tritium fuel cycle models were developed to describe the characteristic parameters of the various elements of the tritium cycle as a tool for evaluating the tritium breeding requirements. In this paper, a modified tritium fuel cycle model with more detailed subsystems and an updated mean residence time calculation method was developed based on ITER tritium model. The tritium inventory in fueling system and in plasma, supposed to be important for part of the initial startup tritium inventory, was considered in the updated mean residence time method. Based on the model, the tritium fuel cycle analysis of CFETR (Chinese Fusion Engineering Testing Reactor) was carried out. The most important two parameters, the minimum initial startup tritium inventory (I{sub m}) and the minimum tritium breeding ratio (TBR{sub req}) were calculated. The tritium inventories in steady state and tritium release of subsystems were obtained.

  11. Tritium breeding in fusion reactors

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1982-10-01

    Key technological problems that influence tritium breeding in fusion blankets are reviewed. The breeding potential of candidate materials is evaluated and compared to the tritium breeding requirements. The sensitivity of tritium breeding to design and nuclear data parameters is reviewed. A framework for an integrated approach to improve tritium breeding prediction is discussed with emphasis on nuclear data requirements

  12. Installation of the Tritium Purification System

    Energy Technology Data Exchange (ETDEWEB)

    Labik, G.; Golian, T.; Satkofsky, J.; Sichta, P.; Crook, D.; Dudek, L.; Coward, G.; Parsells, R. [Princeton Plasma Physics Lab., NJ (United States)

    1995-12-31

    This paper describes and details the design, the tasks, and the considerations for the mechanical and electrical installation of the TFTR Tritium Purification System (TPS) at the Princeton Plasma Physics Laboratory (PPPL). Canadian Fusion Fuels Technology Project (CFFTP) designed, fabricated, assembled and tested the Tritium Purification System in Ontario, Canada. After system tests were accepted by Princeton Plasma Physics Laboratory, the assembled components were disassembled into a set of subassemblies and were shipped to PPPL. The subassemblies were reassembled at PPPL and installed primarily in the Decon Facility. The original site selection was within the TFTR tritium processing area and that selection impacted the column design. The Decon Facility was later chosen to permit a better layout of equipment and improved access for installation personnel. Selection of the Decon Facility site resulted in longer line runs for most of the process streams including the tritium product line. The initial review of the proposed installation was conducted during September of 1994 and the System Integrated Test began during April 1995, subsequent to a successful Operational Readiness Assessment conducted during March of 1995.

  13. Tritium permeation through iron

    International Nuclear Information System (INIS)

    Hagi, Hideki; Hayashi, Yasunori

    1989-01-01

    An experimental method for measuring diffusion coefficients and permeation rates of tritium in metals around room temperature has been established, and their values in iron have been obtained by using the method. Permeation rates of tritium and hydrogen through iron were measured by the electrochemical method in which a tritiated aqueous solution was used as a cathodic electrolyte. Tritium and hydrogen were introduced from one side of a membrane specimen by cathodic polarization, while at the other side of the specimen the permeating tritium and hydrogen were extracted by potentiostatical ionization. The amount of permeated hydrogen was obtained by integrating the anodic current, and that of tritium was determined by measuring the radioactivity of the electrolyte sampled from the extraction side. Diffusion coefficients of tritium (D T ) and hydrogen (D H ) were determined from the time lag of tritium and hydrogen permeation. D T =9x10 -10 m 2 /s and D H =4x10 -9 m 2 /s at 286 K for annealed iron specimens. These values of D T and D H were compared with the previous data of the diffusion coefficients of hydrogen and deuterium, and the isotope effect in diffusion was discussed. (orig.)

  14. Tritium technology. A Canadian overview

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    2002-01-01

    An overview of the various tritium research and operational activities in Canada is presented. These activities encompass tritium processing and recovery, tritium interactions with materials, and tritium health and safety. Many of these on-going activities form a sound basis for the tritium use and handling aspects of the ITER project. Tritium management within the CANDU heavy water reactor, associated detritiation facilities, research and development facilities, and commercial industry and improving the understanding of tritium behaviour in humans and the environment remain the focus of a long-standing Canadian interest in tritium. While there have been changes in the application of this knowledge and experience over time, the operating experience and the supporting research and development continue to provide for improved plant and facility operations, an improved understanding of tritium safety issues, and improved products and tools that facilitate tritium management. (author)

  15. Experience in handling concentrated tritium

    International Nuclear Information System (INIS)

    Holtslander, W.J.

    1985-12-01

    The notes describe the experience in handling concentrated tritium in the hydrogen form accumulated in the Chalk River Nuclear Laboratories Tritium Laboratory. The techniques of box operation, pumping systems, hydriding and dehydriding operations, and analysis of tritium are discussed. Information on the Chalk River Tritium Extraction Plant is included as a collection of reprints of papers presented at the Dayton Meeting on Tritium Technology, 1985 April 30 - May 2

  16. Problems of anthropogenic tritium limitation

    Directory of Open Access Journals (Sweden)

    Kochetkov О.A.

    2013-12-01

    Full Text Available This article contains the current situation in respect to the environmental concentrations of anthropogenic and natural tritium. There are presented and analyzed domestic standards for НТО of all Radiation Safety Standards (NRB, as well as the regulations analyzed for tritium in drinking water taken in other countries today. This article deals with the experience of limitation of tritium and focuses on the main problem of rationing of tritium — rationing of organically bound tritium.

  17. Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

    International Nuclear Information System (INIS)

    Hsu, R.H.; Oji, L.N.

    1997-01-01

    Under the Tritium Facility Modernization ampersand Consolidation (TFM ampersand C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM ampersand C Project also provides for a new replacement R ampersand D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H

  18. Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Oji, L.N.

    1997-11-14

    Under the Tritium Facility Modernization {ampersand} Consolidation (TFM{ampersand}C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM{ampersand}C Project also provides for a new replacement R&D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H.

  19. Tritium protective clothing

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, T. P.; Easterly, C. E.

    1979-06-01

    Occupational exposures to radiation from tritium received at present nuclear facilities and potential exposures at future fusion reactor facilities demonstrate the need for improved protective clothing. Important areas relating to increased protection factors of tritium protective ventilation suits are discussed. These areas include permeation processes of tritium through materials, various tests of film permeability, selection and availability of suit materials, suit designs, and administrative procedures. The phenomenological nature of film permeability calls for more standardized and universal test methods, which would increase the amount of directly useful information on impermeable materials. Improvements in suit designs could be expedited and better communicated to the health physics community by centralizing devlopmental equipment, manpower, and expertise in the field of tritium protection to one or two authoritative institutions.

  20. Tritium protective clothing

    International Nuclear Information System (INIS)

    Fuller, T.P.; Easterly, C.E.

    1979-06-01

    Occupational exposures to radiation from tritium received at present nuclear facilities and potential exposures at future fusion reactor facilities demonstrate the need for improved protective clothing. Important areas relating to increased protection factors of tritium protective ventilation suits are discussed. These areas include permeation processes of tritium through materials, various tests of film permeability, selection and availability of suit materials, suit designs, and administrative procedures. The phenomenological nature of film permeability calls for more standardized and universal test methods, which would increase the amount of directly useful information on impermeable materials. Improvements in suit designs could be expedited and better communicated to the health physics community by centralizing devlopmental equipment, manpower, and expertise in the field of tritium protection to one or two authoritative institutions

  1. PRODUCTION OF TRITIUM

    Science.gov (United States)

    Jenks, G.H.; Shapiro, E.M.; Elliott, N.; Cannon, C.V.

    1963-02-26

    This invention relates to a process for the production of tritium by subjecting comminuted solid lithium fluoride containing the lithium isotope of atomic mass number 6 to neutron radiation in a self-sustaining neutronic reactor. The lithium fiuoride is heated to above 450 deg C. in an evacuated vacuum-tight container during radiation. Gaseous radiation products are withdrawn and passed through a palladium barrier to recover tritium. (AEC)

  2. Tritium at Jefferson Lab

    Science.gov (United States)

    Bane, Jason; Jefferson Lab Hall A Collaboration Collaboration

    2017-09-01

    Jefferson Lab's recently upgraded accelerator will provide the perfect opportunity to increase the quality and quantity of the electron scattering world data on tritium. Tritium, the radioactive isotope of hydrogen with a half-life of 12 years, was last used in a large scale electron scattering experiment a few decades ago. This Fall Jefferson Lab will play host to a set of very exciting electron scattering experiments involving tritium. A 25 cm aluminum cell will be filled with 1 kCi of tritium with an internal pressure of approximately 200 psi at 295 kelvin. The tritium target will first see a 10.6 GeV beam to probe the deep inelastic scattering region to study the down to up quark ratio and the EMC effect. Then the beam will be set to 4.3 GeV to investigate SRCs and momentum distributions in the quasi-elastic scattering regime. If time permits, elastic scattering will be used to extract the ratio of the charge radius of tritium and helium3.

  3. Tritium in nuclear power plants

    International Nuclear Information System (INIS)

    Badyaev, V.V.; Egorov, Yu.A.; Sklyarov, V.P.; Stegachev, G.V.

    1981-01-01

    The problem of tritium formation during NPP operation is considered on the basis of available published data. Tritium characteristics are given, sources of the origin of natural and artificial tritium are described. NPP contribution to the total tritium amount in the environment is determined, as well as contribution of each process in the reactor to the quantity of tritium, produced at the NPP. Thermal- and fast-neutron reactions with tritium production are shown, their contribution to the total amount of tritium in a coolant is estimated, taking into account the type of reactor. Data on tritium content in NPP wastes and in the air of working premises are presented. Methods for sampling and sample preparation to measurements as well as the appropriate equipment are considered. Design of the gas-discharge counter of internal filling, used for measuring tritium activity in samples is described [ru

  4. Tritium Systems Test Assembly: design for major device fabrication review

    International Nuclear Information System (INIS)

    Anderson, J.L.; Sherman, R.H.

    1977-06-01

    This document has been prepared for the Major Device Fabrication Review for the Tritium Systems Test Assembly (TSTA). The TSTA is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for fusion reactor systems. The principal objectives for TSTA are: (a) demonstrate the fuel cycle for fusion reactor systems; (b) develop test and qualify equipment for tritium service in the fusion program; (c) develop and test environmental and personnel protective systems; (d) evaluate long-term reliability of components; (e) demonstrate long-term safe handling of tritium with no major releases or incidents; and (f) investigate and evaluate the response of the fuel cycle and environmental packages to normal, off-normal, and emergency situations. This document presents the current status of a conceptual design and cost estimate for TSTA. The total cost to design, construct, and operate TSTA through FY-1981 is estimated to be approximately $12.2 M

  5. Tritium Assay and Dispensing of KEPRI Tritium Lab

    International Nuclear Information System (INIS)

    Sohn, S. H.; Song, K. M.; Lee, S. K.; Lee, K.W.; Ko, B. W.

    2009-01-01

    The Wolsong Tritium Removal Facility(WTRF) has been constructed to reduce tritium levels in the heavy water systems and environmental emissions at the site. The WTRF was designed to process 100 kg/h of heavy water with the overall tritium extraction efficiency of 97% per single pass and to produce ∼700 g of tritium as T2 per year at the feed concentration of 0.37 TBq/kg. The high purity tritium greater than 99% is immobilized as a metal hydride to secure its long term storage. The recovered tritium will be made available for industrial uses and some research applications in the future. Then KEPRI is constructing the tritium lab. to build-up infrastructure to support tritium research activities and to support tritium control and accountability systems for tritium export. This paper describes the initial phases of the tritium application program including the laboratory infrastructure to support the tritium related R and D activities and the tritium controls in Korea

  6. Hazards of exposure to tritium and tritium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, R.C.; Kornberg, H.A.

    1954-01-01

    Experimental data pertinent to the evaluation of hazards involved in the exposure of personnel to tritium and tritium oxide are reviewed. Conclusions are drawn and recommendations made with regard to the control of these hazards.

  7. Tritium emissions reduction facility (TERF)

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Hedley, W.H.

    1993-01-01

    Tritium handling operations at Mound include production of tritium-containing devices, evaluation of the stability of tritium devices, tritium recovery and enrichment, tritium process development, and research. In doing this work, gaseous process effluents containing 400,000 to 1,000,000 curies per year of tritium are generated. These gases must be decontaminated before they can be discharged to the atmosphere. They contain tritium as elemental hydrogen, as tritium oxide, and as tritium-containing organic compounds at low concentrations (typically near one ppm). The rate at which these gases is generated is highly variable. Some tritium-containing gas is generated at all times. The systems used at Mound for capturing tritium from process effluents have always been based on the open-quotes oxidize and dryclose quotes concept. They have had the ability to remove tritium, regardless of the form it was in. The current system, with a capacity of 1.0 cubic meter of gas per minute, can effectively remove tritium down to part-per-billion levels

  8. Tritium in the Channel; Le tritium en Manche

    Energy Technology Data Exchange (ETDEWEB)

    Masson, M.; Fievet, B.; Bailly-Du-Bois, P. [Laboratoire de Radioecologie de Cherbourg-Octeville, IRSN /DEI /SECRE, 50 (France); Olivier, A.; Tenailleau, L. [Groupe d' Etudes Atomiques, EAMEA, 50 - Cherbourg (France)

    2009-07-01

    After having recalled that sea waters entering the Channel exhibit a natural concentration of tritium, the authors outline that spent nuclear fuel reprocessing plants are now the main sources of tritium for marine ecosystems as some oceanographic campaigns showed it. If data about the presence of tritium in water are numerous, data concerning the presence of tritiated water and of organically bound tritium in organisms are much less frequent. However, some surveys have been performed along the Channel French coasts

  9. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1992-01-01

    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 CY-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

  10. Tritium Systems Test Facility

    International Nuclear Information System (INIS)

    Cafasso, F.A.; Maroni, V.A.; Smith, W.H.; Wilkes, W.R.; Wittenberg, L.J.

    1978-01-01

    This TSTF proposal has two principal objectives. The first objective is to provide by mid-FY 1981 a demonstration of the fuel cycle and tritium containment systems which could be used in a Tokamak Experimental Power Reactor for operation in the mid-1980's. The second objective is to provide a capability for further optimization of tritium fuel cycle and environmental control systems beyond that which is required for the EPR. The scale and flow rates in TSTF are close to those which have been projected for a prototype experimental power reactor (PEPR/ITR) and will permit reliable extrapolation to the conditions found in an EPR. The fuel concentrations will be the same as in an EPR. Demonstrations of individual components of the deuterium-tritium fuel cycle and of monitoring, accountability and containment systems and of a maintenance methodology will be achieved at various times in the FY 1979-80 time span. Subsequent to the individual component demonstrations--which will proceed from tests with hydrogen (and/or deuterium) through tracer levels of tritium to full operational concentrations--a complete test and demonstration of the integrated fuel processing and tritium containment facility will be performed. This will occur near the middle of FY 1981. Two options were considered for the TSTF: (1) The modification of an existing building and (2) the construction of a new facility

  11. Packaging design criteria for the Hanford Ecorok Packaging

    International Nuclear Information System (INIS)

    Mercado, M.S.

    1996-01-01

    The Hanford Ecorok Packaging (HEP) will be used to ship contaminated water purification filters from K Basins to the Central Waste Complex. This packaging design criteria documents the design of the HEP, its intended use, and the transportation safety criteria it is required to meet. This information will serve as a basis for the safety analysis report for packaging

  12. Handling of tritium at TFTR

    International Nuclear Information System (INIS)

    Pierce, C.W.; Howe, H.J.; Yemin, L.; Lind, K.

    1977-01-01

    Some of the engineering approaches taken at TFTR for the tritium control systems are discussed as the requirements being placed on the tritium systems by the operating scenarios of the Tokamak. The tritium control systems presently being designed for TFTR will limit the annual release to the environment to less than 100 curies

  13. Development of tritium technology at the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    Anderson, J.L.; Bartlit, J.R.

    1982-01-01

    The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for large scale fusion reactor systems starting with the Fusion Engineering Device (FED) or the International Tokamak Reactor (INTOR). This paper briefly describes the fuel cycle and safety systems at TSTA including the Vacuum Facility, Fuel Cleanup, Isotope Separation, Transfer Pumping, Emergency Tritium Cleanup, Tritium Waste Treatment, Tritium Monitoring, Data Acquisition and Control, Emergency Power and Gas Analysis systems. Discussed in further detail is the experimental program proposed for the startup and testing of these systems

  14. Monitoring of tritium

    Science.gov (United States)

    Corbett, James A.; Meacham, Sterling A.

    1981-01-01

    The fluid from a breeder nuclear reactor, which may be the sodium cooling fluid or the helium reactor-cover-gas, or the helium coolant of a gas-cooled reactor passes over the portion of the enclosure of a gaseous discharge device which is permeable to hydrogen and its isotopes. The tritium diffused into the discharge device is radioactive producing beta rays which ionize the gas (argon) in the discharge device. The tritium is monitored by measuring the ionization current produced when the sodium phase and the gas phase of the hydrogen isotopes within the enclosure are in equilibrium.

  15. ARIES-I tritium system

    International Nuclear Information System (INIS)

    Sze, D.K.; Tam, S.W.; Billone, M.C.; Hassanein, A.M.; Martin, R.

    1990-09-01

    A key safety concern in a D-T fusion reactor is the tritium inventory. There are three components in a fusion reactor with potentially large inventories, i.e., the blanket, the fuel processing system and the plasma facing components. The ARIES team selected the material combinations, decided the operating conditions and refined the processing systems, with the aiming of minimizing the tritium inventories and leakage. The total tritium inventory for the ARIES-I reactor is only 700 g. This paper discussed the calculations and assumptions we made for the low tritium inventory. We also addressed the uncertainties about the tritium inventory. 13 refs., 2 figs., 3 tabs

  16. Ship and Shoot

    Science.gov (United States)

    Woods, Ron

    2012-01-01

    Ron Woods shared incredibly valuable insights gained during his 28 years at the Kennedy Space Center (KSC) packaging Flight Crew Equipment for shuttle and ISS missions. In particular, Woods shared anecdotes and photos from various processing events. The moral of these stories and the main focus of this discussion were the additional processing efforts and effects related to a "ship-and-shoot" philosophy toward flight hardware.

  17. SHIPPING OF RADIOACTIVE ITEMS

    CERN Document Server

    TIS/RP Group

    2001-01-01

    The TIS-RP group informs users that shipping of small radioactive items is normally guaranteed within 24 hours from the time the material is handed in at the TIS-RP service. This time is imposed by the necessary procedures (identification of the radionuclides, determination of dose rate, preparation of the package and related paperwork). Large and massive objects require a longer procedure and will therefore take longer.

  18. Tritium retention in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Dylla, H.F.; Wilson, K.L. (eds.)

    1988-04-01

    This report discusses the materials physics related to D-T operation in TFTR. Research activities are described pertaining to basic studies of hydrogenic retention in graphite, hydrogen recycling phenomena, first-wall and limiter conditioning, surface analysis of TFTR first-wall components, and estimates of the tritium inventory.

  19. Properties of tritium and its compounds

    International Nuclear Information System (INIS)

    Belovodskij, L.F.; Gaevoj, V.K.; Grishmanovskij, V.I.

    1985-01-01

    Ways of tritium preparation and different aspects of its application are considered. Physicochemical properties of this isotope and some compounds of it - tritium oxides, lithium, titanium, zirconium, uranium tritides, tritium organic compounds - are discussed. In particular, diffusion of tritium and its oxide through different materials, tritium oxidation processes, decomposition of tritium-containing compounds under the action of self-radiation are considered. Main radiobiological tritium properties are described

  20. Shipping Fairways

    Data.gov (United States)

    Department of Homeland Security — Various shipping zones delineate activities and regulations for marine vessel traffic. Traffic lanes define specific traffic flow, while traffic separation zones...

  1. Tritium in the aquatic environment

    Energy Technology Data Exchange (ETDEWEB)

    Blaylock, B.G.; Hoffman, F.O.; Frank, M.L.

    1986-02-01

    Tritium is of environmental importance because it is released from nuclear facilities in relatively large quantities and because it has a half life of 12.26 y. Most of the tritium released into the atmosphere eventually reaches the aqueous environment, where it is rapidly taken up by aquatic organisms. This paper reviews the current literature on tritium in the aquatic environment. Conclusions from the review, which covered studies of algae, aquatic macrophytes, invertebrates, fish, and the food chain, were that aquatic organisms incorporate tritium into their tissue-free water very rapidly and reach concentrations near those of the external medium. The rate at which tritium from tritiated water is incorporated into the organic matter of cells is slower than the rate of its incorporation into the tissue-free water. If organisms consume tritiated food, incorporation of tritium into the organic matter is faster, and a higher tritium concentration is reached than when the organisms are exposed to only tritiated water alone. Incorporation of tritium bound to molecules into the organic matter depends on the chemical form of the ''carrier'' molecule. No evidence was found that biomagnification of tritium occurs at higher trophic levels. Radiation doses from tritium releases to large populations of humans will most likely come from the consumption of contaminated water rather than contaminated aquatic food products.

  2. Tritium in the aquatic environment

    International Nuclear Information System (INIS)

    Blaylock, B.G.; Hoffman, F.O.; Frank, M.L.

    1986-02-01

    Tritium is of environmental importance because it is released from nuclear facilities in relatively large quantities and because it has a half life of 12.26 y. Most of the tritium released into the atmosphere eventually reaches the aqueous environment, where it is rapidly taken up by aquatic organisms. This paper reviews the current literature on tritium in the aquatic environment. Conclusions from the review, which covered studies of algae, aquatic macrophytes, invertebrates, fish, and the food chain, were that aquatic organisms incorporate tritium into their tissue-free water very rapidly and reach concentrations near those of the external medium. The rate at which tritium from tritiated water is incorporated into the organic matter of cells is slower than the rate of its incorporation into the tissue-free water. If organisms consume tritiated food, incorporation of tritium into the organic matter is faster, and a higher tritium concentration is reached than when the organisms are exposed to only tritiated water alone. Incorporation of tritium bound to molecules into the organic matter depends on the chemical form of the ''carrier'' molecule. No evidence was found that biomagnification of tritium occurs at higher trophic levels. Radiation doses from tritium releases to large populations of humans will most likely come from the consumption of contaminated water rather than contaminated aquatic food products

  3. Universal tritium transmitter

    International Nuclear Information System (INIS)

    Cordaro, J. V.; Wood, M.

    2008-01-01

    At the Savannah River Site and throughout the National Nuclear Security Agency (NNSA) tritium is measured using Ion or Kanne Chambers. Tritium flowing through an Ion Chamber emits beta particles generating current flow proportional to tritium radioactivity. Currents in the 1 x 10 -15 A to 1 x 10 -6 A are measured. The distance between the Ion Chamber and the electrometer in NNSA facilities can be over 100 feet. Currents greater than a few micro-amperes can be measured with a simple modification. Typical operating voltages of 500 to 1000 Volts and piping designs require that the Ion Chamber be connected to earth ground. This grounding combined with long cable lengths and low currents requires a very specialized preamplifier circuit. In addition, the electrometer must be able to supply 'fail safe' alarm signals which are used to alert personnel of a tritium leak, trigger divert systems preventing tritium releases to the environment and monitor stack emissions as required by the United States federal Government and state governments. Ideally the electrometer would be 'self monitoring'. Self monitoring would reduce the need for constant checks by maintenance personnel. For example at some DOE facilities monthly calibration and alarm checks must be performed to ensure operation. NNSA presently uses commercially available electrometers designed specifically for this critical application. The problems with these commercial units include: ground loops, high background currents, inflexibility and susceptibility to Electromagnetic Interference (EMI) which includes RF and Magnetic fields. Existing commercial electrometers lack the flexibility to accommodate different Ion Chamber designs required by the gas pressure, type of gas and range. Ideally the electrometer could be programmed for any expected gas, range and high voltage output. Commercially available units do not have 'fail safe' self monitoring capability. Electronics used to measure extremely low current must have

  4. 21 CFR 820.130 - Device packaging.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Device packaging. 820.130 Section 820.130 Food and... QUALITY SYSTEM REGULATION Labeling and Packaging Control § 820.130 Device packaging. Each manufacturer shall ensure that device packaging and shipping containers are designed and constructed to protect the...

  5. Environmental monitoring for tritium at tritium separation facility

    International Nuclear Information System (INIS)

    Varlam, C.; Stefanescu, I.; Steflea, D.; Lazar, R.E.

    2001-01-01

    The Cryogenic Pilot is an experimental project in the nuclear energy national research program, which has the aim of developing technologies for tritium and deuterium separation by cryogenic distillation. The experimental installation is located 15 km near the highest city of the area and 1 km near Olt River. An important chemical activity is developed in the area and the Experimental Cryogenic Pilot's, almost the entire neighborhood are chemical plants. It is necessary to emphasize this aspect because the sewerage system is connected with the other three chemical plants from the neighborhood. This is the reason that we progressively established elements of an environmental monitoring program well in advance of tritium operation in order to determine baseline levels. The first step was the tritium level monitoring in environmental water and waste water of industrial activity from neighborhood. In this work, a low background liquid scintillation is used to determine tritium activity concentration according to ISO 9698/1998. We measured drinking water, precipitation, river water, underground water and waste water. The tritium level was between 10 TU and 27 TU that indicates there is no source of tritium contamination in the neighborhood of Cryogenic Pilot. In order to determine baseline levels we decide to monitories monthly each location. In this paper a standard method is presented which it is used for tritium determination in water sample, the precautions needed in order to achieve reliable results, and the evolution of tritium level in different location near the Experimental Pilot Tritium and Deuterium Cryogenic Separation.(author)

  6. Evaluation of tritium release properties of advanced tritium breeders

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, T. [Breeding Functional Materials Development Group, Department of Blanket Systems Research, JAEA, Rokkasho Fusion Institute, Omotedate, Rokkasho-mura (Japan); Ochiai, K. [Fusion Neutronics Group, Department of Blanket Systems Research, JAEA, Tokai-mura (Japan); Edao, Y.; Kawamura, Y. [Tritium Technology Group, Department of Blanket Systems Research, JAEA, Tokai-mura (Japan)

    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. Considering 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)

  7. Tritium concentration monitor

    International Nuclear Information System (INIS)

    Shono, Kosuke.

    1991-01-01

    A device for measuring the concentration of tritium in gaseous wastes in a power plant and a nuclear fuel reprocessing plant is reduced in the size and improved in performance. The device of the present invention pressurizes a sampling gas and cools it to a dew point. Water content in the sampling gas cooled to the dew point is condensated and recovered to a fine tube-like water content recovering container. The concentration of the recovered condensates is measured by a tritium density analyzer. With such procedures, since the specimen is pressurized, the dew point can be elevated. Accordingly, the size of the cooling device can be decreased, enabling to contribute to the reduction of the size of the entire device. Further, since the water content recovering device is formed as a fine tube, the area of contact between the specimen gas and the liquid condensated water can be reduced. Accordingly, evaporation of the liquid condensates can be prevented. (I.S.)

  8. Tritium monitoring : present status

    International Nuclear Information System (INIS)

    Rathnakaran, M.; Singh, A.N.

    1993-01-01

    The report summarizes the present status of techniques employed for the monitoring of tritium in water, air and other samples. A brief mention of the work done by numerous workers in the field, critical comments about the work and a fairly exhaustive list of references about the work done during the last 4 decades has been presented. On-line monitoring on real time basis in nuclear reactors is also discussed. (author). 83 refs., 10 refs., 2 tabs

  9. Effects of tritium in elastomers

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1982-01-01

    Elastomers are used as flange gaskets in the piping system of the Savannah River Plant tritium facilities. A number of elastomers is being examined to identify those compounds more radiation-resistant than the currently specified Buna-N rubber and to study the mechanism of tritium radiation damage. Radiation resistance is evaluated by compression set tests on specimens exposed to about 1 atm tritium for several months. Initial results show that ethylene-propylene rubber and three fluoroelastomers are superior to Buna-N. Off-gassing measurements and autoradiography show that retained surface absorption of tritium varies by more than an order of magnitude among the different elastomer compounds. Therefore, tritium solubility and/or exchange may have a role in addition to that of chemical structure in the damage process. Ongoing studies of the mechanism of radiation damage include: (1) tritium absorption kinetics, (2) mass spectroscopy of radiolytic products, and (3) infrared spectroscopy

  10. Effects of tritium in elastomers

    Energy Technology Data Exchange (ETDEWEB)

    Zapp, P.E.

    1982-01-01

    Elastomers are used as flange gaskets in the piping system of the Savannah River Plant tritium facilities. A number of elastomers is being examined to identify those compounds more radiation-resistant than the currently specified Buna-N rubber and to study the mechanism of tritium radiation damage. Radiation resistance is evaluated by compression set tests on specimens exposed to about 1 atm tritium for several months. Initial results show that ethylene-propylene rubber and three fluoroelastomers are superior to Buna-N. Off-gassing measurements and autoradiography show that retained surface absorption of tritium varies by more than an order of magnitude among the different elastomer compounds. Therefore, tritium solubility and/or exchange may have a role in addition to that of chemical structure in the damage process. Ongoing studies of the mechanism of radiation damage include: (1) tritium absorption kinetics, (2) mass spectroscopy of radiolytic products, and (3) infrared spectroscopy.

  11. Metabolism of organically bound tritium

    International Nuclear Information System (INIS)

    Travis, C.C.

    1984-01-01

    The classic methodology for estimating dose to man from environmental tritium ignores the fact that organically bound tritium in foodstuffs may be directly assimilated in the bound compartment of tissues without previous oxidation. We propose a four-compartment model consisting of a free body water compartment, two organic compartments, and a small, rapidly metabolizing compartment. The utility of this model lies in the ability to input organically bound tritium in foodstuffs directly into the organic compartments of the model. We found that organically bound tritium in foodstuffs can increase cumulative total body dose by a factor of 1.7 to 4.5 times the free body water dose alone, depending on the bound-to-loose ratio of tritium in the diet. Model predictions are compared with empirical measurements of tritium in human urine and tissue samples, and appear to be in close agreement. 10 references, 4 figures, 3 tables

  12. A prototype wearable tritium monitor

    International Nuclear Information System (INIS)

    Surette, R. A.; Dubeau, J.

    2008-01-01

    Sudden unexpected changes in tritium-in-air concentrations in workplace air can result in significant unplanned exposures. Although fixed area monitors are used to monitor areas where there is a potential for elevated tritium in air concentrations, they do not monitor personnel air space and may require some time for acute tritium releases to be detected. There is a need for a small instrument that will quickly alert staff of changing tritium hazards. A moderately sensitive tritium instrument that workers could wear would bring attention to any rise in tritium levels that were above predetermined limits and help in assessing the potential hazard therefore minimizing absorbed dose. Hand-held instruments currently available can be used but require the assistance of a fellow worker or restrict the user to using only one hand to perform some duties. (authors)

  13. Toxicity of tritium

    International Nuclear Information System (INIS)

    Dobson, R.L.

    1979-01-01

    Among radionuclides of importance in atomic energy, 3 H has relatively low toxicity. The main health and environmental worry is the possibility that significant biological effects may follow from protracted exposure to low concentrations in water. To examine this possible hazard and measure toxicity at low tritium concentrations, chronic exposure studies were done on mice and monkeys. During vulnerable developmental periods animals were exposed to 3 HOH, and mice were exposed also to 60 Co gamma irradiation and energy-related chemical agents. The biological endpoint measured was the irreversible loss of female germ cells. Effects from tritium were observed at surprisingly low concentrations where 3 H was found more damaging than previously thought. Comparisons between tritium and gamma radiation showed the relative biological effectiveness (RBE) to be greater than 1 and to reach approximately 3 at very low exposures. For perspective, other comparisons were made: between radiation and chemical agents, which revealed parallels in action on germ cells, and between pre- and postnatal exposure, which warn of possible special hazard to the fetus from both classes of energy-related byproducts

  14. New Packaging for Amplifier Slabs

    Energy Technology Data Exchange (ETDEWEB)

    Riley, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Thorsness, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Suratwala, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Steele, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rogowski, G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-03-18

    The following memo provides a discussion and detailed procedure for a new finished amplifier slab shipping and storage container. The new package is designed to maintain an environment of <5% RH to minimize weathering.

  15. Tritium inventory tracking and management

    International Nuclear Information System (INIS)

    Eichenberg, T.W.; Klein, A.C.

    1990-01-01

    This investigation has identified a number of useful applications of the analysis of the tracking and management of the tritium inventory in the various subsystems and components in a DT fusion reactor system. Due to the large amounts of tritium that will need to be circulated within such a plant, and the hazards of dealing with the tritium an electricity generating utility may not wish to also be in the tritium production and supply business on a full time basis. Possible scenarios for system operation have been presented, including options with zero net increase in tritium inventory, annual maintenance and blanket replacement, rapid increases in tritium creation for the production of additional tritium supplies for new plant startup, and failures in certain system components. It has been found that the value of the tritium breeding ratio required to stabilize the storage inventory depends strongly on the value and nature of other system characteristics. The real operation of a DT fusion reactor power plant will include maintenance and blanket replacement shutdowns which will affect the operation of the tritium handling system. It was also found that only modest increases in the tritium breeding ratio are needed in order to produce sufficient extra tritium for the startup of new reactors in less than two years. Thus, the continuous operation of a reactor system with a high tritium breeding ratio in order to have sufficient supplies for other plants is not necessary. Lastly, the overall operation and reliability of the power plant is greatly affected by failures in the fuel cleanup and plasma exhaust systems

  16. Tritium removal and retention device

    International Nuclear Information System (INIS)

    Boyle, R.F.; Durigon, D.D.

    1980-01-01

    A device is provided for removing and retaining tritium from a gaseous medium, and also a method of manufacturing the device. The device, consists of an inner core of zirconium alloy, preferably Zircaloy-4, and an outer adherent layer of nickel which acts as a selective and protective window for passage of tritium. The tritium then reacts with or is absorbed by the zirconium alloy, and is retained until such time as it is desirable to remove it during reprocessing. (auth)

  17. An overview of tritium production

    International Nuclear Information System (INIS)

    He Kaihui; Huang Jinghua; Feng Kaiming

    2002-01-01

    The characteristics of three types of proposed tritium production facilities, fissile type, accelerator production tritium (APT), and fusion type, are presented. The fissile reactors, especially commercial light water reactor, use comparatively mature technology and are designed to meet current safety and environmental guidelines. Conversely, APT shows many advantages except its rather high cost, while fusion reactors appear to offer improved safety and environmental impact, in particular, tritium production based on the fusion-based neutron source. However, its cost keeps unknown

  18. Tritium monitor and collection system

    Science.gov (United States)

    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.

  19. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1983-08-01

    This report undertakes a preliminary assessment of the prospects of tritium-assisted D-D fuel cycle fusion breeders. Two well documented fusion power reactor designs - the STARFIRE (D-T fuel cycle) and the WILDCAT (Cat-D fuel cycle) tokamaks - are converted into fusion breeders by replacing the fusion electric blankets with 233 U producing fission suppressed blankets; changing the Cat-D fuel cycle mode of operation by one of the several tritium-assisted D-D-based modes of operation considered; adjusting the reactor power level; and modifying the resulting plant cost to account for the design changes. Three sources of tritium are considered for assisting the D-D fuel cycle: tritium produced in the blankets from lithium or from 3 He and tritium produced in the client fission reactors. The D-D-based fusion breeders using tritium assistance are found to be the most promising economically, especially the Tritium Catalyzed Deuterium mode of operation in which the 3 He exhausted from the plasma is converted, by neutron capture in the blanket, into tritium which is in turn fed back to the plasma. The number of fission reactors of equal thermal power supported by Tritium Catalyzed Deuterium fusion breeders is about 50% higher than that of D-T fusion breeders, and the profitability is found to be slightly lower than that of the D-T fusion breeders

  20. Tritium removal and retention device

    International Nuclear Information System (INIS)

    1976-01-01

    The patent discloses an apparatus comprising a two-layered composite with an internal core of zirconium or zirconium alloy which retains tritium, and an adherent nickel outer layer which acts as a protective and selective window for passage of the tritium. The invention provides a device to remove and store tritium from a gaseous medium as well as a method for manufacturing the device. It specifically provides a device which may be incorporated in the fuel rod of a nuclear reactor to minimize release of tritium to the reactor coolant

  1. Overview of tritium fast-fission yields

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1981-03-01

    Tritium production rates are very important to the development of fast reactors because tritium may be produced at a greater rate in fast reactors than in light water reactors. This report focuses on tritium production and does not evaluate the transport and eventual release of the tritium in a fast reactor system. However, if an order-of-magnitude increase in fast fission yields for tritium is confirmed, fission will become the dominant production source of tritium in fast reactors

  2. Alternative containers for low-level wastes containing large amounts of tritium

    International Nuclear Information System (INIS)

    Gause, E.P.; Lee, B.S.; MacKenzie, D.R.; Wiswall, R. Jr.

    1984-11-01

    High-activity tritiated waste generated in the United States is mainly composed of tritium gas and tritium-contaminated organic solvents sorbed onto Speedi-Dri which are packaged in small glass bulbs. Low-activity waste consists of solidified and adsorbed liquids. In this report, current packages for high-activity gaseous and low-activity adsorbed liquid wastes are emphasized with regard to containment potential. Containers for low-level radioactive waste containing large amounts of tritium need to be developed. An integrity may be threatened by: physical degradation due to soil corrosion, gas pressure build-up (due to radiolysis and/or biodegradation), rapid permeation of tritium through the container, and corrosion from container contents. Literature available on these points is summarized in this report. 136 references, 20 figures, 40 tables

  3. JET experiments with tritium and deuterium–tritium mixtures

    NARCIS (Netherlands)

    Horton, L.; Batistoni, P.; Boyer, H.; Challis, C.; Ciric, D.; Donne, A. J. H.; Eriksson, L. G.; Garcia, J.; Garzotti, L.; Gee, S.; Hobirk, J.; Joffrin, E.; Jones, T.; King, D. B.; Knipe, S.; Litaudon, X.; Matthews, G. F.; Monakhov, I.; Murari, A.; Nunes, I.; Riccardo, V.; Sips, A. C. C.; Warren, R.; Weisen, H.; Zastrow, K. D.

    2016-01-01

    Extensive preparations are now underway for an experiment in the Joint European Torus (JET) using tritium and deuterium–tritium mixtures. The goals of this experiment are described as well as the progress that has been made in developing plasma operational scenarios and physics reference pulses for

  4. IN-SITU TRITIUM BETA DETECTOR

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Berthold; L.A. Jeffers

    1998-04-15

    The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye. Based on the results achieved, it is premature to initiate Phase 2 and commit to a prototype

  5. IN-SITU TRITIUM BETA DETECTOR

    International Nuclear Information System (INIS)

    Berthold, J.W.; Jeffers, L.A.

    1998-01-01

    The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye. Based on the results achieved, it is premature to initiate Phase 2 and commit to a prototype

  6. Tritium contaminated waste management at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Jalbert, R.A.; Carlson, R.V.

    1987-01-01

    The Tritium Systems Test Assembly (TSTA) at Los Alamos continues to move toward full operation of an integrated, full-sized, computer-controlled fusion fuel processing loop. Concurrent nonloop experiments further the development of advanced tritium technologies and handling methods. Since tritium operations began in June 1984, tritium contaminated wastes have been produced at TSTA that are roughly typical in kind and amount of those to be produced by tritium fueling operations at fusion reactors. Methods of managing these wastes are described, including information on some methods of decontamination so that equipment can be reused. Data are given on the kinds and amounts of wastes and the general level of contamination. Also included are data on environmental emissions and doses to personnel that have resulted from TSTA operations. Particular problems in waste managements are discussed

  7. Test Ship

    Data.gov (United States)

    Federal Laboratory Consortium — The U. S. Navy dedicated the decommissioned Spruance Class destroyer ex-PAUL F. FOSTER (EDD 964), Test Ship, primarily for at sea demonstration of short range weapon...

  8. Tritium transport and control in the FED

    International Nuclear Information System (INIS)

    Rogers, M.L.

    1981-01-01

    The tritium systems for the FED have three primary purposes. The first is to provide tritium and deuterium fuel for the reactor. This fuel can be new tritium or deuterium delivered to the plant site, or recycled DT from the reactor that must be processed before it can be recycled. The second purpose of the FED tritium systems is to provide state-of-the-art tritium handling to limit worker radiation exposure and to minimize tritium losses to the environment. The final major objective of the FED tritium systems is to provide an integrated system test of the tritium handling technology necessary to support the fusion reactor program. Every effort is being made to incorporate available information from the Tritium System Test Assembly (TSTA) at Los Alamos National Laboratory, the Tokamak Fusion Test Reactor (TFTR) tritium systems, and the tritium handling information generated within DOE for the past 20 years

  9. Ontario Hydro diversifies into tritium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    A report is given on a plant which is to be built at the Darlington Candu reactor site in Canada for the extraction of tritium from heavy water. As tritium is used as a fuel in fusion research the market for it is expected to grow. The design of the system is outlined with the help of a flow diagram. (U.K.)

  10. The tritium operations experience on TFTR

    Energy Technology Data Exchange (ETDEWEB)

    von Halle, A.; Gentile, C. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Anderson, J.L. [Los Alamos National Lab., NM (United States)] [and others

    1994-09-01

    The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grains of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes- the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the US Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described.

  11. Tritium practices past and present

    International Nuclear Information System (INIS)

    Gede, V.P.; Gildea, P.D.

    1980-01-01

    History of the production and use of tritium, as well as handling techniques, are reviewed. Handling techniques first used at Lawrence Livermore National Laboratory made use of glass vacuum systems and relatively crude ion chambers for monitoring airborne activity. The first use of inert atmosphere glove boxes demonstrated that uptake through the skin could be a serious personnel exposure problem. Growing environmental concerns in the early 1970's resulted in the implementation by the Atomic Energy Commission of a new criteria to limit atmospheric tritium releases to levels as low as practicable. An important result of the new criteria was the development of containment and recovery systems to capture tritium rather than vent it to the atmosphere. The Sandia National Laboratories, Livermore, Tritium Research Laboratory containment and decontamination systems are presented as a typical example of this technology. The application of computers to control systems is expected to provide the greatest potential for change in future tritium handling practices

  12. Tritium behaviour in higher plants

    International Nuclear Information System (INIS)

    Guenot, J.

    1984-05-01

    Vine grapes and potato seedlings have been exposed in situ to tritiated water vapor and 14 C labeled carbon dioxide. Leaves sampling was done during and after the exposition. Measurements allowed to distinguish the three forms of tritium in leaves, i.e. tissue free water tritium (TFWT) and organically bound tritium (OBT), in exchangeable position or not. The results lead to a description of the dynamical behaviour of tritium between these three compartments. It has been shown that 20% of organically bound hydrogen is readily exchangeable thus being in permanent isotopic equilibium with tissue free water. Moreover, the activity of nonexchangeable OBT appears to be strongly related to the organic 14 C, which shows that photosynthesis is responsible of tritium incorporation in organic nonexchangeable position, and occurs with a 20% discrimination in favor of protium. In contrast with the other two compartments, this fixation is almost irreversible, which is a fact of importance from a radiological point of view [fr

  13. Tritium metabolism in rat tissues

    International Nuclear Information System (INIS)

    Takeda, H.

    1982-01-01

    As part of a series of studies designed to evaluate the relative radiotoxicity of various tritiated compounds, metabolism of tritium in rat tissues was studied after administration of tritiated water, leucine, thymidine, and glucose. The distribution and retention of tritium varied widely, depending on the chemical compound administered. Tritium introduced as tritiated water behaved essentially as body water and became uniformly distributed among the tissues. However, tritium administered as organic compounds resulted in relatively high incorporation into tissue constituents other than water, and its distribution differed among the various tissues. Moreover, the excretion rate of tritium from tissues was slower for tritiated organic compounds than for tritiated water. Administrationof tritiated organic compounds results in higher radiation doses to the tissues than does administration of tritiated water. Among the tritiated compounds examined, for equal radioactivity administered, leucine gave the highest radiation dose, followed in turn by thymidine, glucose, and water. (author)

  14. Tritium-surface interactions

    International Nuclear Information System (INIS)

    Kirkaldy, J.S.

    1983-06-01

    The report deals broadly with tritium-surface interactions as they relate to a fusion power reactor enterprise, viz., the vacuum chamber, first wall, peripherals, pumping, fuel recycling, isotope separation, repair and maintenance, decontamination and safety. The main emphasis is on plasma-surface interactions and the selection of materials for fusion chamber duty. A comprehensive review of the international (particularly U.S.) research and development is presented based upon a literature review (about 1 000 reports and papers) and upon visits to key laboratories, Sandia, Albuquerque, Sandia, Livermore and EGβG Idaho. An inventory of Canadian expertise and facilities for RβD on tritium-surface interactions is also presented. A number of proposals are made for the direction of an optimal Canadian RβD program, emphasizing the importance of building on strength in both the technological and fundamental areas. A compendium of specific projects and project areas is presented dealing primarily with plasma-wall interactions and permeation, anti-permeation materials and surfaces and health, safety and environmental considerations. Potential areas of industrial spinoff are identified

  15. Tritium-fueled betacells

    International Nuclear Information System (INIS)

    Walko, R.J.; Lincoln, R.C.; Baca, W.E.; Goods, S.H.; Negley, G.H.

    1991-01-01

    Betavoltaic power sources operate by converting the nuclear decay energy of beta-emitting radioisotopes into electricity. Since they are not chemically driven, they could operate at temperatures which would either be too hot or too cold for typical chemical batteries. Further, for long lived isotopes, they offer the possibility of multi-decade active lifetimes. In this paper two approaches are investigated: direct and indirect conversion. Direct conversion cells consist of semiconductor diodes similar to photovoltaic cells. Beta particles directly bombard these cells, generating electron-hole pairs in the semiconductor which are converted to useful power. When using low power flux beta emitters, wide bandgap semiconductors are required to achieve useful power. When using low power flux beta emitters, wide bandgap semiconductors are required to achieve useful conversion efficiencies. The combination of tritium, as the beta emitter, and gallium phosphide (GaP), as the semiconductor converter, was evaluated. Indirect conversion betacells first convert the beta energy to light with a phosphor, and then to electricity with photovoltaic cells. An indirect conversion power source using a tritium radioluminescent (RL) light is being investigated. The authors analysis indicates that this approach has the potential for significant volume and cost savings over the direct conversion method

  16. 10 CFR 71.127 - Handling, storage, and shipping control.

    Science.gov (United States)

    2010-01-01

    ... handling, storage, shipping, cleaning, and preservation of materials and equipment to be used in packaging... MATERIAL Quality Assurance § 71.127 Handling, storage, and shipping control. The licensee, certificate... 10 Energy 2 2010-01-01 2010-01-01 false Handling, storage, and shipping control. 71.127 Section 71...

  17. Overview of the tritium system of Ignitor

    International Nuclear Information System (INIS)

    Rizzello, C.; Tosti, S.

    2008-01-01

    Among the recent design activities of the Ignitor program, the analysis of the tritium system has been carried out with the aim to describe the main equipments and the operations needed for supplying the deuterium-tritium mixtures and recovering the plasma exhaust. In fact, the tritium system of Ignitor provides for injecting deuterium-tritium mixtures into the vacuum chamber in order to sustain the fusion reaction: furthermore, it generally manages and controls the tritium and the tritiated materials of the machine fuel cycle. Main functions consist of tritium storage and delivery, tritium injection, tritium recovery from plasma exhaust, treatment of the tritiated wastes, detritiation of the contaminated atmospheres, tritium analysis and accountability. In this work an analysis of the designed tritium system of Ignitor is summarized

  18. RH Packaging Operations Manual

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2003-01-01

    This procedure provides operating instructions for the RH-TRU 72-B Road Cask, Waste Shipping Package. In this document, ''Packaging'' refers to the assembly of components necessary to ensure compliance with the packaging requirements (not loaded with a payload). ''Package'' refers to a Type B packaging that, with its radioactive contents, is designed to retain the integrity of its containment and shielding when subject to the normal conditions of transport and hypothetical accident test conditions set forth in 10 CFR Part 71. Loading of the RH 72-B cask can be done two ways, on the RH cask trailer in the vertical position or by removing the cask from the trailer and loading it in a facility designed for remote-handling (RH). Before loading the 72-B cask, loading procedures and changes to the loading procedures for the 72-B cask must be sent to CBFO at sitedocuments at wipp.ws for approval

  19. Tritium, biography of an element

    International Nuclear Information System (INIS)

    Keller, C.

    1980-01-01

    Tritium is the lightest radioactive atom, an isotope of hydrogen. In science it has many uses, particularly for marking organic molecules in order to find out about biochemical and medical processes. But also the traces of tritium contained in rain or sea water are used for investigations; they range from establishing the vintage of old wines to ascertaining sea water mixtures. Tritium will become important in large-scale technology if it should become possible to construct fusion reactors, since it is one of the fuels. (orig.) [de

  20. The organically bound tritium: an analyst vision

    International Nuclear Information System (INIS)

    Ansoborlo, E.; Baglan, N.

    2009-01-01

    The authors report the work of a work group on tritium analysis. They recall the different physical forms of tritium: gas (HT, hydrogen-tritium), water vapour (HTO or tritiated water) or methane (CH3T), but also in organic compounds (OBT, organically bound tritium) which are either exchangeable or non-exchangeable. They evoke measurement techniques and methods, notably to determine the tritium volume activity. They discuss the possibilities to analyse and distinguish exchangeable and non-exchangeable OBTs

  1. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

    It is well known that tritium is used as the fuel for nuclear fusion reactors. The neutrons produced by the nuclear fusion reaction of deuterium and tritium react with lithium in blankets, and tritium is produced. The blankets reproduce the tritium consumed in the D-T reaction. Tritium circulates through the main cooling system and the fuel supply and evacuation system, and is accumulated. Tritium is a radioactive substance emitting β-ray with 12.6 year half-life, and harmful to human bodies. It is an isotope of hydrogen, and apt to diffuse and leak. Especially at high temperature, it permeates through materials, therefore it is important to evaluate the release of tritium into environment, to treat leaked tritium to reduce its release, and to select the method of containing tritium. The permeability of tritium and its solubility in structural materials are discussed. The typical blanket-cooling systems of nuclear fusion reactors are shown, and the tungsten coating of steam generator tubes and tritium recovery system are adopted for reducing tritium leak. In case of the Tokamak type reactor of JAERI, the tritium recovery system is installed, in which the tritium gas produced in blankets is converted to tritium steam with a Pd-Pt catalytic oxidation tower, and it is dehydrated and eliminated with a molecular sieve tower, then purified and recovered. (Kako, I.)

  2. Normalized Tritium Quantification Approach (NoTQA) a Method for Quantifying Tritium Contaminated Trash and Debris at LLNL

    International Nuclear Information System (INIS)

    Dominick, J.L.; Rasmussen, C.L.

    2008-01-01

    Several facilities and many projects at LLNL work exclusively with tritium. These operations have the potential to generate large quantities of Low-Level Radioactive Waste (LLW) with the same or similar radiological characteristics. A standardized documented approach to characterizing these waste materials for disposal as radioactive waste will enhance the ability of the Laboratory to manage them in an efficient and timely manner while ensuring compliance with all applicable regulatory requirements. This standardized characterization approach couples documented process knowledge with analytical verification and is very conservative, overestimating the radioactivity concentration of the waste. The characterization approach documented here is the Normalized Tritium Quantification Approach (NoTQA). This document will serve as a Technical Basis Document which can be referenced in radioactive waste characterization documentation packages such as the Information Gathering Document. In general, radiological characterization of waste consists of both developing an isotopic breakdown (distribution) of radionuclides contaminating the waste and using an appropriate method to quantify the radionuclides in the waste. Characterization approaches require varying degrees of rigor depending upon the radionuclides contaminating the waste and the concentration of the radionuclide contaminants as related to regulatory thresholds. Generally, as activity levels in the waste approach a regulatory or disposal facility threshold the degree of required precision and accuracy, and therefore the level of rigor, increases. In the case of tritium, thresholds of concern for control, contamination, transportation, and waste acceptance are relatively high. Due to the benign nature of tritium and the resulting higher regulatory thresholds, this less rigorous yet conservative characterization approach is appropriate. The scope of this document is to define an appropriate and acceptable

  3. Tritium release reduction and radiolysis gas formation

    International Nuclear Information System (INIS)

    Batifol, G.; Douche, Ch.; Sejournant, Ch.

    2008-01-01

    At CEA Valduc, the usual tritiated waste container is the steel drum. It allows good release reduction performance for middle activity waste but in some cases tritium outgassing from the waste drums is too high. It was decided to over-package each drum in a tighter container called the over-drum. According to good safety practices it was also decided to measure gas composition evolution into the over-drum in order to defect hydrogen formation over time. After a few months, a significant release reduction was observed. Additionally there followed contamination reduction in the roof storage building rainwater. However hydrogen was also observed in some over-drums, in addition to other radiolysis products. Catalyst will be added to manage the hydrogen risk in the over-drums. (authors)

  4. 9 CFR 381.144 - Packaging materials.

    Science.gov (United States)

    2010-01-01

    ... to health. All packaging materials must be safe for the intended use within the meaning of section..., from the packaging supplier under whose brand name and firm name the material is marketed to the... distinguishing brand name or code designation appearing on the packaging material shipping container; must...

  5. Tritium management for fusion reactors

    International Nuclear Information System (INIS)

    Rouyer, J.L.; Djerassi, H.

    1985-01-01

    To determine a waste management strategy, one has to identify first the wastes (quantities, activities, etc.), then to define options, and to compare these options by appropriate criteria and evaluations. Two European Associations are working together, i.e., Studsvik and CEA, on waste treatment and tritium problems. A contribution to fusion specific tritiated waste management strategy is presented. It is demonstrated that the best strategy is to retain tritium (outgas and recover, or immobilize it) so that residual tritium releases are kept to a minimum. For that, wastes are identified, actual regulations are described and judged inadequate without amendments for fusion problems. Appropriate criteria are defined. Options for treatment and disposal of tritiated wastes are proposed and evaluated. A tritium recovery solution is described

  6. Radiotoxicity of tritium in mammals

    International Nuclear Information System (INIS)

    Silini, G.; Metalli, P.; Vulpis, G.

    1972-12-01

    Basic data relative to tritium, its physicochemical behaviour in environment, its major sources of contamination and its metabolism through the mammalian organisms are reviewed. After considering the radiotoxicity of tritium particularly at the cellular and whole-body level the conclusion is drawn that the major uncertainties regard the fraction of tritium incorporated into the nuclei of some tissues. This fraction is eliminated very slowly and is capable of modifying the genetic structures of the nucleus. A more refined analysis of radiobiological phenomena and a better knowledge of the dose effect relationship should permit the extrapolation of the data to the low doses of tritium contamination. This extrapolation is of great interest in the field of public health for the elaboration of the relevant radioprotection standards

  7. Tritium transport around nuclear facilities

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.; Sweet, C.W.

    1981-01-01

    The transport and cycling of tritium around nuclear facilities is reviewed with special emphasis on studies at the Savannah River Laboratory, Aiken, South Carolina. These studies have shown that the rate of deposition from the atmosphere, the site of deposition, and the subsequent cycling are strongly influenced by the compound with which the tritium is associated. Tritiated hydrogen is largely deposited in the soil, while tritiated water is deposited in the greatest quantity in the vegetation. Tritiated hydrogen is converted in the soil to tritiated water that leaves the soil slowly, through drainage and transpiration. Tritiated water deposited directly to the vegetation leaves the vegetation more rapidly after exposure. Only a small part of the tritium entering the vegetation becomes bound in organic molecules. However, it appears tht the existence of soil organic compounds with tritium concentrations greater than the equilibrium concentration in the associated water can be explained by direct metabolism of tritiated hydrogen in vegetation

  8. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.; Talbot, J.B.

    1980-01-01

    When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

  9. Design and test about de tritium system to filling tritium glove box

    International Nuclear Information System (INIS)

    Lei, Jiarong; Du, Yang; Yang, Yong

    2008-01-01

    In order to deal tritium permeated from inflating tritium system at the scene of inflating tritium, dealing waste tritium gas system was designed according to demand and action of dealing waste tritium gas from inflating tritium, and the data of character and volume about appliance of catalyst reaction and drying agent was calculated. Through the test at the scene of inflating tritium, it is result that dealing waste tritium gas system's efficiency reaches above 85% average in circulatory system, so that it can be used in practice extensively. (author)

  10. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

    Wall, W.R.

    1981-03-01

    This instruction describes the tritium accountability procedures practiced by the Tritium Research Laboratory, Building 968 at Sandia National Laboratories, Livermore. The accountability procedures are based upon the Sandia National Laboratories, Livermore, Nuclear Materials Operations Manual, SAND78-8018. The Nuclear Materials Operations Manual describes accountability techniques which are in compliance with the Department of Energy Manual, Code of Federal Regulations, and Sandia National Laboratories Instructions

  11. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

    Wall, W.R.; Cruz, S.L.

    1985-08-01

    This instruction describes the tritium accountability procedures practiced by the Tritium Research Laboratory, at Sandia National Laboratories, Livermore. The accountability procedures are based upon the Sandia National Laboratories, Livermore, Nuclear Materials Operations Manual, SAND83-8036. The Nuclear Materials Operations Manual describes accountability techniques which are in compliance with the Department of Energy 5630 series Orders, Code of Federal Regulations, and Sandia National Laboratories Instructions

  12. Separation of Tritium from Wastewater

    International Nuclear Information System (INIS)

    JEPPSON, D.W.

    2000-01-01

    A proprietary tritium loading bed developed by Molecular Separations, Inc (MSI) has been shown to selectively load tritiated water as waters of hydration at near ambient temperatures. Tests conducted with a 126 (micro)C 1 tritium/liter water standard mixture showed reductions to 25 (micro)C 1 /L utilizing two, 2-meter long columns in series. Demonstration tests with Hanford Site wastewater samples indicate an approximate tritium concentration reduction from 0.3 (micro)C 1 /L to 0.07 (micro)C 1 /L for a series of two, 2-meter long stationary column beds Further reduction to less than 0.02 (micro)C 1 /L, the current drinking water maximum contaminant level (MCL), is projected with additional bed media in series. Tritium can be removed from the loaded beds with a modest temperature increase and the beds can be reused Results of initial tests are presented and a moving bed process for treating large quantities of wastewaters is proposed. The moving bed separation process appears promising to treat existing large quantities of wastewater at various US Department of Energy (DOE) sites. The enriched tritium stream can be grouted for waste disposition. The separations system has also been shown to reduce tritium concentrations in nuclear reactor cooling water to levels that allow reuse. Energy requirements to reconstitute the loading beds and waste disposal costs for this process appear modest

  13. Primer on tritium safe handling practices

    Energy Technology Data Exchange (ETDEWEB)

    1994-12-01

    This Primer is designed for use by operations and maintenance personnel to improve their knowledge of tritium safe handling practices. It is applicable to many job classifications and can be used as a reference for classroom work or for self-study. It is presented in general terms for use throughout the DOE Complex. After reading it, one should be able to: describe methods of measuring airborne tritium concentration; list types of protective clothing effective against tritium uptake from surface and airborne contamination; name two methods of reducing the body dose after a tritium uptake; describe the most common method for determining amount of tritium uptake in the body; describe steps to take following an accidental release of airborne tritium; describe the damage to metals that results from absorption of tritium; explain how washing hands or showering in cold water helps reduce tritium uptake; and describe how tritium exchanges with normal hydrogen in water and hydrocarbons.

  14. Applications of tritium in industry and research

    International Nuclear Information System (INIS)

    Murthy, T.S.; Iyengar, T.S.

    1990-01-01

    As a naturally occuring isotope and as an injected tracer tritium has been found to be useful in meteorology, cosmology, geohydrology, biology, agriculture, and medical sciences both in aqueous and organic forms. In selfluminous compounds, paints and plastics the radioisotopic power of tritium (0.26 w/g) is found to be useful. Several biochemically significant tritium labelled compounds have been produced for use in industry and research. Tritium loaded consumer products are extensively used all over the world. In gas chromatographs and for neutron research tritium targets are found to be useful. This review summarises the various aspects of tritium as a tracer. (author). 7 refs., 1 tab., 1 fig

  15. Type A radioactive liquid sample packaging family

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1995-11-01

    Westinghouse Hanford Company (WHC) has developed two packagings that can be used to ship Type A quantities of radioactive liquids. WHC designed these packagings to take advantage of commercially available items where feasible to reduce the overall packaging cost. The Hedgehog packaging can ship up to one liter of Type A radioactive liquid with no shielding and 15 cm of distance between the liquid and the package exterior, or 30 ml of liquid with 3.8 cm of stainless steel shielding and 19 cm of distance between the liquid and the package exterior. The One Liter Shipper can ship up to one liter of Type A radioactive liquid that does not require shielding

  16. Studies on chemical phenomena of high concentration tritium water and organic compounds of tritium from viewpoint of the tritium confinement

    International Nuclear Information System (INIS)

    Yamanishi, Toshihiko; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Hara, Masanori; Sugiyama, Takahiko; Okuno, Kenji

    2009-01-01

    As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated two research programs on chemical phenomena of high concentration tritium water and organic compounds of tritium from view point of the tritium confinement have been conducted by the C01 team. The results are summarized as follows: (1) Chemical effects of the high concentration tritium water on stainless steels as structural materials of fusion reactors were investigated. Basic data on tritium behaviors at the metal-water interface and corrosion of metal in tritium water were obtained. (2) Development of the tritium confinement and extraction system for the circulating cooling water in the fusion reactor was studied. Improvement was obtained in the performance of a chemical exchange column and catalysts as major components of the water processing system. (J.P.N.)

  17. Type B Drum packages

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1995-11-01

    The Type B Drum package is a container in which a single drum containing Type B quantities of radioactive material will be packaged for shipment. The Type B Drum containers are being developed to fill a void in the packaging and transportation capabilities of the US Department of Energy (DOE), as no double containment packaging for single drums of Type B radioactive material is currently available. Several multiple-drum containers and shielded casks presently exist. However, the size and weight of these containers present multiple operational challenges for single-drum shipments. The Type B Drum containers will offer one unshielded version and, if needed, two shielded versions, and will provide for the option of either single or double containment. The primary users of the Type B Drum container will be any organization with a need to ship single drums of Type B radioactive material. Those users include laboratories, waste retrieval facilities, emergency response teams, and small facilities

  18. Evaluation of tritium analysis techniques for a continuous tritium monitor

    International Nuclear Information System (INIS)

    Fernandez, S.J.; Girton, R.C.

    1978-04-01

    Present methods for tritium monitoring are evaluated and a program is proposed to modify the existing methods or develop new instrumentation to establish a state-of-the-art monitoring capability for nuclear fuel reprocessing plants. The capabilities, advantages, and disadvantages of the most popular counting and separation techniques are described. The following criteria were used to evaluate present methods: specificity, selectivity, precision, insensitivity to gamma radiation, and economy. A novel approach is explored to continuously separate the tritium from a complex mixture of stack gases. This approach, based on the different permeabilities of the stack gas constituents, is integrated into a complete monitoring system. This monitoring system is designed to perform real time tritium analysis. A schedule is presented for development and demonstration of the completed system

  19. Turkey Point tritium. Progress report

    International Nuclear Information System (INIS)

    Ostlund, H.G.; Dorsey, H.G.

    1976-01-01

    In 1972-73 the Florida Power and Light Company (FPL) began operation of two nuclear reactors at Turkey Point on lower Biscayne Bay. One radioactive by-product resulting from the operation of the nuclear reactors, tritium, provides a unique opportunity to study transport and exchange processes on a local scale. Since the isotope in the form of water is not removed from the liquid effluent, it is discharged to the cooling canal system. By studying its residence time in the canal and the pathways by which it leaves the canals, knowledge of evaporative process, groundwater movement, and bay exchange with the ocean can be obtained. Preliminary results obtained from measurement of tritium levels, both in the canal system and in the surrounding environment are discussed. Waters in lower Biscayne Bay and Card and Barnes Sounds receive only a small portion of the total tritium produced by the nuclear plant. The dominating tritium loss most likely is through evaporation from the canals. The capability of measuring extremely low HTO levels allows the determination of the evaporation rate experimentally by measuring the tritium levels of air after having passed over the canals

  20. Tritium Management Loop Design Status

    Energy Technology Data Exchange (ETDEWEB)

    Rader, Jordan D. [ORNL; Felde, David K. [ORNL; McFarlane, Joanna [ORNL; Greenwood, Michael Scott [ORNL; Qualls, A L. [ORNL; Calderoni, Pattrick [Idaho National Laboratory (INL)

    2017-12-01

    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 a 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.

  1. Arctic Shipping

    DEFF Research Database (Denmark)

    Hansen, Carsten Ørts; Grønsedt, Peter; Lindstrøm Graversen, Christian

    maritime industries (including shipping, offshore energy, ports, and maritime service and equipment suppliers) as well as addresses topics that cut across maritime industries (regulation and competitiveness). The topics and narrower research questions addressed in the initiative were developed in close...... at CBS and create the foundation for CBS as a stronger partner for the maritime industries, as well as for other universities and business school with a devotion to maritime economics research. The competitiveness initiative comprises a number of PhD projects and five short term mapping projects...

  2. Tritium in the environment. Knowledge synthesis

    International Nuclear Information System (INIS)

    2009-01-01

    This report first presents the nuclear and physical-chemical properties of tritium and addresses the notions of bioaccumulation, bio-magnification and remanence. It describes and comments the natural and anthropic origins of tritium (natural production, quantities released in the environment in France by nuclear tests, nuclear plants, nuclear fuel processing plants, research centres). It describes how tritium is measured as a free element (sampling, liquid scintillation, proportional counting, enrichment method) or linked to organic matter (combustion, oxidation, helium-3-based measurement). It discusses tritium concentrations noticed in different parts of the environment (soils, continental waters, sea). It describes how tritium is transferred to ecosystems (transfer of atmospheric tritium to ground ecosystems, and to soft water ecosystems). It discusses existing models which describe the behaviour of tritium in ecosystems. It finally describes and comments toxic effects of tritium on living ground and aquatic organisms

  3. Recommended radiological controls for tritium operations

    International Nuclear Information System (INIS)

    Mansfield, G.

    1992-01-01

    This informal report presents recommendations for an adequate radiological protection program for tritium operations. Topics include hazards analysis, facility design, personnel protection equipment, training, operational procedures, radiation monitoring, to include surface and airborne tritium contamination, and program management

  4. Tritium in metals: Techniques of preparation

    International Nuclear Information System (INIS)

    Laesser, R.; Klatt, K.H.; Mecking, P.; Wenzl, H.

    1982-08-01

    In order to study the behavior of tritium in metals, an all metal apparatus has been built for the safe handling of 100 mg of tritium. Samples of palladium, vanadium, niobium, and tantalum were loaded with tritium, deuterium or hydrogen. Some details of the phase diagrams could be established by DTA and by measurement of the lattice parameters. The diffusion of tritium in V, Nb, and Ta was studied with the Gorsky-effect. (TWO)

  5. Tritium decontamination of machine components and walls

    International Nuclear Information System (INIS)

    Hircq, B.; Wong, K.Y.; Jalbert, R.A.; Shmayda, W.T.

    1991-01-01

    Tritium decontamination techniques for machine components and their application at tritium handling facilities are reviewed. These include commonly used methods such as vacuuming, purging, thermal desorption and isotopic exchange as well as less common methods such as chemical/electrochemical etching, plasma discharge cleaning, and destructive methods. Problems associated with tritium contamination of walls and use of protective coatings are reviewed. Tritium decontamination considerations at fusion facilities are discussed

  6. Design of the Target Fabrication Tritium Laboratory

    International Nuclear Information System (INIS)

    Sherohman, J.W.; Roberts, D.H.; Levine, B.H.

    1982-01-01

    The design of the Target Fabrication Tritium Laboratory for deuterium-tritium fuel processing for laser fusion targets has been accomplished with the intent of providing redundant safeguard systems. The design of the tritium laboratory is based on a combination of tritium handling techniques that are currently used by experienced laboratories. A description of the laboratory in terms of its interrelated processing systems is presented to provide an understanding of the design features for safe operation

  7. Linear accelerator for tritium production

    International Nuclear Information System (INIS)

    Garnett, R.W.; Billen, J.H.; Chan, K.C.D.

    1995-01-01

    For many years now, Los Alamos National Laboratory has been working to develop a conceptual design of a facility for accelerator production of tritium (API). The APT accelerator will produce high energy protons which will bombard a heavy metal target, resulting in the production of large numbers of spallation neutrons. These neutrons will be captured by a low-Z target to produce tritium. This paper describes the latest design of a room-temperature, 1.0 GeV, 100 mA, cw proton accelerator for tritium production. The potential advantages of using superconducting cavities in the high-energy section of the linac are also discussed and a comparison is made with the baseline room-temperature accelerator

  8. Implanted-tritium permeation experiment

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Longhurst, G.R.; Miller, L.G.; Watts, K.D.; Kershner, C.J.; Rogers, M.L.

    1981-01-01

    Recent theoretical investigations have pointed to considerable uncertainty in estimating the amount of tritium which will permeate the first wall of a fusion reactor and enter the primary coolant system due in part to the implantation of energetic ions. An experiment is being planned to study this problem in a small test reactor where the 3 He(n,p) 3 T reaction is used to generate protons and tritons for implantation in and permeation of a simulated first wall. By comparing the amount of tritium moving through the wall in the presence of implantation with that in its absence while maintaining the time background partial pressure and temperature, the efflct of implantation on tritium permeation will be determined. The experiment offers an interesting and important complement to similar experiments based on plasmas or ion beams

  9. Tritium calorimeter setup and operation

    CERN Document Server

    Rodgers, D E

    2002-01-01

    The LBNL tritium calorimeter is a stable instrument capable of measuring tritium with a sensitivity of 25 Ci. Measurement times range from 8-hr to 7-days depending on the thermal conductivity and mass of the material being measured. The instrument allows accurate tritium measurements without requiring that the sample be opened and subsampled, thus reducing personnel exposure and radioactive waste generation. The sensitivity limit is primarily due to response shifts caused by temperature fluctuation in the water bath. The fluctuations are most likely a combination of insufficient insulation from ambient air and precision limitations in the temperature controller. The sensitivity could probably be reduced to below 5 Ci if the following improvements were made: (1) Extend the external insulation to cover the entire bath and increase the top insulation. (2) Improve the seal between the air space above the bath and the outside air to reduce evaporation. This will limit the response drift as the water level drops. (...

  10. Tritium oxidation and exchange: preliminary studies

    International Nuclear Information System (INIS)

    Phillips, J.E.; Easterly, C.E.

    1978-05-01

    The radiological hazard resulting from an exposure to either tritium oxide or tritium gas is discussed and the factors contributing to the hazard are presented. From the discussion it appears that an exposure to tritium oxide vapor is 10 4 to 10 5 times more hazardous than exposure to tritium gas. Present and future sources of tritium are briefly considered and indicate that most of the tritium has been and is being released as tritium oxide. The likelihood of gaseous releases, however, is expected to increase in the future, calling to task the present general release assumption that 100% of all tritium released is as oxide. Accurate evaluation of the hazards from a gaseous release will require a knowledge of the conversion rate of tritium gas to tritium oxide. An experiment for determining the conversion rate of tritium gas to tritium oxide is presented along with some preliminary data. The conversion rates obtained for low initial concentrations (10 -4 to 10 -1 mCi/ml) indicate the conversion may proceed more rapidly than would be expected from an extrapolation of previous data taken at higher concentrations

  11. 10 CFR 30.55 - Tritium reports.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Tritium reports. 30.55 Section 30.55 Energy NUCLEAR..., Inspections, Tests, and Reports § 30.55 Tritium reports. (a)-(b) [Reserved] (c) Except as specified in paragraph (d) of this section, each licensee who is authorized to possess tritium shall report promptly to...

  12. Toxicity and dosimetry of tritium

    International Nuclear Information System (INIS)

    Myers, D.K.; Johnson, J.R.

    1991-01-01

    Tritium doses to the general public are very low (currently about 0.2 μSv per year). Radiation doses from tritium to members of the public living in the vicinity of a CANDU power station are higher but rarely exceed 20 μSv per year or 1% of normal exposures to radiation from all natural sources, but doses to some radiation workers can approach ten mSv per year. The relative biological effectiveness (RBE) of tritium beta rays varies appreciably depending upon the biological endpoint. Observed RBE values at low doses and low dose-rates are usually about 2 to 3 when tritium beta rays are compared to 60 Co gamma rays but are closer to 1 than to 2 when compared to 200 kVp X-rays. This conclusion is supported by microdosimetric considerations of the quality of tritium beta rays, 60 Co gamma rays and X-rays. Since X-rays have traditionally been accepted as reference radiation by the International Commission on Radiological Protection, it seems reasonable that the quality factor (Q) assigned to tritium beta rays should be close to one. Recommended procedures in Canada for estimation of effective dose equivalents from exposures to HTO and HT assume that Q = 1 and that body water represents 67% of the mass of soft tissue; they take into account conversions of HTO to appear to be reasonable for radiation protection purposes when the source of exposure is HTO or HT, but will not be adequate for exposures to other tritiated compounds. (modified author abstract) (137 refs., 11 figs., 12 tabs.)

  13. Tritium turnover in succulent plants

    International Nuclear Information System (INIS)

    Krishnamoorthy, T.M.; Gogate, S.S.; Soman, S.D.

    1977-01-01

    Measurements of turnover rates for tissue free water tritium (TFWT) and tissue bound tritium (TBT) were carried out in three succulent plants, Opuntia sp., E. Trigona and E. Mili using tritiated water as tracer. The estimated half-times were 52, 57.5 and 80 days for TFWT and 212, 318 and 132 days for TBT in the stems of the above plants respectively. Opuntia sp. showed significant incorporation of TBT, 10% of TFWT on weight basis, while the other two plants showed lesser incorporation, 2-3% of TFWT. However, the leaves of E. Mili indicated the same level of fixation of TBT as the stem of Opuntia sp. (author)

  14. Tritium processing in JT-60U

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Masaki, Kei

    1997-01-01

    Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7 x 10 19 D-D fusion neutrons were produced by D (d, p) T reactions in plasma, which are expected to produce ∼31 GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium Distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ∼ 7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm 3 for radiation work permit requirements. (author)

  15. Packaging fluency

    DEFF Research Database (Denmark)

    Mocanu, Ana; Chrysochou, Polymeros; Bogomolova, Svetlana

    2011-01-01

    Research on packaging stresses the need for packaging design to read easily, presuming fast and accurate processing of product-related information. In this paper we define this property of packaging as “packaging fluency”. Based on the existing marketing and cognitive psychology literature...... on packaging design and processing fluency, our aim is to define and conceptualise packaging fluency. We stress the important role of packaging fluency since it is anticipated that a fluent package would influence the evaluative judgments for a product. We conclude this paper by setting the research agenda...

  16. Tritium compatibility of alumina and Fosterite

    Energy Technology Data Exchange (ETDEWEB)

    Coffin, D.O.

    1979-09-01

    Many pressure measurements are required to control processing of the fuel gases associated with fusion power reactors. Since most pressure transducers respond to changes in pressure sensitive electrical parameters, insulators will be required to withstand chronic exposures to concentrated tritium. For this investigation samples of alumina and Fosterite were exposed to concentrated tritium gas for 11 weeks. Gas phase impurities were then analyzed for clues that would indicate decomposition of the exposed materials. The only gaseous impurity resulting from these tritium exposures was tritio-methane, which is always produced when tritium is stored in stainless steel containers. There was no evidence that either alumina or Fosterite decomposed in the presence of tritium.

  17. Synthesis of plant hormones labelled by tritium

    International Nuclear Information System (INIS)

    Sidorov, G.V.; Myasoedov, N.F.

    1999-01-01

    Reaction of solid-phase catalytic hydrogenation, isotopic exchange with enriched tritium water, catalytic heterogenous isotopic exchange with gaseous tritium, hydrogenolysis as applied to synthesis of plants labelled by tritium were studied. Auxins, cytokinins, gibberellins, fusicoccins - representatives of the basic hormones of plants - were objects of investigations. In dependence on synthesis method compounds labelled by tritium were prepared with molar radioactivity from 5 up to 155 Ci/mmol. Order of universal approaches to synthesis of plant hormones labelled by tritium was formulated [ru

  18. Management of tritium at nuclear facilities

    International Nuclear Information System (INIS)

    1984-01-01

    This report presents extending summaries of the works of the participants to an IAEA co-ordinated research programme, ''Handling Tritium - bearing effluents and wastes''. The subjects covered include production of tritium in nuclear power plants (mainly heavy water and light water reactors), as well as at reprocessing plants; removal and enrichment of tritium at nuclear facilities; conditioning methods and characteristics of immobilized tritium of low and high concentration; some potential methods of storage and disposal of tritium. In addition to the conclusions of this three-years work, possible activities in the field are recommended

  19. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  20. Tritium behavior intentionally released in the room

    International Nuclear Information System (INIS)

    Kobayashi, K.; Hayashi, T.; Iwai, Y.; Yamanishi, T.; Willms, R. S.; Carlson, R. V.

    2008-01-01

    To construct a fusion reactor with high safety and acceptability, it is necessary to establish and to ensure tritium safe handling technology. Tritium should be well-controlled not to be released to the environment excessively and to prevent workers from excess exposure. It is especially important to grasp tritium behavior in the final confinement area, such as the room and/or building. In order to obtain data for actual tritium behavior in a room and/or building, a series of intentional Tritium Release Experiments (TREs) were planned and carried out within a radiologically controlled area (main cell) at Tritium System Test Assembly (TSTA) in Los Alamos National Laboratory (LANL) under US-JAPAN collaboration program. These experiments were carried out three times. In these experiments, influence of a difference in the tritium release point and the amount of hydrogen isotope for the initial tritium behavior in the room were suggested. Tritium was released into the main cell at TSTA/LANL. The released tritium reached a uniform concentration about 30 - 40 minutes in all the experiments. The influence of the release point and the amount of hydrogen isotope were not found to be important in these experiments. The experimental results for the initial tritium behavior in the room were also simulated well by the modified three-dimensional eddy flow analysis code FLOW-3D. (authors)

  1. Weapons engineering tritium facility overview

    Energy Technology Data Exchange (ETDEWEB)

    Najera, Larry [Los Alamos National Laboratory

    2011-01-20

    Materials provide an overview of the Weapons Engineering Tritium Facility (WETF) as introductory material for January 2011 visit to SRS. Purpose of the visit is to discuss Safety Basis, Conduct of Engineering, and Conduct of Operations. WETF general description and general GTS program capabilities are presented in an unclassified format.

  2. Generation of gaseous tritium standards

    International Nuclear Information System (INIS)

    Hohorst, F.A.

    1994-09-01

    The determination of aqueous and non-aqueous tritium in gaseous samples is one type of determination often requested of radioanalytical laboratories. This determination can be made by introducing the sample as a gas into a sampling train containing two silica gel beds separated by.a catalytic oxidizer bed. The first bed traps tritiated water. The sample then passes into and through the oxidizer bed where non-aqueous tritium containing species are oxidized to water and other products of combustion. The second silica gel bed then traps the newly formed tritiated water. Subsequently, silica gel is removed to plastic bottles, deionized water is added, and the mixture is permitted to equilibrate. The tritium content of the equilibrium mixture is then determined by conventional liquid scintillation counting (LSC). For many years, the moisture content of inert, gaseous samples has been determined using monitors which quantitatively electrolyze the moisture present after that moisture has been absorbed by phosphorous pentoxide or other absorbents. The electrochemical reaction is quantitative and definitive, and the energy consumed during electrolysis forms the basis of the continuous display of the moisture present. This report discusses the experimental evaluation of such a monitor as the basis for a technique for conversion of small quantities of SRMs of tritiated water ( 3 HOH) into gaseous tritium standards ( 3 HH)

  3. Tritium pellet injection sequences for TFTR

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Milora, S.L.; Attenberger, S.E.; Singer, C.E.; Schmidt, G.L.

    1983-01-01

    Tritium pellet injection into neutral deuterium, beam heated deuterium plasmas in the Tokamak Fusion Test Reactor (TFTR) is shown to be an attractive means of (1) minimizing tritium use per tritium discharge and over a sequence of tritium discharges; (2) greatly reducing the tritium load in the walls, limiters, getters, and cryopanels; (3) maintaining or improving instantaneous neutron production (Q); (4) reducing or eliminating deuterium-tritium (D-T) neutron production in non-optimized discharges; and (5) generally adding flexibility to the experimental sequences leading to optimal Q operation. Transport analyses of both compression and full-bore TFTR plasmas are used to support the above observations and to provide the basis for a proposed eight-pellet gas gun injector for the 1986 tritium experiments

  4. Experimental pilot plant for tritium and deuterium separation. Future perspectives

    International Nuclear Information System (INIS)

    Cristescu, I; Bornea, Anisia; Brad, S; Constantin, N; Cristescu, Ioana; Pearsica, Claudia; Poenariu, V; Sofalca, N; Stefan, L; Zamfirache, M

    1998-01-01

    Experimental Pilot Plant for Tritium and Deuterium Separation at Rm. Valcea has the goal of establishing the technological data required to design and operate an industrial-scale for detritiation of heavy water contaminated by various tritium concentrations. The technology developed at Rm. Valcea is based upon the catalytic isotopic exchange heavy water - deuterium, followed of cryogenic distillation of a mixture between molecular species of hydrogen and its isotopes. In the paper we present the main steps of commissioning and we analyse the plant's performances regarding tritium and deuterium separation. The links between the detritiation plant and a CANDU type reactor are presented. The experimental pilot plant contains five modules that can work independently or coupled between them: a. In the isotopic exchange module the transfer of deuterium and tritium from heavy water into the hydrogen flow is realized by means of a catalytic isotope exchange process at 80 o C. The mixed catalytic packing was made and tested. It contains a catalyst of Pt/C/PTFE type and B7 packing; b. In the preliminary purification module the purification of hydrogen is carried out to remove the oxygen and water, which can affect the good functioning of the cryogenic distillation module; c. In the cryogenic distillation module the mixture of hydrogen isotopes is separated; The deuterium concentration at the bottom of the column is up to 99.9% D/(D+H); The column is filled with ordered package and the condenser's temperature is 22 K; d. In the catalytic burning module, deuterated hydrogen is catalytically combined with oxygen and heavy water results with the concentration of 19.9% D/(D+H). The hydrogen burning takes place on mixed catalytic package with 10% hydrophobe catalyst of Pt/C/PTFE type and 90% package; e. The module of water isotopic distillation under vacuum allows heavy water concentration to raise from 49.9% D/(D+H) to 99.8% D/(D+H) on ordered package.(authors)

  5. Microelectronic packaging

    CERN Document Server

    Datta, M; Schultze, J Walter

    2004-01-01

    Microelectronic Packaging analyzes the massive impact of electrochemical technologies on various levels of microelectronic packaging. Traditionally, interconnections within a chip were considered outside the realm of packaging technologies, but this book emphasizes the importance of chip wiring as a key aspect of microelectronic packaging, and focuses on electrochemical processing as an enabler of advanced chip metallization.Divided into five parts, the book begins by outlining the basics of electrochemical processing, defining the microelectronic packaging hierarchy, and emphasizing the impac

  6. Weapons Engineering Tritium Facility, Building 205, Technical Area 16: Los Alamos National Laboratory, Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    1991-04-01

    The Weapons Engineering Tritium Facility (WETF) was planned by the US Department of Energy (DOE) to retain at Los Alamos National Laboratory the capability of repackaging small quantities of tritium to exacting specifications. Small quantities of tritium are required for energy research and development activities and for research on nuclear weapons test devices carried out as part of the laboratory mission. The WETF is an improved design proposed to replace an aging Los Alamos facility where tritium has been repackaged for many years. This Environmental Assessment evaluates the environmental consequences to be expected from operating the new facility, for which construction was completed in 1984, compared with those from continuing to operate the old facility. The document was prepared for compliance with NEPA. In operation, the WETF will incorporate state-of-the-art systems for containing tritium in glove boxes and capturing any tritium released into the glove box exhaust system and the laboratory atmosphere. Liquid discharges from the WETF would contain less than 1% of the tritium found in effluents from the present facility. Effluent streams would be surface discharges and would not enter the aquifer from which municipal water supplies are drawn. The quantity of solid radioactive waste generated at the WETF would be approximately the same as that generated at the present facility. The risk to the public from normal tritium-packaging operations would be significantly less from the WETF than from the present facility. The proposed action will reduce the adverse environmental impacts caused by tritium repackaging by substantially reducing the amount of tritium that escapes to the environment. 35 refs., 3 figs., 21 tabs

  7. On line tritium measurement; La mesure du tritium en ligne

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2011-07-01

    Berthold Technologies has developed a new beacon able to measure the activity of tritium in the atmosphere. The real-time measurement will allow the operator to be warned of the exceeding of radiation thresholds. The air sample to be measured is mixed with a counting gas (generally argon/methane) and pumped through a proportional counter. The device counts the electric impulses due to the interaction between the beta particles generated by the tritium and the counting gas. The detection threshold is 500 Bq/m{sup 3} for a counting time of 1 hour. The device also allows the operator to get an emission spectrum of the air sample. (A.C.)

  8. Estimation of Biological Effects of Tritium.

    Science.gov (United States)

    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.

  9. The LLNL portable tritium processing system

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The end of the Cold War significantly reduced the need for facilities to handle radioactive materials for the US nuclear weapons program. The LLNL Tritium Facility was among those slated for decommissioning. The plans for the facility have since been reversed, and it remains open. Nevertheless, in the early 1990s, the cleanup (the Tritium Inventory Removal Project) was undertaken. However, removing the inventory of tritium within the facility and cleaning up any pockets of high-level residual contamination required that we design a system adequate to the task and meeting today's stringent standards of worker and environmental protection. In collaboration with Sandia National Laboratory and EG ampersand G Mound Applied Technologies, we fabricated a three-module Portable Tritium Processing System (PTPS) that meets current glovebox standards, is operated from a portable console, and is movable from laboratory to laboratory for performing the basic tritium processing operations: pumping and gas transfer, gas analysis, and gas-phase tritium scrubbing. The Tritium Inventory Removal Project is now in its final year, and the portable system continues to be the workhorse. To meet a strong demand for tritium services, the LLNL Tritium Facility will be reconfigured to provide state-of-the-art tritium and radioactive decontamination research and development. The PTPS will play a key role in this new facility

  10. Tritium in Exit Signs | RadTown USA | US EPA

    Science.gov (United States)

    2017-08-07

    Many exit signs contain tritium to light the sign without batteries or electricity. Using tritium in exit signs allows the sign 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. If a tritium exit sign is broken, leave the area immediately and notify the building maintenance staff.

  11. The tritium content of precipitation and groundwater at Yola, Nigeria ...

    African Journals Online (AJOL)

    Tritium is a radioactive isotope of hydrogen which occurs in precipitation. In groundwater studies tritium measurements give information on the time of recharge to the system; the tritium content of precipitation being used to estimate the input of tritium to the groundwater system. At Yola, the tritium ontents in precipitation and ...

  12. Safety evaluation report for packaging (onsite) concrete-lined waste packaging

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

    The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site

  13. Simulation of tritium behavior after intended tritium release in ventilated room

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko; Kobayashi, Kazuhiro; Nishi, Masataka

    2001-01-01

    At the Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI), Caisson Assembly for Tritium Safety study (CATS) with 12 m 3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate tritium behavior in the case where a tritium leak event should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak event should happen in a ventilated room. The RNG model was found to be valid for eddy flow calculation in the 50 m 3 /h ventilated Caisson with acceptable engineering precision. The calculated initial and removal tritium concentration histories after intended tritium release were consistent with the experimental observations in the 50 m 3 /h ventilated Caisson. It is found that the flow near a wall plays an important role for the tritium transport in the ventilated room. On the other hand, tritium behavior intentionally released in the 3,000 m 3 of tritium handling room was investigated experimentally under a US-Japan collaboration. The tritium concentration history calculated with the same method was consistent with the experimental observations, which proves that the present developed method can be applied to the actual scale of tritium handling room. (author)

  14. Green Shipping Practices of Shipping Firms

    Directory of Open Access Journals (Sweden)

    Young-Tae Chang

    2017-05-01

    Full Text Available The primary objective of this study is to provide an empirical research using structural equation modeling to identify the factors that motivate shipping firms to adopt green shipping practices (GSP. Furthermore, it also examines if adopting GSP can enhance the shipping firms’ environmental and productivity performance. The findings show that shipping firms are motivated to adopt GSP mostly by industrial norms set by institutionalized associations. They are also motivated by customers’ demand for environmental friendliness and their own strategy to make good image. Unlike our expectation, government regulations and international environmental laws are not significant in influencing shipping firms to adopt GSP. Moreover, adoption of green shipping practices can improve the environmental and productivity performance of the shipping firms.

  15. Behaviour of tritium in the environment

    International Nuclear Information System (INIS)

    1979-01-01

    Full text: There is considerable interest in the behaviour of radionuclides of global character that may be released to the environment through the development of nuclear power. Tritium is of particular interest due to its direct incorporation into water and organic tissue. Although there has been a large decrease (more than ten times) in tritium concentration since the stopping of nuclear weapons tests in the atmosphere, the construction in the near future of many water reactors and in the far future of fusion reactors could increase the present levels. Progress has been made during recent years in the assessment of tritium distribution, in detection methods and in biological studies While several meetings have given scientists an opportunity to present papers on tritium, no specific symposium on this topic has been organized by the IAEA since 1961. Thus the purpose of the meeting was to review recent advances and to report on the practical aspects of tritium utilization and monitoring. The symposium was jointly organized with OECD/NEA, in co-operation with the US Department of Energy and the Lawrence Livermore Laboratory. Papers were presented on distribution of tritium, evaluation of future discharges, measurement of tritium, tritium in the aquatic environment, tritium in the terrestrial environment, tritium in man and monitoring of tritium Very interesting papers were given on distribution of tritium and participants got a good idea of the circulation of this radionuclide Some new data were provided on tritium pollution from luminous compounds and we learnt that the tritium release of the Swiss luminous compounds industry is of the same order of magnitude as the tritium release of Windscale. Projections indicate that, in the USA, the total quantity of tritium contained in discarded digital watches will be equal, approximately ten years in the future, to the release of nuclear power reactors Whereas nuclear reactor discharges are controlled there is no control

  16. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2002-03-04

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT Shipping Package, and directly related components. This document complies with the minimum requirements as specified in TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event there is a conflict between this document and the SARP or C of C, the SARP and/or C of C shall govern. C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SAR P charges the WIPP Management and Operation (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize these operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs.

  17. Green shipping management

    CERN Document Server

    Lun, Y H Venus; Wong, Christina W Y; Cheng, T C E

    2016-01-01

    This book presents theory-driven discussion on the link between implementing green shipping practices (GSP) and shipping firm performance. It examines the shipping industry’s challenge of supporting economic growth while enhancing environmental performance. Consisting of nine chapters, the book covers topics such as the conceptualization of green shipping practices (GSPs), measurement scales for evaluating GSP implementation, greening capability, greening and performance relativity (GPR), green management practice, green shipping network, greening capacity, and greening propensity. In view of the increasing quest for environment protection in the shipping sector, this book provides a good reference for firms to understand and evaluate their capability in carrying out green operations on their shipping activities.

  18. Radiological training for tritium facilities

    International Nuclear Information System (INIS)

    1996-12-01

    This program management guide describes a recommended implementation standard for core training as outlined in the DOE Radiological Control Manual (RCM). The standard is to assist those individuals, both within DOE and Managing and Operating contractors, identified as having responsibility for implementing the core training recommended by the RCM. This training may also be given to radiological workers using tritium to assist in meeting their job specific training requirements of 10 CFR 835

  19. Radiological training for tritium facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    This program management guide describes a recommended implementation standard for core training as outlined in the DOE Radiological Control Manual (RCM). The standard is to assist those individuals, both within DOE and Managing and Operating contractors, identified as having responsibility for implementing the core training recommended by the RCM. This training may also be given to radiological workers using tritium to assist in meeting their job specific training requirements of 10 CFR 835.

  20. HiPER Tritium factory elements

    Science.gov (United States)

    Guillaume, Didier

    2011-06-01

    HiPER will include a Tritium target factory. This presentation is an overview. We start from process ideas to go to first sketch passing through safety principles. We will follow the Tritium management process. We need first a gas factory producing the right gas mixture from hydrogen, Deuterium and Tritium storage. Then we could pass through the target factory. It is based on our LMJ single shot experiment and some new development like the injector. Then comes pellet burst and vapour recovery. The Tritium factory has to include the waste recovery, recycling process with gas purification before storage. At least, a nuclear plant is not a classical building. Tritium is also very special... All the design ideas have to be adapted. Many facilities are necessary, some with redundancy. We all have to well known these constraints. Tritium budget will be a major contributor for a material point of view as for a financial one.

  1. Modeling tritium transport in the environment

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.

    1986-01-01

    A model of tritium transport in the environment near an atmospheric source of tritium is presented in the general context of modeling material cycling in ecosystems. The model was developed to test hypotheses about the process involved in tritium cycling. The temporal and spatial scales of the model were picked to allow comparison to environmental monitoring data collected in the vicinity of the Savannah River Plant. Initial simulations with the model showed good agreement with monitoring data, including atmospheric and vegetation tritium concentrations. The model can also simulate values of tritium in vegetation organic matter if the key parameter distributing the source of organic hydrogen is varied to fit the data. However, because of the lack of independent conformation of the distribution parameter, there is still uncertainty about the role of organic movement of tritium in the food chain, and its effect on the dose to man

  2. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

    Skinner, C.H.; Coad, J.P.; Federici, G.

    2003-01-01

    Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating

  3. Purification of tritium-free water

    International Nuclear Information System (INIS)

    Hussain, S.D.

    1982-10-01

    Ground water which has been out of contact with the atmosphere for a long time as compared to the half life of tritium (12.43 years) does not contain any measureable amount of tritium. Such water is called tritium-free water. It may contain dissolved and suspended impurities and has to be purified before it can be used for the preparation of blanks and standards required in the routine measurement of low level tritium in water samples. The purification of tritium-free water by distillation in a closed system has been described. The quality of processed tritium-free water was precisely checked at International Atomic Energy Agency (IAEA) Vienna and found satisfactory. (authors)

  4. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

  5. Tritium production in fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.

    1981-08-01

    The present analyses on the possibilities of extracting tritium from the liquid and solid fusion reactor blankets show up many problems. A consistent ensemble of materials and devices for extracting the heat and the tritium has not yet been integrated in a fusion reactor blanket project. The dimensioning of the many pipes required for shifting the tritium can only be done very approximately and the volume taken up by the blanket is difficult to evaluate, etc. The utilization of present data leads to over-dimensioning the installations by prudence and perhaps rejecting the best solutions. In order to measure the parameters of the most promising materials, work must be carried out on well defined samples and not only determine the base physical-chemical coefficients, such as thermal conductivity, scattering coefficients, Sievert parameters, but also the kinetic parameters conventional in chemical engineering, such as the hourly space rates of degassing. It is also necessary to perform long duration experiments under radiation and at operating temperatures, or above, in order to study the ageing of the bodies employed [fr

  6. Regulating tritium in drinking water

    International Nuclear Information System (INIS)

    Fluke, R.

    1994-01-01

    This article incorporates an article by E. Koehl from an internal Ontario Hydro publication, and a letter from the Joint Committee of Health and Safety of the Royal Society of Canada and the Canadian Academy of Engineering, submitted to the Ontario Minister of the Environment and Energy. The Advisory Committee on Environmental Standards had recommended that the limit for tritium in Ontario drinking water be reduced from 40,000 to 100 Bq/L, with a further reduction to 20 in five years. Some facts and figures are adduced to show that the effect of tritium in drinking water in Ontario is negligible compared to the effect of background radiation. The risk from tritium to the people of Ontario is undetectably small, and the attempt to estimate this risk by linear extrapolation is extremely dubious. Regulation entails social and economic costs, and the government ought to ensure that the benefits exceed the costs. The costs translate into nothing less than wasted opportunity to save lives in other ways. 3 refs

  7. Tritium handling systems for TFTR and PITR

    International Nuclear Information System (INIS)

    Lind, K.E.; Yemin, L.; Rossmassler, R.; Phillips, H.O.; Youssef, N.S.; Levine, J.D.; Howe, H.J.; Pierce, C.W.

    1978-01-01

    Operation of the Tokamak Fusion Test Reactor (TFTR) and the proposed Princeton Ignition Test Reactor (PITR) will involve the generation and burning of a deuterium-tritium plasma. Systems associated with both the TFTR and PITR will be designed to minimize the potential release of tritium and other radioisotopes under both routine operational and accidental conditions. Systems have been proposed for the control and processing of large tritium inventories

  8. Tritium radioluminescent devices, Health and Safety Manual

    Energy Technology Data Exchange (ETDEWEB)

    Traub, R.J.; Jensen, G.A.

    1995-06-01

    This document consolidates available information on the properties of tritium, including its environmental chemistry, its health physics, and safe practices in using tritium-activated RL lighting. It also summarizes relevant government regulations on RL lighting. Chapters are divided into a single-column part, which provides an overview of the topic for readers simply requiring guidance on the safety of tritium RL lighting, and a dual-column part for readers requiring more technical and detailed information.

  9. Ontario approves limited sales of tritium

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    The Ontario government announced at the end of August 1989 that it had authorized the limited sale of tritium to help establish a self-powered lighting industry, and for certain other specified purposes including fusion research. To ensure that the tritium will be used only for peaceful purposes, sales will be restricted and subject to strict controls. All sales of tritium will be made public

  10. The movement of tritium in ecological systems

    International Nuclear Information System (INIS)

    Polevoy, Y; Laichter, Y.

    1988-11-01

    This literature survey summarizes the interaction of tritium gas and tritiated water with various components of the ecological system. The intake of tritium gas and tritiated water in plants and soil is described as well as the location of the highest measurable concentration. This information may serve as a basis for risk assessment from tritium to man through the food chain and enables effective tracing of its concentration in the environment. (author)

  11. Degradation of elastomers by tritium beta radiation

    International Nuclear Information System (INIS)

    Zapp, P.E.; Tuer, G.L. Jr.

    1984-01-01

    Based on its tritium radiation resistance, ethylene propylene rubber has been selected as a candidate for replacement of nitrile rubber in the SRP tritium facilities. A specification for flange gasket material has been developed for ethylene propylene such that its mechanical properties are similar to those of nitrile rubber. In-process testing of ethylene propylene and nitrile gaskets will be conducted in the tritium facilities under identical exposure conditions

  12. Effluent Treatment Facility tritium emissions monitoring

    International Nuclear Information System (INIS)

    Dunn, D.L.

    1991-01-01

    An Environmental Protection Agency (EPA) approved sampling and analysis protocol was developed and executed to verify atmospheric emissions compliance for the new Savannah River Site (SRS) F/H area Effluent Treatment Facility. Sampling equipment was fabricated, installed, and tested at stack monitoring points for filtrable particulate radionuclides, radioactive iodine, and tritium. The only detectable anthropogenic radionuclides released from Effluent Treatment Facility stacks during monitoring were iodine-129 and tritium oxide. This paper only examines the collection and analysis of tritium oxide

  13. Tritium radioluminescent devices, Health and Safety Manual

    International Nuclear Information System (INIS)

    Traub, R.J.; Jensen, G.A.

    1995-06-01

    This document consolidates available information on the properties of tritium, including its environmental chemistry, its health physics, and safe practices in using tritium-activated RL lighting. It also summarizes relevant government regulations on RL lighting. Chapters are divided into a single-column part, which provides an overview of the topic for readers simply requiring guidance on the safety of tritium RL lighting, and a dual-column part for readers requiring more technical and detailed information

  14. Tritium in precipitation of Vostok (Antarctica): conclusions on the tritium latitude effect.

    Science.gov (United States)

    Hebert, Detlef

    2011-09-01

    During the Antarctic summer of 1985 near the Soviet Antarctic station Vostok, firn samples for tritium measurements were obtained down to a depth of 2.40 m. The results of the tritium measurements are presented and discussed. Based on this and other data, conclusions regarding the tritium latitude effect are derived.

  15. Tritium loss in molten flibe systems

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Scott Willms, R.

    2000-01-01

    An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF.BeF 2 , commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons

  16. Development of tritium handing technology(II)

    International Nuclear Information System (INIS)

    Chung, H. S.; Ahn, D. H.; Kim, K. R.; Yook, D. S.; Song, K. M.; Son, S. H.; Lee, K. J.; Jung, H. Y.; Song, M. C.

    2004-02-01

    The buildup rate of tritium in heavy water moderator and coolant of pressurized heavy water reactors in Wolsong Nuclear Power Plant is about 4MCi/a. The control of tritium is of increasing concern to the power reactor industry and general public in Korea. Metal tritides have the advantage of significantly decreasing the volume required to store tritium without increasing the pressure of storage vessel. Titanium hydride was safely used for the long-term storage of tritium. The experimental thermodynamic P-C-T data show that titanium soaks up hydrogen isotope gas at ambient temperature and modest pressures

  17. Tritium resources available for fusion reactors

    Science.gov (United States)

    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

  18. Development of tritium handing technology(II)

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. S.; Ahn, D. H.; Kim, K. R. [KAERI, Taejon (Korea, Republic of); Yook, D. S.; Song, K. M.; Son, S. H. [KEPRI, Taejon (Korea, Republic of); Lee, K. J.; Jung, H. Y.; Song, M. C. [KAIST, Taejon (Korea, Republic of)

    2004-02-01

    The buildup rate of tritium in heavy water moderator and coolant of pressurized heavy water reactors in Wolsong Nuclear Power Plant is about 4MCi/a. The control of tritium is of increasing concern to the power reactor industry and general public in Korea. Metal tritides have the advantage of significantly decreasing the volume required to store tritium without increasing the pressure of storage vessel. Titanium hydride was safely used for the long-term storage of tritium. The experimental thermodynamic P-C-T data show that titanium soaks up hydrogen isotope gas at ambient temperature and modest pressures.

  19. Tritium control in helium-cooled blankets

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Maya, I.; Kessel, C.; Roelant, D.; Schultz, K.R.

    1985-06-01

    As a part of the Blanket Comparison and Selection Study (BCSS), GA Technologies was responsible for the design of helium-cooled, solid- and liquid-metal breeder blankets. Conceptual blanket designs were developed, including the consideration of the generation, transport, and extraction of tritium. Evaluations were made of the inventory and leakage of tritium for helium-cooled Li 2 O and LiAlO 2 and liquid lithium breeder blankets for tokamak and tandem mirror reactors. To facilitate the evaluation, a solid breeder tritium code TRIT4 was developed. The results from this study indicate that tritium inventories and leakages are acceptable for the proposed helium-cooled blankets. An assumption made in the tritium leakage calculations was that tritium is released to the helium purge and coolant streams as T 2 and remains in that form. If oxidation to T 2 O is possible, significant reduction in the tritium leakage will be possible. We conclude that more experimental data on breeder material properties and tritium permeation behavior are needed. However, we are certain that an adequate number of different techniques are available to control the breeder tritium inventory and leakage to an acceptable level in helium-cooled solid- and lithium-breeder blankets

  20. Measurement of tritium concentration in urine

    International Nuclear Information System (INIS)

    Sekiyama, Shigenobu; Deshimaru, Takehide

    1979-01-01

    Concerning the safety management of the advanced thermal reactor ''Fugen'', the internal exposure management for tritium is important, because heavy water is used as the moderator in the reactor, and tritium is produced in the heavy water. Tritium is the radioactive nuclide with the maximum β-ray energy of 18 keV, and the radiation exposure is limited to the internal exposure in human bodies, as tritium is taken in through the skin and by breathing. The tritium concentration in urine of the operators of the Fugen plant was measured. As for tritium measurement, the analysis of raw urine, the analysis after passing through mixed ion exchange resin and the analysis after distillation are applied. The scintillator, the liquid scintillation counter, the ion exchange resin and the distillator are introduced. The preliminary survey was conducted on the urine sample, the scintillator the calibration, etc. The measuring condition, the measurement of efficiency, and the limitation of detection with various background are explained, with the many experimental data and the calculating formula. Concerning the measured tritium concentration in urine, the tritium concentrations in distilled urine, raw urine and the urine refined with ion exchange resin were compared, and the correlation formulae are presented. The actual tritium concentration value in urine was less than 50 pci/ml. The measuring methods of raw urine and the urine refined with ion exchange resin are adequate as they are quick and accurate. (Nakai, Y.)

  1. Tritium proof-of-principle pellet injector

    International Nuclear Information System (INIS)

    Fisher, P.W.

    1991-07-01

    The tritium proof-of-principle (TPOP) experiment was designed and built by Oak Ridge National Laboratory (ORNL) to demonstrate the formation and acceleration of the world's first tritium pellets for fueling of future fusion reactors. The experiment was first used to produce hydrogen and deuterium pellets at ORNL. It was then moved to the Tritium Systems Test Assembly at Los Alamos National Laboratory for the production of tritium pellets. The injector used in situ condensation to produce cylindrical pellets in a 1-m-long, 4-mm-ID barrel. A cryogenic 3 He separator, which was an integral part of the gun assembly, was capable of lowering 3 He levels in the feed gas to <0.005%. The experiment was housed to a glovebox for tritium containment. Nearly 1500 pellets were produced during the course of the experiment, and about a third of these were pure tritium or mixtures of deuterium and tritium. Over 100 kCi of tritium was processed through the experiment without incident. Tritium pellet velocities of 1400 m/s were achieved with high-pressure hydrogen propellant. The design, operation, and results of this experiment are summarized. 34 refs., 44 figs., 3 tabs

  2. Tritium stripping by a catalytic exchange stripper

    International Nuclear Information System (INIS)

    Heung, L.K.; Gibson, G.W.; Ortman, M.S.

    1991-01-01

    A catalytic exchange process for stripping elemental tritium from gas streams has been demonstrated. The process uses a catalyzed isotopic exchange reaction between tritium in the gas phase and protium or deuterium in the solid phase on alumina. The reaction is catalyzed by platinum deposited on the alumina. The process has been tested with both tritium and deuterium. Decontamination factors (ration of inlet and outlet tritium concentrations) as high as 1000 have been achieved, depending on inlet concentration. The test results and some demonstrated applications are presented

  3. Tritium in the environment: origins and analysis

    International Nuclear Information System (INIS)

    Baglan, N.

    2009-01-01

    The author recalls the chemical reactions at the origin of tritium formation, that tritium has been introduced in the atmosphere by nuclear tests, and that it is now produced by nuclear reactors. He also outlines that tritium is mainly released in waters (oceans, seas, rivers) or in the atmosphere. Some high concentrations may therefore occur. He discusses measurements of activity in rain waters and surface waters, and outlines the impact of the end of atmospheric nuclear tests. He discusses the technical challenges of low level tritium analysis and activity measurement

  4. The shipping man adventures in ship finance

    CERN Document Server

    McCleery, Matthew

    2013-01-01

    When restless New York City hedge fund manager Robert Fairchild watches the Baltic Dry Cargo Index plunge 97%, registering an all-time high and a 25-year low within the span of just six months, he decides to buy a ship. Immediately fantasizing about naming a vessel after his wife, carrying a string of worry beads and being able to introduce himself as a "shipowner" at his upcoming college reunion, Fairchild immediately embarks on an odyssey into the most exclusive, glamorous and high stakes business in the world. From pirates off the coast of Somalia and on Wall Street to Greek and Norwegian shipping magnates, the education of Robert Fairchild is an expensive one. In the end, he loses his hedge fund, but he gains a life - as a Shipping Man. Part fast paced financial thriller, part ship finance text book, The Shipping Man is 310 pages of required reading for anyone with an interest in capital formation for shipping.

  5. MEMS packaging

    CERN Document Server

    Hsu , Tai-Ran

    2004-01-01

    MEMS Packaging discusses the prevalent practices and enabling techniques in assembly, packaging and testing of microelectromechanical systems (MEMS). The entire spectrum of assembly, packaging and testing of MEMS and microsystems, from essential enabling technologies to applications in key industries of life sciences, telecommunications and aerospace engineering is covered. Other topics included are bonding and sealing of microcomponents, process flow of MEMS and microsystems packaging, automated microassembly, and testing and design for testing.The Institution of Engineering and Technology is

  6. Phase 1 Final Report for In-Situ Tritium Beta Detector

    Energy Technology Data Exchange (ETDEWEB)

    Berthold, J.W.; Jeffers, L.A.

    1998-04-15

    The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye.

  7. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.; Kopasz, J.P.; Tam, S.W.

    1994-01-01

    In an operating fusion reactor,, the tritium breeding blanket will reach a condition in which the tritium release rate equals the production rate. The tritium release rate must be fast enough that the tritium inventory in the blanket does not become excessive. Slow tritium release will result in a large tritium inventory, which is unacceptable from both economic and safety viewpoints As a consequence, considerable effort has been devoted to understanding the tritium release mechanism from ceramic breeders and beryllium neutron multipliers through theoretical, laboratory, and in-reactor studies. This information is being applied to the development of models for predicting tritium release for various blanket operating conditions

  8. Determination of tritium by counting; Dosage du tritium par comptage

    Energy Technology Data Exchange (ETDEWEB)

    Schott, R.; Froment, G.; Pinson, J.; Genty, C. [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes

    1968-07-01

    Ionisation chamber assay of tritium in any gaseous mixture is a simple, fast and accurate method. We used the method of relative determination by comparison to a standard rather than the method of absolute assay in which case the constants are known with too little accuracy. The efficiency of the chamber was studied in connection to the pressure inside the chamber and its total volume. The calibration is linear in the range we are taking into account (1 to 80 millicuries). The reproducibility of the method is good: 13 runs gave a coefficient of variation of 1.6 per cent. The relative accuracy was found equal to {+-} 1.3 per cent. To end the paper, we describe in detail the apparatus and the ways of proceedings. (authors) [French] Le comptage du tritium par chambre d'ionisation est une methode simple, rapide et precise pour determiner la teneur en tritium d'un melange gazeux quelconque. Nous avons prefere utiliser la methode de determination relative par rapport a un etalon car, dans le cas d'une determination absolue, les constantes sont connues avec une trop grande incertitude. L'efficacite de la chambre a ete etudiee en fonction de la variation de la pression d'argon a l'interieur de la chambre et du volume total, de cette derniere. L'etalonnage s'est revele lineaire dans le domaine de mesures qui nous interessaient (1 a 80 millicuries). La reproductibillte de la methode est tres bonne, le coefficient de variation pour une serie de 13 essais etant de 1,6 pour cent, quant a la precision relative, elle a ete evaluee a {+-} 1,3 pour cent. Pour terminer, nous donnons une description detaillee de l'appareillage utilise et du mode operatoire suivi. (auteurs)

  9. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

  10. The effective cost of tritium for tokamak fusion power reactors with reduced tritium production systems

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Evans, K.

    1983-01-01

    If sufficient tritium cannot be produced and processed in tokamak blankets then at least two alternatives are possible. Tritium can be purchased; or reactors with reduced tritium (RT) content in the plasma can be designed. The latter choice may require development of magnet technology etc., but the authors show that the impact on the cost-of-electricity may be mild. Cost tradeoffs are compared to the market value of tritium. Adequate tritium production in fusion blankets is preferred, but the authors show there is some flexibility in the deployment of fusion if this is not possible

  11. Transfer and incorporation of tritium in mammals

    International Nuclear Information System (INIS)

    Hoek, J. van den; Juan, N.B.

    1979-01-01

    The metabolism of tritium in mammals has been studied in a number of laboratories which have participated in the IAEA Co-ordinated Research Programme on the Behaviour of Tritium in the Environment. The results of these studies are discussed and related to data obtained elsewhere. The animals studied are small laboratory and domestic animals. Tritium has been administered as THO, both in single and long-term dosing experiments, and also as organically bound tritium. The biological half-life of tritium in the body water pool has been determined in different species. The following values have been found: 1.1 days in mice; 13.2 days in kangaroo rats; 3.8 days in pigs; 4.1 days in lactating versus 8.3 in non-lactating goats and 3.1-4.0 days in lactating cows and steers. Much attention has been paid to the incorporation of tritium into organic constituents, both in the animal organism (organs, tissues) and in the secretions of the animal after continuous administration of tritium, mostly as THO. When compared with tritium levels in body water, and expressed as the ratio of specific activities, values of 0.25 and 0.40 have been found in mice liver and testis respectively. In cow's milk, these ratios vary from 0.30 for casein to 0.60 for lactose. The transfer of tritium into milk after continuous ingestion of THO by a lactating cow is about 1.50% of the daily ingested tritium per litre of milk. Some results of experiments, utilizing organically bound tritium, are also presented. (author)

  12. Tritium sorption on protective coatings for concrete

    International Nuclear Information System (INIS)

    Miller, J.M.; Senohrabek, J.A.; Allsop, P.A.

    1992-11-01

    Because of the high sorption level of tritium on unprotected concrete, a program to examine the effectiveness of various concrete coatings and sealants in reducing tritium sorption was undertaken, and various exposure conditions were examined. Coatings of epoxy, polyurethane, bituminous sealant, bituminous sealant covered with polyvinylidene chloride wrap, alkyd paint, and sodium silicate were investigated with tritium (HTO) vapor concentration, humidity and contact time being varied. An exposure to HT was also carried out, and the effect of humidity on the tritium desorption rate was investigated. The relative effectiveness of the coatings was in the order of bituminous sealant + wrap > bituminous sealant > solvent-based epoxy > 100%-solids epoxy > alkyd paint > sodium silicate. The commercially available coatings for concrete resulted in tritium sorption being reduced to less than 7% of unprotected concrete. This was improved to ∼0.1% with the use of the Saran wrap (polyvinylidene chloride). The amount of tritium sorbed was proportional to tritium concentration. The total tritium sorbed decreased with an increase in humidity. A saturation effect was observed with increasing exposure time for both the coated and unprotected samples. Under the test conditions, complete saturation was not achieved within the maximum 8-hour contact time, except for the solvent-based epoxy. The desorption rate increased with a higher-humidity air purge stream. HT desorbed more rapidly than HTO, but the amount sorbed was smaller. The experimental program showed that HTO sorption by concrete can be significantly reduced with the proper choice of coating. However, tritium sorption on concrete and proposed coatings will continue to be a concern until the effects of the various conditions that affect the adsorption and desorption of tritium are firmly established for both chronic and acute tritium release conditions. Material sorption characteristics must also be considered in

  13. Application of tritium behavior simulation code (TBEHAVIOR) to an actual-scale tritium handling room

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Hayashi, Takumi; Kobayashi, Kazuhiro; Yamanishi, Toshihiko

    2007-11-01

    It is essential from the viewpoint of fusion safety to confine and remove tritium in a room since tritium handling room is placed as 'final barrier' of fusion plant to prevent the environmental discharge of tritium. At the Tritium Process Laboratory (TPL) of Japan Atomic Energy Agency (JAEA), the application of our original three-dimensional TBEHAVIOR code to the tritium behavior in a room of 3000 m 3 was verified. The Renormalization Group Theory (RNG) model was selected as Low-Reynolds model for practical calculation time as well as to reasonable precision in evaluation of velocity from the engineering viewpoint. A series of evaluated results indicated that a flow adjacent to a wall surface plays an important role for tritium transport in a ventilated room. Evaluation of attenuating behavior is further important since the ventilation is normally stopped for the tritium confinement in the case of tritium leakage. We demonstrated that an attenuating behavior can also be evaluated well by the TBEHAVIOR code. Even an attenuating or stagnant flow of less than 10mm/s in a room mixed tritium concentration uniform promptly. The presence of apparatuses in a room did not generally affect tritium behavior. Although the effect of buoyancy was limited to the initial period after the leak, the spread of tritium was promoted by buoyancy. It led to the shortening of elapsed time until the concentration became uniform. (author)

  14. Doses due to tritium releases by NET - data base and relevant parameters on biological tritium behaviour

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.

    1990-12-01

    This study gives an overview on the current knowledge about the behaviour of tritium in plants and in food chains in order to evaluate the ingestion pathway modelling of existing computer codes for dose estimations. The tritium uptake and retention by plants standing at the beginning of the food chains is described. The different chemical forms of tritium, which may be released into the atmosphere (HT, HTO and tritiated organics), and incorporation of tritium into organic material of plants are considered. Uptake and metabolism of tritiated compounds in animals and man are reviewed with particular respect to organically bound tritium and its significance for dose estimations. Some basic remarks on tritium toxicity are also included. Furthermore, a choice of computer codes for dose estimations due to chronic or accidental tritium releases has been compared with respect to the ingestion pathway. (orig.) [de

  15. Tritium removal using vanadium hydride

    International Nuclear Information System (INIS)

    Hill, F.B.; Wong, Y.W.; Chan, Y.N.

    1978-01-01

    The results of an initial examination of the feasibility of separation of tritium from gaseous protium-tritium mixtures using vanadium hydride in cyclic processes is reported. Interest was drawn to the vanadium-hydrogen system because of the so-called inverse isotope effect exhibited by this system. Thus the tritide is more stable than the protide, a fact which makes the system attractive for removal of tritium from a mixture in which the light isotope predominates. The initial results of three phases of the research program are reported, dealing with studies of the equilibrium and kinetics properties of isotope exchange, development of an equilibrium theory of isotope separation via heatless adsorption, and experiments on the performance of a single heatless adsorption stage. In the equilibrium and kinetics studies, measurements were made of pressure-composition isotherms, the HT--H 2 separation factors and rates of HT--H 2 exchange. This information was used to evaluate constants in the theory and to understand the performance of the heatless adsorption experiments. A recently developed equilibrium theory of heatless adsorption was applied to the HT--H 2 separation using vanadium hydride. Using the theory it was predicted that no separation would occur by pressure cycling wholly within the β phase but that separation would occur by cycling between the β and γ phases and using high purge-to-feed ratios. Heatless adsorption experiments conducted within the β phase led to inverse separations rather than no separation. A kinetic isotope effect may be responsible. Cycling between the β and γ phases led to separation but not to the predicted complete removal of HT from the product stream, possibly because of finite rates of exchange. Further experimental and theoretical work is suggested which may ultimately make possible assessment of the feasibility and practicability of hydrogen isotope separation by this approach

  16. Environmental tritium monitoring around Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Yamato, A.

    1982-01-01

    The environmental tritium monitoring in the sea near Tokai Reprocessing Plant has been performed since 1977, the year of having started the hot test operation of the plant. On the other hand, atmospheric tritium monitoring was started almost at the same time as a research program instead of a routine program. This paper is a review for tritium monitoring in the sea and in the air around the Tokai Reprocessing Plant. The plant is located in Tokai Village, Ibaraki Prefecture, on the Pacific coast. It is based on the Purex process, and the nominal capacity is 210 tons per year. Around the TRP, there are four uranium fabrication facilities, five research reactors, two power reactors and other research facilities. About 173,000 inhabitants are within 10 km range from the plant. The authorized discharge limit of tritium is 200 Ci per day and 51,100 Ci per year in the sea. That in the atmosphere is 50 Ci per day and about 15,000 Ci per year. The tritium from the TRP was discharged mainly into the sea. The sea water samples were distilled, and the tritium concentration was measured by liquid scintillation counting. During three years of the hot operation of TRP, discharged tritium was about 7,000 Ci into the sea and about 140 Ci into the atmosphere. The tritium level has been maintained, and its significant increase was not observed. (Kako, I.)

  17. Enantiospecific tritium labeling of 28-homocastasterone

    Czech Academy of Sciences Publication Activity Database

    Elbert, Tomáš; Patil, Mahadeo Rajshekhar; Marek, Aleš

    2017-01-01

    Roč. 60, č. 3 (2017), s. 176-182 ISSN 0362-4803 R&D Projects: GA AV ČR IAA400550801 Institutional support: RVO:61388963 Keywords : 28-homocastasterone * brassinosteroids * enantiospecific reaction * tritium dehalogenation * tritium labeling Subject RIV: CC - Organic Chemistry OBOR OECD: Organic chemistry Impact factor: 1.745, year: 2016

  18. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2009-06-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  19. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2008-09-11

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the pplication." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  20. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2008-01-01

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the 'RH-TRU 72-B cask') and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous

  1. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2006-01-01

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 Code of Federal Regulations (CFR) 1.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these

  2. RH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2008-01-12

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the "RH-TRU 72-B cask") and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a

  3. RH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2006-11-07

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to

  4. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2009-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  5. Tritium labeling of detonation nanodiamonds.

    Science.gov (United States)

    Girard, Hugues A; El-Kharbachi, Abdelouahab; Garcia-Argote, Sébastien; Petit, Tristan; Bergonzo, Philippe; Rousseau, Bernard; Arnault, Jean-Charles

    2014-03-18

    For the first time, the radioactive labeling of detonation nanodiamonds was efficiently achieved using a tritium microwave plasma. According to our measurements, the total radioactivity reaches 9120 ± 120 μCi mg(-1), with 93% of (3)H atoms tightly bonded to the surface and up to 7% embedded into the diamond core. Such (3)H doping will ensure highly stable radiolabeled nanodiamonds, on which surface functionalization is still allowed. This breakthrough opens the way to biodistribution and pharmacokinetics studies of nanodiamonds, while this approach can be scalable to easily treat bulk quantities of nanodiamonds at low cost.

  6. Methane generated from graphite--tritium interaction

    International Nuclear Information System (INIS)

    Coffin, D.O.; Walthers, C.R.

    1979-01-01

    When hydrogen isotopes are separated by cryogenic distillation, as little as 1 ppM of methane will eventually plug the still as frost accumulates on the column packings. Elemental carbon exposed to tritium generates methane spontaneously, and yet some dry transfer pumps, otherwise compatible with tritium, convey the gas with graphite rotors. This study was to determine the methane production rate for graphite in tritium. A pump manufacturer supplied graphite samples that we exposed to tritium gas at 0.8 atm. After 137 days we measured a methane synthesis rate of 6 ng/h per cm 2 of graphite exposed. At this rate methane might grow to a concentration of 0.01 ppM when pure tritium is transferred once through a typical graphite--rotor transfer pump. Such a low methane level will not cause column blockage, even if the cryogenic still is operated continuously for many years

  7. The organic tritium in the environment

    International Nuclear Information System (INIS)

    Kirchmann, R.

    1979-01-01

    Sources, organization process, and biological availability of organic tritium released in the environment, transfer of organic tritium in the environment from methane or soil to plants and from food to mammals, transfer of tritium in aquatic ecosystems, and dose to man resulting of the ingestion of tritiated food were reviewed and discussed. Some data about transfer of organic tritium in terrestrial and aquatic ecosystems reported by literatures were summarized and were supplied with recent data on biological accumulation of organic tritium in the food chain. It was stressed that more research must be done in future because data available were still insufficient. Last, some research programs in progress or planned were stated. (Tsunoda, M.)

  8. Handling of tritium-bearing wastes

    International Nuclear Information System (INIS)

    1981-01-01

    The generation of nuclear power and reprocessing of nuclear fuel results in the production of tritium and the possible need to control the release of tritium-contaminated effluents. In assessing the need for controls, it is necessary to know the production rates of tritium at different nuclear facilities, the technologies available for separating tritium from different gaseous and liquid streams, and the methods that are satisfactory for storage and disposal of tritiated wastes. The intention in applying such control technologies and methods is to avoid undesirable effects on the environment, and to reduce the radiation burden on operational personnel and the general population. This technical report is a result of the IAEA Technical Committee Meeting on Handling of Tritium-bearing Effluents and Wastes, which was held in Vienna, 4 - 8 December 1978. It summarizes the main topics discussed at the meeting and appends the more detailed reports on particular aspects that were prepared for the meeting by individual participants

  9. Disposal of tritium-exposed metal hydrides

    International Nuclear Information System (INIS)

    Nobile, A.; Motyka, T.

    1991-01-01

    A plan has been established for disposal of tritium-exposed metal hydrides used in Savannah River Site (SRS) tritium production or Materials Test Facility (MTF) R ampersand D operations. The recommended plan assumes that the first tritium-exposed metal hydrides will be disposed of after startup of the Solid Waste Disposal Facility (SWDF) Expansion Project in 1992, and thus the plan is consistent with the new disposal requiremkents that will be in effect for the SWDF Expansion Project. Process beds containing tritium-exposed metal hydride powder will be disposed of without removal of the powder from the bed; however, disposal of tritium-exposed metal hydride powder that has been removed from its process vessel is also addressed

  10. Uptake of atmospheric tritium by market foods

    International Nuclear Information System (INIS)

    Inoue, Y.; Tanaka-Miyamoto, K.; Iwakura, T.

    1992-01-01

    In this paper uptake of tritium by market foods from tritiated water vapor in the air is investigated using cereals and beans purchased in Deep River, Canada. The concentrations of tissue free water tritium (TFWT) and organically bound tritium (OBT) range from 12 to 79% and from 10 to 38% respectively, of that estimated for atmospheric water vapor of the sampling month. The specific activity ratios of OBT to TFWT were constant for cereals, but variable for beans. The elevated OBT was shown to be the result of isotopic exchange of labile hydrogen by the fact that washing the foods with tritium free-water reduced their tritium contents to levels characteristic of their production sites

  11. Experiences with decontaminating tritium-handling apparatus

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

    1992-01-01

    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 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

  12. Management of Tritium in European Spallation Source

    DEFF Research Database (Denmark)

    Ene, Daniela; Andersson, Kasper Grann; Jensen, Mikael

    2015-01-01

    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 shown that the annual release 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...

  13. Recent environmental tritium levels in Japan

    International Nuclear Information System (INIS)

    Iwakura, T.; Inoue, Y.; Tanaka, K.; Kasida, Y.

    1982-01-01

    Data of the tritium surveillance program are summarized for the period of 1967 through 1980. Samples of surface water, tap water, coastal sea water and ground water were collected from environs of commercial nuclear power plants and nuclear facilities, and were analyzed by liquid scintillation counting. Although the results show some differences in tritium concentrations in water samples from various part of the country, there is a general tendency of the concentration in surface waters to decline as a function of time. This implies that environmental waters in Japan generally have not been influenced by the discharged effluents of the facilities or the stations with regard to tritium contamination and that the tritium content of precipitation still plays the dominant role in reflecting annual variation of tritium concentration in surface waters. (J.P.N.)

  14. Diurnal variations of tritium uptake by plants

    International Nuclear Information System (INIS)

    Hettinger, M.; Diabate, S.; Strack, S.

    1991-02-01

    The influence of the diurnal cycle is important for the behaviour of environmental tritium in the vegetation. A mathematical model has been used to calculate the deposition of tritium in plants as a function of diurnal variations of climatic parameters. The necessary physiological parameters (relationship of net photosynthesis and growth) were derived from growth experiments for tomatoes and maize. In chamber experiments, tomato and maize plants were exposed to tritium with natural diurnal variations of the climatic conditions. Within the range of standard deviations the measured concentrations of tritium in tissue free water of tomatoes correspond well to the estimated values. Furthermore, the incorporation into non-exchangeable organically bound tritium (OBT nx) can be sufficiently modelled and explained. There are deviations from the estimated concentrations in some parts of maize leaves. (orig.) [de

  15. Commercial nuclear ships

    International Nuclear Information System (INIS)

    Poirier d'Ange d'Orsay, P.; Andrieu, J.L.

    1975-01-01

    Following discussion of the arguments in favour of nuclear merchant ships and development of nuclear ship propulsion, the rentability of nuclear ships is dealt with in a comparative study with conventional ships regarding fuel costs, investments and environmental impact. The structure and components of a nuclear ship are described, with emphasis on its safety equipment. The problem of insurance is examined, particularly in connection with the difficulty in applying insurance for liability due to the present uncertainties of the legal position at international level. French legislation does not solve the problem of visiting foreign nuclear ships. The risk concept at operational, design and construction stages is also discussed. Unresolved difficulties, especially concerning the construction of nuclear ships, uncertainties in competitivity analyses between conventional and nuclear propulsion, and implementation of specific and complete legislation on nuclear ships are also mentioned, particularly in the context of their visits to ports [fr

  16. Guide to ship sanitation

    National Research Council Canada - National Science Library

    2011-01-01

    "The third edition of the Guide to Ship Sanitation presents the public health significance of ships in terms of disease and highlights the importance of applying appropriate control measures"--Back cover...

  17. An Arsenal Ship Design

    Science.gov (United States)

    1996-12-01

    and dental services underway. 5. Disbursing Office In addition to being paperless, the Arsenal Ship is also a cashless ship. The ship does not have...1, Society of Naval Architects & Marine Engineers, 1988. 57. Gillmer, T.C., and Johnson, B., Introduction to Naval Architecture, Naval Institute...Vending Machines Cashless ship (i.e. Use MARC card) O 10.4.3 Provide Crew Entertainment O 10.4.3.1 Closed Circuit TV Shipboard Multi-Media Resource Center

  18. Tritium behavior in an aquatic ecosystem

    International Nuclear Information System (INIS)

    Komatsu, K.

    1982-01-01

    Tritium behavior in aquatic organisms through a model food chain was investigated. In this model food chain, tritium in water reaches bacteria or Japanese killifish via diatoms and brine shrimps. Tritium accumulation in these organisms as organic bound form was expressed as the R value which is defined as the ratio of tritium specific activity in lyophilized organisms (μCi/gH) to that in water (μCi/gH). The maximum R values were 0.5 in diatoms: Chaetoceros gracilis, 0.2 in bacteria: Escherichia coli, 0.5 in brine shrimps: Artemia salina, and 0.32 in Japanese killifish: Oryzias latipes under the growing condition in which tritium accumulation was due to tritium in tritiated water and not tritiated foods. Brine shrimps and Japanese killifish were grown from larve to adult in tritiated sea water and were fed on tritiated foods (model food chain). Their R values were 0.70 and 0.67, respectively. Bacteria, which grew in tritiated water by adding the hydrolysate of tritiated brine shrimps, showed a maximum R value at 0.32. The R values of each organ of Japanese killifish and of DNA and the nucleotides purified from brine shrimps growing in tritiated water with or without tritiated food were measured to estimate the tritium distribution in the body or various molecules of the organisms. These results did not indicate concentration of tritium in specific organs or compounds. The changes of specific activity of tritium in these organisms were measured when they were transferred to non-tritiated water. These retention of tritium was not only different among the tissues but also depended on whether or not the organisms were reared with tritiated foods. (author)

  19. Tritium decay helium-3 effects in tungsten

    Directory of Open Access Journals (Sweden)

    M. Shimada

    2017-08-01

    Full Text Available Tritium (T implanted by plasmas diffuses into bulk material, especially rapidly at elevated temperatures, and becomes trapped in neutron radiation-induced defects in materials that act as trapping sites for the tritium. The trapped tritium atoms will decay to produce helium-3 (3He atoms at a half-life of 12.3 years. 3He has a large cross section for absorbing thermal neutrons, which after absorbing a neutron produces hydrogen (H and tritium ions with a combined kinetic energy of 0.76 MeV through the 3He(n,HT nuclear reaction. The purpose of this paper is to quantify the 3He produced in tungsten by tritium decay compared to the neutron-induced helium-4 (4He produced in tungsten. This is important given the fact that helium in materials not only creates microstructural damage in the bulk of the material but alters surface morphology of the material effecting plasma-surface interaction process (e.g. material evolution, erosion and tritium behavior of plasma-facing component materials. Effects of tritium decay 3He in tungsten are investigated here with a simple model that predicts quantity of 3He produced in a fusion DEMO FW based on a neutron energy spectrum found in literature. This study reveals that: (1 helium-3 concentration was equilibrated to ∼6% of initial/trapped tritium concentration, (2 tritium concentration remained approximately constant (94% of initial tritium concentration, and (3 displacement damage from 3He(n,HT nuclear reaction became >1 dpa/year in DEMO FW.

  20. Experimental pilot plant for tritium and deuterium separation. Results and future prospects

    International Nuclear Information System (INIS)

    Cristescu, I.; Bornea, Anisia; Brad, S.; Cristescu, Loana; Sofalca, N.; Stefan, L.; Zamfirache, M.

    1999-01-01

    Experimental Pilot Plant for tritium and deuterium separation at Rm. Valcea has the aim of establishing the technological data required in designing and operating an industrial plant for heavy water detritiation as function of tritium containment in heavy water. The technology developed at Rm. Valcea is based upon catalytic isotopic exchange in heavy water-deuterium, followed by cryogenic distillation of a mixture between molecular species of hydrogen and its isotopes. The experimental pilot plant contains five modules that can work independently or coupled between them: a - in the isotopic exchange module the transfer of deuterium and tritium is made from heavy water in a hydrogen flow by catalytic isotope exchange. The mixed catalytic packing was made and tested by means of the catalyst Pt/C/PTFE and B7 packing; b - in the preliminary purification module the hydrogen purification is realized by removing oxygen and water, which can affect the good functioning of the cryogenic distillation module; c - in the cryogenic distillation module the mixture of hydrogen isotopes is separated. The deuterium concentration at the bottom D/(D+H) of the column is up to 99.9%. The column is filed with ordered package and the condenser temperature is 22 K.; d - in the catalytic burn module, deuterated hydrogen is catalytically combined with oxygen and heavy water results with a concentration D/(D+H) of 19.9%. The hydrogen burning takes place on mixed catalytic package with 10% hydrophobic catalyst of Pt/C/PTFE type and 90% package; e - the module of water isotopic distillation under vacuum allows heavy water concentration rise from 49.9% to 99.8% on ordered package. In the paper we present the main steps of starting the operation of this experimental pilot plant and analyze the performances with respect to tritium and deuterium separation. (authors)

  1. TRITIUM RESERVOIR STRUCTURAL PERFORMANCE PREDICTION

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P.S.; Morgan, M.J

    2005-11-10

    The burst test is used to assess the material performance of tritium reservoirs in the surveillance program in which reservoirs have been in service for extended periods of time. A materials system model and finite element procedure were developed under a Savannah River Site Plant-Directed Research and Development (PDRD) program to predict the structural response under a full range of loading and aged material conditions of the reservoir. The results show that the predicted burst pressure and volume ductility are in good agreement with the actual burst test results for the unexposed units. The material tensile properties used in the calculations were obtained from a curved tensile specimen harvested from a companion reservoir by Electric Discharge Machining (EDM). In the absence of exposed and aged material tensile data, literature data were used for demonstrating the methodology in terms of the helium-3 concentration in the metal and the depth of penetration in the reservoir sidewall. It can be shown that the volume ductility decreases significantly with the presence of tritium and its decay product, helium-3, in the metal, as was observed in the laboratory-controlled burst tests. The model and analytical procedure provides a predictive tool for reservoir structural integrity under aging conditions. It is recommended that benchmark tests and analysis for aged materials be performed. The methodology can be augmented to predict performance for reservoir with flaws.

  2. Issues Associated with Tritium Legacy Materials

    International Nuclear Information System (INIS)

    Mills, Michael

    2008-01-01

    This paper highlights some of the issues associated with the treatment of legacy materials linked to research into tritium over many years and also of materials used to contain or store tritium. The aim of the work is to recover tritium where practicable, and to leave the residual materials passively safe, either for disposal or for continued storage. A number of materials are currently stored at AWE which either contain tritium or have been used in tritium processing. It is essential that these materials are characterised such that a strategy may be developed for their safe stewardship, and ultimately for their treatment and disposal. Treatment processes for such materials are determined by the application of best practicable means (BPM) studies in accordance with the requirements of the Environment Agency of England and Wales. Clearly, it is necessary to understand the objectives of legacy material treatment / processing and the technical options available before a definitive BPM study is implemented. The majority of tritium legacy materials with which we are concerned originate from the decommissioning of a facility that was operational from the late 1950's through to the late 1990's when, on post-operative clear-out (POCO), the entire removable and transportable tritium inventory was moved to new, purpose built facilities. One of the principle tasks to be undertaken in the new facilities is the treatment of the legacy materials to recover tritium wherever practicable, and render the residual materials passively safe for disposal or continued storage. Where tritium recovery was not reasonably or technically feasible, then a means to assure continued safe storage was to be devised and implemented. The legacy materials are in the following forms: - Uranium beds which may or may not contain adsorbed tritium gas; - Tritium gas stored in containers; - Tritide targets for neutron generation; - Tritides of a broad spectrum of metals manufactured for research / long

  3. PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Smith, P.; Sheetz, S.

    2013-09-30

    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 of 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

  4. Packaging microservices

    DEFF Research Database (Denmark)

    Montesi, Fabrizio; Thrane, Dan Sebastian

    2017-01-01

    We describe a first proposal for a new packaging system for microservices based on the Jolie programming language, called the Jolie Package Manager (JPM). Its main features revolve around service interfaces, which make the functionalities that a service provides and depends on explicit. For the f......We describe a first proposal for a new packaging system for microservices based on the Jolie programming language, called the Jolie Package Manager (JPM). Its main features revolve around service interfaces, which make the functionalities that a service provides and depends on explicit...

  5. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2007-12-13

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  6. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2006-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  7. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

  8. Optimization of tritium management within the ITER project

    International Nuclear Information System (INIS)

    Cortes, P.; Elbez-Uzan, J.; Glugla, M.; Rosanvallon, S.; Ciattaglia, S.; Iseli, M.; Rodriguez-Rodrigo, L.

    2009-01-01

    The authors describe the tritium cycle existing within the ITER project and which has been considered since its beginning. They indicate how confinement systems ensure tritium confinement, how tritium is recovered and processed. They indicate the different tritium management optimization ways which have been identified and integrated into the ITER design

  9. 10 CFR 39.55 - Tritium neutron generator target sources.

    Science.gov (United States)

    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. Tritium distributing in stainless steel determined by chemical etchin

    International Nuclear Information System (INIS)

    Xiong Yifu; Luo Deli; Chen Changan; Chen Shicun; Jing Wenyong

    2009-01-01

    The depth distribution of tritium in stainless steel was measured by chemical etching. The results show that the method can more quantitatively evaluate the tritium distributing in stainless steel. The maximum amount of tritium which distributed in crystal lattice of stainless steel is limitted by its solubility at room temperature. The other form of tritium in stainless steel is gaseous tritium that are trapped by defects, impurities, fractures, etc. within it. The gaseous tritium is several times more than the solid-dissolved tritium. (authors)

  11. Tritium monitor with improved gamma-ray discrimination

    Science.gov (United States)

    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.

  12. Tritium Decay Helium-3 Effects in Tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Merrill, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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 microstructural 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

  13. Tritium Room Air Monitor Operating Experience Review

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader; B. J. Denny

    2008-09-01

    Monitoring the breathing air in tritium facility rooms for airborne tritium is a radiological safety requirement and a best practice for personnel safety. Besides audible alarms for room evacuation, these monitors often send signals for process shutdown, ventilation isolation, and cleanup system actuation to mitigate releases and prevent tritium spread to the environment. Therefore, these monitors are important not only to personnel safety but also to public safety and environmental protection. This paper presents an operating experience review of tritium monitor performance on demand during small (1 mCi to 1 Ci) operational releases, and intentional airborne inroom tritium release tests. The tritium tests provide monitor operation data to allow calculation of a statistical estimate for the reliability of monitors annunciating in actual tritium gas airborne release situations. The data show a failure to operate rate of 3.5E-06/monitor-hr with an upper bound of 4.7E-06, a failure to alarm on demand rate of 1.4E-02/demand with an upper bound of 4.4E-02, and a spurious alarm rate of 0.1 to 0.2/monitor-yr.

  14. Thermal release of tritium from SS316

    International Nuclear Information System (INIS)

    Torikai, Y.; Naoe, S.; Penzhorn, R.D.; Akaishi, K.; Watanabe, K.; Matsuyama, M.

    2007-01-01

    In an effort to improve current understanding of the mechanisms controlling the long-term release of tritium incorporated thermally into stainless steel SS316 and to develop reliable as well as economically feasible techniques for the conditioning of tritium-containing metallic wastes, a systematic investigation is underway in Toyama under carefully controlled conditions. The release rate of tritium from SS316 at ambient pressure was determined experimentally in a flow system at several constant temperatures within the range 287-573 K for rather extended periods of time. Under these conditions HTO was found to constitute by far the most important tritium-containing species being released, i.e. approx. 99 %. Much data has accumulated in recent years with a variety of specimens, i.e. type of stainless steel and specimen dimension, loaded with tritium under different pressure and temperature conditions. Dynamic behavior of long-term tritium release has been successfully modeled using a onedimensional diffusion equation and assuming that the release rate is governed by the tritium flux at the metal surface boundary. The implications of the results for interim storage and thermal conditioning of stainless steel waste will be discussed. (orig.)

  15. Cernavoda NPP - Management of internal tritium exposures

    International Nuclear Information System (INIS)

    Chitu, Catalina; Popescu, Ion; Samson, Liliana; Simionov, Vasile

    2010-01-01

    Full text: During normal operation of a CANDU nuclear power plant significant tritium quantities are generated. Through design solutions that have been implemented we manage to control the tritium losses from the reactor systems and keep them as low as possible. Special dryers are designed and are used to remove moisture from different ventilation systems of a CANDU reactor in order to maintain tritium in air concentration and gaseous tritium emissions below the limits established by the national authorities. Vapor Recovery System is designed to control tritium in air concentration and to recover heavy water loss from PHT and Moderator Systems and to control the air circulation, providing atmosphere separation between different areas of the Reactor Building. Cernavoda NPP developed a special strategy in order to control workers' internal exposures to tritium and dedicated programs are running to implement this strategy: improvement of radiation protection procedures; increasing equipment performances; leakages prevention through maintenance program; finalization of the de-tritiation facility. This paper presents the evolution of workers tritium exposure and emphasizes the results of the ALARA policy promoted by Cernavoda NPP management. (authors)

  16. Risks of tritium and their mitigation

    International Nuclear Information System (INIS)

    Ichimasa, Y.; Shiba, H.; Ichimasa, M.; Chikuuti, M.; Akita, Y.

    1992-01-01

    In this study, the effects of an antibacterial drug, norfloxacin, and an antibiotic, clindamycin, on in vivo oxidation of tritium gas in rats were investigated. Wistar strain male rats were used. They were provided with a standard diet, water ad libitum, and maintained in glass metabolic cages of approximately 20 liters capacity. The air flow and temperature were controlled. To investigate the availability of norfloxacin and clindamycin on the inhibition effects of the oxidation of tritium gas, two types of the experiments were conducted one was that, before the exposure to tritium gas for 2 hours, norfloxacin or clindamycin was administrated to rats three times a day for 4 days, and the other was administration of a drug after tritium gas exposure. After the exposure to tritium gas, blood, the liver, urine and feces samples were collected from rats and the radioactivity of them was determined after combustion using a sample oxidizer. In the case of norfloxacin, tritium concentration in rat body decreased one fifth of that in non-treated rats. On the other hand, administration of clindamycin shortened the biological half-life of tritium in urine to three fifth of that of non-treated rats. (author)

  17. Quick management of accidental tritium exposure cases.

    Science.gov (United States)

    Singh, Vishwanath P; Badiger, N M; Managanvi, S S; Bhat, H R

    2012-07-01

    Removal half-life (RHL) of tritium is one of the best means for optimising medical treatment, reduction of committed effective dose (CED) and quick/easy handling of a large group of workers for medical treatment reference. The removal of tritium from the body depends on age, temperature, relative humidity and daily rainfall; so tritium removal rate, its follow-up and proper data analysis and recording are the best techniques for management of accidental acute tritium exposed cases. The decision of referring for medical treatment or medical intervention (MI) would be based on workers' tritium RHL history taken from their bodies at the facilities. The workers with tritium intake up to 1 ALI shall not be considered for medical treatment as it is a derived limit of annual total effective dose. The short-term MI may be considered for tritium intake of 1-10 ALI; however, if the results show intake ≥100 ALI, extended strong medical/therapeutic intervention may be recommended based on the severity of exposure for maximum CED reduction requirements and annual total effective dose limit. The methodology is very useful for pressurized heavy water reactors (PHWRs) which are mainly operated by Canada and India and future fusion reactor technologies. Proper management will optimise the cases for medical treatment and enhance public acceptance of nuclear fission and fusion reactor technologies.

  18. Isotopic fractionation of tritium in biological systems.

    Science.gov (United States)

    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.

  19. Quick management of accidental tritium exposure cases

    International Nuclear Information System (INIS)

    Singh, V. P.; Badiger, N. M.; Managanvi, S. S.; Bhat, H. R.

    2008-01-01

    Removal half-life (RHL) of tritium is one of the best means for optimising medical treatment, reduction of committed effective dose (CED) and quick/easy handling of a large group of workers for medical treatment reference. The removal of tritium from the body depends on age, temperature, relative humidity and daily rainfall; so tritium removal rate, its follow-up and proper data analysis and recording are the best techniques for management of accidental acute tritium exposed cases. The decision of referring for medical treatment or medical intervention (MI) would be based on workers' tritium RHL history taken from their bodies at the facilities. The workers with tritium intake up to 1 ALI shall not be considered for medical treatment as it is a derived limit of annual total effective dose. The short-term MI may be considered for tritium intake of 1-10 ALI; however, if the results show intake ≥100 ALI, extended strong medical/therapeutic intervention may be recommended based on the severity of exposure for maximum CED reduction requirements and annual total effective dose limit. The methodology is very useful for pressurized heavy water reactors (PHWRs) which are mainly operated by Canada and India and future fusion reactor technologies. Proper management will optimise the cases for medical treatment and enhance public acceptance of nuclear fission and fusion reactor technologies. (authors)

  20. Conceptual design of tritium accountancy system for LLCB TBM

    International Nuclear Information System (INIS)

    Patel, Rudreksh; Sircar, Amit

    2017-01-01

    Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) will be tested in ITER for performance evaluation of high grade of heat extraction and tritium breeding. The bred tritium in the breeder materials is extracted and recovered by Tritium Extraction System (TES), whereas tritium permeated from breeder materials to helium coolants, viz., primary coolant and secondary coolant, is recovered by Coolant Purification System (CPS). This recovered tritium has to be accounted before transferring it to tritium plant (i.e., ITER inner fuel). This tritium accountancy is performed by Tritium Accountancy System (TAS). In addition to tritium accountancy, TAS also provides necessary data for the validation of design and modelling tools.In this work, we have presented conceptual design of TAS. It also describes operational philosophy, process parameters, process flow diagram, and interface details with ITER tritium plant. (author)

  1. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua

    2003-01-01

    A summary of the tritium system design studies for the engineering outline design of a fusion experimental breeder (FEB-E) is presented. This paper is divided into three sections. In first section, the geometry, loading features and tritium concentrations in liquid lithium of tritium breeding zones of blanket are described. The tritium flow chart corresponding to the tritium fuel cycle system has been constructed, and the inventories in ten subsystems are calculated using SWITRIM code in section 2. Results show that the necessary initial tritium storage to start up FEB-E with fusion power of 143 MW is about 319 g. In final section, the tritium leakage issues under different operation circumstances have been analyzed. It was found that the potential danger of tritium leakage could be resulted from the exhausted gas of the diverter system. It is important to elevate the tritium burnup fraction and reduce the tritium throughput. (authors)

  2. Tritium extraction technologies and DEMO requirements

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D., E-mail: david.demange@kit.edu [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Antunes, R.; Borisevich, O.; Frances, L. [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rapisarda, D. [Laboratorio Nacional de Fusión, EURATOM-CIEMAT, 28040 Madrid (Spain); Santucci, A. [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy)

    2016-11-01

    Highlights: • We detail the R&D plan for tritium technology of the European DEMO breeding blanket. • We study advanced and efficient extraction techniques to improve tritium management. • We consider inorganic membranes and catalytic membrane reactor for solid blankets. • We consider permeator against vacuum and vacuum sieve tray for liquid blankets. - Abstract: The conceptual design of the tritium extraction system (TES) for the European DEMO reactor is worked out in parallel for four different breeding blankets (BB) retained by EUROfusion. The TES design has to be tackled in an integrated manner optimizing the synergy with the directly interfacing inner fuel cycle, while minimizing the tritium permeation into the coolant. Considering DEMO requirements, it is most likely that only advanced technologies will be suitable for the tritium extraction systems of the BB. This paper overviews the European work programme for R&D on tritium technology for the DEMO BB, summaries the general first outcomes, and details the specific and comprehensive R&D program to study experimentally immature but promising technologies such as vacuum sieve tray or permeator against vacuum for tritium extraction from PbLi, and advanced inorganic membranes and catalytic membrane reactor for tritium extraction from He. These techniques are simple, fully continuous, likely compact with contained energy consumption. Several European Laboratories are joining their efforts to deploy several new experimental setups to accommodate the tests campaigns that will cover small scale experiments with tritium and inactive medium scale tests so as to improve the technology readiness level of these advanced processes.

  3. Optimization of tritium management within the ITER project; Optimisation de la gestion du tritium dans le projet ITER

    Energy Technology Data Exchange (ETDEWEB)

    Cortes, P.; Elbez-Uzan, J.; Glugla, M.; Rosanvallon, S.; Ciattaglia, S.; Iseli, M.; Rodriguez-Rodrigo, L. [ITER Organization, CS 9 0046, 13067 St Paul lez Durance Cedex (France)

    2009-07-01

    The authors describe the tritium cycle existing within the ITER project and which has been considered since its beginning. They indicate how confinement systems ensure tritium confinement, how tritium is recovered and processed. They indicate the different tritium management optimization ways which have been identified and integrated into the ITER design

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

    International Nuclear Information System (INIS)

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

    1988-01-01

    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 useful agents for labeling peptides and proteins

  5. Conversion of tritium gas to tritiated water

    International Nuclear Information System (INIS)

    Papagiannakopoulos, P.J.; Easterly, C.E.

    1979-05-01

    The mechanisms of conversion of tritium gas to tritiated water (HTO) have been examined for several tritium gaseous mixtures. The physical and chemical processes involved in the self-radiolysis of such mixtures have been analyzed and the kinetics involved in the formation of HTO has been presented. It has been determined that the formation of the H and/or OH free radicals, as intermediate species, are of significance in the formation of HTO. Therefore, the problem of reducing the rate of formation of tritiated water in a mixture of gaseous tritium with atmospheric components is one of finding an effective scavenger for the H and/or OH free radicals

  6. Preparation of honey sample for tritium monitoring

    International Nuclear Information System (INIS)

    Chen Bingru; Wang Chenlian; Wang Weihua

    1989-01-01

    The method of preparation of honey sample for tritium monitoring was described. The equipments consist of an air and honey supply system, a quartz combustor with CM-type monolithic combustion catalyst and a condensation system. In the equipments, honey sample was converted into cooling water by the distilling, cracking and carbonizing procedures for tritium counting. The recovery ratio is 99.0 ± 4.5 percent for tritiated water and 96.0 ± 2.0 for tritiated organic compounds. It is a feasible preparing method for the total tritium monitoring in honey sample

  7. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

    Tritium release from the candidate ceramic materials, Li 2 O, LiA10 2 , Li 2 SiO 3 , Li 4 SiO 4 and Li 2 ZrO 3 , is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

  8. Electrolytic gettering of tritium from air

    International Nuclear Information System (INIS)

    Souers, P.C.; Tsugawa, R.T.; Stevens, C.G.

    1983-01-01

    We have removed 90% of 1 part-per-million tritium gas in air of 25% to 35% humidity by the dc electrical action of the solid proton electrolyte hydrogen uranyl phosphate (HUP). Gettering takes 5 to 24 hours for a 1 cm 2 HUP disc at 2 to 4 V in a static, 1200 cc gas volume. Hydrogen gas may be used to flush captured tritium through the HUP. Liquid water leaches out the tritium but water vapor is ineffective. This technique promises an alternative to the conventional catalyst/zeolite method

  9. Management of tritium contaminated wastes national strategies and practices at some European countries, USA and Canada

    International Nuclear Information System (INIS)

    Mannone, F.

    1992-01-01

    The European Tritium Handling Experiment Laboratory (ETHEL) is the Commission of European Communities facility designed for handling multigram quantities of tritium for safety inherent R and D purposes. Tritium contamined wastes in gaseous, liquid and solid forms will be generated in ETHEL during the experiments as well as during the maintenance operations. All such wastes must be adequately managed under the safest operating conditions to minimize the releases of tritium to the environment and the consequent radiological risks to workers and general population. This safety requirement can be met by carefully defining strategies and practices to be applied for the safe management of these wastes. To this end an adequate background information must be collected which is the intent of this report. Through an exhaustive literature survey current strategies and practices applied in Europe, USA and Canada for managing tritiated wastes from specific tritium handling laboratories and plant have been assessed. For some countries, where only tritium bearing wastes simultaneously contaminated with nuclear fission products are generated, the attention has been focused on the strategies and practices currently applied for managing fission wastes. Operational criteria for waste collection, sorting, classification, conditioning and packaging as well as acceptance criteria for their storage or disposal have been identified. Waste storage or disposal options already applied in various countries or still being investigated in terms of safety have also been considered. Even if the radwaste management strategy is submitted to a nearly continuing process of review, some general comments resulting from the assessment of the present waste management scenario are presented. 60 refs., 16 figs., 13 tabs

  10. Tritium containing polymers having a polymer backbone substantially void of tritium

    Science.gov (United States)

    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.

  11. The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

    International Nuclear Information System (INIS)

    Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

    2003-01-01

    Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project

  12. Kinetic behaviour of tritium water in the simulated paddy

    International Nuclear Information System (INIS)

    Wang Shouxiang; Chen Chuanqun; Zhang Yongxi; Sun Zhiming; Huang Dan; Hu Bingmin

    1993-02-01

    The dynamic process of movement and transportation for tritium water in the simulated paddy have been investigated. The results showed that after the paddy water was filled the tritium water not only was transferred to the components of the system but also was rapidly dissipated to the outside of the system. The states of tritium in the rice and soil were tritium of free water and bound tritium. The tritium concentration in the early paddy water and late paddy water was monotonously decreasing with the time increasing. The tritium concentration of free water in the overground part of early rice was increasing at begin, and reached to maximum at the 8th day, then went down. The tritium concentration in bound tritium was monotonously increasing, and at the 22th day it was higher than free water tritium. The tritium concentration in early rice root was lower than in the overground part. The tritium concentrations of free water tritium and bound tritium in the overground part and root of late rice were decreasing at the sampling period. The total tritium in the soil of early rice could reach the maximum value within one day and then it decreased, but in the soil of late rice it was monotonously decreasing with the time

  13. Tritium metabolism in cow's milk after administration of tritiated water and of organically bound tritium

    International Nuclear Information System (INIS)

    Hoek, J. van den

    1982-01-01

    Tritium was administered as THO and as organically bound tritium (OBT) to lactating cows. Urine and milk samples were collected and analyzed for tritium content. Plateau levels in milk water and in milk fat, lactose and casein were reached in about 20 days after feeding either THO or OBT. Comparison of the specific activity (pCi 3 H/g H) of the various milk constituents with the specific activity of the body water showed that, after administration of THO, the highest tritium incorporation occurred in lactose (0.58), followed by milk fat (0.36) and casein (0.27). Tritium incorporation in milk dry matter (0.45) is considerably higher than in most tissue components of several mammalian species after continuous ingestion of THO as reported in the literature. After feeding OBT, the highest tritium incorporation occurred in milk fat and to a lesser extent in casein. Tritium levels in lactose were surprisingly low and the reason for this is not clear. They were similar to those in milk water. Tritium levels in milk and urine water showed systematic differences during administration of OBT and after this was stopped. There was more tritium in milk water until the last day of OBT feeding and this situation was reversed after this. (author)

  14. Overview of tritium processing development at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1986-01-01

    The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been operating with tritium since June 1984. Presently there are some 50 g of tritium in the main processing loop. This 50 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation system and to integrate the fuel cleanup system into the main fuel processing loop. In January 1986 two major experiments were conducted. During these experiments the fuel cleanup system was integrated, through the transfer pumping system, with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems leaves only the main vacuum system to be integrated into the TSTA fuel processing loop. In September 1986 another major tritium experiment was performed in which the integrated loop was operated, the tritium inventory increased to 50 g and additional measurements on the performance of the distillation system were taken. In the period June 1984 through September 1986 the TSTA system has processed well over 10 8 Ci of tritium. Total tritium emissions to the environment over this period have been less than 15 Ci. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flows have been uncovered

  15. Hazardous Material Packaging and Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Hypes, Philip A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-02-04

    This is a student training course. Some course objectives are to: recognize and use standard international and US customary units to describe activities and exposure rates associated with radioactive material; determine whether a quantity of a single radionuclide meets the definition of a class 7 (radioactive) material; determine, for a given single radionuclide, the shipping quantity activity limits per 49 Code of Federal Regulations (CFR) 173.435; determine the appropriate radioactive material hazard class proper shipping name for a given material; determine when a single radionuclide meets the DOT definition of a hazardous substance; determine the appropriate packaging required for a given radioactive material; identify the markings to be placed on a package of radioactive material; determine the label(s) to apply to a given radioactive material package; identify the entry requirements for radioactive material labels; determine the proper placement for radioactive material label(s); identify the shipping paper entry requirements for radioactive material; select the appropriate placards for a given radioactive material shipment or vehicle load; and identify allowable transport limits and unacceptable transport conditions for radioactive material.

  16. Cleanup Verification Package for the 118-B-1, 105-B Solid Waste Burial Ground

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    This cleanup verification package documents completion of remedial action, sampling activities, and compliance criteria for the 118-B-1, 105-B Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-B Reactor and P-10 Tritium Separation Project and also received waste from the 105-N Reactor. The burial ground received reactor hardware, process piping and tubing, fuel spacers, glassware, electrical components, tritium process wastes, soft wastes and other miscellaneous debris

  17. Safety Evaluation for Packaging (onsite) T Plant Canyon Items

    International Nuclear Information System (INIS)

    OBRIEN, J.H.

    2000-01-01

    This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments

  18. Safety Evaluation for Packaging (onsite) T Plant Canyon Items

    Energy Technology Data Exchange (ETDEWEB)

    OBRIEN, J.H.

    2000-07-14

    This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments.

  19. Methods of tritium recovery from molten lithium

    International Nuclear Information System (INIS)

    Farookhi, R.; Rogers, J.E.

    1968-01-01

    It is important to keep the tritium inventory in a blanket of a thermonuclear reactor at a low level both to eliminate possible hydriding of structural components and to reduce inventory cost. Removing the tritium from a lithium blanket by fractional distillation, flash vaporization, and fractional crystallization was investigated. No definitive data are available either on the vapor-liquid equilibrium between lithium and tritium at low T 2 concentrations, or on the rate of formation and decomposition of lithium tritide. The final distinction between the recovery systems discussed in this report will depend on such data, but presently distillation appears to be the best alternate to the diffusion scheme proposed by A.P. Fraas. The capital cost of equipment necessary to remove tritium by distillation appears to be greater than 10 million dollars for a 5000 MW system, whereas the capital cost associated with the diffusion process has been estimated to be 4 million dollars

  20. Technologies for immobilization and disposal of tritium

    International Nuclear Information System (INIS)

    Coppari, N.R.

    1996-01-01

    This study was done within a program one of whose objectives was to know the state of the technology development for tritium separation in the moderator circuit at HWR and to define the possible technologies to be applied to the Argentine nuclear power plants. Within this framework the strategies adopted by each country and the available technologies for a safe disposal of tritium, not only in its gaseous state tritium but also as tritiated water were analyzed. It is considered that if the selected separation method is such that the tritium is in its gaseous state, the hydride formation for long periods of immobilization should be studied. whereas if it were triated water immobilization should be studied to choose the technology between cementation and drying agents, in both cases the final disposal site will have to be selected. (author). 8 refs

  1. Survey of pumps for tritium gas

    International Nuclear Information System (INIS)

    Dowell, T.M.

    1983-05-01

    This report considers many different types of pumps for their possible use in pumping tritium gas in the low, intermediate and high vacuum ranges. No one type of pump is suitable for use over the wide range of pumping pressure required in a typical pumping system. The favoured components for such a system are: bellows pump (low vacuum); orbiting scroll pump (intermediate vacuum); magnetically suspended turbomolecular pump (high vacuum); cryopump (high vacuum). Other pumps which should be considered for possible future development are: mound modified vane pump; SRTI wobble pump; roots pump with canned motor. It is proposed that a study be made of a future tritium pumping system in a Canadian tritium facility, e.g. a tritium laboratory

  2. Tritium proof-of-principle injector experiment

    International Nuclear Information System (INIS)

    Fisher, P.W.; Milora, S.L.; Combs, S.K.; Carlson, R.V.; Coffin, D.O.

    1988-01-01

    The Tritium Proof-of-Principle (TPOP) pellet injector was designed and built by Oak Ridge National Laboratory (ORNL) to evaluate the production and acceleration of tritium pellets for fueling future fision reactors. The injector uses the pipe-gun concept to form pellets directly in a short liquid-helium-cooled section of the barrel. Pellets are accelerated by using high-pressure hydrogen supplied from a fast solenoid valve. A versatile, tritium-compatible gas-handling system provides all of the functions needed to operate the gun, including feed gas pressure control and flow control, plus helium separation and preparation of mixtures. These systems are contained in a glovebox for secondary containment of tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). 18 refs., 3 figs

  3. Tritium Measurements in Slovenia - Chronology Till 2004

    International Nuclear Information System (INIS)

    Logar, Jasmina Kozar; Vaupotic, Janja; Kobal, Ivan

    2005-01-01

    Almost all the analyses of tritium in Slovenia have been performed by the tritium laboratory at the Jozef Stefan Institute. Nearly 90 % of its measurements have been covered by two national programs, both approved by the Slovenian Nuclear Safety Administration: the radioactive monitoring program in the environs of Krsko Nuclear Power Plant (KNPP) and the program of global radioactive contamination monitoring in the environment. These programs include samples of groundwaters, surface waters, precipitation and drinking waters, as well as liquid and gaseous effluents from KNPP. Tritium was determined in some research projects and in hydrological studies of thermal waters, groundwater and coalmine waters. Tritium in the Karst region was mapped as well as the springs of entire territory of Slovenia. Around 5500 samples have been analyzed up to 2004

  4. Tritium handling facility at KMS Fusion Inc

    International Nuclear Information System (INIS)

    Bowman, C.C.; Vis, V.A.

    1990-01-01

    The tritium facility at KMS Fusion, Inc. supports the inertial confinement fusion research program. The main function of the facility is to fill glass and polymer Microshell (TM) capsules (small fuel containers) to a maximum pressure of 100 atm with tritium (T 2 ) or deuterium--tritium (DT). The recent upgrade of the facility allows us to fill Microshell capsules to a maximum pressure of 200 atm. A second fill port allows us to run long term fills of Macroshell (TM) capsules (large fuel containers) concurrently. The principle processes of the system are: (1) storage of the tritium as a uranium hydride; (2) pressure intensification using cryogenics; and (3) filling of the shells by permeation at elevated temperatures. The design of the facility was centered around a NRC license limit of 6000 Ci

  5. Tritium Inventory in ARIES-AT

    International Nuclear Information System (INIS)

    Longhurst, Glen R.

    2001-01-01

    This report documents an investigation into the tritium inventory expected in the ARIES-AT fusion reactor. ARIES-AT features silicon carbide fibers in a silicon carbide matrix as its primary construction. It uses the same fusion power core as the previous ARIES-RS. Based on experimental results of several researchers, consideration was given to swelling, sputtering, film coatings, erosion, and implantation. Estimates were made of tritium inventory using the TMAP4 code. About 700 g of tritium may be expected in the machine, two thirds of which would reside in the first wall. Under assumed accident conditions that involve first wall temperatures up to 1000 C, evolution of retained tritium may be expected to vary from 0.8 to nearly 40 percent depending on the temperature of the first wall

  6. Investigation of tritium in the aquatic environment

    International Nuclear Information System (INIS)

    Cohen, L.K.

    1977-01-01

    The behavior, cycling and distribution of tritium in an aquatic ecosystem was studied in the field and in the laboratory from 1969 through 1971. Field studies were conducted in the Hudson River Estuary, encompassing a 30 mile region centered about the Indian Point Nuclear Plant. Samples of water, bottom sediment, rooted emergent aquatic plants, fish, and precipitation were collected over a year and a half period from more than 15 locations. Specialized equipment and systems were built to combust and freeze-dry aquatic media to remove and recover the loose water and convert the bound tritium into an aqueous form. An electrolysis system was set up to enrich the tritium concentrations in the aqueous samples to improve the analytical sensitivity. Liquid scintillation techniques were refined to measure the tritium activity in the samples. Over 300 samples were analyzed during the course of the study

  7. FDMH - The tritium model in RODOS

    International Nuclear Information System (INIS)

    Galeriu, D.; Mateescu, G.; Melintescu, A.; Turcanu, C.; Raskob, W.

    2000-01-01

    Under the auspices of its RTD (Research and Technological Development) Framework Programmes, the European Commission has supported the development of the RODOS (Real-time On-line DecisiOn Support) system for off-site emergency management. The project started in 1989 focusing on PWR/LWR type accidents and using experience from the Chernobyl accident. In 1996 it was realised that tritium should be included in the list of radionuclides, as large tritium sources exist in Europe and to allow a potential expansion of the RODOS system for application on future fusion reactor accidents. The National Institute for Physics and Nuclear Engineering (IFIN-HH) in Romania - in close co-operation with the Research Centre Karlsruhe (FZK) - was charged to develop the tritium module, based on previous experience in environmental tritium modelling and the operation of CANDU reactor-based NPP in Romania (with potential tritium accidents). Tritium, being an isotope of hydrogen, is incorporated immediately in the life cycle and its transport into the biosphere differs considerably from other radionuclides treated by the RODOS system. Concentrations in the individual compartments may change very rapidly (hours) under varying environmental conditions and conversion to organic forms by biochemical and metabolic processes takes place in plants and animals. Consequently, the tritium code in RODOS was developed as a separate module and harmonisation in data sets and interfaces with other food chain modules integrated in RODOS was ensured. Presently, the tritium module - FDMH- is integrated and documented in the RODOS system, delivering time dependent tritium concentration (as tritiated water or organically bound tritium) in plant and animal products, inhalation dose and ingestion dose for various groups of population, after an accident emitting tritiated water and for up to 2520 locations around the source. FDMH incorporates many improved techniques in radiological assessment and makes

  8. Review of tritium behavior in HTGR systems

    International Nuclear Information System (INIS)

    Gainey, B.W.

    1976-01-01

    The available experimental evidence from laboratory and reactor studies pertaining to tritium production, capture, release, and transport within an HTGR leading to release to the environment is reviewed. Possible mechanisms for release, capture, and transport are considered and a simple model was used to calculate the expected tritium release from HTGRs. Comparison with Federal regulations governing tritium release confirm that expected HTGR releases will be well within the allowable release limits. Releases from HTGRs are expected to be somewhat less than from LWRs based on the published LWR operating data. Areas of research deserving further study are defined but it is concluded that a tritium surveillance at Fort St. Vrain is the most immediate need

  9. Calibrations of a tritium extraction facility

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Oliver, B.M.; Farrar, H. IV.

    1983-01-01

    A tritium extraction facility has been built for the purpose of measuring the absolute tritium concentration in neutron-irradiated lithium metal samples. Two independent calibration procedures have been used to determine what fraction, if any, of tritium is lost during the extraction process. The first procedure compares independently measured 4 He and 3 H concentrations from the 6 Li(n,α)T reaction. The second procedure compared measured 6 Li(n,α)T/ 197 Au (n,γ) 198 Au thermal neutron reaction rate ratios with those obtained from Monte Carlo calculations using well-known cross sections. Both calibration methods show that within experimental errors (approx. 1.5%) no tritium is lost during the extraction process

  10. Tritium glovebox stripper system seismic design evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Grinnell, J. J. [Savannah River Site (SRS), Aiken, SC (United States); Klein, J. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    The use of glovebox confinement at US Department of Energy (DOE) tritium facilities has been discussed in numerous publications. Glovebox confinement protects the workers from radioactive material (especially tritium oxide), provides an inert atmosphere for prevention of flammable gas mixtures and deflagrations, and allows recovery of tritium released from the process into the glovebox when a glovebox stripper system (GBSS) is part of the design. Tritium recovery from the glovebox atmosphere reduces emissions from the facility and the radiological dose to the public. Location of US DOE defense programs facilities away from public boundaries also aids in reducing radiological doses to the public. This is a study based upon design concepts to identify issues and considerations for design of a Seismic GBSS. Safety requirements and analysis should be considered preliminary. Safety requirements for design of GBSS should be developed and finalized as a part of the final design process.

  11. Tritium means of detection and of protection

    International Nuclear Information System (INIS)

    Sutra-Fourcade, Y.

    1967-01-01

    The report is an attempt to correlate present data concerning tritium, especially from the health physics points of view. The various detection and measurement methods are reviewed in turn: measurement of tritium in the atmosphere, in liquids and on surfaces. The operation of various types of apparatus is analyzed and the sensitivity limits deduced from laboratory tests are given. Otter sections are devoted to the means of protection which can be used against inhalation of tritium (ventilation, protective clothing) and to calculations of the changes in atmospheric pollution in a given place and of the time spent in a contaminated zone. The last part deals with the decontamination of equipment contaminated with tritium. (author) [fr

  12. Tritium Issues in Next Step Devices

    Energy Technology Data Exchange (ETDEWEB)

    C.H. Skinner; G. Federici

    2001-09-05

    Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

  13. Tritium Issues in Next Step Devices

    International Nuclear Information System (INIS)

    C.H. Skinner; G. Federici

    2001-01-01

    Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

  14. Performance Monitoring of Ships

    DEFF Research Database (Denmark)

    Hansen, Søren Vinther

    The purpose of the research project is to establish a reliable index in the performance evaluation of ships. During operation the ship will experience added resistance due to fouling of hull and propeller. The added resistance will lead to increased fuel consumption and thus increased emissions...... to the environment. The monitoring of the ship’s performance can be used as decision support in determining when actions to improve performance should be taken. The performance evaluation is based on a model of the ship and the added resistance from wind and waves during operation. Logged data on board the ship...... is used as input to the system and by comparing model and ship behaviour, an index describing the ship’s performance is generated. The work in this thesis is based on data logged through the automation system on board a PostPanmax container ship where data have been logged through a year. A routine...

  15. Shipping Information Pipeline

    DEFF Research Database (Denmark)

    Jensen, Thomas; Vatrapu, Ravi

    2015-01-01

    systems researchers engaged in the design and development of a prototype for an innovative IT-artifact called Shipping Information Pipeline which is a kind of “an internet” for shipping information. The instrumental aim is to enable information seamlessly to cross the organizational boundaries......This paper presents a design science approach to solving persistent problems in the international shipping eco system by creating the missing common information infrastructures. Specifically, this paper reports on an ongoing dialogue between stakeholders in the shipping industry and information...... and national borders within international shipping which is a rather complex domain. The intellectual objective is to generate and evaluate the efficacy and effectiveness of design principles for inter-organizational information infrastructures in the international shipping domain that can have positive...

  16. Desactivation of tritium waters by rectification methods

    International Nuclear Information System (INIS)

    Egorov, A.I.; Tyunis, V.M.

    2002-01-01

    Results of experiments into the basic rectification processes dedicated to tritium separation from reactor, technological and waste waters are presented. Coefficients of separation for rectification of water (1.028), ammonia (1.05), azeotrope H 2 O - HTO - HNO 3 (1.098) and D 2 O - DTO - DNO 3 (1.039) are performed. Operating schemes of tritium separating units are reviewed [ru

  17. Extraction of tritium from lithium aluminate

    International Nuclear Information System (INIS)

    Yunker, W.

    1976-09-01

    Complete data are presented for the extraction of tritium from neutron irradiated lithium aluminate pellets. Two methods were used: (a) thermal/vacuum extraction at temperatures up to 980 0 C and (b) dissolution of the pellets in sodium tetraborate at 850 0 C followed by vacuum extraction. Approximately 1 percent of the tritium was in a noncondensable (at -195 0 C and 10 -3 torr) form. Extraction efficiency was greater than 95 percent

  18. Extraction of tritium from lithium aluminate

    Energy Technology Data Exchange (ETDEWEB)

    Yunker, W [ed.

    1976-09-01

    Complete data are presented for the extraction of tritium from neutron irradiated lithium aluminate pellets. Two methods were used: (a) thermal/vacuum extraction at temperatures up to 980/sup 0/C and (b) dissolution of the pellets in sodium tetraborate at 850/sup 0/C followed by vacuum extraction. Approximately 1 percent of the tritium was in a noncondensable (at -195/sup 0/C and 10/sup -3/ torr) form. Extraction efficiency was greater than 95 percent.

  19. Enantiospecific tritium labeling of 28-homocastasterone.

    Science.gov (United States)

    Elbert, Tomáš; Patil, Mahadeo R; Marek, Aleš

    2017-03-01

    A regiospecific and enantiospecific synthesis of tritium-labeled 28-homocastasterone is reported. Appropriate chlorocarbonate, efficiently synthesized from the starting 28-homocastasterone in an overall yield of 46%, undergoes catalytic tritium dechlorination by the T 2 /Pd[0]/Et 3 N system, providing 28-[3β- 3 H]homocastasterone, in a good yield, radiochemical purity (>97%), and with a high specific activity (5.8 Ci/mmol). Copyright © 2016 John Wiley & Sons, Ltd.

  20. Energy Metabolism and Human Dosimetry of Tritium

    International Nuclear Information System (INIS)

    Galeriu, D.; Takeda, H.; Melintescu, A.; Trivedi, A.

    2005-01-01

    In the frame of current revision of human dosimetry of 14 C and tritium, undertaken by the International Commission of Radiological Protection, we propose a novel approach based on energy metabolism and a simple biokinetic model for the dynamics of dietary intake (organic 14 C, tritiated water and Organically Bound Tritium-OBT). The model predicts increased doses for HTO and OBT comparing to ICRP recommendations, supporting recent findings

  1. Y-12 defense programs. Nuclear Packaging Systems testing capabilities

    International Nuclear Information System (INIS)

    1995-06-01

    The Nuclear Packaging Systems (NPS) Department can manage/accomplish any packaging task. The NPS organization is responsible for managing the design, testing, certification, procurement, operation, refurbishment, maintenance, and disposal of packaging used to transport radioactive materials, other hazardous materials, and general cargoes on public roads and within the Oak Ridge Y-12 Plant. Additionally, the NPS Department has developed a Quality Assurance plan for all packaging, design and procurement of nonweapon shipping containers for radioactive materials, and design and procurement of performance-oriented packaging for hazardous materials. Further, the NPS Department is responsible for preparation and submittal of Safety Analysis Reports for Packaging (SARP). The NPS Department coordinates shipping container procurement and safety certification activities that have lead-times of up to two years. A Packaging Testing Capabilities Table at the Oak Ridge complex is included as a table

  2. Leveraging Available Data to Support Extension of Transportation Packages Service Life

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, K.; Abramczyk, G.; Bellamy, S.; Daugherty, W.; Hackney, B.; Hoffman, E.; Skidmore, E.; Stefek, T.

    2012-06-12

    Data obtained from testing shipping package materials have been leveraged to support extending the service life of select shipping packages while in nuclear materials transportation. Increasingly, nuclear material inventories are being transferred to an interim storage location where they will reside for extended periods of time. Use of a shipping package to store nuclear materials in an interim storage location has become more attractive for a variety of reasons. Shipping packages are robust and have a qualified pedigree for their performance in normal operation and accident conditions within the approved shipment period and storing nuclear material within a shipping package results in reduced operations for the storage facility. However, the shipping package materials of construction must maintain a level of integrity as specified by the safety basis of the storage facility through the duration of the storage period, which is typically well beyond the one year transportation window. Test programs have been established to obtain aging data on materials of construction that are the most sensitive/susceptible to aging in certain shipping package designs. The collective data are being used to support extending the service life of shipping packages in both transportation and storage.

  3. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  4. Tritium evolution from various morphologies of palladium

    International Nuclear Information System (INIS)

    Tuggle, D.G.; Claytor, T.N.; Taylor, S.F.

    1994-01-01

    The authors have been able to extend the tritium production techniques to various novel morphologies of palladium. These include small solid wires of various diameters and a type of pressed powder wire and a plasma cell. In most successful experiments, the amount of palladium required, for an equivalent tritium output, has been reduced by a factor of 100 over the older powder methods. In addition, they have observed rates of tritium production (>5 nCi/h) that far exceed most of the previous results. Unfortunately, the methods that they currently use to obtain the tritium are poorly understood and consequently there are numerous variables that need to be investigated before the new methods are as reliable and repeatable as the previous techniques. For instance, it seems that surface and/or bulk impurities play a major role in the successful generation of any tritium. In those samples with total impurity concentrations of >400 ppM essentially no tritium has been generated by the gas loading and electrical simulation methods

  5. A model for global cycling of tritium

    International Nuclear Information System (INIS)

    Killough, G.G.; Kocher, D.C.

    1988-01-01

    Dynamic compartment models are widely used to describe global cycling of radionuclides for purposes of dose estimation. In this paper, we present a new global tritium model that reproduces environmental time-series data on concentrations in precipitation, ocean surface waters, and surface fresh waters in the northern hemisphere, concentrations of atmospheric tritium in the soutehrn hemisphere, and the latitude dependence of tritium in both hemispheres. Named TRICYCLE for Tritium CYCLE, the model is based on the global hydrologic cycle and includes hemisphereic stratospheric compartments, disaggregation of the troposphere and ocean surface waters into eight latitudezones, consideration of the different concentrations of atmospheric tritium over land and over the ocean, and a diffusive model for transport in the ocean. TRICYCLE reproduces the environmental data if we assume that about 50% of the tritium from atmospheric weapons testing was injected directly into the northern stratosphere as HTO. The models latitudinal disaggregation permits taking into account the distribution of population. For a unfiormaly distributed release of HTO into the worldwide troposphere, TRICYCLE predicts a collective dose commitment to the world population that exceeds the corresponding prediction by the NCRP model by about a factor of 3. 11 refs., 5 figs., 1 tab

  6. Separation of tritium from aqueous effluents

    International Nuclear Information System (INIS)

    Bruggeman, A.; Leysen, R.; Meynendonckx, L.; Parmentier, C.; Bellien, H.; Smets, D.; Stevens, J.

    1984-01-01

    This report describes the further development of the so-called ELEX process, carried out from 1 July 1980 until 31 December 1982. The ELEX process is the combination of electrolysis with the catalytic tritium exchange between hydrogen and water in order to accumulate the tritium in the liquid phase. The experimental study of the catalytic tritium exchange between hydrogen and liquid water was continued and the overall exchange rate could be substantially increased. An alternative process based on bithermal exchange of tritium has been evaluated. In the 10 mol h -1 mini-pilot bench scale detritiation unit the ELEX process was successfully demonstrated by detritiating up to now more than 1m 3 of water containing up to 100 mCi tritium per dm 3 , which is the feed concentration to be expected for application of the process in a reprocessing plant. A 280 mol h -1 pilot detritiation installation now being constructed is described. This installation will realize a volume reduction factor of 100 and a process decontamination factor of 100. The maximum total tritium inventory will be about 1000 Ci. The plant consists mainly of a 80 kW electrolyser and a 10 cm diameter exchange column and can be considered as the ultimate step before industrial application of the ELEX process

  7. Experiments on tritium behavior in beryllium, (1)

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi; Ishizuka, Etsuo; Matsumoto, Mikio; Inada, Seiji; Sezaki, Katsuji; Saito, Minoru; Kato, Mineo.

    1989-06-01

    In JMTR, it was observed that the tritium concentration of the primary coolant increases with the reactor operation at 50 MW. As one of the tritium generation sources, we paid attention to a neutron reflector made of beryllium because the tritium generation rate in the beryllium is bigger than other components in the reactor core. On the other hand, the irradiation test of blanket materials (i.e. tritium breeding materials and neutron multipling materials) are planned for development of the fusion reactor in JMTR and the beryllium will be also irradiated as a neutron multiplier with tritium breeding materials. Therefore, as the irradiated specimens, we used a hot-pressed beryllium disk fabricated by the same method as the neutron reflector or the neutron multiplier and conducted the irradiation tests in JMTR. The purpose of these tests are to clarify the tritium behavior in the hot-pressed beryllium. In this paper, from a viewpoint of the fabrication of capsules for neutron irradiation, the specifications of the irradiated specimens and capsules are summarized. Additionally, the results on the puncture test of the container of the irradiation specimens are described. (author)

  8. Liner Shipping Fleet Repositioning

    DEFF Research Database (Denmark)

    Tierney, Kevin; Jensen, Rune Møller

    2011-01-01

    Liner shipping fleet repositioning consists of moving vessels between services in a liner ship- ping network in order to better orient the overall network to the world economy, and to ensure the proper maintenance of vessels. Thus, fleet repositioning involves sailing and loading activities subject......, can be accurately modeled. Numerous liner shipping fleet repositioning problems are solved each year by the world’s shipping firms without the assistance of any decision support, even though humans can require between two to three days to find a reasonable solution. Finding optimal repositionings...

  9. Buckling of Ship Structures

    CERN Document Server

    Shama, Mohamed

    2013-01-01

    Buckling of Ship Structures presents a comprehensive analysis of the buckling problem of ship structural members. A full analysis of the various types of loadings and stresses imposed on ship plating and primary and secondary structural members is given. The main causes and consequences of the buckling mode of failure of ship structure and the methods commonly used to control buckling failure are clarified. This book contains the main equations required to determine the critical buckling stresses for both ship plating and the primary and secondary stiffening structural members. The critical buckling stresses are given for ship plating subjected to the induced various types of loadings and having the most common boundary conditions encountered in ship structures.  The text bridges the gap existing in most books covering the subject of buckling of ship structures in the classical analytical format, by putting the emphasis on the practical methods required to ensure safety against buckling of ship structur...

  10. Tritium behavior in the Caisson, a simulated fusion reactor room

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Yamada, Masayuki; Suzuki, Takumi; O'hira, Shigeru; Nakamura, Hirofumi; Shu, Weimin; Yamanishi, Toshihiko; Kawamura, Yoshinori; Isobe, Kanetsugu; Konishi, Satoshi; Nishi, Masataka

    2000-01-01

    In order to confirm tritium confinement ability in the deuterium-tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called 'Caisson', at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak-tight vessel of 12 m 3 , simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code

  11. Studies on steps affecting tritium residence time in solid blanket

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

    For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

  12. Design options to minimize tritium inventories at Savannah River

    International Nuclear Information System (INIS)

    Klein, J.E.; Wilson, J.; Heroux, K.J.; Poore, A.S.; Babineau, D.W.

    2016-01-01

    Highlights: • La-Ni-Al alloys are used as tritium storage materials and retain He-3. • La-Ni-Al He-3 effects decrease useable process tritium inventory. • Use of Pd or depleted uranium beds decreases process tritium inventories. • Reduced inventory tritium facilities will lower public risk. - Abstract: Large quantities of tritium are stored and processed at the Savannah River Site (SRS) Tritium Facilities. In many design basis accidents (DBAs), it is assumed the entire tritium inventory of the in-process vessels are released from the facility and the site for inclusion in public radiological dose calculations. Pending changes in public dose calculation methodologies are driving the need for smaller in-process tritium inventories to be released during DBAs. Reducing the in-process tritium inventory will reduce the unmitigated source term for public dose calculations and will also reduce the production demand for a lower inventory process. This paper discusses process design options to reduce in-process tritium inventories. A Baseline process is defined to illustrate the impact of removing or replacing La-Ni-Al alloy tritium storage beds with palladium (Pd) or depleted uranium (DU) storage beds on facility in-process tritium inventories. Elimination of La-Ni-Al alloy tritium storage beds can reduce in-process tritium inventories by over 1.5 kg, but alternate process technologies may needed to replace some functions of the removed beds.

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

    Science.gov (United States)

    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.

  14. Results of observations of the tritium concentration in water fractions in the disposition regions of tritium laboratories

    International Nuclear Information System (INIS)

    Koval, G.N.; Kuzmina, A.I.; Kolomiets, N.F.; Svarichevskaya, E.V.; Rogosin, V.N.; Svyatun, O.V.

    1995-01-01

    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 quantity 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

  15. Method and apparatus for controlling accidental releases of tritium

    Science.gov (United States)

    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.

  16. Method and apparatus for controlling accidental releases of tritium

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1980-01-01

    An improvement in a tritium control system based on a catalytic oxidation reactor is provided 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

  17. Analysis of in-pile tritium release experiments

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Tam, S.W.; Johnson, C.E.

    1992-01-01

    The objective of this work is to characterize tritium release behavior from lithium ceramics and develop insight into the underlying tritium release mechanisms. Analysis of tritium release data from recent laboratory experiments with lithium aluminate has identified physical processes which were previously unaccounted for in tritium release models. A new model that incorporates the recent data and provides for release from multiple sites rather than only one site was developed. Calculations of tritium release using this model are in excellent agreement with the tritium release behavior reported for the MOZART experiment

  18. Procedures for the retention of gaseous tritium released from a tritium enrichment plant

    International Nuclear Information System (INIS)

    Gutowski, H.; Bracha, M.

    1987-01-01

    General aim of the study is the comparison of two alternative processes for the retention of gaseous tritium which is released during normal operation and emergency operation in a tritium-enrichment-plant. Two processes for the retention of tritium were compared: 1. Oxidation-process. The hydrogen-gas containing HT will be burnt on an oxidation catalyst to H 2 O and HTO. In a subsequent step the water will be removed from the process by condensation, freezing and adsorption. 2. TROC-process (Tritium Removal by Organic Compounds). The tritium is added to an organic compound (acid) via catalyst. This reaction is irreversible and leads to solid products. (orig./RB) [de

  19. FDNH - the tritium module in RODOS

    International Nuclear Information System (INIS)

    Galeriu, D.; Melintescu, A.; Turcanu, C. O.; Raskob, W.

    2001-01-01

    Under the auspices of its RTD (Research and Technological Development) Framework Programmes, the European Commission has supported the development of the RODOS (Real-time On-line Decision Support) system for off-site emergency management. The project started in 1989 focusing on PWR/LWR type accidents and using experience from the Chernobyl accident. In 1997 it was realised that tritium should be included in the list of radionuclides, as large tritium sources exists in Europe and to allow a potential expansion of the RODOS system for application on future fusion reactor accidents. The National Institute for Physics and Nuclear Engineering (IFIN-HH) in Romania - in close co-operation with the Research Centre Karlsruhe (FZK) - was charged to develop the tritium module, based on previous experience in environmental tritium modelling and the operation of CANDU reactors in Romania (with potential tritium accidents). At present, the Food and Dose Module Hydrogen -(FDMH) - for tritium applications - is integrated and documented in the RODOS system. It calculates the time dependent tritium concentration (as tritiated water or organically bound tritium) in crops (as much as 22 different species) and up to 12 animal products, inhalation doses and ingestion dose from up to 34 diet items for various groups of the population and for up to 2520 locations around the source, following an accidental emission of tritiated water. FDMH incorporates many improved techniques in radiological assessment and makes intensively use of interdisciplinary research. It is developed in a modular structure with a variable time grid according to the physical processes. Differing from other models, using generic transfer parameters or parameters fitted on individual experiments, FDMH derives tritium transfer rates based on physical and physiological process analysis, using scientifically accepted results from interdisciplinary research on, among others, land-atmosphere interaction, water cycle in the

  20. Simulation study of intentional tritium release experiments in the caisson assembly for tritium safety at the TPL/JAERI

    International Nuclear Information System (INIS)

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

    2001-01-01

    At the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI), Caisson assembly for tritium safety study (CATS) with 12 m 3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate the tritium behavior in the case, where the tritium leak accident should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak accident should happen in a ventilated room. As for the understanding of initial tritium behavior until the tritium concentration become steady, the precise estimation of local flow rate in a room and time-dependent release behavior from the leak point are essential to predict the tritium behavior by simulation code. The three-dimensional eddy flow model considering, tritium-related phenomena was adopted to estimate the local flow rate in the 50 m 3 /h ventilated Caisson. The time-dependent tritium release behavior from the sample container was calculated by residence time distribution function. The calculated tritium concentrations were in good agreement with the experimental observations. The primary removal tritium behavior was also investigated by another code. Tritium gas concentrations decreased logarithmically to the time by ventilation. These observations were understandable by the reason that the flow in the ventilated Caisson was regarded as the perfectly mixing flow. The concentrations of tritiated water measured, and indications of tritium concentration by tritium monitors became gradually flat. This phenomena called 'tritium soaking effect' was found to be reasonably explained by considering the contribution of the exhaustion velocity by ventilation system, and the adsorption and desorption reaction rate of tritiated water on the wall material which is SUS 304. The calculated tritium concentrations were in good agreement with the experimental observations

  1. A compact tritium AMS system

    International Nuclear Information System (INIS)

    Roberts, M.L.; Hamm, R.W.; Dingley, K.H.; Chiarappa-Zucca, M.L.; Love, A.H.

    2000-01-01

    Tritium ( 3 H) is a radioisotope that is extensively utilized in biological and environmental research. For biological research, 3 H is generally quantified by liquid scintillation counting requiring gram-sized samples and counting times of several hours. For environmental research, 3 H is usually quantified by 3 He in-growth which requires gram-sized samples and in-growth times of several months. In contrast, provisional studies at LLNL's Center for Accelerator Mass Spectrometry have demonstrated that accelerator mass spectrometry (AMS) can be used to quantify 3 H in milligram-sized biological samples with a 100 to 1000-fold improvement in detection limits when compared to scintillation counting. This increased sensitivity is expected to have great impact on the biological and environmental research community. However, in order to make the 3 H AMS technique more broadly accessible, smaller, simpler, and less expensive AMS instrumentation must be developed. To meet this need, a compact, relatively low cost prototype 3 H AMS system has been designed and built based on an LLNL ion source/sample changer and an AccSys Technology radio frequency quadrupole (RFQ) linac. With the prototype system, 3 H/ 1 H ratios ranging from 1x10 -10 to 1x10 -13 have be measured from milligram-sized samples. With improvements in system operation and sample preparation methodology, the sensitivity limit of the system is expected to increase to approximately 1x10 -15

  2. Tritium technology and safety at JET

    International Nuclear Information System (INIS)

    Bell, A.C.

    1994-01-01

    D-T plasma operation has always been envisaged since the beginning of the JET Project and both the original design and subsequent modifications have been designed to take account of the requirements of D-T operation. A limited tritium experiment was carried out in November 1991 which generated 1.7 MW of fusion power. In addition to the physics objectives, this experiment was intended to provide results which would be important for the technology to be used in full D-T phase, such as tritium accounting and hold-up. Because of the limited usage of tritium it was possible to use a open-quotes once-throughclose quotes system in which around 99% of the tritium was recovered. It is currently planned to have a daily throughput of around l0g of tritium per day in the full D-T phase, introduced through neutral beam and/or gas puffing. As it would be neither environmentally acceptable nor cost-effective to discharge even 1% of this to the atmosphere, a tritium recycling plant, known as the Active Gas Handling System (AGHS) has been constructed and is currently being commissioned. It was necessary to take several issues into consideration in the design of the AGHS to ensure that it and the JET machine would be capable of being licensed for handling tritium. These were ensuring that open-quotes Best Practicable Meansclose quotes were used to limit routine discharges to the environment; ensuring that routine radiation exposure of the JET workforce would be minimised; and ensuring that the risk to the workforce and the public arising from accidents would be acceptably low. The technology involved, waste management and regulatory issues are discussed further in the paper

  3. Tritium surveillance around nuclear facilities in Japan

    International Nuclear Information System (INIS)

    Inoue, Y.; Kasida, Y.

    1978-01-01

    In order to measure the tritium levels in the environmental water around the nuclear facilities, the tritium surveillance program began in 1967 locally at Tsuruga and Mihama districts. Nowadays it has been expanded to the ten commercial nuclear power stations and three nuclear facilities. For samples whose tritium concentration is believed less than about 100 pCi/l, they were electrolytically enriched, and then counted by the liquid scintillation counter. Some of samples believed higher than 100 pCi/l were analysed without any enrichment by the low background liquid scintillation counters, Aloka LB 600 or Aloka LB 1. The results of each station are listed in Table. The sampling points corresponding to each results are shown in Figure. Tritium from the effluent was not reflected in all the land water and the tap water around the nuclear power stations and the nuclear facilities. Tritium concentration in rivers, streams, and reservoirs (pools) decreased exponentially from about 600 pCi/l in 1967 to about 150 pCi/l in 1972 at Tsuruga and Mihama, and 360 pCi/l in 1968 to 120 pCi/l in 1973 at Genkai, with the half life of about 2.5 years in both cases. After around 1972, tritium levels of river system in all districts of Japan kept nearly constant up to the end of 1975 and they were in the range from 100 to 300 pCi/l corresponding to the districts. Thereafter, it seems to start to decrease again in 1976. Sea water sampled at the intake of the station or on the seashore far from the outlet was regarded not to be influenced by the effluent from the nuclear reactors or facilities. Tritium concentration in these coastal waters decreased from 100 - 300 pCi/l in 1971 to 30 - 40 pCi/l in 1972 in Fukushima, Ibaraki and Fukui prefectures. (author)

  4. Crushing Strength of Ship Structures

    DEFF Research Database (Denmark)

    Cerup-Simonsen, Bo; Abramowicz, W.; Høstgaard-Brene, C.N.S.

    1999-01-01

    The crushing response of ship structures is of primary importance to the designers and practicing engineers concerned with accidental loading and accident reconstruction of marine vehicles. Ship to-ship collisions, ship-harbor infrastructure interaction or ship-offshore structure interaction...... are just few examples of accident scenarios where the detailed knowledge of the crushing response of ship structure is crucial for a reliable engineering analysis....

  5. Tritium and helium-3 in metals

    International Nuclear Information System (INIS)

    Lasser, R.

    1989-01-01

    The book surveys recent results on the behaviour of tritium and its decay product helium-3 metals. In contrast to many earlier books which discuss the properties of the stable hydrogen isotopes without mentioning tritium, this book reviews mainly the results on tritium in metals. Due to the difficulties in preparing metal tritide samples, very important quantities such as diffusivity, superconductivity, solubility, etc. have only been determined very recently. The book not only presents the measured tritium data, but also the isotopic dependency of the different physical properties by comparing H, D and T results. A chapter is devoted to helium-3 in metals. Aspects such as helium release, generation of helium bubbles, swelling, and change of the lattice parameter upon aging are discussed. The book provides the reader with up-to-date information and deep insight into the behaviour of H, D, T and He-3 in metals. Further important topics such a tritium production, its risks, handling and discharge to the environment are also addressed

  6. Selection of fluids for tritium pumping systems

    International Nuclear Information System (INIS)

    Chastagner, P.

    1984-02-01

    The degradation characteristics of three types of vacuum pump fluids, polyphenyl ethers, perfluoropolyethers and hydrocarbon oils were reviewed. Fluid selection proved to be a critical factor in the long-term performance of tritium pumping systems and subsequent tritium recovery operations. Thermal degradation and tritium radiolysis of pump fluids produce contaminants which can damage equipment and interfere with tritium recovery operations. General characteristics of these fluids are as follows: polyphenyl ether has outstanding radiation resistance, is very stable under normal diffusion pump conditions, but breaks down in the presence of oxygen at anticipated operating temperatures. Perfluoropolyether fluids are very stable and do not react chemically with most gases. Thermal and mechanical degradation products are inert, but the radiolysis products are very corrosive. Most of the degradation products of hydrogen oils are volatile and the principal radiolysis product is methane. Our studies show that polyphenyl ethers and hydrocarbon oils are the preferred fluids for use in tritium pumping systems. No corrosive materials are formed and most of the degradation products can be removed with suitable filter systems

  7. Tritium in the Savannah River Site environment

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E. Jr.; Bauer, L.R.; Hayes, D.W.; Marter, W.L.; Zeigler, C.C.; Stephenson, D.E.; Hoel, D.D.; Hamby, D.M.

    1991-05-01

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found.

  8. Experiences with decontaminating tritium-handling apparatus

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

    1991-07-01

    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 4x10 4 --10 6 disintegrations per minute (dpm)/cm 2 to 2x10 2 --4x10 4 dpm/cm 2 . 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 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

  9. Tritium in the Savannah River Site environment

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.; Bauer, L.R.; Hayes, D.W.; Marter, W.L.; Zeigler, C.C.; Stephenson, D.E.; Hoel, D.D.; Hamby, D.M.

    1991-05-01

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found

  10. Tritium and radon risks for humans

    International Nuclear Information System (INIS)

    Mauna, Traian; Mauna, Andriesica

    2008-01-01

    Full text: The gaseous and liquid releases into environment from the two CANDU type units of Cernavoda NPP now in operation has more tritium contents than other kind of western power reactors. CANDU type reactor uses heavy water as moderator and primary circuit heat transfer agent. In normal operation deuterium go to tritium by neutron capture, the molecule of tritiated heavy water can escape from nuclear systems in very small amounts and so it is released into environment. After release the tritium follows the way of water into environment. One year ago the antinuclear NGO led a hard attack against Units 3 and 4 during the procedure of public acceptance request. This attack tried to demonstrate the great risk for humans of the tritium released by Cernavoda NPP. Obviously this risk is very low as demonstrated by many years reactor operation. SNN as owner of Cernavoda NPP ensures by all kind of information channels about the radioactive potential risk for humans. By the other hand, ironically, the antinuclear NGO makes nothing to inform the people about radon risk magnitude in some areas. This is a well-known fact but the radon concentration in dwellings can be decreased by some improved building procedures. The radon is the first natural cause of lung cancer. The environmental NGO and Romanian authorities do not have an information service about radon hazard data neither in dwellings or in uranium mining areas. The paper compares the properties and risks for tritium and radon. (authors)

  11. Radiation-induced tritium labelling and product analysis

    Energy Technology Data Exchange (ETDEWEB)

    Peng, C.T. (California Univ., San Francisco, CA (United States). Dept. of Pharmaceutical Chemistry)

    1993-05-01

    By-products formed in radiation-induced tritium labelling are identified by co-chromatography with authentic samples or by structure prediction using a quantitative structure-retention index relationship. The by-products, formed from labelling of steroids, polynuclear aromatic hydrocarbons, 7-membered heterocyclic ring structures, 1,4-benzodiazepines, 1-haloalkanes, etc. with activated tritium and adsorbed tritium, are shown to be specifically labelled and anticipated products from known chemical reactions. From analyses of the by-products, one can conclude that the hydrogen abstraction by tritium atoms and the substitution by tritium ions are the mechanisms of labelling. Classification of the tritium labelling methods, on the basis of the type of tritium reagent, clearly shows the active role played by tritium atoms and ions in radiation-induced methods. (author).

  12. Mapping of tritium in drinking water from various Indian states

    International Nuclear Information System (INIS)

    Shah, Chirag A.; Baburajan, A.; Ravi, P.M.; Tripathi, R.M.

    2015-01-01

    The tritium in fresh water used for drinking purpose across five state of India was analyzed for tritium activity. The tritium data obtained were compared with the monitoring data of tritium in drinking water sources at Tarapur site, which houses a number of nuclear facilities. It is observed that the tritium activity in the water sample from various out station locations were in the range of < 0.48 to 1.33 Bq/l. The tritium value obtained in the drinking water sources at Tarapur was found to be in the range of 0.91 to 3.10 Bq/l. The monitoring of tritium in drinking water from Tarapur and from various out station location indicate that the level is negligible compared to the USEPA limit of 10000 Bq/l and the contribution of operation nuclear facilities to the tritium activity in drinking water source at Tarapur is insignificant. (author)

  13. Modelization of tritium transfer into the organic compartments of algae

    International Nuclear Information System (INIS)

    Bonotto, S.; Gerber, G.B.; Arapis, G.; Kirchmann, R.

    1982-01-01

    Uptake of tritium oxide and its conversion into organic tritium was studied in four different types of algae with widely varying size and growth characteristics (Acetabularia acetabulum, Boergesenia forbesii, two strains of Chlamydomonas and Dunaliella bioculata). Water in the cell and the vacuales equilibrates rapidly with external tritium water. Tritium is actively incorporated into organically bound form as the organisms grow. During the stationary phase, incorporation of tritium is slow. There exists a discrimination against the incorporation of tritium into organically bound form. A model has been elaborated taking in account these different factors. It appears that transfer of organic tritium by algae growing near the sites of release would be significant only for actively growing algae. Algae growing slowly may, however, be useful as cumulative indicators of discontinuous tritium release. (author)

  14. Tritium activity in Australian rainwater 1962-1986

    International Nuclear Information System (INIS)

    Calf, G.E.

    1988-07-01

    Atmospheric tritirecipitated in rainfall has been measured since 1962 in samples from a network of Australian monitoring stations. Atmospheric thermonuclear bomb tests have injected into the atmosphere about 350 kg of tritium, mostly into the northern hemisphere's troposphere and stratosphere, causing atmospheric concentrations of tritium to reach peak values of 50 to 100 tritium units (TU) in the southern hemisphere. Tritium from underground testing has not contributed significantly to the amounts in the atmosphere or in surface waters. In Australia, peak concentrations of precipitated tritium were observed in 1962, 1964 and 1969. The tritium activity in rainwater for all Australian stations decreased after 1969. Radioactive decay of bomb tritium will decrease the tritium budget to the natural level by the early 1990s in Australian precipitation

  15. Tritium in surface water of the Yenisei river Basin

    International Nuclear Information System (INIS)

    Bondareva, L.G.; Bolsunovsky, A.Ya.

    2005-01-01

    The paper reports an investigation of the tritium content in the surface waters of the Yenisei River basin near the Mining-and-Chemical Combine (MCC). In 2001-2003 the maximum tritium concentration in the Yenisei River did not exceed 4±1 Bq/L. It has been found that there are surface waters containing enhanced tritium, up to 168 Bq/L, as compared with the background values for the Yenisei River. There are two possible sources of tritium input. First, the last operating reactor of the MCC, which still uses the Yenisei water as coolant. Second, tritium may come from the deep aquifers at the Severny testing site. For the first time tritium has been found in two aquatic plant species of the Yenisei River with maximal tritium concentration 304 Bq/Kg wet weight. Concentration factors of tritium for aquatic plants are much higher than 1

  16. Tritium Safety-Related Studies at TPL of JAERI

    Science.gov (United States)

    O'hira, S.; Hayashi, T.; Okuno, K.

    1997-09-01

    Activities regarding tritium safety technology in the Tritium Process Laboratory (TPL) at Tokai Establishment of Japan Atomic Energy Research Institute are reviewed. Research and development of a new tritium removal system is being carried out by using a gas separation membrane which enable to make the ITER atmosphere detritiation system more compact and cost-effective. Techniques of gas flowing calorimetry and laser Raman spectroscopy are applied to develop new tritium accountancy methods. Studies of tritium-material interaction, such as plasma material interactions, radiochemical reaction of tritium in gas phase, radiolysis of tritiated water, and waste processing are being carried out under ITER/EDA and U.S.-Japan collaboration. Tritium release experiments for research of tritium behavior in confinements and environment and demonstration of safety related components are planned.

  17. [Psychopathology service on ships].

    Science.gov (United States)

    Nowosielski, Radosław; Mazurek, Tomasz; Florkowski, Antoni

    2010-06-01

    The aim of this study was to describe the specific engineering services and suitability of candidates for the psychophysical performance. Navy ships are equipped with equipment and weapons are controlled by electronic devices ship and crew. Advanced technology puts high demands on operator. For the ship's staff are recruited soldiers of the psychophysical characteristics predisposing to this kind of action. The paper uses personal experience to work in military units of the Navy, and data from the literature. Terms of sailing ships off the summer season are defined as difficult. The crew during a combat mission felt the risks associated with movements of the ship in difficult meteorological conditions, and associated with the implementation of the task. The development of ship's technical equipment, working in isolated groups, functioning within a limited space, noise, vibration, electromagnetic waves heighten the emotional burden on crew members. Military service on Navy ships require high psycho-physical predisposition, resistance to stress. The crucial factor is proper selection among the candidates based on psychiatric and psychological counseling for military and medical jurisprudence. Also plays a significant role for training doctors and specialists in psychoprophylaxy of military units in the field of mental hygiene.

  18. Tritium handling safety and operating experience at the Tritium Systems Test Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, R.V.

    1989-01-01

    The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory is a facility designed to develop and demonstrate, in full scale, technologies necessary for safe and efficient operation of tokamak fusion reactors. TSTA currently consists of systems for pumping DT gas mixtures; for removing impurities; for separating the isotopes of hydrogen; for storage of hydrogen isotopes; for gas analysis; and for assuring safety by the necessary control, monitoring, and detritiation of effluent gaseous streams. TSTA also has several small scale experiments to develop and test new equipment and processes necessary for fusion reactors. Tritium was introduced into TSTA in June 1984. Current inventory is approximately 100 grams. Approximately 10{sup 9} Curies of tritium have been processed in closed loop operation at TSTA. Total tritium releases from the facility stack have been less than 75 Curies. Total operating personnel exposures are less than 500 person-mrem. Exposures to the general public from TSTA tritium releases are extremely small (less than 10{sup {minus}2} mrem). Total tritium buried as waste is less than 36,000 Curies. In this paper, data on component reliability, failure types and rates, and waste quantities are presented. Operational experience under normal, abnormal, and emergency conditions is presented. The DOE requirements for the operation of a tritium facility like TSTA include personnel training, emergency preparedness, radiation protection, safety analysis, and preoperational appraisals. 4 refs., 3 figs., 3 tabs.

  19. Source function for tritium transport models in the Pacific

    International Nuclear Information System (INIS)

    Fine, R.A.; Ostlund, H.G.

    1977-01-01

    An empirically fitted function describes surface Pacific Ocean tritium concentrations as varying exponentially with latitude, the r.m.s. fit to observations is 18%. The oceanic tritium concentration maximum in the North Pacific, which resulted from nuclear weapons testing, lagged the rain data by two to three years occurring in 1965--66. Tritium-salinity correlations are consistent with climatology. Tritium-longitude correlations are consistent with surface water circulation

  20. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    International Nuclear Information System (INIS)

    Gentile, C.A.; LaMarche, P.H.

    1995-01-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During '' time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems

  1. Technology and component development for a closed tritium cycle

    International Nuclear Information System (INIS)

    Hircq, B.; Penzhorn, R.D.; Haange, R.; Naruse, Y.

    1991-01-01

    A brief summary on recent advances in the field of tritium technology concerning the most important subsystems of the fuel cycle of a fusion reactor, i.e. the plasma exhaust pumping system, the exhaust gas clean up system, the isotope separation, the tritium storage and the tritium extraction from a blanket is provided. Experimental results, single component developments, and technical tests including those with relevants amounts of tritium that constitute the basis of proposed integral process concepts are described. 48 refs

  2. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    Energy Technology Data Exchange (ETDEWEB)

    Gentile, C.A.; LaMarche, P.H. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Anderson, J.L. [Los Alamos National Lab., NM (United States)

    1995-07-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During `` time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems.

  3. Non-labile tritium in Savannah River Plant pine trees

    International Nuclear Information System (INIS)

    Sanders, S.M. Jr.

    1976-06-01

    Non-labile tritium bound in cellulose of pine trees was measured to learn about the effects and fate of tritium contributed to the environment by the Savannah River Plant (SRP). An estimation of the regional inventory and the distance tritium can be observed from SRP was desired because tritium is a major component of the radioactivity released by SRP, and as the oxide, it readily disperses in the environment

  4. Distribution of tritium in a chronically contaminated lake

    International Nuclear Information System (INIS)

    Blaylock, B.G.; Frank, M.L.

    1978-01-01

    White Oak Lake located on the U.S. Department of Energy's Oak Ridge Reservation receives a continuous input of tritium from operating facilities and waste disposal operations at the Oak Ridge National Laboratory. The purpose of this paper was (1) to determine the distribution and concentration of tritium in an aquatic environment which has received releases of tritium significantly greater than expected releases from nuclear power plants, and (2) to determine the effect of fluctuating tritium concentrations in ambient water on the concentration of tritium in fish. Aquatic biota from White Oak Lake were analyzed for tissue water tritium and tissue bound tritium. Except for one plant species, the ratio of tissue water tritium to lake water tritium ranged from 0.80 to 1.02. The tissue water tritium in Gambusia affinis, the mosquito fish, followed closely the significant changes in tritium concentration in lake water. The turnover of tissue water tritium was very rapid; Gambusia from White Oak Lake eliminated 50% of their tissue water tritium in 14 min. The ratio of the specific activity of the tissue bound tritium to the specific activity of the lake water was greatest for the larger species of fish but never exceeded unity. The radiation dose to man from tritium which could be acquired through the aquatic food chain was relatively small when compared to other pathways. The whole body dose to a hypothetical individual taking in concentrations of tritium measured in White Oak Lake was 1.8 mrem/yr from eating fish and 10.0 mrem/yr from drinking water

  5. Effective and Safe Ships

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup; Amdahl, Jørgen; Rutgersson, Olle

    1996-01-01

    A Joint Nordic Research project "Effecive and Safe Ships" is presented. The project is aiming to develop methods and tools for quantitative evaluation fo ship safety. This report is the report of the preliminary phase where the plan for the main project is developed. The objectives of the project...... are described. The areas of safety that are presented are collisions, groundings, cargo shifting and ship fire. The areas of lack of knowledge are identified and the tasks within each subproject suggested. The proposal for the main project is presented in the summary. Background material of the state-of-the art...

  6. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    Clements, E.P.

    1997-01-01

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

  7. Continuous aqueous tritium monitor applying membrane technology

    International Nuclear Information System (INIS)

    Sigg, R.A.

    1992-01-01

    Efforts to develop continuous, real-time tritium monitors are focusing on liquid scintillation (LS) counting because its sensitivity for tritium can be below Environmental Protection Agency guidelines (20 pCi/ml). A flow-through LS counter should have a counting efficiency = 0.48, instrumental backgrounds = 5 count/min counting cell volume = 5 ml, and sample water loading capacities = 15 wt%. The system would offer minimum quantifiable activities at the ±10% level of <18 pCi/ml in a 10-min count or <1 pCi/ml if data are averaged for 1 day. The latter value is below the tritium background of surface waters worldwide. However, good time response in a flow-through LS monitor requires flow rates of at least several milliliters per minute. Such a single-pass system would consume large volumes of expensive LS cocktail, and it would generate notable amounts of mixed organic waste

  8. Low-exposure tritium radiotoxicity in mammals

    International Nuclear Information System (INIS)

    Dobson, R.L.

    1982-01-01

    Studies of tritium radiotoxicity involving chronic 3 H0H exposures in mammals demonstrate in both mice and monkeys that biological effects can be measured following remarkably low levels of exposure - levels in the range of serious practical interest to radiation protection. These studies demonstrate also that deleterious effects of 3 H beta radiation do not differ significantly from those of gamma radiation at high exposures. In contrast, however, at low exposures tritium is significantly more effective than gamma rays, rad for rad, by a factor approaching 3. This is important for hazard evaluation and radiation protection because knowledge concerning biological effects of chronic low-level radiation exposure has come mainly from gamma-ray data; and predictions based on gamma-ray data will underestimate tritium effects - especially at low exposures - unless the RBE is fully taken into account

  9. Advancement Of Tritium Powered Betavoltaic Battery Systems

    Energy Technology Data Exchange (ETDEWEB)

    Staack, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Gaillard, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hitchcock, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peters, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Colon-Mercado, H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Teprovich, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coughlin, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Neikirk, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fisher, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-14

    Due to their decades-long service life and reliable power output under extreme conditions, betavoltaic batteries offer distinct advantages over traditional chemical batteries, especially in applications where frequent battery replacement is hazardous, or cost prohibitive. Although many beta emitting isotopes exist, tritium is considered ideal in betavoltaic applications for several reasons: 1) it is a “pure” beta emitter, 2) the beta is not energetic enough to damage the semiconductor, 3) it has a moderately long half-life, and 4) it is readily available. Unfortunately, the widespread application of tritium powered betavoltaics is limited, in part, by their low power output. This research targets improving the power output of betavoltaics by increasing the flux of beta particles to the energy conversion device (the p-n junction) through the use of low Z nanostructured tritium trapping materials.

  10. Tritium releases, birth defects and infant deaths

    International Nuclear Information System (INIS)

    1991-01-01

    The AECB has published a report 'Tritium releases from the Pickering Nuclear Generating Station and Birth Defects and Infant Mortality in Nearby Communities 1971-1988' (report number INFO-0401). This presents the results of a detailed analysis of deaths and birth defects occurring in infants born to mothers living in the area (25 Km radius) of the Pickering nuclear power plant, over an 18-year period. The analysis looked at the frequency of these defects and deaths in comparison to the general rate for Ontario, and also in relation to airborne and waterborne releases of tritium from the power plant. The overall conclusion was that the rates of infant death and birth defects were generally not higher in the study population than in all of Ontario. There was no prevalent relationship between these deaths and defects and tritium releases measured either at the power plant or by ground monitoring stations t some distance from the facility

  11. Tritium monitoring system for near ambient measurements

    International Nuclear Information System (INIS)

    Falter, K.G.; Bauer, M.L.

    1992-01-01

    This paper describes the current status of research on an improved tritium measurement system at the Oak Ridge National Laboratory (ORNL) for the U.S. Navy. Present tritium-in-air monitoring systems installed by the Navy can reliably measure to less than 10 μCi/m 3 , but medical and safety issues are pushing measurement needs to below 1 μCi/m 3 , which is equivalent to 1-10 nCi/ml in liquid samples, using calcium metal converter. A significant effort has been expended over the past 10 years by the Navy RADIAC Development Program at ORNL on various schemes to improve the detection of tritium in both air and liquid at near ambient levels. One such scheme includes a liquid flow-through system based on an NE102 sponge scintillator with dual photomultiplier tubes for tube noise rejection

  12. Safety analysis of tritium recycling system

    International Nuclear Information System (INIS)

    Yang Yong; Zhang Dong; Xing Shixiong

    2009-04-01

    Safety of a tritium recycling system is analysed according to the structure of the system. The method of accident tree is used to analyse the leakage probability of the system. The result show that the leakage probability of the system failure is 1.1 x 10 -3 and the leakage probability of human fault is 7.2 x 10 -3 , which is are in safe limit. But the leakage probability of human fault is higher than system failure. The MCA will occur because of tritium waste emission cell breakage or misplay, in this case, all tritium in the system will leak, which is about 5.84 TBq. The maximal effective individual dose is 1.24 x 10 -3 mSv, the maximal effective close of the collectivity is 15.33 Person·mSv. (authors)

  13. Tritium in water ecosystems of Ural

    International Nuclear Information System (INIS)

    Chebotina, M.Ya.; Nikolin, O.A.

    2005-01-01

    The paper provides the data on tritium monitoring in water ecosystems of the Ural region. The study area comprises the Beloyarsk Atomic Power Plant (cooling reservoir and the Olkhovsk bog-river ecosystem), a territory around the 'Mayak' Enterprise, and control territory, for comparison, located in the North of Sverdlovsk region. It was found that a large area in the Ural region, particularly near the power plant and the 'Mayak,, was characterized by increased tritium content in water as compared with technogenic background is typical for control areas. It may be considered that nearly all the tritium within the study area including control ones are of anthropogenic origin taking into account the act that the global background level for the radionuclide is 1 Bq/l.(author)

  14. Experiment of hydrogen embrittlement of tritium storage vessel material

    International Nuclear Information System (INIS)

    Jung, Hai Yong; Lee, Kun Jai; Chung, H.; Paek, S.

    2000-01-01

    The tritium storage is one of the most important problems for the safety of tritium removal facility. In current, many researches for tritium immobilization have been carried out. The research for tritium storage could be divided into two parts, one is for the metal getter of tritium and another is for the integrity of tritium storage vessel. Especially, the integrity of tritium storage vessel is up to the tritium embrittlement of vessel material, for tritium vessel is mostly made of metal material. In this work, the evaluation of the tritium embrittlement for the tritium storage vessel material is performed with the equipment that is made for high temperature and high vacuum. However, tritium is the radioactivity material, so hydrogen is used for this work. In this work, three metals were chosen for the vessel candidate material, carbon steel, austenitic stainless steel (SUS) 304 and 316L. The experiment was carried out for the several conditions of temperature and pressure. The property change of metal was investigated through the tensile test. Austenitic stainless steel has a high resistance for the hydrogen embrittlement from the result. But the obvious gap between SUS 304 and SUS 316L is not revealed, because the experiment condition may be not sufficient to show the difference between SUS 304 and SUS 316L

  15. Design and operations at the National Tritium Labelling Facility

    International Nuclear Information System (INIS)

    Morimoto, H.; Williams, P.G.

    1991-09-01

    The National Tritium Labelling Facility (NTLF) is a multipurpose facility engaged in tritium labeling research. It offers to the biomedical research community a fully equipped laboratory for the synthesis and analysis of tritium labeled compounds. The design of the tritiation system, its operations and some labeling techniques are presented

  16. Operating experience and procedures at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Carlson, R.V.; Binning, K.E.; Cole, S.P.; Jenkins, E.M.; Wilhelm, R.C.; Cole, S.P.

    1988-01-01

    Operating procedures are important for the safe and efficient operation of the Tritium Systems Test Assembly (TSTA). TSTA has been operating for four years with tritium in a safe and efficient manner. The inventory of tritium in the process loop is 100 grams and several milestone runs have been completed. This paper describes the methods used to operate TSTA. 3 refs., 1 fig

  17. Preliminary measures of tritium content in the Hague (Cotentin) area

    International Nuclear Information System (INIS)

    Umezu, T.

    1971-01-01

    Tritium content of about thirty samples (seawater, algae and rainwater) taken during April-July 1973 in La Hague (Cotentin) area was determined with a liquid scintillation detector after electrolytic enrichment. Tritium content was 15 to 130 tritium units not much higher than concentration generally observed in environment nowadays [fr

  18. Tritium contamination of concrete walls and floors in tritium-handling laboratory

    International Nuclear Information System (INIS)

    Kawano, T.; Kuroyanagi, M.; Tabei, T.

    2006-01-01

    A tritium handling laboratory was constructed at the National Institute for Fusion Science about twenty years ago and it was recently closed down. We completed the necessary work that is legally required in Japan at the laboratory, when the use of radioisotopes is discontinued, involving measurements of radioactive contamination. We mainly used smear and direct-immersion methods for the measurements. In applying the smear method, we used a piece of filter paper to wipe up the tritium staining the surfaces. The filter paper containing the tritium was placed directly into a dedicated vial, a scintillation cocktail was then poured over it, and the tritium was measured with a liquid scintillation counter. With the direct-immersion method, a piece of concrete was placed directly into a vial containing a scintillation cocktail, and the tritium in the concrete was measured with a liquid scintillation counter. As well as these measurements, we investigated water-extraction and heating-cooling methods for measuring tritium contamination in concrete. With the former, a piece of concrete was placed into water in a tube to extract the tritium, the water containing the extracted tritium was then poured into a dedicated vial containing a scintillation cocktail, and the tritium contamination was measured. With the latter, a piece of concrete was placed into a furnace and heated to 800 degrees centigrade to vaporize the tritiated water into flowing dry air. The flowing air was then cooled to collect the vaporized tritiated water in a tube. The collected water was placed in a vial for scintillation counting. To evaluate the direct-immersion method, ratios were determined by dividing the contamination measured with the heating-cooling method by that measured with the direct-immersion method. The average ratio was about 2.5, meaning a conversion factor from contamination obtained with the direct-immersion method to that with the heating-cooling method. We also investigated the

  19. Tritium Research Laboratory safety analysis report

    International Nuclear Information System (INIS)

    Wright, D.A.

    1979-03-01

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment

  20. Tritium Research Laboratory safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Wright, D.A.

    1979-03-01

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment. (ERB)

  1. Tritium metrology within different media: focus on organically bound tritium (OBT); Metrologie du tritium dans differentes matrices: cas du tritium organiquement lie (TOL)

    Energy Technology Data Exchange (ETDEWEB)

    Baglan, N. [CEA Bruyeres-le-Chatel, DIF, 91 (France); Ansoborlo, E. [CEA Marcoule, DEN/DRCP/CETAMA, 30 (France); Cossonnet, C. [IRSN, DEI/STEME/LMRE, 91 - Orsay (France); Fouhal, L. [CEA Cadarache, DEN/D2S/LANSE, 13 - Saint-Paul-lez-Durance (France); Deniau, I.; Mokili, M. [SUBATECH/IN2P3/CNRS, 44 - Nantes (France); Henry, A. [AREVA-NC/DQSSE/PR - La Hague, 50 - Beaumont-Hague, (France); Fourre, E. [CEA Saclay, DSM/DRECAM/LSCE, 91 - Gif-sur-Yvette (France); Olivier, A. [GEA-Marine nationale, 50 - Cherbourg (France)

    2010-07-15

    The measurement of tritium in its various forms (mainly gas (HT), water (HTO) or solid (hydrides)), is an important key step for evaluating health and environmental risks and finally, dosimetry assessment. In vegetable or animal samples, tritium is often associated with the free water fraction, but may be included in the organic form as organically bound tritium (OBT). In this case, 2 forms exist: (i) a fraction called exchangeable or labile (E-OBT), bound to oxygen and nitrogen atoms, and (ii) a so-called non-exchangeable fraction (NE-OBT) bound to carbon atoms. The main technique for tritium analysis is liquid scintillation, which enables one to measure concentrations in the range of several Bq.L{sup -1}. The standards (AFNOR, ISO) published to date relate only to tritium analysis in water. Only one CETAMA method addresses OBT analysis in biological environments. This method has been tested since 2001 through intercomparison circuits on grass samples collected from the environment. Regarding tritium analysis in water, the strengths are reliability of this analysis at low concentrations (order of Bq.L{sup -1}), robustness and simplicity, and weaknesses are linked to problems of background, conservation and contamination of samples. Concerning OBT analysis, the analysis is reliable for values around 50 Bq.kg{sup -1} of fresh sample. The weaknesses are problems of contamination, reproducibility, analysis time (2 to 6 days) and lack of reference materials. The difficulty to date is the separation between E-OBT and NE-OBT, that will need experimental validation. (authors)

  2. Packaged low-level waste verification system

    Energy Technology Data Exchange (ETDEWEB)

    Tuite, K.; Winberg, M.R.; McIsaac, C.V. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-12-31

    The Department of Energy through the National Low-Level Waste Management Program and WMG Inc. have entered into a joint development effort to design, build, and demonstrate the Packaged Low-Level Waste Verification System. Currently, states and low-level radioactive waste disposal site operators have no method to independently verify the radionuclide content of packaged low-level waste that arrives at disposal sites for disposition. At this time, the disposal site relies on the low-level waste generator shipping manifests and accompanying records to ensure that low-level waste received meets the site`s waste acceptance criteria. The subject invention provides the equipment, software, and methods to enable the independent verification of low-level waste shipping records to ensure that the site`s waste acceptance criteria are being met. The objective of the prototype system is to demonstrate a mobile system capable of independently verifying the content of packaged low-level waste.

  3. Decommissioning of a tritium-contaminated laboratory

    International Nuclear Information System (INIS)

    Harper, J.R.; Garde, R.

    1981-11-01

    A tritium laboratory facility at the Los Alamos National Laboratory, Los Alamos, New Mexico, was decommissioned in 1979. The project involved dismantling the laboratory equipment and disposing of the equipment and debris at an on-site waste disposal/storage area. The laboratory was constructed in 1953 and was in service for tritium research and fabrication of lithium tritide components until 1974. The major features of the laboratory included some 25 meters of gloveboxes and hoods, associated vacuum lines, utility lines, exhaust ducts, electrodryers, blowers, and laboratory benches. This report presents details on the decommissioning, health physics, waste management, environmental surveillance, and costs for the operation

  4. Tritium transport in lithium ceramics porous media

    International Nuclear Information System (INIS)

    Tam, S.W.; Ambrose, V.

    1991-01-01

    A random network model has been utilized to analyze the problem of tritium percolation through porous Li ceramic breeders. Local transport in each pore channel is described by a set of convection-diffusion-reaction equations. Long range transport is described by a matrix technique. The heterogeneous structure of the porous medium is accounted for via Monte Carlo methods. The model was then applied to an analysis of the relative contribution of diffusion and convective flow to tritium transport in porous lithium ceramics. 15 refs., 4 figs

  5. Polymeric media for tritium fixation. Supplement I

    International Nuclear Information System (INIS)

    Franz, J.A.; Burger, L.L.

    1976-01-01

    Procedures for the fixation of tritium as TH or THO in two different polymeric media are described. The complete procedure for THO fixation in a polyureylene-polyurethane polumer, including polymer molding procedures and leach tests is presented. The catalytic tritiation of polystyrene under very mild conditions using a rhodium catalyst is also described. Thermal stabilities and cost estimates for the polymers examined under this program are discussed. Organic polymers were found to have attractive features for the fixation and storage of concentrated tritium wastes due to the convenience of fixation procedures and favorable properties of the resulting media

  6. Measurement of tritium in dial painting industry

    International Nuclear Information System (INIS)

    Sawant, J.V.; Rudran, Kamala

    1995-01-01

    Tritium in the form of polystyrene is used coated on zinc sulphide as the active component for the manufacture of self-luminous paint. To study the radiological implication of airborne tritium in the luminous paint industry air monitoring study was conducted by cold strip method and Andersen method. Airborne particulate in different locations in luminous paint (LP) building and background areas were observed to be associated with activity median aerodynamic diameter (AMAD) of 1.8 to 5.0 um. Dose to soft tissue and lungs and effective whole body dose were evaluated. (author). 6 refs., 2 figs., 4 tabs., 2 ills

  7. Tritium projectiles for fueling magnetic fusion plasmas

    International Nuclear Information System (INIS)

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

    1995-01-01

    As part of the International Thermonuclear Engineering Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet (cylindrical projectile of frozen hydrogenic gas at a temperature in the range 6--16 K) injection system to test the mechanical and thermal properties of extruded tritium, a radioactive isotope of hydrogen. 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 and accelerate hydrogenic pellets sized for the ITER device. The TPOP-2 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; evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory that requires secondary and room containment systems. In initial tests with deuterium feed at ORNL, up to 13 pellets 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. The pellets, typically 7.4 mm in diameter and up to 11 mm in length, 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 in which tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge

  8. Development of an environmental tritium model; ETDOSE

    Energy Technology Data Exchange (ETDEWEB)

    Andoh, Mariko; Takahashi, Tomoyuki; Amano, Hikaru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    ETDOSE is a simple computer code to calculate atmospheric distribution patterns of tritium for an acute and a chronic release of HT and HTO. This code calculates tritium concentrations in air, soil, plant free water and OBT, and estimates dose impact by inhalation of air and ingestion of food. Participation in IAEA`s model validation program BIOMASS has been done using this code for BIOMASS Scenario 1. This paper shows the outline of ETDOSE and preliminary results of model comparison in BIOMASS program. (author)

  9. Tritium and the environment 3H

    International Nuclear Information System (INIS)

    2002-01-01

    After a presentation of Tritium properties (radioactivity, presence in solids, liquids and gases), this document gives comments and assessments of its natural origin, its presence in relationship with nuclear tests, nuclear power plants, processing plants and other industrial installations, its presence in waters and oceans, in the atmosphere and soils, in plants, animals and mankind. It indicates the main exposure ways (ingestion, inhalation, skin) and its biological effects. It gives an overview of the Tritium-containing waste issue, and indicates some related standards

  10. Report of the Task Group on operation Department of Energy tritium facilities

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

    This report discusses the following topics on the operation of DOE Tritium facilities: Environment, Safety, and Health Aspects of Tritium; Management of Operations and Maintenance Functions; Safe Shutdown of Tritium Facilities; Management of the Facility Safety Envelope; Maintenance of Qualified Tritium Handling Personnel; DOE Tritium Management Strategy; Radiological Control Philosophy; Implementation of DOE Requirements; Management of Tritium Residues; Inconsistent Application of Requirements for Measurement of Tritium Effluents; Interdependence of Tritium Facilities; Technical Communication among Facilities; Incorporation of Confinement Technologies into New Facilities; Operation/Management Requirements for New Tritium Facilities; and Safety Management Issues at Department of Energy Tritium Facilities.

  11. Civilian nuclear ships

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1993-03-01

    This report contains a review of the information available on nuclear powered ships, built for civilian purposes. In the introduction a short discussion is given of the reasons for the limited use of nuclear ships for these purposes. In the second section a brief review is presented of data for the three experimental/merchant ships build by the United States, Germany and Japan, i.e. NS Savannah, NS Otto Hahn and NS Mutsu. In the third section the Soviet/Russian icebreaker NS Lenin is considered. Its design, operational experience and the introduction of a new nuclear propulsion plant is reviewed. In the fourth section the newer Soviet/Russian icebreakers with nuclear propulsion are considered. Finally design of the Soviet/Russian icebreaker transport/container ship NS Sevmorput is reviewed in the fifth section. The future Russian plans for nuclear vessels for the arctic waters are briefly discussed. (au)

  12. Ship construction and welding

    CERN Document Server

    Mandal, Nisith R

    2017-01-01

    This book addresses various aspects of ship construction, from ship types and construction materials, to welding technologies and accuracy control. The contents of the book are logically organized and divided into twenty-one chapters. The book covers structural arrangement with longitudinal and transverse framing systems based on the service load, and explains basic structural elements like hatch side girders, hatch end beams, stringers, etc. along with structural subassemblies like floors, bulkheads, inner bottom, decks and shells. It presents in detail double bottom construction, wing tanks & duct keels, fore & aft end structures, etc., together with necessary illustrations. The midship sections of various ship types are introduced, together with structural continuity and alignment in ship structures. With regard to construction materials, the book discusses steel, aluminum alloys and fiber reinforced composites. Various methods of steel material preparation are discussed, and plate cutting and form...

  13. Optimization in liner shipping

    DEFF Research Database (Denmark)

    Brouer, Berit Dangaard; Karsten, Christian Vad; Pisinger, David

    2017-01-01

    Seaborne trade is the lynchpin in almost every international supply chain, and about 90% of non-bulk cargo worldwide is transported by container. In this survey we give an overview of data-driven optimization problems in liner shipping. Research in liner shipping is motivated by a need for handling...... still more complex decision problems, based on big data sets and going across several organizational entities. Moreover, liner shipping optimization problems are pushing the limits of optimization methods, creating a new breeding ground for advanced modelling and solution methods. Starting from liner...... shipping network design, we consider the problem of container routing and speed optimization. Next, we consider empty container repositioning and stowage planning as well as disruption management. In addition, the problem of bunker purchasing is considered in depth. In each section we give a clear problem...

  14. Nuclear ship accidents

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1993-05-01

    In this report available information on 28 nuclear ship accident and incidents is considered. Of these 5 deals with U.S. ships and 23 with USSR ships. The ships are in almost all cases nuclear submarines. Only events that involve the nuclear propulsion plants, radiation exposures, fires/explosions and sea water leaks into the submarines are considered. Comments are made on each of the events, and at the end of the report an attempt is made to point out the weaknesses of the submarine designs which have resulted in the accidents. It is emphasized that much of the available information is of a rather dubious nature. consequently some of the assessments made may not be correct. (au)

  15. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  16. 78 FR 77773 - Shipping Coordinating Committee; Notice of Committee Meeting

    Science.gov (United States)

    2013-12-24

    ... ships operating in polar waters --Review and modernization of the Global Maritime Distress and Safety... of relevant IMO instruments in relation to the carriage of dangerous goods in packaged form by sea...-point fuels (IGF Code) Members of the public may attend this meeting up to the seating capacity of the...

  17. Performance-oriented packaging: A guide to identifying, procuring, and using. Procurement and use of packaging using HM-181 regulations

    International Nuclear Information System (INIS)

    1994-09-01

    This document addresses procurement of nonbulk packaging used for nonradioactive, hazardous materials that are transported by highway. The basic procedure is the same when making shipments by other modes. However, other sections of the regulations may affect the packaging requirements when shipping by other modes. The variation in requirements is extensive when air transportation is selected. A packaging engineer or transportation specialist should be contacted for guidance wen transport is by a mode other than highway

  18. Towards Real Time Simulation of Ship-Ship Interaction

    DEFF Research Database (Denmark)

    Lindberg, Ole; Bingham, Harry B.; Engsig-Karup, Allan Peter

    2012-01-01

    We present recent and preliminary work directed towards the development of a simplified, physics-based model for improved simulation of ship-ship interaction that can be used for both analysis and real-time computing (i.e. with real-time constraints due to visualization). The goal is to implement...... accurate (realistic) and much faster ship-wave and ship-ship simulations than are currently possible. The coupling of simulation with visualization should improve the visual experience such that it can be perceived as more realistic in training. Today the state-of-art in real-time ship-ship interaction...... (GPUs) can provide the basis for efficient simulations in combination with an accurate free-surface model for Ship-Ship simulation. Another area of application is the determination of wave disturbances from a ship in a coastal environment, channels and harbours. The model proposed in the following can...

  19. SHIPPING REQUESTS ON EDH

    CERN Multimedia

    Logistics Group

    2001-01-01

    Users are informed that as from 1 September 2001 all Shipping Requests must be made on EDH using the appropriate electronic form. The submission of user requests directly into EDH will help rationalise the activities of the Shipping Service (Import & Export), with requests being automatically forwarded to hierarchical supervisors thereby improving the processing speed and facilitating the follow-up. Thank you for your collaboration.

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

    Science.gov (United States)

    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.

  1. Recent situations around nuclear ships

    International Nuclear Information System (INIS)

    Mizuno, Hiroshi

    1978-01-01

    The philosophy when the safety standard for nuclear ships is drawn up and the international rules specifically for nuclear ships are summarized. As for the safety standard for nuclear ships, the safety requirements for ordinary ships, for the ships transporting nuclear reactors, for ordinary nuclear reactors, and for the reactors moving around the seas must be included. As for the international rules for nuclear ships, there are chapter 8 ''Nuclear ships'' in the International Convention on the Safety of Life at Sea, 1960 and 1974, and Safety Consideration in the Use of Ports and Approaches by Nuclear Merchant Ships. Also there are national rules and standards in Japan and foreign countries. One of the means to explore the practicality of nuclear ships is the investigation of the economy. At this time, the social merits and demerits of nuclear ships must be compared with conventional ships by taking total expenses into account without omission. When oil is depleted, the age of nuclear ships will not necessarily begin, and the will be still some competitors. The investigations concerning the economy of nuclear ships have been carried out in various countries. The present state of the development of nuclear ships in Japan and foreign countries is explained. Many conferences and symposia have been held concerning nuclear ships, and those held recently are enumerated. The realization of nuclear ship age cannot be anticipated from existing papers and shipbuilding projects. (Kako, I.)

  2. Outer Dynamics of Ship Collisions

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup

    1996-01-01

    The purpose of these notes is to present analysis procedures for the motion of ships during ship-ship collisions and for ship collisons with offshore structures. The aim is to estimate that part of the lost kinetic energy which will have to be absorbed by rupture and plastic damage of the colliding...

  3. Outer Dynamics of Ship Collisions

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup

    1996-01-01

    The purpose is to present analysis procedures for the motion of ships during ship-ship collisions and for ship collisions with offshore structures. The aim is to estimate that part of the lost kinetic energy which will have to be absorbed by rupture and plastic damage of the colliding structures....

  4. Tritium dynamics in soils and plants at a tritium processing facility in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Mihok, S.; St-Amanat, N.; Kwamena, N.O. [Canadian Nuclear Safety Commission (Canada); Clark, I.; Wilk, M.; Lapp, A. [University of Ottawa (Canada)

    2014-07-01

    The dynamics of tritium released as tritiated water (HTO) have been studied extensively with results incorporated into environmental models such as CSA N288.1 used for regulatory purposes in Canada. The dispersion of tritiated gas (HT) and rates of oxidation to HTO have been studied under controlled conditions, but there are few studies under natural conditions. HT is a major component of the tritium released from a gaseous tritium light manufacturing facility in Canada (CNSC INFO-0798). To support the improvement of models, a garden was set up in one summer near this facility in a spot with tritium in air averaging ∼ 5 Bq/m{sup 3} HTO (passive diffusion monitors). Atmospheric stack releases (575 GBq/week) were recorded weekly. HT releases occur mainly during working hours with an HT:HTO ratio of 2.6 as measured at the stack. Soils and plants (leaves/stems and roots/tubers) were sampled for HTO and organically-bound tritium (OBT) weekly. Active day-night monitoring of air was conducted to interpret tritium dynamics relative to weather and solar radiation. The experimental design included a plot of natural grass/soil, contrasted with grass (sod) and Swiss chard, pole beans and potatoes grown in barrels under different irrigation regimes (in local topsoil at 29 Bq/L HTO, 105 Bq/L OBT). All treatments were exposed to rain (80 Bq/L) and atmospheric releases of tritium (weekdays), and reflux of tritium from soils (initial conditions of 284 Bq/L HTO, 3,644 Bq/L OBT) from 20 years of operations. Three irrigation regimes were used for barrel plants to mimic home garden management: rain only, low tritium tap water (5 Bq/L), and high tritium well water (mean 10,013 Bq/L). This design provided a range of plants and starting conditions with contrasts in initial HTO/OBT activity in soils, and major tritium inputs from air versus water. Controls were two home gardens far from any tritium sources. Active air monitoring indicated that the plume was only occasionally present for

  5. Incorporation of tritium in milk lipids after feeding organically bound tritium to cows

    International Nuclear Information System (INIS)

    Rochalska, M.; Hoek, J. van den

    1982-01-01

    Hay labelled with organically bound tritium was given to two cows for a period of 26 to 28 days. During hay feeding and at different times thereafter, lipids (fatty acids, cholesterol, glycerol, choline phospholipids, other phospholipids, flycolipids and gangliosides) were isolated from milk fat, and their total and specific activities were determined. During tritium administration, fatty acids and cholesterol contained the highest total activity, but the specific activity was highest in cholesterol and choline phospholipids. Activity decreased most rapidly for fatty acids and cholesterol, so that at 56 and 182 days after termination of 3 H feedings, phospholipids and glycolipids made an important contribution to lipid activity in milk. Regression analysis of the values for tritium activity in milk fat samples after stopping tritium administration, showed that three components with different half lives could be distinguished. The differences in metabolic behaviour of the various lipids in milk fat are mainly concerned with their relative participation in these components. (author)

  6. Development of nuclear micro-battery with solid tritium source

    International Nuclear Information System (INIS)

    Lee, Sook-Kyung; Son, Soon-Hwan; Kim, KwangSin; Park, Jong-Wan; Lim, Hun; Lee, Jae-Min; Chung, Eun-Su

    2009-01-01

    A micro-battery powered by tritium is being developed to utilize tritium produced from the Wolsong Tritium Removal Facility. The 3D p-n junction device has been designed and fabricated for energy conversion. Titanium tritide is adopted to increase tritium density and safety. Sub micron films or nano-powders of titanium tritide is applied on silicon semiconductor device to reduce the self absorption of beta rays. Until now protium has been used instead of tritium for safety. Hydrogen was absorbed up to atomic ratio of ∼1.3 and ∼1.7 in titanium powders and films, respectively.

  7. Quantitative determination of tritium in metals and oxides

    International Nuclear Information System (INIS)

    Vance, D.E.; Smith, M.E.; Waterbury, G.R.

    1979-04-01

    Metallic samples are analyzed for tritium by heating the sample at 1225 K in a moist oxygen stream. The volatile products are trapped and the tritium is quantitatively determined by scintillation spectroscopy. The method is used to determine less than 1 ppb of tritium in 100-mg samples of lithium, iron, nickel, cerium, plutonium, and plutonium dioxide. Analysis of 18 cuts of a tritium-zirconium, copper foil standard over a 3-yr period showed a tritium content of 45 ppM and a standard deviation of 6 ppM

  8. Tritium experience in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Brooks, J.N. [Argonne National Lab., IL (United States); Hogan, J. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  9. TFTR D-T experience with tritium radioactivity during maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Gilbert, J.; Ascione, G.; Birckbichler, D.; Elwood, S.; Flournoy, R.; Stencel, J.; Tilson, C. [Princeton Plasma Physics Lab., NJ (United States)

    1995-10-01

    The Tokamak Fusion Test Reactor facility began operations with trace tritium consisting of the delivery, storage, injection, and subsequent processing of tritium gas in support of the D-T fusion program. The tritium is transferred throughout the facility using vacuum pumping systems and expansion volumes. This paper discusses the operational health physics program with regard to the performance of maintenance on tritium contaminated systems. Data and findings are provided from the maintenance situations ranging from work on small volume piping to large volume neutral beam systems. Results and comparisons of the tritium contamination levels, airborne radioactivity levels, and oil concentrations are presented for these systems. 4 refs., 1 fig.

  10. Tritium time series in precipitation of Rm. Valcea, Romania.

    Science.gov (United States)

    Varlam, Carmen; Duliu, Octavian G; Faurescu, Ionut; Vagner, Irina; Faurescu, Denisa

    2016-01-01

    Following tritium concentration records in precipitation for the period 1999-2013 and tritium concentration behaviour during this period for the Ramnicu Valcea (Rm. Valcea) location, the tritium level of individual precipitations of the late spring and summer for the 2009-2013 period was investigated. Despite good correlation between monthly mean tritium concentrations and monthly mean precipitations over the 15-year period of observations (Pearson coefficient 0.87), the individual precipitations had no linear correlation between the tritium concentration and the amount of precipitation.

  11. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  12. Elements of thought on the health risk associated to tritium

    International Nuclear Information System (INIS)

    2009-01-01

    This report addresses and analyses the health problematic set by tritium and assesses the robustness of the radiation protection system with respect to this radionuclide by highlighting the lack of scientific knowledge on biological effects, and researches to be promoted. After a presentation of epidemiologic and dosimetric approaches of the radiological risk assessment, the authors discuss results and knowledge gained by epidemiologic studies on the risk associated to tritium for mankind, and discuss the knowledge on biological effects of tritium and on the relative biological effectiveness of tritium. The report finally discusses the possibility of reconsidering the radiation weighting factor in the case of tritium

  13. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    Washington TRU Solutions, LLC

    2003-01-01

    The purpose of this program guidance document is to provide technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the SARP and/or C of C shall govern. The C of C states: ''...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, ''Operating Procedures,'' of the application.'' It further states: ''...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, ''Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC approved, users need to be familiar with 10 CFR (section) 71.11, ''Deliberate Misconduct.'' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions. Following these instructions assures that operations are safe and meet the requirements of the SARP. This document is available on the Internet at: ttp://www.ws/library/t2omi/t2omi.htm. Users are responsible for ensuring they are using the current revision and change notices. Sites may prepare their own document using the word

  14. Improving tritium exposure reconstructions using accelerator mass spectrometry

    Science.gov (United States)

    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

  15. Tritium release experiments with CATS and numerical simulation

    Energy Technology Data Exchange (ETDEWEB)

    Munakata, Kenzo, E-mail: kenzo@gipc.akita-u.ac.jp [Faculty of Engineering and Resource Sciences, Akita University, Tegata-gakuen-cho 1-1, Akita 010-8502 (Japan); Wajima, Takaaki; Hara, Keisuke; Wada, Kohei [Faculty of Engineering and Resource Sciences, Akita University, Tegata-gakuen-cho 1-1, Akita 010-8502 (Japan); Takeishi, Toshiharu; Shinozaki, Yohei; Mochizuki, Kazuhiro; Katekari, Kenichi [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581 (Japan); Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko [Tritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2010-12-15

    In D-T fusion power plants, large amounts of tritium would be handled. Tritium is the radioisotope of protium, and is easily taken into the human body, and thus the behavior of tritium accidentally released in fusion power plants should be studied for the safety design and radioprotection of workers. Therefore, it is necessary to investigate the behavior of tritium released into large rooms with objectives, since complex flow fields should exist in such rooms and they could influence the ventilation of the air containing released tritium. Thus, tritium release experiments were conducted using Caisson Assembly for Tritium Safety Study (CATS) in TPL/JAEA. Some data were taken for tritium behavior in the ventilated area and response of tritium monitors. In the experiments, approximately 17 GBq of tritium was released into Caisson with the total volume of 12 m{sup 3}, and the room was ventilated at the rate of 12 m{sup 3}/h after release of tritium. It was found that placement of an objective in the vessel substantially affects decontamination efficiency. With regard to an experimental result, numerical calculation was performed and the experimental result and the result of numerical calculation were compared, which indicates that experimental results are qualitatively reproduced by numerical calculation. However, further R and D needs to be carried out for quantitative reproduction of the experimental results.

  16. Tritium burning in inertial electrostatic confinement fusion facility

    Energy Technology Data Exchange (ETDEWEB)

    Ohnishi, Masami, E-mail: onishi@kansai-u.ac.jp [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Yamamoto, Yasushi; Osawa, Hodaka [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Hatano, Yuji; Torikai, Yuji [Hydrogen Isotope Science Center, University of Toyama, Gofuku, Toyama 930-8555 (Japan); Murata, Isao [Faculty of Engineering Environment and Energy Department, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Masuda, Kai [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hotta, Eiki [Interdisciplinary Graduate School of Science and Engineering, Tokyo Institute of Technology, 4259 Nagatsuda-cho, Midori-ku, Yokohama 226-8503 (Japan)

    2016-11-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10{sup 8}(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  17. Preliminary Tritium Management Design Activities at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Logsdon, Randall J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McFarlane, Joanna [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritium mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.

  18. Tritium concentrations in some European foods

    International Nuclear Information System (INIS)

    Hisamatsu, Shun'ichi; Takizawa, Yukio

    1989-01-01

    Tritium concentrations in some European foods imported into Japan are reported here. The specific activity ratio of so-called tissue-bound 3 H to free water 3 H was found to be lower than that previously reported by European and American research groups. This discrepancy might be caused by the difference in sampling year and/or by analytical problems. (author)

  19. Release of gaseous tritium during reprocessing

    International Nuclear Information System (INIS)

    Bruecher, H.; Hartmann, K.

    1983-01-01

    About 50% of the tritium put through an LWR reprocessing plant is obtained as tritium-bearing water, HTO. Gaseous tritium, HT has a radiotoxicity which is by 4 orders of magnitude lower than that of HTO. A possibility for the removal of HTO could therefore be its conversion into the gas phase with subsequent emission of the HT into the atmosphere. However, model computations which are, in part, supported by experimental data reveal that the radiation exposure caused by HT release is only by about one order of magnitude below that caused by HTO. This is being attributed to the relatively quick reoxidation of HT by soil bacteria. Two alternatives for producing HT from HTO (electrolysis; voloxidation with subsequent electrolysis) are presented and compared with the reference process of deep-well injection of HTO. The authors come to the conclusion that tritium removal by HT release into the atmosphere cannot be recommended at present under either radiological or economic aspects. (orig.) [de

  20. Permeability of protective coatings to tritium

    International Nuclear Information System (INIS)

    Braun, J.M.

    1987-10-01

    The permeability of four protective coatings to tritium gas and tritiated water was investigated. The coatings, including two epoxies, one vinyl and one urethane, were selected for their suitability in CANDU plant service in Ontario Hydro. Sorption rates of tritium gas into the coatings were considerably larger than for tritiated water, by as much as three to four orders of magnitude. However, as a result of the very large solubility of tritiated water in the coatings, the overall permeability to tritium gas and tritiated water are comparable, being somewhat larger for HTO. Marked differences were also evident among the four coatings, the vinyl proving to be unique in behaviour and morphology. Because of a highly porous surface structure water condensation takes place at high relative humidities, leading to an abnormally high retention of free water. Desorption rates from the four coatings were otherwise quite similar. Of practical importance was the observation that more effective desorption of tritiated water could be carried out at relatively high humidities, in this case 60%. It was believed that isotopic exchange was responsible for this phenomenon. It appears that epoxy coatings having a high pigment-to-binder ratio are most suited for coating concrete in tritium handling facilities

  1. Frustrated Lewis pairs-assisted tritium labeling

    Czech Academy of Sciences Publication Activity Database

    Marek, Aleš; Široká, Sabina; Elbert, Tomáš

    2016-01-01

    Roč. 14, č. 5 (2016), s. 219 ISSN 2336-7202. [Sjezd českých a slovenských chemických společností /68./. 04.09.2016-07.09.2016, Praha] Institutional support: RVO:61388963 Keywords : frustrated Lewis pairs * one-pot synthesis * tritium -labeling Subject RIV: CC - Organic Chemistry

  2. Stereo and regioselectivity in ''Activated'' tritium reactions

    International Nuclear Information System (INIS)

    Ehrenkaufer, R.L.E.; Hembree, W.C.; Wolf, A.P.

    1988-01-01

    To investigate the stereo and positional selectivity of the microwave discharge activation (MDA) method, the tritium labeling of several amino acids was undertaken. The labeling of L-valine and the diastereomeric pair L-isoleucine and L-alloisoleucine showed less than statistical labeling at the α-amino C-H position mostly with retention of configuration. Labeling predominated at the single β C-H tertiary (methyne) position. The labeling of L-valine and L-proline with and without positive charge on the α-amino group resulted in large increases in specific activity (greater than 10-fold) when positive charge was removed by labeling them as their sodium carboxylate salts. Tritium NMR of L-proline labeled both as its zwitterion and sodium salt showed also large differences in the tritium distribution within the molecule. The distribution preferences in each of the charge states are suggestive of labeling by an electrophilic like tritium species(s). 16 refs., 5 tabs

  3. Carbon 14 and tritium radioactivity of alcohols

    International Nuclear Information System (INIS)

    Guerain, J.; Tourliere, S.

    1975-01-01

    The method of measuring carbon 14 radioactivity of alcohols has been perfected in order to establish the correct determination of synthetic alcohol added to fermentation alcohol. The specific carbon and tritium activity of alcohol of different origins have been determined for 1973 and 1974. The Suess effect and nuclear fall-out are observed [fr

  4. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    International Nuclear Information System (INIS)

    Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

    1995-01-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

  5. Tritium forms discrimination in ryegrass under constant tritium exposure: From seed germination to seedling autotrophy.

    Science.gov (United States)

    Renard, H; Maro, D; Le Dizès, S; Escobar-Gutiérrez, A; Voiseux, C; Solier, L; Hébert, D; Rozet, M; Cossonnet, C; Barillot, R

    2017-10-01

    Uncertainties remain regarding the fate of atmospheric tritium after it has been assimilated in grasslands (ryegrass) in the form of TFWT (Tissue Free Water Tritium) or OBT (Organically Bound Tritium). One such uncertainty relates to the tritium forms discrimination during transfer from TFWT to OBT resulting from photosynthesis (OBT photo ), corresponding to the OBT photo /TFWT ratio. In this study, the OBT/TFWT ratio is determined by experiments in the laboratory using a ryegrass model and hydroponic cultures, with constant activity of tritium in the form of tritiated water (denoted as HTO) in the "water" compartment (liquid HTO) and "air" compartment (HTO vapour in the air). The OBT photo /TFWT ratio and the exchangeable OBT fraction are measured for three parts of the plant: the leaf, seed and root. Plant growth is modelled using dehydrated biomass measurements taken over time in the laboratory and integrating physiological functions of the plant during the first ten days after germination. The results suggest that there is no measurable discrimination of tritium in the plant organic matter produced by photosynthesis. Copyright © 2017 Elsevier Ltd. All rights reserved.

  6. Tokamak fusion reactors with less than full tritium breeding

    International Nuclear Information System (INIS)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed

  7. Plasma wall interaction and tritium retention in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Amarescu, E.; Ascione, G.

    1996-01-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced an impressive number of record breaking results including core fusion power, ∼ 2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently, the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. The authors present recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues

  8. Plasma wall interaction and tritium retention in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Amarescu, E.; Ascione, G.

    1997-01-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, ∝2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented. (orig.)

  9. Investigation of tritium in groundwater at Site 300

    International Nuclear Information System (INIS)

    Buddemeier, R.W.

    1985-01-01

    In 1984, landfill monitoring wells at Site 300, a Lawrence Livermore National Laboratory (LLNL) explosive test site, revealed the presence of groundwater contaminated with tritium. These tritium levels were in excess of the State of California drinking water standard. A major investigation was initiated that included a search of records concerning tritium use, disposal, and previous analyses, and a survey of tritium levels in soil, vegetation, and water in contaminated and potentially contaminated areas. Over 50 boreholes were drilled for this investigation to characterize the local hydrogeology and tritium distributions, and a network of soil moisture and groundwater monitoring points was installed. This report presents the work completed through the end of September 1985: the records search; records for drilling completed as part of this study; characterization of the geology, hydrology, and tritium distributions in the contaminated area; and an initial assessment of the probable tritium sources, pathways, and migration rates. 19 refs

  10. Investigation of tritium in groundwater at Site 300

    Energy Technology Data Exchange (ETDEWEB)

    Buddemeier, R.W.

    1985-12-30

    In 1984, landfill monitoring wells at Site 300, a Lawrence Livermore National Laboratory (LLNL) explosive test site, revealed the presence of groundwater contaminated with tritium. These tritium levels were in excess of the State of California drinking water standard. A major investigation was initiated that included a search of records concerning tritium use, disposal, and previous analyses, and a survey of tritium levels in soil, vegetation, and water in contaminated and potentially contaminated areas. Over 50 boreholes were drilled for this investigation to characterize the local hydrogeology and tritium distributions, and a network of soil moisture and groundwater monitoring points was installed. This report presents the work completed through the end of September 1985: the records search; records for drilling completed as part of this study; characterization of the geology, hydrology, and tritium distributions in the contaminated area; and an initial assessment of the probable tritium sources, pathways, and migration rates. 19 refs.

  11. Plasma wall interaction and tritium retention in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H.; Amarescu, E.; Ascione, G. [and others

    1997-02-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, {proportional_to}2 MW/m{sup 3}, comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented. (orig.).

  12. Plasma wall interaction and tritium retention in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C. H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Amarescu, E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ascione, G. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Synakowski, E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    1996-05-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced an impressive number of record breaking results including core fusion power, ~ 2 MW/m³, comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently, the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. The authors present recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues.

  13. Measurement of Tritium Surface Distribution on TFTR Bumper Limiter Tiles

    International Nuclear Information System (INIS)

    Sugiyama, K.; Tanabe, T.; Skinner, C.H.; Gentile, C.A.

    2004-01-01

    The tritium surface distribution on graphite tiles used in the Tokamak Fusion Test Reactor (TFTR) bumper limiter and exposed to TFTR deuterium-tritium (D-T) discharges from 1993 to 1997 was measured by the Tritium Imaging Plate Technique (TIPT). The TFTR bumper limiter shows both re-/co-deposition and erosion. The tritium images for all tiles measured are strongly correlated with erosion and deposition patterns, and long-term tritium retention was found in the re-/co-depositions and flakes. The CFC tiles located at erosion dominated areas clearly showed their woven structure in their tritium images owing to different erosion yields between fibers and matrix. Significantly high tritium retention was observed on all sides of the erosion tiles, indicating carbon transport via repetition of local erosion/deposition cycles

  14. Accidents in nuclear ships

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10 -3 per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au)

  15. Impact analysis of shipping casks

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.

    1989-01-01

    Shipping casks are being used in the United States Department of Energy to transport irradiated experiments, reactor fuel, radioactive waste, etc. One of the critical requirements in shipping cask analysis is the necessity to withstand severe impact environments. It is still conventional to develop the design and to verify the design requirements by hand calculations. Full three dimensional computations of impact scenarios have been performed but they are too expensive and time consuming for design purposes. Typically, on the order of more than an hour of CRAY time is required for a detailed, three dimensional analysis. The paper describes how simpler two- and three-dimensional models can be used to provide an intermediate level of detail between full three dimensional finite element calculations and hand calculations. The regulation that is examined here is: 10 CFR-71.73 hypothetical accident conditions, free drop. Free drop for an accident condition of a Class I package (approximate weight of 22,000 lb) is defined as a 30 foot drop onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. Three free drop scenarios are analyzed to assess the integrity of the cask when subjected to large bending and axial stresses. These three drop scenarios are: (1) a thirty foot axial drop on either end, (2) a thirty foot oblique angle drop with the cask having several different orientations from the vertical with impact on the top end cask corner, and (3) a thirty foot side drop with simultaneous impact on the strength of the various components that comprise the cask. The predicted levels of deformation and stresses in the cask will be used to assess the potential damage level. 5 refs., 5 figs., 1 tab

  16. System certification: An alternative to package certification?

    International Nuclear Information System (INIS)

    Luna, R.E.; Jefferson, R.J.

    1992-01-01

    One precept of the current radioactive material transportation regulations is that the package is the primary protection for the public. A packaging is chosen to provide containment, shielding, and criticality control suitable to the quantity and characteristics of the radionuclide being transported. Occasionally, radioactive materials requiring transport are not of a mass or size that would allow the materials to be shipped in an appropriate packaging. This is a particular problem for materials that should be shipped in a Type B package, but because such packages are designed and certified for specific contents, the package is usually fairly expensive, available in relatively small numbers, and often requires a fairly long period to achieve certification or amended certification for new contents. Where the shipment to be made is relatively infrequent, there may be economic and time penalties that may hamper shipment or force the shipper into uneconomic or high risk options. However, there is recognition of such situations in the International Atomic Energy Agency (IAEA) regulations under the provisions for Special Arrangement

  17. Development of methodology for certification of Type B shipping containers using analytical and testing techniques

    International Nuclear Information System (INIS)

    Sharp, R.R.; Varley, D.T.

    1993-01-01

    The use of multidisciplinary teams to develop Type B shipping containers improves the quality and reliability of these reusable packagings. Including the people involved in all aspects of the design, certification and use of the package leads to more innovative, user-friendly containers. Concurrent use of testing and analysis allows engineers to more fully characterize a shipping container's responses to the environments given in the regulations, and provides a strong basis for certification. The combination of the input and output of these efforts should provide a general methodology that designers of Type B radioactive material shipping containers can utilize to optimize and certify their designs. (J.P.N.)

  18. Finite element analysis of container ship's cargo hold using ANSYS and POSEIDON software

    Science.gov (United States)

    Tanny, Tania Tamiz; Akter, Naznin; Amin, Osman Md.

    2017-12-01

    Nowadays ship structural analysis has become an integral part of the preliminary ship design providing further support for the development and detail design of ship structures. Structural analyses of container ship's cargo holds are carried out for the balancing of their safety and capacity, as those ships are exposed to the high risk of structural damage during voyage. Two different design methodologies have been considered for the structural analysis of a container ship's cargo hold. One is rule-based methodology and the other is a more conventional software based analyses. The rule based analysis is done by DNV-GL's software POSEIDON and the conventional package based analysis is done by ANSYS structural module. Both methods have been applied to analyze some of the mechanical properties of the model such as total deformation, stress-strain distribution, Von Mises stress, Fatigue etc., following different design bases and approaches, to indicate some guidance's for further improvements in ship structural design.

  19. A Low-Level Real-Time In Situ Monitoring System for Tritium in Groundwater and Vadose Zone

    Science.gov (United States)

    Santo, J. T.; Levitt, D. G.

    2002-12-01

    Tritium is a radioactive isotope of hydrogen produced as a by-product of the nuclear fuel cycle. It is also an integral part of the nuclear weapons industry and has been released into the environment through both the production and testing of nuclear weapons. There are many sites across the DOE complex where tritium has been released into the subsurface through the disposal of radioactive waste and at the Nevada Test Site, through the underground testing of nuclear weapons. Numerous DOE facilities have an on-going regulatory need to be able to monitor tritium concentrations in groundwater within deep hydrologic zones and in the shallower non-saturated vadose zone beneath waste disposal pits and shafts and other release sites. Typical access to groundwater is through deep monitoring wells and situated in remote locations. In response to this need, Science and Engineering Associates, Inc. (SEA) and its subcontractor, the University of Nevada Las Vegas (UNLV) Harry Reid Center (HRC) for Environmental Studies has conducted the applied research and engineering and produced a real time, in situ monitoring system for the detection and measurement of low levels of tritium in the groundwater and in the shallower vadose zone. The monitoring system has been deployed to measure tritium in both the vadose zone near a subsurface radioactive waste package and the groundwater in a deep hydrologic reservoir at the Nevada Test Site. The monitoring system has been designed to detect tritium in the subsurface below federal and/or state regulatory limits for safe drinking water and has been successfully demonstrated. The development effort is being funded through the U.S. Department of Energy, National Energy Technology Laboratory and the DOE Nevada Operations Office Advanced Monitoring Systems Initiative (AMSI).

  20. Tritium concentrations in natural waters in Japan before use of a large quantity of tritium on its fusion program

    International Nuclear Information System (INIS)

    Kaji, Toshio; Momoshima, Noriyuki; Takashima, Yoshimasa.

    1989-01-01

    To clarify environmental tritium levels in Japan before use of a large quantity of tritium on its fusion program, the authors analyzed the tritium concentrations in various water samples, such as rain, river, lake, coastal sea and deep sea waters in Japan. The tritium concentrations in rain water were high at higher latitude. The definite differences of the tritium concentrations due to the weather conditions or seasons were not observed. The average tritium concentration in river water was 51.5 pCi/l in 1982 and that in lake water was 63.5 pCi/l in 1983. The vertical profiles of the tritium concentrations in the representative lakes were almost homogeneous except surface water. The average tritium concentrations in coastal seawater were about 20 pCi/l in both 1982 and 1983. The tendency of the increased tritium level with latitude as reported in literature was not observed by these experiments. Tritium levels in natural water in small isolated islands were lower than those at other places. In the Japan Sea, it was recognized that tritium was distributed down to around 2000 m in depth. This means that the more active vertical mixing of water masses than that in the Pacific Ocean is taking place. (author)