International Nuclear Information System (INIS)

The TVR heavy water research reactor was deployed at Moscow Institute of Theoretical and Experimental Physics. In 1990, the final batch of the spent nuclear fuel from this reactor was shipped to Production Association (PA) 'Mayak' for reprocessing. The SNF removal was a stage of the reactor decommissioning activities. The designs of the TVR reactor and its fuel elements are similar to the RA reactor designs. Two ways of the RA reactor SNF transportation to PA 'Mayak' have been considered: in aluminum barrels and in additional canisters using respectively TUK-32 and TUK-19 shipping casks. The practical experience and the equipment used to prepare for the TVR reactor SNF removal can be helpful to the RA reactor personnel in finding the best way to perform these engineering operations. (author)

International Science & Technology Center (ISTC)

Safe Transference of Spent Fuel Assemblies from Near-Reactor Storage Pools to Long-term "Dry" Storage

International Nuclear Information System (INIS)

Peculiarities of Kurchatov Institute WWR-2 and TR research reactors spent fuel treating and transportation for radiochemical processing are stated. Spent fuels were performed as fuel assemblies of different forms and containing similar fuel elements: EhK-10 with 10% enrichment UO2-Mg fuel kernels or S-36 with 36% enrichment U-Al alloys. Spent fuel storage conditions are described. Features of developed procedures for identification of fuel assemblies by type of fuel elements are given. Transport package TUK-19 for loading and transportation of spent fuel for processing was chosen. Details of spent fuel loading in TUK-19 that is conducted by personnel under protective sheet of ...

Energy Technology Data Exchange (ETDEWEB)

No abstract prepared.

International Nuclear Information System (INIS)

... Russian Federation Balakin, IM Dobrotvorskij, VV OAO SverdNIIkhimmash,

International Nuclear Information System (INIS)

Russian (Dec 1973). USSR Chmovzh, VE Lipshtejn, RA Zalkind, I.Ya. Dik,

Science.gov (United States)

This plan describes the Quality Assurance Program which will be applied to spent fuel handling and packaging tests conducted at the Area 25 E-MAD facility, Nevada Test Site (NTS).

International Nuclear Information System (INIS)

The condition of spent nuclear fuel (SNF) in wet storage at ten Soviet-designed research reactors has been assessed in the light of international experience in order to identify any associated safety issues. These reactors use Al-clad UO2-Al or U-Al alloy dispersion fuels of ?20% enrichment that were fabricated in Russia; the reactors have been in operation since 1955-70. Although originally sent for reprocessing, much of the SNF generated over the last 25-30 years has been stored in fuel storage pools (FSPs) of variable water quality. The external condition of wet-stored SNF assemblies from the reactors surveyed varied from significant failure due to galvanic corrosion that was driven by poor water quality, through gradual pitting caused by slightly impure water, to a stable condition of no observable change in the oxidized Al alloy surface of the irradiated fuel. SNF stability in ...

International Nuclear Information System (INIS)

Approximately 8% of the spent nuclear fuel owned by the US Department of Energy is clad with aluminum alloys. The spent fuel must be either reprocessed or temporarily stored in wet or dry storage systems until a decision is made on final disposition in a repository. There are corrosion issues associated with the aluminum cladding regardless of the disposition pathway selected. This paper discusses those issues and provides data and analysis to demonstrate that control of corrosion induced degradation in aluminum clad spent fuels can be achieved through relatively simple engineering practices.

International Nuclear Information System (INIS)

The possibility of separation in plasma-mass-separator POMS-E-3 spent nuclear fuel (SNF) in 3 fractions: transuranic elements, and two groups of fission products. New scheme of compact energy-mass analyzer for monitoring the separation process spent nuclear fuel in the POMS-E- 3 offered.

Energy Technology Data Exchange (ETDEWEB)

There are 4 CANDU reactors at Wolsung site together with a spent fuel dry storage associated with unit 1. These CANDU reactors, classified as On-Load Reactor (OLR) for Safeguards application, change 16- 24 fuel bundles with fresh fuel in everyday. Especially, the spent fuel bundles are transferred from spent fuel bays to dry storage throughout a year because of the insufficient capacity of spent fuel pond. Safeguards inspectors verify the spent fuel transfer to meet safeguards purposes according to the safeguards criteria by means of inspector's presence during the transfer campaign. For the verification, 60-80 person-days of inspection (PDIs) are needed during approximately 3 months for each unit. In order to reduce the ...

Energy Technology Data Exchange (ETDEWEB)

An inspection system was developed for spent fuel assemblies of PWR so that to check their general state, perform dimensional control and measure oxide layer thickness of peripheral rods. (orig./HP)

International Nuclear Information System (INIS)

A pair of bundles is placed in a shuttle tube, which is enclosed by a carriage tube kept inside the Shuttle Transfer Station (STS). It takes about 90 minutes for the spent fuel bundles to travel from the reactor channel to Transfer Magazine (TM). Subsequently, the dry transfer operation takes about 4 minutes. An emergency air cooling flow rate of 600 m3/hr is supplied to cool the spent fuel bundles after they reach the STS, following lapse of 4 minutes of spent fuel dry transfer, in case the bundles are not submerged in the light water in STS. A thermal and hydraulic safety analysis has been done to estimate the maximum sheath temperature, if the spent fuel bundles are stuck at a location of 30 mm below the normal location aligned to the TM port. This position of the spent fuel ...

International Nuclear Information System (INIS)

We provide a detailed overview of an on-going, multinational test programme that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolised materials plus volatilised fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy/density device. The programme participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research programme. Sandia National Laboratories has the lead role for conducting this research programme; test programme support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. We ...

International Nuclear Information System (INIS)

The authors provide a detailed overview of an on-going, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy-density device. The program participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research program. Sandia National Laboratories has the lead role for conducting this research program; test program support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. The ...

International Nuclear Information System (INIS)

Power density measurements in the critical facility RA-8 are presented. These measurements were the first systematic use of the reactor. A measurement system was designed, built and proved for this goal. Power profiles are showed and the results are compared with calculated values. (author)

International Nuclear Information System (INIS)

Scenarios are developed for spent fuel that is taken out of the nuclear reactor and stored for 20 years before reprocessing, and three storage techniques which use a water pool, dry cask or vault are compared with respect to their costs. The storage price (storage cost per kilogram of spent fuel), which is employed as the economic index, is calculated on the assumption that all the charge is paid when the spent fuel is brought in the storage facilities. Four scenarios are assumed for spent PWR and BWR fuels to be stored in at-reactor (AT) or away-from-reactor (AFR) facilities. The capital costs cover the buildings of the storage facilities, equipment, decommissioning, casks for transportation and storage (for cask storage) and casks for transportation (for water pool or vault storage). Costs for operation and maintenance ...

Energy Technology Data Exchange (ETDEWEB)

In Sweden, spent nuclear fuel will be encapsulated and placed in a deep geological repository. In this procedure, reliable and accurate spent fuel data such as discharge burnup, cooling time and residual heat must be available. The gamma scanning method was proposed in earlier work as a fast and reliable method for the experimental determination of such spent fuel data. This thesis is focused on the recent achievements in the development of a pilot gamma scanning system and its application in measuring spent fuel residual heat. The achievements include the development of dedicated spectroscopic data-acquisition and analysis software and the use of a specially designed calorimeter for calibrating the gamma scanning system. The pilot system is described, including an evaluation of the performance of the spectrum analysis ...

Energy Technology Data Exchange (ETDEWEB)

Norwegian authorities regard with some disquiet the possibility of a criticality accident in a ship propulsion reactor core at the Kola coast. Along this coast, in land storages, floating storages and in submarines taken out of service, the total number of spent fuel reactor cores amount to two hundred. The total Cs-137 radioactivity in spent ship propulsion reactor fuel at the Kola peninsula can be assessed to 600,000 TBq. A worst case release may amount to more than 5,000 TBq Cs-137, a quantity which under unfavourable conditions might cause serious contamination locally and even across the border to Norway.

International Nuclear Information System (INIS)

The objective was to find out whether international research has now provided sufficiently reliable models to assess the corrosion behavior of spent fuel in groundwater and by this to contribute to answering the question whether the highly radioactive used fuel from nuclear reactors can be disposed of safely in a geological repository. Principal project results are described in the paper

Energy Technology Data Exchange (ETDEWEB)

The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and ...

International Nuclear Information System (INIS)

German 2001 p. 253-258 Germany Hages, P. Stratmann, W. STEAG encotec

International Nuclear Information System (INIS)

This document provides the detailed design requirements for the Spent Nuclear Fuel Project Cold Vacuum Drying Facility. Process, safety, and quality assurance requirements and interfaces are specified.

Energy Technology Data Exchange (ETDEWEB)

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this ...

Energy Technology Data Exchange (ETDEWEB)

The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission ...

Energy Technology Data Exchange (ETDEWEB)

Thermal-fluid flow analysis and demonstration test were performed for a spent fuel storage system. The commercial computational fluid dynamics (CFD) code, FLUENT was used for the numerical analysis. Effective thermal conductivities of a spent fuel assembly and a fuel basket were derived to optimize a thermal analysis model. Also, a porous model, which can simplify a complex configuration of a fuel assembly, was used in the thermal analysis. Demonstration test were performed to verify the thermal analysis method and procedure using a half scaled-down model and an electrically heated dummy fuel. The numerical analysis results were compared with the experimental data. Thermal analyses of the storage system were carried out for normal and off-normal conditions by using the verified analysis method.

International Nuclear Information System (INIS)

Economic analysis was performed at KAERI with the assistance of US DOE to compare single reactor fuel cycle costs for a once-through option and a thermal recycle option to operate 1 GWe of a PWR plant for its lifetime. A reference fuel cycle cost was first calculated for each option with best estimated reference input data. Then a sensitivity analysis was performed changing each single value of such fuel cycle component costs as yellow cake price, enrichment charges, spent fuel storage cost, reprocessing cost, spent fuel disposal cost and reprocessing waste disposal cost. Savings due to thermal recycle in requirements of uranium, conversion, and enrichment were examined using formulas suggested by US DOE, while MOX fabrication penalty was accounted for. As a result of the reference fuel cycle cost analysis, it is ...

Energy Technology Data Exchange (ETDEWEB)

Continued public acceptance of commercial nuclear power is contingent to a large degree on the US Department of Energy (DOE) establishing an integrated waste management system for spent nuclear fuel. As part of the from-reactor transportation segment of this system, the B W Fuel Company (BWFC) is under contract to the DOE to develop a spent-fuel cask that is compatible with both rail and barge modes of transportation. Innovative design approaches were the keys to achieving a cask design that maximizes payload capacity and cask performance. The result is the BR-100, a 100-ton rail/barge cask with a capacity of 21 PWR or 52 BWR ten-year cooled, intact fuel assemblies. 3 figs.

International Nuclear Information System (INIS)

This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate ...

International Nuclear Information System (INIS)

The identification, and hence origin determination, of unknown nuclear material that might be found undeclared away from designated locations in the nuclear fuel cycle, is an important task in the frame of nuclear forensics. Material with forensic importance can be found at the microscopic level as particles in environmental samples indicating possible clandestine production of fissile material, and as bulky samples in the case of illicit trafficking of nuclear material. The objective of this work is to present, at a theoretical level, an isotopic finger-printing methodology which would determine the origin of unknown nuclear material with forensic importance. This is demonstrated for the case when the unknown nuclear material is spent nuclear fuel. The methodology is based on multivariate statistics, such as cluster and factor analysis, complemented by spent fuel isotopic ...

International Nuclear Information System (INIS)

The report presents the results of the study carried out to estimate the accident risk involved in the transport of spent fuel from Rajasthan Atomic Power Station near Kota to the fuel reprocessing plant at Tarapur. The technique of probabilistic safety analysis is used. The fuel considered is the Indian pressurised heavy water reactor fuel with a minimum cooling period of 485 days. The spent fuel is transported in a cuboidal, naturally-cooled shipping cask over a distance of 822 km by rail. The Indian rail accident statistics are used to estimate the basic rail accident frequency. The possible ways in which a release of radioactive material can occur from the spent fuel cask are identified by the fault tree analysis technique. The release sequences identified are classified into eight accident ...

Energy Technology Data Exchange (ETDEWEB)

The authors provide the intact criticality information that supports the disposal of spent nuclear fuel (SNF) from the US Department of Energy's (DOE's) Fast Flux Test Facility (FFTF) in the potential Monitored Geologic Repository at Yucca Mountain. FFTF is one of more than 250 forms of DOE-owned SNF. Because of the variety of the DOE SNF, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The FFTF fuel is representative of the mixed-oxide fuel (MOX) group.

International Nuclear Information System (INIS)

This paper describes the CANDU fuel development activities in Canada during 1997 through 1998. The activities include those of the Fuel Technology Program sponsored by the CANDU Owners Group. The goal of the Fuel Technology Program is to maintain and improve the reliability, economics and safety of CANDU fuel in operating reactors. These activities, therefore, concentrate on the present designs of 28-element and 37-element fuel bundles. The Canadian fuel development activities also include those of the Advanced Fuel and Fuel Cycle Technology Program at AECL. These activities concentrate on the development of advanced fuel designs and advanced fuel cycles, which among other advantages, can reduce the capital and fuelling costs, maintain operating margins in aging reactors, improve ...

Energy Technology Data Exchange (ETDEWEB)

The first Russian reprocessing facility, known as RT-1 (located at PA-Mayak in Ural region) was started on the radiochemical plant base in 1977. Nowadays RT-1 is the sole operating reprocessing plant in Russia. The main features of RT-1 is its broad spectrum of reprocessing spent nuclear fuels (SNF). The following spent fuel types are reprocessed: -) SNF from PWR reactors (WWER-440) and FB reactor (BN-600); -) SNF of transport ship reactors; -) production reactors SNF; -) research reactor spent nuclear fuel. The world-known technological processes are used at RT-1, but there are the following distinctive features. First, the universality of the three technological lines which allows not only the reprocessing of various SNF kinds, but also to implement the combined reprocessing of some types of them. Secondly, extraction of neptunium during SNF reprocessing which ...

International Nuclear Information System (INIS)

The paper discusses the safety problems connected with the conversion to dense storage of RBMK-1000 spent fuel in reactor cooling pools and independent storage facilities. Recourse to dense storage has been made for a number of reasons, among which are the absence of spent fuel shipments from the nuclear power plant site, prolongation of storage time and a partial change in storage conditions. Increasing the storage density per unit volume of the storage facility and turning to new technical procedures (as against the basic design) call for further investigation of safety problems. The safety assessment of the dense storage mode includes: (1) Selecting a list of initiating events for design basis and unforeseeable accidents; (2) Assessing dense storage safety under normal as well as design basis accident conditions; (3) Safety analysis and development of measures to compensate for unforeseen accidents. ...

Energy Technology Data Exchange (ETDEWEB)

A algorithm was developed to determine the characteristic parameters of PWR spent fuel, such as burnup, cooling time and initial enrichment of {sup 235}U by use of gamma-ray activity ratios of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 106}Ru/{sup 137}Cs from the high resolution gamma-ray spectroscopy and ORIGEN-S calculations. For the verification of the method developed, gamma-ray measurements of Kori-1 and Kori-2 nuclear fuel rods were carried out using HPGe gamma ray scanning system. As a results, it is revealed that the measured values are in a good agreement with the operator declared values within the about {+-} 5% errors. Besides, the under-water burnup measuring device has been designed to measure the gamma-ray from the PWR spent fuel assembly. This device will be set up in the pool of Post-Irradiation Examination Facility(PIEF), and used in determination of ...

International Nuclear Information System (INIS)

Spent nuclear wastes are inevitable issues to use nuclear power as a sustainable energy. Therefore, every country has their fuel cycles which are best for their environmental and/or political circumstances for the use of nuclear energy. These days agreements are made that spent nuclear fuels should be recycled to minimize waste volume and its toxicity all around the world. Republic of Korea also has a plan to recycle the spent nuclear fuels by using Gen-IV concept burner reactors and pyro-process plants. Not many options of national nuclear strategies are exist because Korea has too many people for its limited land space. KAERI already has been proposing a national fuel cycle concept called 'KIEP-21' that encompasses all the requirements of the advanced nuclear fuel cycle such as reduction of volume, toxicity, HLW heat ...

Energy Technology Data Exchange (ETDEWEB)

This study is one element of a larger engineering study effort by WHC to examine the feasibility of irradiated fuel and sludge consolidation in the KW Basin in response to TPA Milestone (target date) M-34-00-T03. The study concludes that up to 11,500 fuel storage canisters could be accommodated in the KW Basin with modifications. These modifications would include provisions for multi-tiered canister storage involving the fabrication and installation of new storage racks and installation of additional decay heat removal systems for control of basin water temperature. The ability of existing systems to control radionuclide concentrations in the basin water is examined. The study discusses requirements for spent nuclear fuel inventory given the proposed multi-tiered storage arrangement, the impact of the consolidated mass on the KW Basin structure, and criticality issues associated with multi-tiered ...

International Nuclear Information System (INIS)

The Nuclear Regulatory Commission (NRC), working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of various spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. The thermal performance of three different spent fuel cask designs (HOLTEC HI-STAR 100, TransNuclear TN-68, and NAC-LWT) was evaluated with the ANSYSR and COBRA-SFS analysis codes, utilizing boundary conditions for the tunnel fire obtained using NIST's Fire Dynamics Simulator (FDS) code. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of ...

International Nuclear Information System (INIS)

This work assesses the performance of a modified two-dimensional gamma scan system in spent fuel pin studies. The techniques for a two-dimensional gamma scan studied have been developed at the Hot Cell of Institute of Nuclear Energy Research (INER). Samples are acquired from the spent fuel pin, TPC-SP-C1, which was irradiated in a commercial reactor core (the first of its kind in Taiwan) for 2 years and then deposited in a cooling pool for 10 years. The spent fuel pin was then transferred into INER for further examination. The gamma scanning system was driven by a step motor which had an accuracy within 0.1 mm in both X-Y directions. Data obtained from this system are presented in both an isotopic distribution and contour plot. Results in this study closely correspond to those in other investigations, thereby confirming the effectiveness of this modified system. ...

International Nuclear Information System (INIS)

The reference concept 'direct disposal of spent fuel' was developed as an alternative to spent fuel reprocessing and vitrified HLW disposal. The technical facilities necessary for the implementation of this reference concept - the so called POLLUX-concept, e.g. interim storages for casks containing spent fuel, a pilot conditioning facility, and a special cask 'POLLUX' for final disposal have been built. With view to a geological salt formation all handling procedures for the repository were tested aboveground in a test facility at a 1:1 scale. To optimise the concept all operational steps are reviewed for possible improvement. Most promising are a concept using canisters (BSK 3) instead of POLLUX casks, and the direct disposal of transport and storage casks (DIREGT-concept) which is the most recent one and has been designed for the direct disposal of large ...

International Nuclear Information System (INIS)

HFH Herbst has designed and built a unique machine for storage of spent highly radioactive nuclear fuel rods within two years for the Swedish SKB. The vehicle (total weight 98 t) can be operated underground without a driver. Herbst was able to bring to this project almost 30 years of experience in the complementation of vehicle projects for the nuclear industry. The Procyon 1 already proved its efficiency impressively in several hundred storage processes and operates with absolute reliability. (orig.)

International Nuclear Information System (INIS)

Epoxy-resin based neutron shielding material, NS-4-FR, is used for spent fuel transport and/or storage cask. In this paper the outline of thermal aging test performed to evaluate the heating effect on this neutron shielding material, NS-4-FR, is introduced. The test is consisted of two kinds of thermal aging test, one is 'Basic Test' and the other is 'Block Heating Test'. The former is cooperatively performed by ten Japanese Electrical Power Companies, and the latter is done by GESC and NOF Corporation. (authors)

Energy Technology Data Exchange (ETDEWEB)

This report presents the results of the 2001 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Fort St. Vrain Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the predominant radiation exposure pathway, direct and scattered radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

Energy Technology Data Exchange (ETDEWEB)

The {sup 252}Cf-source-driven frequency analysis method can be used for measuring the subcritical neutron multiplication factor of arrays of LWR fuel and as little as a single PWR fuel assembly. These measurements can be used to verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. In addition, the data can be used to validate calculational methods for criticality safety. These measurements provide parameters that have a higher sensitivity to changes in fissile mass than neutron multiplication factor and thus serve as a better test of calculational methods. The analysis have also shown that measurement of the cross power spectral density (CPSD) between detectors on one side of a single fuel assembly and an internal or ...

International Nuclear Information System (INIS)

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate ...

Energy Technology Data Exchange (ETDEWEB)

The COBRA-SFS computer code was used to perform thermal-hydraulic analyses of consolidated spent fuel stored in casks. The ability of the COBRA-SFS code to model consolidated fuel was evaluated by comparing predictions with experimental data obtained from electrically heated rod bundles by Ridihalgh, Eggers, and Associates and Eggers Ridihalgh Partners, Inc. under sponsorship of the Electric Power Research Institute. The calculations agreed with the measured temperatures well within the bounds of experimental error. Based on the evaluation results, best-estimate temperature predictions were performed for consolidated fuel in several cask designs. Results are presented for the REA 2023 BWR cask, the CASTOR-1C BWR cask, and the Concrete Sealed Storage Cask designed for Monitored Retrievable Storage (MRS). The cask simulation results indicate that consolidation of spent ...

International Nuclear Information System (INIS)

The RESF is utilized for storage of spent fuel under emergency conditions as well as for testing of FM heads. It receives cooling supply from the PHT Pressurizing pumps and after removal of decay heat from the spent fuel it goes to the D2O Storage Tank. The geometry of the RESF system is such that it can not sustain the thermosyphon loop during SBO, due to high frictional forces. To achieve the sustained thermosyphon, modifications in the design were suggested viz., removal of the steam trap and the relief valve above it and replacement by a solenoid valve (SV-16). In the event of SBO, SV-16 will open on 'RESF channel temperature high' signal and connect to FT D2O tank. The tank, being at atmospheric pressure and at lower elevation, will provide higher cooling flow rate through the RESF channel. D2O is periodically removed from the FT D2O tank by operating a Class-II pump intermittently. The analysis ...

International Nuclear Information System (INIS)

The results of calculational comparative study into radiotoxicity and residual power release time dependences for spent uranium-plutonium and thorium-uranium nuclear fuels from the WWER-1000 reactors during ling- time (up to 300 thousand years) storage are discussed. It is shown that the total radiotoxicities for actinides from uranium, uranium-plutonium and thorium-uranium spent fuels at storage period begin amount to 5.2 x 10"1"4, 1.3 x 10"1"5 and 1.5 x 10"1"4 kg of water per 1 t of discharged fuel, respectively. Radiotoxicity of actinides in uranium-plutonium fuel is revealed to be by the factor of 2.5 greater than that for ordinary uranium fuel because of greater accumulation of "2"3"9Pu, "2"4"0Pu, "2"4"1Pu and "2"4"4Cm. Radiotoxicity of actinides for thorium-uranium fuel calculated taking into account "2"3"4U is ...

Energy Technology Data Exchange (ETDEWEB)

In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning ...

International Nuclear Information System (INIS)

Correct prediction of the fission products inventory in irradiated nuclear fuels is essential for accurate estimation of fuel burnup, establishing proper requirements for spent fuel transportation and storage, materials accountability and nuclear forensics. Such prediction is impossible without accurate knowledge of neutron induced fission yields. The uncertainty of the fission yields reported in the ENDF/B-VII.0 library is not uniform across all of the data and much of the improvement is desired for certain fissioning isotopes and fission products. We discuss our measurements of cumulative fission yields in nuclear fuels irradiated in thermal and fast reactor spectra using Inductively Coupled Plasma Mass Spectrometry. (author)

Energy Technology Data Exchange (ETDEWEB)

Assessment on the back end fuel cycle, in PEACER (Proliferation-resistant Environmental-friendly Accident-tolerant Continuable and Economical Reactor) that was designated as a new transmutation concept, was performed. Recovery system of uranium and TRU for PEACER is based on pyroprocessing. In the assessment of Long-Lived Fission Products (LLFP) wastes, initially {sup 90}Sr and {sup 137}Cs are dominant contributor nuclides until 30 years and especially {sup 90}Sr and {sup 137}Cs have the highest activity and decay heat than other LLFP. In this study, recovery of {sup 90}Sr and {sup 137}Cs is recommended for reducing of wastes loading. The acceptable decontamination factor is investigated by the toxicity of PEACER spent fuel. The acceptable decontamination factor is about 1.02E+05 for the actinides from PEACER spent fuel after 10 years cooling, 4.26E+05 after 100 years cooling, ...

International Nuclear Information System (INIS)

The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel ...

Energy Technology Data Exchange (ETDEWEB)

The issues of the geothermal resource at Brady's Hot Springs are dealt with: the prospective supply of feedstocks to the ethanol plant, the markets for the spent grain by-products of the plant, the storage, handling and transshipment requirements for the feedstocks and by-products from a rail siding facility at Fernley, the probable market for fuel ethanol in the region, and an assessment of the economic viability of the entire undertaking.

International Nuclear Information System (INIS)

A simplified two step thermal analysis method has been develop to evaluate (1) the mean temperature of the CANDU fuel bundles within a fuel basket in a given spent fuel dry storage canister by HEATING5 code with additional input data and heat transfer correlations in the step-1 analysis and (2) the maximum fuel rod temperature within a CANDU 37-element fuel bundle by MAXROT code developed here for step-2 analysis. In addition, the results of sample analysis is performed to examine the parametric effects of the site-specific ambient conditions on the maximum fuel temperature within a canister are presented. The comparison between the results of step-1 analysis and the mock-up test, in particular, is quite satisfactory. In essence, the two-step thermal analysis method proposed here is a code package that used the HEATING5 and MAXROT codes, ...

Energy Technology Data Exchange (ETDEWEB)

A general criterion has been suggested for deep geological repositories containing spent fuel - the repositories should impose no greater radiological risk than due to naturally occurring uranium deposits. The following analysis investigates the rationale of that suggestion and determines whether current expectations of spent-fuel repository performance are consistent with such a criterion. In this study, reference spent-fuel repositories were compared to natural uranium-ore deposits. Comparisons were based on intrinsic characteristics, such as radionuclide inventory, depth, proximity to aquifers, and regional distribution, and actual and potential radiological consequences that are now occurring from some ore deposits and that may eventually occur from repositories and other ore deposits. The comparison results show that the repositories are quite comparable to the natural ore deposits and, in some ...

Energy Technology Data Exchange (ETDEWEB)

The ability of the /sup 252/Cf-source-driven neutron noise analysis method to measure subcriticality has been demonstrated in a variety of experimental configurations of fissile materials. Calculations for an approximately 4-m-dia configuration of light water reactor (LWR) fuel elements indicated the feasibility of measuring the subcriticality of large, loosely coupled arrays of LWR fuel elements by this same method. These analysis suggested application to the initial loading of both pressurized and boiling water reactors, zero-power testing of reactors (such as shutdown margin measurements after initial loading), light water reactor refueling, and safe storage of LWR spent fuel. In the fuel storage application, direct measurement of subcriticality in the actual fuel storage facilities provides the parameter which is directly related to criticality safety.

International Nuclear Information System (INIS)

During the phases of separation, treatment, conditioning and storage of radioactive waste, destructive and nondestructive methods for their characterization are needed. In order to satisfy this necessity, in the frame of the National Program of Research and Development, the 'Laboratory for characterization of spent nuclear fuel and high/medium level radioactive waste- LABORAD' was created. The purpose of the project was to accredit the analysis methods available in the laboratory, and also to develop new methods for the characterization of the radioactive waste. A special attention was paid to the high-level radioactive waste and spent nuclear fuel characterization that require special facilities for handling. These facilities (e.g. hot cells, remote handlers, transport container) are already available in our institute. Experimental results and performances obtained during validation of the methods are ...

International Nuclear Information System (INIS)

A major objective of the DOE Advanced Fuel Cycle Initiative, AFCI, is to explore technologies that may reduce the long-term environmental burden of nuclear energy through more efficient disposal of waste materials. In this work, the potential impact of the AFCI technology and its beneficial effects on waste management and its ability to meet waste management objectives are demonstrated. In addition, practical scenarios to improve permanent disposal utilization and/or reduce the temporary spent nuclear fuel (SNF) storage inventory by closing the fuel cycle through transition to fast reactor (FR) converters are also discussed. (authors)

International Nuclear Information System (INIS)

One of the alternative ideas to solve the spent fuel issues, the partitioning and transmutation (P and T) technology has been developed for decades. Moreover, the concept of LILW production from P and T are proposed by Bowman. A PEACER (Proliferationresistant, Environmental-friendly Accident-tolerant, Continuable and Economical Reactor), based on pyrochemical process and Pb-Bi coolant transmutation reactor, has been conceptually designed to be able to convert all PWR spent fuel into low and intermediate level waste for near-surface disposal. In this study, the acceptance criteria for near-surface disposal facility is derived by the methodology for establishment of acceptance criteria. Then acceptable TRU decontamination factor (DF) and LLFP removal efficiency in order to meet acceptance criteria is evaluated

Energy Technology Data Exchange (ETDEWEB)

Pilot-scale pyrolysis units have been in operation since 1980 to test the efficiency of thermal treatment of transuranic (TRU) solid waste to retrieve the TRUs and to reduce the volume of wastes such as spent solvent, spent resin, and others. These wastes are generated by reprocessing, fuel production, and utilities. NUKEM has developed a criticality-safe, ring-slab reactor to decompose solid TRU waste. The plant processes 25 kg/h with a polyvinyl chloride content up to 70%. The overall throughput (inactive) up to the spring of 1982 was 2000 kg. The decontamination factor for the reactor itself is 1000. The liquid wastes, mainly spent solvent, are cracked under nitrogen at 400 to 700/sup 0/C in a reactor that is filled by a packed bed kept in motion by a specially designed agitator. This unit was built for 15 kg/h water equivalent evaporation. Up to 1982 the unit processed 2000 kg of ...

Energy Technology Data Exchange (ETDEWEB)

To resolve the central thermal safety issue for spent fuel dry storage concrete canister design or Wolsung (CANDU) nuclear power plant unit 1, a thermal analysis method has been developed for the complicated geometry of rod bundles and the multi-dimensional and multi-mode heat transfer phenomena. The canister geometry is simplified and combined heat transfer by conduction, convection, and radiation is considered through effective heat transfer coefficients. Mean temperature distributions of the fuel bundles within the fuel basket are obtained by solving the heat transfer problem using an existing computer code HEATING5. The measured steady state temperature distribution within a mock-up of a storage basket is compared to the calculated result. Steady state and/or transient fuel temperature distributions have been calculated for various ambient conditions at the canister exterior ...

Energy Technology Data Exchange (ETDEWEB)

This paper explores the potential of solid oxide fuel cells (SOFCS) as 3--10 kW auxiliary power units for trucks and military vehicles operating on diesel fuel. It discusses the requirements and specifications for such units, and the advantages, challenges, and development issues for SOFCS used in this application. Based on system design and analysis, such systems should achieve efficiencies approaching 40% (lower heating value), with a relatively simple system configuration. The major components of such a system are the fuel cell stack, a catalytic autothermal reformer, and a spent gas burner/air preheater. Building an SOFC-based auxiliary power unit is not straightforward, however, and the tasks needed to develop a 3--10 kW brassboard demonstration unit are outlined.

International Nuclear Information System (INIS)

As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points ...

International Nuclear Information System (INIS)

Supplemental criticality safety analysis of a pool type storage for TRIGA spent fuel at 'Jozef Stefan' Institute in Ljubljana, Slovenia, is presented. Previous results (Ravnik, M, Glumac, B., 1996) have shown that subcriticality is not guaranteed for some postulated accidents. To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods (Glumac, B., Ravnik, M., Logar, M., 1997) to lower the supercriticality probability, when the pitch is decreased to contact (as a consequence of a severe earthquake) in a square arrangement. The criticality analysis for the hexagonal contact pitch is presented in this paper, following the same scenario as outlined above. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. First, the analysis about the influence of the number of triangular fuel piles ...

Energy Technology Data Exchange (ETDEWEB)

The spent fuel management concept of direct ultimate disposal is compatible in its fundamental features with the law of the European Community. This applies to a national law prohibiting spent fuel reprocessing and the handling of plutonium in Germany, imposing restrictions on exporting spent fuel assemblies and importing plutonium and reprocessing remnants, and on power plant operators to employ reprocessing services abroad. Also, a nuclear power phase-out decided by the German government would in principle not mean a breach of the EURATOM treaty. (orig./HP) [Deutsch] Als Alternative zur geltenden Rechtslage werden im Deutschen Rundestag Aenderungen vorgeschlagen, die das Entsorgungskonzept der direkten Endlagerung - d.h. ein Verbot der Wiederaufarbeitung von Brennelementen aus deutschen Kernkraftwerken auch im europaeischen Ausland - vorschreiben und einen ...

International Nuclear Information System (INIS)

The field trial of an unattended monitoring system for SF transfer verification ended in June 2005. The system is aiming to improve the safeguards measures applied during SF transfers and reduce inspection effort currently required by the Operator, NNCA and the IAEA. They were reviewed based on its technical performance and other issues in this paper.

Energy Technology Data Exchange (ETDEWEB)

The 200 Area Interim Storage Area Technical Safety Requirements define administrative controls and design features required to ensure safe operation during receipt and storage of canisters containing spent nuclear fuel. This document is based on the 200 Area Interim Storage Area, Annex D, Final Safety Analysis Report which contains information specific to the 200 Area Interim Storage Area.

International Nuclear Information System (INIS)

We obtained the ratios of "1"3"4Cs/"1"3"7Cs and "1"5"4Eu/"1"3"7Cs in spent PWR fuels with gamma scanning equipment of irradiated material examination facility. The fuel segments, of which the burn-up is about 40 GWD/MTU and its cooling time is 8.4 years, are prepared in Post Irradiated Examination Facility and transported to IMEF. By considering of multi-peaks of "1"3"4Cs and "1"5"4Eu, we obtained the relative efficiency of the gamma scanning system as a function of energy. And finally we obtained the number ratios of radioactive nuclides in 72 fuel pellets, radioactive "1"3"4Cs/"1"3"7Cs and "1"5"4Eu/"1"3"7Cs. (author). 4 refs., 27 tabs., 28 figs.

International Nuclear Information System (INIS)

A method for express determination of "2"3"4"-"2"3"8U, "2"3"8"-"2"4"2Pu, "2"4"1"-"2"4"3Am and "2"4"2"-"2"4"4Cm content in fuel 'hot' particles and spent nuclear fuel is offered. The method is based on precision measurement of a sample #alpha# activity followed by estimate of relative contributions of individual nuclides, or groups of radionuclides, to total activity. Segregation in separate fractions of uranium, plutonium, americium and curium was made with the help of ion-exchange chromatography. Results are presented of "2"3"4"U, "2"3"8U, "2"3"8"Pu,"2"3"9"+"2"4"0Pu, "2"4"1Pu, "2"4"2Pu, "2"4"1"Am, "2"4"2"mAm, "2"4"3Am "2"4"2"Cm and "2"4"4Cm definition in 'hot' particles sampled in Chernobyl NPP surroundings, Opportunities to apply this method for identifying radionuclide content of spent nuclear fuel are discussed.

International Nuclear Information System (INIS)

Assessment on the back end fuel cycle, in PEACER (Proliferation-resistant Environmental-friendly Accident-tolerant Continuable and Economical Reactor) that was designated as a new transmutation concept, was performed. Recovery system of uranium and TRU for PEACER is based on pyro-processing. In the assessment of long-lived fission products (LLFP) wastes, initially "9"0Sr and "1"3"7Cs are dominant contributor nuclides until 30 years and especially "9"0Sr and "1"3"7Cs have the highest activity and decay heat than other LLFP. In this study, recovery of "9"0Sr and "1"3"7Cs is recommended for reducing of wastes loading. The acceptable decontamination factor is investigated by the toxicity of PEACER spent fuel. The acceptable decontamination factor is about 1.02 E+05 for the actinides from PEACER spent fuel after 10 years cooling, 4.26 E+05 after 100 years cooling, 1.97 E+04 after 300 ...

International Nuclear Information System (INIS)

Spent oxide fuels are reduced in a molten salt of CaCl_2-CaF_2 to convert them into metals, then melted in an Fe-U bath disposed in an electrolytic refining vessel and brought into contact with molten Mg, to extract transuranium elements and rare earth elements contained in the Fe-U bath as metals in the molten Mg. Then molten Mg is removed and the residue is brought into contact with KCl-LiCl molten salt and electrolyzed using the Fe-U as an anode. Then, uranium is recovered by deposition on an iron cathode disposed in chloride electrolytes of the electrolytic refining vessel. Uranium and transuranium elements can be thus separated and, for example, depleted uranium for use in blanket fuels can be recovered easily. This can greatly reduce the temporary storage amount of depleted uranium, to eliminate requirement for a large-scaled facility used exclusively for storing uranium and long time management for uranium. (T.M.).

Science.gov (United States)

Ormiston, R.A., Applications of the Induced Power Model and Performance of Conventional and Advanced Rotorcraft, 2010, details ...

International Nuclear Information System (INIS)

A neutron scanning device was developed in order to obtain accurate neutron intensities of high burn-up BWR fuels. This scanning device was calibrated with a "2"5"2Cf source and used to measure axial distributions of neutron intensities of BWR fuels with various enrichments (2.0%-3.4%) irradiated up to 60 GWd/tU at Fukushima Daini Nuclear Power Station Unit 2(2F-2). The measured neutron intensities were approximated well with power law interpolations on the calculated burn-up values. The neutron intensities calculated by the ORIGEN2-86 code showed good agreements with the measured ones within 20%. (author)

Energy Technology Data Exchange (ETDEWEB)

A trans-uranium (TRU) fuel should be manufactured and loaded in transmutation systems in order to transmute the long-lived TRU nuclides into short-lived ones. However, since all of the TRU nuclides are not completely transmuted in one cycle lifetime in transmutation systems, the spent TRU fuel has to be treated to recover the long-lived radionuclides or fuel matrix materials. One concept to manufacture TRU fuel for transmutation is to recover uranium from TRU and molten salt. If this type of fuel is adopted for transmutation, uranium could also be an objective material to be recovered and recycled. Since electrowinning is a promising technology to be employed for the recovery of uranium from fuel materials, some experimental work of electrowinning using anode of Cd-Li alloy was carried out in this study. The basic salt chosen was a mixture ...

International Nuclear Information System (INIS)

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that "2"3"3U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of ...

International Nuclear Information System (INIS)

The Quality Assurance Requirements and Description (QARD) is the principal quality assurance document for the Civilian Radioactive Waste Management Program (Program). It establishes the minimum requirements for the Quality Assurance Program. The QARD contains regulatory requirements and program commitments necessary for the development of an effective quality assurance program. Quality assurance implementing documents must be based on, and consistent with, QARD requirements. The QARD applies to the following: (1) acceptance of spent nuclear fuel and high-level radioactive waste; (2) transport of spent nuclear fuel and high-level radioactive waste; (3) the Monitored Retrievable Storage (MRS) facility through application for an operating license; (4) Mined Geologic Disposal System (MGDS), including the site characterization activities (exploratory studies facility (ESF) and surface based testing), through ...

International Nuclear Information System (INIS)

A procedure is described for the isolation and determination of "9"0Sr, "2"2"6Ra, "2"2"8Ra and "2"1"0Pb in food. These nuclides are highly radiotoxic, above all for babies and children. Samples are dry ashed, alkali metals are removed as carbonates. Separation from other matrix ions and separation of Ra, Sr and Pb can be achieved with a column of Sr-Spec, an immobilized crown ether. For activity measurements membrane filters with SrSO_4, Ba(Ra)SO_4 and PbS are prepared. Ra is determined by gamma-spectrometry, "9"0Sr and "2"1"0Pb are determined by low-level-betacounting. The determination limits are 10 mBq/kg for "9"0Sr and "2"1"0Pb and 50 mBq/kg for "2"2"6Ra and "2"2"8Ra. The procedure is useful for all kinds of foodstuff. (orig.).

Science.gov (United States)

Amongst the minor actinides issued from the spent nuclear fuel, {sup 241}Am is present in high concentration and contributes significantly to the long-term radiotoxicity of nuclear waste. A major uncertainty was present in the transmutation chain of {sup 241}Am when irradiated by a high intensity thermal neutron flux. This uncertainty was brought about by the poor knowledge of the {sup 242gs}Am neutron capture cross section. A dedicated experiment has been performed at the Institut Laue-Langevin in Grenoble, which gives a definitive experimental answer to this problem.

Energy Technology Data Exchange (ETDEWEB)

In the safety assessment of a potential repository for spent nuclear fuel, it is important to assess the consequences of a hypothetical leak of radionuclides through the seabed and into a waterborne transport phase. Radionuclides adsorbed to sediment particles may be transported great distances through the processes of sedimentation and resuspension. This study investigates the transport patterns of sediment particles of two different sizes, released in the Forsmark and Laxemar area. The results show that the closed waters around Forsmark to a higher degree makes the particles stay in the area close to the release points

Energy Technology Data Exchange (ETDEWEB)

This executive summary presents the motivation, structure, objectives, methodologies and results of the first stage of the international DECOVALEX project - DECOVALEX I (1992-1995). The acronym stands for Development of Coupled Models and their Validation against Experiment in Nuclear Waste Isolation, and the project is an international effort to develop mathematical models, numerical methods and computer codes for coupled thermo-hydro-mechanical processes in fractured rocks and buffer materials for geological isolation of spent nuclear fuel and other radioactive wastes, and validate them against laboratory and field experiments. 24 refs.

Energy Technology Data Exchange (ETDEWEB)

On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for ...

Energy Technology Data Exchange (ETDEWEB)

This paper introduces the method of seismic stability analysis for the increase of fuel storage capacity of wet storage stacks by one or two more stack floor at Wolsung Nuclear Unit 2,3,4, which had been originally licensed assuming 16 tray stack-o-storage. As a basic procedure, tipping and sliding stability of the structure is checked at first thru seismic analysis and the resultant load from dynamic analysis is applied for static stress analysis, and the result of which is reviewed for compatability with applicable standard. As a result, sliding and overturning are not expected under design basis earthquakes for increased storage cases of 17 tray and 18 tray stacks. And it is anticipated the result of stress analysis will be acceptable.

British Library Electronic Table of Contents (United Kingdom)

Nuclear energy generates 30% of the electricity in the EU. Still, different countries of EU27 have very different attitudes towards the future use of nuclear energy for electricity generation or other uses. However, independently of the political decision of continuation or phase out of the nuclear energy, all countries using nuclear energy to generate electricity are facing the question of the final management of its spent nuclear fuel and other high level radioactive wastes. Partition and Transmutation are emerging technologies that when integrated in advanced fuel cycles can strongly influence on the final wastes from the nuclear industry and its management. A review of the progress on the understanding of their real potentialities and main conclusions from a large number of internation...

International Nuclear Information System (INIS)

Pyrochemical processing of nuclear fuels using molten salts has attracted much attention because of its potential to be applied for a future spent nuclear fuel management. In the pyrochemical processing, there are a number of steps to electro-refine and electro-win each element of lanthanides and actinides, commonly called trans-uranic elements (TRU). In order to materialize the pyrochemical processing in the nuclear power plant environments, qualitatively and quantitatively monitoring of each elements is necessary. Thus, we have undertaken to develop an on-line observing system of the TRU in LiCl-KCl molten salt media by using electrochemical and spectroscopic methods. In this work, the electrochemical and spectroscopic behaviors of europium as a proxy material for TRU were investigated simultaneously in the LiCl-KCl molten salt

Energy Technology Data Exchange (ETDEWEB)

The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E&E) and Chemistry & Material Sciences (C&MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E&E and C&MS Directorates co-sponsored this Laboratory Directed Research & Development ...

Science.gov (United States)

This paper is dedicated to the treatment of sludge occurring in frame of the Egyptian produced from oil and gas production. The activity levels of three radium isotopes: Ra-226 (of U-series), Ra-228 and Ra-224 (of Th-series) in the solid TENORM waste (sludge) were first evaluated and followed by a sequential treatment for all radium species (fractions) presented in TENORM. The sequential treatment was carried out based on two approaches 'A' and 'B' using different chemical solutions. The results obtained indicate that the activity levels of all radium isotopes (Ra-226, Ra-228 and Ra-224) of the environmental interest in the TENORM waste sludge were elevated with regard to exemption levels established by IAEA [International Atomic Energy Agency (IAEA), International basic safety standards for the protection against ionizing radiation and for the safety of ...

International Nuclear Information System (INIS)

A feature of the Fast Flux Test Facility (FFTF) is that fuel assemblies can be stored within the reactor vessel, either for future use in the core or for temporary storage of spent fuel. Due to the close proximity of fuel assemblies stored in the in-vessel storage (IVS) locations to the ex-vessel flux monitors (EVFMs), neutrons originating in the stored fuel can result in significant contributions to the total EVFM signals. The paper shows EVFM-induced electronic current data during operation at full power for FFTF power cycles 1 through 7. The data were obtained early in each cycle so that signal drift due to core burnup could be ignored. The differences in currents are attributed to differences in cycle-to-cycle core and stored fuel loadings. At the beginning of cycle 1 there were no fuel assemblies stored in IVS. Subsequent to cycle 1, ...

Energy Technology Data Exchange (ETDEWEB)

Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high {sup 235}U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing ...

International Nuclear Information System (INIS)

In the radiochemical reprocessing of spent fuel from nuclear reactors, a considerable amount of liquid, solid, and gaseous waste is formed; this waste is potentially dangerous to humans and requires the development of special and complex technological techniques for its localization and reliable long-term storage. The most hazardous are liquid wastes of high specific activity - water-tailings solutions obtained in the first cycle of extraction after the removal of uranium and plutonium. These solutions contain more than 99.9% of all the other transuranic elements - isotopes of neptunium, americium, and curium. Where necessary, some fission products and actinides may be removed from wastes of high specific activity for subsequent use. The quantity, composition, and activity of these wastes varies within broad limits, depending on the type and power of the reactor, the initial nuclide composition of the fuel, and its specific ...

International Nuclear Information System (INIS)

(Aug 1973). United States Moser, AR Matzner, RA Ryan, MP Jr. Center

Energy Technology Data Exchange (ETDEWEB)

The fertilizer contribution to radio uptake by tomato plants was assessed. Tomato and soil samples from four different plantations were analyzed: two using organic fertilizer and two using phosphate fertilizer. The Ra concentrations in eight of the most used tomato fertilizers were determined. The concentration values ranged from 14 to 221 Bq/kg of Ra-226 and from 25 to 176 Bq/kg of Ra-228. Owing to the low Ra concentration and to the used fertilizer quantities, the fertilizers would not increase significantly the Ra soil concentration. However, the highest soil to plant concentration ratios were found in the plantations using phosphate fertilizers. This outcome pointed out the higher Ra availability in the phosphate fertilizer plantations than in the organic fertilizer plantations, showing a possible positive influence of phosphate fertilizers for the tomato ...

International Nuclear Information System (INIS)

On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB, the U.S. agency responsible for determining the cause of transportation accidents), to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for ...

Energy Technology Data Exchange (ETDEWEB)

In this experiment, we measured the detection efficiency of gamma-scanning system of Irradiated Materials Examination Facility (IMEF) in Korea Atomic Energy Research Institute(KAERI) for the spent PWR fuel and activity known sources with high activity. We measured the absolute detection efficiency of gamma scanning system of IMEF to the {sup 137}Cs-source with 10.4 GBq and {sup 60}Co source with 25.9 GBq High-activity sources as standard sources, and a fuel burned in the KORI -1 reactor of the Kori Nuclear power plant. In analyzing, we used three peeks those were the 661.64 keV peak form {sup 137} Cs 1173.23 and 1332.51 keV{sup 60} Co sources, and 14 peaks of {sup 134}Cs and {sup 154}Eu on the spent fuel gamma spectrum which are in the energy range from 500 to 1,600 keV. We find second order equations for detection efficiency by using our experimental results. The equation show that ...

British Library Electronic Table of Contents (United Kingdom)

For the reuse of a waste salt from an electrorefining process of a spent oxide fuel, a separation of rare earth elements by an oxidative precipitation in a LiCl-KCl molten salt was tested without using precipitate agents. From the results obtained from the thermochemical calculations by HSC Chemistry software, the most stable rare earth compounds in the oxygen-used rare earth chlorides system were oxychlorides (EuOCl, NdOCl, PrOCl) and oxides (CeO2, PrO2), which coincide well with results of the Gibbs free energy of the reaction. In this study, similar to the thermochemical results, regardless of the sparging time and molten salt temperature, oxychlorides and oxides were formed as a precipitant by a reaction with oxygen. The structure of the rare earth precipitates was divided into two sha...

International Nuclear Information System (INIS)

An oxidative precipitation reaction of rare-earth chlorides in a LiCl-KCl molten salt was successfully carried out by a lab-scale apparatus. The conversion efficiency of the used rare earth chlorides into the insoluble precipitates was increased with the sparging time and temperature and was affected on oxygen sparger type. In the conditions of 700 .deg. C molten salt temperature and above 540min sparging time, the final conversion efficiencies were over 99% for all the experimented rare-earth chlorides. The hydrodynamic characteristics of an oxygen-molten salt two phase flow system are very important for a co-oxidative precipitation of rare earth elements

International Nuclear Information System (INIS)

Large quantities of organic chemicals used in reprocessing spent nuclear-fuels at the Hanford Site have accumulated in underground high-level radioactive waste tanks. The organic content of these tanks must he known so that the potential for hazardous reactions between organic components and sodium nitrate/nitrite salts in the waste can he evaluated. The solubilities of organic compounds described in this report will help determine if they are present in the solid phases (salt cake and sludges) as well as the liquid phase (interstitial liquor/supernate) in the tanks. The solubilities of five significant sodium salts of carboxylic acids and aminocarboxylic acids [sodium oxalate, formate, citrate, nitrilotriacetate (NTA) and ethylendiaminetetraacetate (EDTA)] were measured in a simulated supernate solution at 25 degrees C, 30 degrees C, 40 degrees C, and 50 degrees C.

International Nuclear Information System (INIS)

Based on the enhanced cooperation between Republic of Korea (ROK) and International Atomic Energy Agency (IAEA), IAEA installed new safeguards equipment (Unattended Monitoring System, UMS) at Wolsung unit 2 on January, 2005. The UMS was utilized in parallel with traditional human surveillance during the transfer campaign of Wolsung unit 2 (7 March - 22 July). Installation of UMS at all Wolsung units will be decided on the basis of the result of this test. The transfer campaign with UMS at Wolsung units 2 and 4 was started from September 1, 2006 at the same time. Implementation of UMS to the transfer campaign in Wolsung site will be considerably reduced the IAEA's inspection effort and burden of operator.

International Nuclear Information System (INIS)

The aim of the risk assessment consists of a comparative security evaluation for the SNR-300 and the PWR Biblis B. The failure analysis focusses on the reactor core; in addition, possible fission product release from the spent fuel pits is examined. By reliability analyses, the frequency of events leading to incidents is determined together with the probability of core destruction. In the accident analysis, the kind and frequency of failure of the activity barriers, i.e., primary system (reactorvessel) and inner and outer containment are investigated for the various incident sequences. The radionuclide release into the environment is classified into five different release categories. Besides internal failures, external causes (especially earthquakes and plane crashes) are considered under the aspect of their risk contribution. (RF).

Energy Technology Data Exchange (ETDEWEB)

Before disposing of transuranic radioactive wastes in the Waste Isolation Pilot Plant (WIPP), the United States Department of Energy (DOE) must evaluate compliance with applicable long-term regulations of the United States Environmental Protection Agency (EPA). Sandia National Laboratories is conducting iterative performance assessments of the WIPP for the DOE to provide interim guidance while preparing for final compliance evaluations. This volume contains an overview of WIPP performance assessment and a preliminary comparison with the long-term requirements of the Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes (40 CFR 191, Subpart B).

International Nuclear Information System (INIS)

Radioactive waste for legal and regulatory purposes may be defined as material that contains or is contaminated with radio-nuclides at concentrations or activities greater than clearance levels as established by the regulatory body, and for which no use is foreseen. Safe management of radioactive waste is essential to ensure protection of humans and environment. Radioactive waste management policy is a guideline for the safe management of radioactive waste. It expresses the commitment of the country towards the goal. Government should initiate investigation into best long-term option for management of spent nuclear fuels. Process of selecting option and eventual site should involve comprehensive public participation within set time frames (with deep geological disposal as preferred management option).

International Nuclear Information System (INIS)

The foundation conducts an environmental monitoring and surveillance program over an area covering much of the upper Snake River Plain and provide environmental education and support services related to INEL natural resource issues. Also, the foundation, with its university affiliates, conducts ecological and radioecological research on the Idaho National Environmental Research Park. This research benefits major DOE-ID programs including waste management, environmental restoration, spent nuclear fuels, and land management issues. Major accomplishments during CY1995 can be divided into five categories: environmental surveillance program, environmental education, environmental services and support, ecological risk assessment, and research benefitting the DOE-ID mission.

International Nuclear Information System (INIS)

In the early sixties it was decided to investigate salt formations on its suitability to host heat generating radioactive waste in Germany. In the reference repository concept consequently the emplacement of vitrified waste canisters in deep vertical boreholes inside a salt mine was considered whereas spent fuel should be disposed of in self shielding casks (type POLLUX) in horizontal drifts. The POLLUX casks, 65 t heavy carbon steel casks, will be laid down on the floor of a horizontal drift in one of the disposal zones to be constructed in the salt dome at the 870 m level. The space between casks and drift walls will be backfilled with crushed salt. The transport, the handling und the emplacement of POLLUX casks were subject of successfully performed demonstration and in situ tests in the nineties and resulted in an adjustment of the atomic law. The borehole disposal concept comprises the emplacement of unshielded canisters with vitrified HLW ...

Energy Technology Data Exchange (ETDEWEB)

The feasibility was evaluated using MCNP-DSP, an analog Monte Carlo transport cod to simulate source-driven measurements. Models of an isolated Westinghouse 17x17 PWR fuel assembly in a 1500-ppM borated water storage pool were used. In the models, the fuel burnup profile was represented using seven axial burnup zones, each with isotopics estimated by the PDQ code. Four different fuel assemblies with average burnups from fresh to 32 GWd/MTU were modeled and analyzed. Analysis of the fuel assemblies was simulated by inducing fission in the fuel using a {sup 252}Cf source adjacent to the assembly and correlating source fissions with the response of a bank of {sup 3}He detectors adjacent to the assembly opposite the source. This analysis was performed at 7 different axial positions on each of the 4 assemblies, and the source-detector cross-spectrum signature was calculated for each of ...

Energy Technology Data Exchange (ETDEWEB)

Studies of naturally occurring radioactive materials (NORM) distribution of {sup 226}Ra, {sup 228}Ra and {sup 40}K in East Malaysia were carried out as part of a marine coastal environment project. The results of measurements will serve as baseline data and background reference level for Malaysia coastlines. Sediments from 21 coastal locations and 10 near shore locations were collected for analyses. The samples were dried, finely ground, sealed in a container and stored for a minimum of 30 days to establish secular equilibrium between {sup 226}Ra and {sup 228}Ra and their respective radioactive progenies. They were counted using a high-purity germanium (HPGe) spectrometer covering the respective progeny energy peak. For {sup 40}K, the presence of this was measured directly via its 1460 keV energy peak. The concentration of {sup 226}Ra, {sup 228}Ra and {sup 40}K ...

Science.gov (United States)

Thermal models are constructed and analyses are performed of aluminum-based spent nuclear fuel (Al-SNF) in interim dry storage and geologic disposal configurations. Two models are developed, referred to as the interim storage model and the codisposal waste package (WP) model. Time-dependent source terms of Al-SNF forms and the defense high-level waste (DHLW) canisters are also developed for thermal performance analysis of the geologic codisposal WP.The interim storage model is a three-dimensional conduction-convection conjugate model to investigate the natural convection cooling of a sealed dry storage canister with vertical orientation in a dry storage vault. The analysis is made for various decay heat sources (equivalent to 25 to 35 kW/m{sup 3}) using various boundary conditions around the canister wall and with backfilled nitrogen or helium gas. Based on the data obtained from the Savannah River Site experimental work and available from the ...

Energy Technology Data Exchange (ETDEWEB)

Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell ...

Energy Technology Data Exchange (ETDEWEB)

The release of {sup 14}C as {sup 14}CO{sub 2} from partly failed spent fuel containers has been analyzed by the flow of gases into and out of the containers. This flow of gases is driven by pressure differences, which are in turn caused by heating by the spent fuel. In this analysis, the timing and size of holes in the containers are assumed to be given. A better means of predicting the time distribution and sizes of penetrations in nuclear waste containers is needed. For the purposes of far-field transport calculations, we have adopted release rates that are shown to be bonding for the large range of hole sizes studied. The transport of released {sup 14}CO{sub 2} has been analyzed by transport in equivalent porous medium. The peak {sup 14}CO{sub 2} concentration in pore gas at 350 m above the repository does not depend on the time of hole occurrence, although the time of penetration obviously affects ...

Energy Technology Data Exchange (ETDEWEB)

This report compares the performance characteristics of /sup 10/B-lined and fission-neutron detectors in gamma-ray fluxes typical of the fields to be encountered during nondestructive testing of irradiated light-water fuel assemblies stored in water. Using the optimum time constants for each of the /sup 10/B-lined detectors, the 0.25-in.-dia detector had a 5% loss in neutron count sensitivity at 7000 rad/h. Similarly, the 0.5-in.-dia detector had a 7% loss at 13,000 rad/h and the 1-in.-dia detector had a 5% loss in sensitivity at 1000 rad/h. Uranium-235 fission chambers were operated successfully in fields above 100,000 rad/h with no loss in neutron counting sensitivity. Shielding calculations were done to determine the appropriate shield thickness needed for a /sup 10/B-lined neutron detector to operate in a 50,000 rad/h field, typical of light-water-reactor spent-fuel assemblies stored in water. 7 refs., 4 figs., 9 tabs.

Science.gov (United States)

... Zou, S. and RA Dalrymple, ``Coastal Sediment Transport Simulation by Smoothed Particle Hydrodynamics,'' Journal of Waterways, Ports, Coastal ...

British Library Electronic Table of Contents (United Kingdom)

To determine the proportion of rheumatoid arthritis (RA) patients receiving preventive health care according to US Preventive Services Task Force recommendations compared with a community-based population sample, with emphasis on dyslipidemia testing, given the increased risk of cardiovascular disease (CVD) in RA patients. Patients with RA (ICD-9 code 714.0 at ?2 office visits with a rheumatologist) and a primary care physician (PCP) at the Geisinger Health System (GHS) were identified through electronic health records. The records were searched back from 3/31/08 for the length of time required to satisfy each outcome measure. Percentages were compared with population testing rates using the Pearson Chi-square test. Eight hundred and thirty-one RA patients were compared to 169,476 subjects...

Energy Technology Data Exchange (ETDEWEB)

This study has confirmed large seasonal and daily variations of Rn in soil gas, developed models for the effects of temperature and moisture on air-water Rn partition, inhibited Rn diffusion from wet soil into sparse large air-filled pores and effects of diffusion into bedrock, demonstrated that organic matter is a major host for 226Ra in soils and that organic-bound Ra largely determines the proportion of 222Rn emanated to pore space, shown that in contrast 220Rn is emanated mainly from 224Ra in Fe-oxides, detected significant disequilibrium between 226Ra and 238U in organic matter and in some recent glacial soils, demonstrated by computer models that air convection driven by temperature differences is expected in moderately permeable soils on hillsides.

British Library Electronic Table of Contents (United Kingdom)

Purpose Evaluation of the all-trans retinoic acid (t-RA) effects on in vitro maturation (IVM) and in vitro fertilization (IVF) of immature mouse oocytes in the presence and absence of granulosa cell monolayer. Methods Denuded oocytes isolated from mice ovaries and matured in IVM medium alone (Control I), IVM medium in the presence of granulosa cells (Control II), IVM medium with t-RA (Experimental I) and IVM medium simultaneously with t-RA and granulosa cells (Experimental II). After 24?h, matured oocytes were fertilized in T6 medium and their development was followed until the blastocyst stage. Metaphase II oocytes ploidy were evaluated by chromosome counting. Results The t-RA group compared to the control groups showed no obvious abnormalities. Additionally maturation and embryo developm...

International Nuclear Information System (INIS)

Plutonium and minor actinides produced as by-products of the UO_2 nuclear cycle could be considered as waste or energy source depending on the strategy selected in the nuclear energy programme. Considering Pu and Minor Actinides as a source, they can be burned in existing water reactor for diminishing the radiotoxicity of the spent fuel, it is necessary to use 'inactive' materials as matrix like ThO_2. ThO_2 matrix has demonstrated its Pu burning efficiency and higher corrosion resistance than UO_2. Uranium-plutonium mixed oxide (MOX) fuel efficiency is low because the presence of U in MOX results in the creation of some new Pu under irradiation. The dissolution behaviour of irradiated (Th,Pu)O_2 pellets with burn-up of 38.8 MWd/kg Th has been studied in carbonated (20 mM HCO_3"-), deionised and granite ground water solution in a hot cell. The dissolution behaviour of Th, Pu, U and Np was studied in order to find out ...

Science.gov (United States)

... matrices were written in the form R(t) = R'(t) Mpp T V(t) = V (t) MppT (C-3) 49 Page 59. ... GW~/EE/64.-9 where ra(!L) rb(L) 0 MPPT= (C-4) [rb(L) ra(L) 0 ...

International Nuclear Information System (INIS)

One of the limiting contributors to the heat load constraint for a long term spent fuel repository is the decay of americium-241. A possible option to reduce the heat load produced by Am-241 is to eliminate it via transmutation in a light water reactor thermal neutron environment, in particular, by taking advantage of the large thermal fission cross section of Am-242 and Am-242m. In this study we employ lattice loading optimization techniques to define the loadings and arrangements of fuel pins with blended americium and uranium oxide in boiling water reactor bundles, specifically, by defining the incineration of pre-loaded americium as an objective function to maximize americium transmutation. Subsequently, the viability of these optimized lattices is tested by assembling them into bundles with Am-spiked fuel pins and by loading these bundles into realistic three-dimensional BWR core-wide simulations ...

International Nuclear Information System (INIS)

The ability of "9"9"mTc-IgG scintigraphy to predict the development of rheumatoid arthritis (RA) in 47 patients with arthralgia was investigated. "9"9"mTc-IgG scintigraphy and the serum test for rheumatoid factor (RF), measured at the beginning of a year long study, were compared for their ability to predict RA. During the study 8 patients developed RA. The specificity and positive predictive values of RF in predicting RA were 79% and 50% respectively, and for "9"9"mTc-IgG scintigraphy 97% and 88%. The sensitivity and negative predictive values of RF were 100% and of "9"9"mTc-IgG-scintigraphy 88% and 97%, respectively. In conclusion, "9"9"mTc-IgG scintigraphy has additional value to RF with respect to the prediction of the development of RA in patients with arthralgia. 27 refs.

International Nuclear Information System (INIS)

Radiochemical results ("2"3"4U, "2"3"8U, "2"2"6Ra and "2"2"8Ra activities; "2"3"4U/"2"3"8U, "2"2"8Ra/"2"2"6Ra and "2"2"6Ra/"2"3"8U activity ratios) are reported for 31 drinking water samples collected from 8 wells, 12 hot mineral springs, 5 cold springs, and 6 tap water samples. Activities of the Ra isotopes were measured by gamma spectrometry using a low background and high efficiency well type germanium detector. The U isotopes were counted in an alpha spectrometer. The results show that "2"3"8U activity varies between 4.46 and about 308 mBq l"-"1 in wells, 0.59 and 17.88 mBq l"-"1 in hot springs, 2.19 and 3.21 mBq l"-"1 in cold springs, and between 2.5 and 15.70 mBq l"-"1 in tap water samples. "2"3"4U/"2"3"8U activity ratio is generally higher than 1 which is the equilibrium activity ratio value in closed systems. It varies in the ranges 0.79-2.11 in wells, ...

Energy Technology Data Exchange (ETDEWEB)

The objective of this paper is to perform an impact analysis of the Interim Storage Cask (ISC) of the Fast Flux Test Facility (FFTF) for a 4-ft end drop. The ISC is a concrete cask used to store spent nuclear fuels. The analysis is to justify the impact force calculated by General Atomics (General Atomics, 1994) using the ILMOD computer code. ILMOD determines the maximum force developed by the concrete crushing which occurs when the drop energy has been absorbed. The maximum force, multiplied by the dynamic load factor (DLF), was used to determine the maximum g-level on the cask during a 4-ft end drop accident onto the heavily reinforced FFTF Reactor Service Building`s concrete surface. For the analysis, this surface was assumed to be unyielding and the cask absorbed all the drop energy. This conservative assumption simplified the modeling used to qualify the cask`s structural integrity for this accident condition.

International Nuclear Information System (INIS)

For the reuse of a waste salt from an electrorefining process of a spent oxide fuel, a separation of rare earth elements by an oxidative precipitation in a LiCl-KCl molten salt was tested without using precipitate agents. From the results obtained from the thermochemical calculations by HSC Chemistry software, the most stable rare earth compounds in the oxygen-used rare earth chlorides system were oxychlorides (EuOCl, NdOCl, PrOCl) and oxides (CeO2, PrO2), which coincide well with results of the Gibbs free energy of the reaction. In this study, similar to the thermochemical results, regardless of the sparging time and molten salt temperature, oxychlorides and oxides were formed as a precipitant by a reaction with oxygen. The structure of the rare earth precipitates was divided into two shapes: small cubic (oxide) and large plate-like (tetragonal) structures. The conversion efficiencies of the rare earth elements to their molten salt-insoluble ...

International Nuclear Information System (INIS)

It was agreed that both of the Republic Of Korea (ROK) and the International Atomic Energy Agency (IAEA) cooperate with the new approach using the Unattended Monitoring Systems (UMS), the mailbox declaration and the short notice inspections, which can decrease the agency inspection efforts and operator' s burden. As a result of the positive cooperation of ROK for applying new approach, the IAEA could accomplish a successful implementation of new safeguards approach during last transfer campaigns at Wolsung site. As of the end of Feb. 2007, the IAEA installed the UMS at Wolsung units 2, 3 and 4. The transfer campaigns at Wolsung units 2 and 4 are performing from April, 2007. The IAEA will install the UMS at Wolsung unit 1 by the end of 2007. Then agency will inspect all of transfer campaign at Wolsung units by using new approach with UMS, mailbox declaration and SNI. It is based on the SSAC full cooperation with IAEA to meet agency safeguards goals.

Energy Technology Data Exchange (ETDEWEB)

The Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico, has been constructed to be a repository for transuranic (TRU) radioactive wastes generated from the US defense activities. In order to use WIPP as a repository for permanent disposal of TRU waste, the US Department of Energy (DOE) has to demonstrate compliance with the Standards for the Management and Disposal of Spent Nuclear Fuel, High Level and Transuranic Radioactive Wastes'' promulgated by the US Environmental Protection Agency (EPA) in the US Code of Federal Regulations 40 CFR Part 191. The DOE initially plans to perform experiments with a small quantity of waste at WIPP and would like to bring additional quantities for operational demonstration'', before determining whether WIPP is to be a repository for permanent disposal. There are serious problems in pursuing this course of action from an operational point of view. It would ...

International Nuclear Information System (INIS)

At Tarapur Atomic Power Station 3 and 4, 540 MWe Pressurised Heavy Water Reactors, core being large in size requires a continuous in core monitoring for local flux disturbances. Nearly 200 Self Powered Neutron Detectors (SPNDs) of the Straight Individually Replaceable (SIR) type are distributed in the reactor core. For purpose of reactor regulation and protection, cobalt SPNDs that have a prompt response for changes in power is used for in-core flux mapping, vanadium SPNDs that provide accurate measure of neutron flux, even though having slow response is used In core SPNDs are placed in Vertical Flux Units (VFU) and Horizontal Flux Units (HFUs). These SPNDs were to be replaced at regular intervals to meet the design intent. Cobalt SPNDs have dose rates of the order of 1000 Gy/h and the Mineral Insulated (MI) cables of Vanadium SPNDs have dose rates of the order of 100 Gy/h. So far 3 Cobalt SPNDs were removed from HFUs and are being stored in lead shielding inside ...

International Nuclear Information System (INIS)

A pyrochemical partitioning process has been conceptually designed so that the transmutation of spent LWR fuels in PEACER can produce mainly low-level waste (Class C waste) for near-surface burial. Chloride salt technology developed for IFR has been employed as the baseline. Electrorefining, reductive extraction and salt recycling steps are used to construct overall flowsheet in order to support PEACER operation. The decontamination factor for transuranic elements was estimated based on both thermodynamic models and reported experimental data. It is expected that overall decontamination factor can be as high as 10"5 for transuranic elements. Final wastes from pyrochemical processing for PEACER are noble metals, alkaline earth metal, and lanthanides. The final wastes are stabilized by mixing with zeolite and glass-frits such that concentration limit for class C waste can be met. The volume of Class C waste is estimated to be small enough to make ...

Energy Technology Data Exchange (ETDEWEB)

Uranyl nitrate solution from 200-Area processing of spent SRP fuel tubes is now sent to Oak Ridge Y-12 for conversion of uranium metal. However, after implementation of the powder metallurgy (P/M) process, U{sub 3}O{sub 8} powder will be needed at SRP but not uranium metal. U{sub 3}O{sub 8} powder for fabrication and irradiation tests was produced during development of P/M at SRL by firing UO{sub 3}, obtained from Y-12, at 800{degrees}C for 6 hours in a low grade nitrogen atmosphere. The UO{sub 3} powder was produced by denitration of unsulfated uranyl nitrate solution. The stoichiometry, particle size distribution, surface area and density of the Y-12 and SRL powders were measured. A comparison was then made between SRL U{sub 3}O{sub 8} produced at 800{degrees}C in nitrogen and in air and U{sub 3}O{sub 8} produced at Y-12 at other heating temperatures.

Energy Technology Data Exchange (ETDEWEB)

The Oklo natural fossil fission reactors in Gabon, Equatorial Africa, have been studied as a natural analogue for spent nuclear fuel in a geological environment. For these studies, it is important to know what has happened to these reactors since they formed. This review is focussed on existing geological and geochronological information concerning the Oklo reactors and the surrounding ore. A sequence of geological and geochemical events in the Oklo area, as described in the literature, is given. The data and the studies behind this established geochronology are discussed and evaluated. Of the regional geology, special attention is given to the dating of the Francevillian sediments, and the intrusion of a dolerite dyke swarm. The processes that led to the mineralisation at Oklo, the subsequent formation of the nuclear reactors and later migration of fission products are described. Further discussion concerns the studies of the dolerite dyke ...

Energy Technology Data Exchange (ETDEWEB)

This research focuses on the development and application of high order statistical analyses applied to measurements performed with subcritical fissile systems driven by an introduced neutron source. The signatures presented are derived from counting statistics of the introduced source and radiation detectors that observe the response of the fissile system. It is demonstrated that successively higher order counting statistics possess progressively higher sensitivity to reactivity. Consequently, these signatures are more sensitive to changes in the composition, fissile mass, and configuration of the fissile assembly. Furthermore, it is shown that these techniques are capable of distinguishing the response of the fissile system to the introduced source from its response to any internal or inherent sources. This ability combined with the enhanced sensitivity of higher order signatures indicates that these techniques will be of significant utility in a variety of applications. Potential ...

Energy Technology Data Exchange (ETDEWEB)

The rates of hydrogen release from simulated high-level liquid waste (SHLLW) during irradiation of {sup 60}Co {gamma}-rays have been studied using a laboratory-scale apparatus. The SHLLW used was made up of 28 different metal elements, the concentrations of which were determined from a computed data (ORIGEN-2) for a Purex spent fuel (burn-up of 4.5 GWd/t, cooling time of 4 years) except for those of corrosion products (Fe, Cr and Ni) and of chemical process additive (P). The nitrate ion concentration of the SHLLW was 4.97 M. The G-value of hydrogen release was 0.0164 for stirred conditions, but for nonstirred conditions with dose rate : 2.8 kGy/h and with the SHLLW solution height more than 8 cm, the following relation was found between the G-value and the solution height (d cm); G(H{sub 2}) = 0.100 d{sup -1.6}. The G-values of oxygen and nitrogen releases were also determined. (author)

Energy Technology Data Exchange (ETDEWEB)

One option in establishing advanced technology for disposing of high-level radioactive waste (HLW) is to remove TRU elements from HLW and then transform them into short-lived radionuclides. Tertiary pyridine-type anion exchange resin was investigated as a separation medium for spent fuel reprocessing using a HCl solution. Distribution coefficients of typical elements, along with uranium, were measured in the pyridine resin--HCl solution system, where HCl concentration was varied from 1 to 9 M. The results reveal: elements of the alkali metal, alkaline earth and lanthanide groups are not adsorbed, the adsorption patterns of the transition metals differ according to HCl concentration, and uranium is adsorbed in the high HCl concentration region. Furthermore, the present report describes the experimental results obtained for the adsorption behaviour of uranium, americium, curium and various lanthanide elements on tertiary pyridine-type anion ...

Energy Technology Data Exchange (ETDEWEB)

A pyrochemical process has been introduced and utilized so that the transmutation of spent PWR fuel in PEACER can produce mainly low and intermediate level waste for near surface disposal. Major radioactive nuclides from PEACER pyroprocessing are composed of TRU and LLFP. In this study, the requirement for the final waste from PEACER is evaluated based on the methodology for establishment of waste acceptance criteria. Also, sensitivity analysis for several input parameters is conducted in order to determine acceptable decontamination factor (DF) and LLFP removal efficiency and to find out input parameter that extremely have an effect on DF. As a result of the study, LLFP removal efficiency, especially Sr-90 and Tc-99, is proved to be a major nuclide which contributes to annual dose by human intrusion scenario rather than TRU DF. More than 98.5% of LLFP have to be removed to meet below dose constraint within the DF more than 5.0E + 03. Besides, ...

International Nuclear Information System (INIS)

The thermal environment of a hypothetical large indoor hydrocarbon pool fire is more complex compared to outdoor fires and can be more severe for engulfed objects. In order to analyze potential thermal environments for interim storage of spent fuel casks or low-level radioactive waste packages engulfed in pool fires numerical simulations with the CFD fire code FDS3 were carried out for different storage configurations. In addition, data of indoor pool fire experiments were used to validate the model for this type of application. A series of pool fire experiments under different ventilation conditions and varied pool surface (1 m"2 - 4 m"2) inside a compartment of 3.6 m x 3.6 m x 5.7 m was conducted at iBMB (Institut fuer Baustoffe, Massivbau und Brandschutz) of Braunschweig University of Technology, Germany. The instrumentation included thermocouples, heatflux and pressure gauges, bi-directional flow probes and gas concentration measurements. A ...

International Nuclear Information System (INIS)

As the nuclear world is changing, the non-proliferation and safeguard systems have to change along the global development. Nuclear security as well as safety must be involved in all phases. Thus, modernization of thinking is a must. State system of accounting for and control of nuclear material (SSAC) is a basis, but now it is time to move ahead. Safeguards is not any more only to verify the declared nuclear materials but it is to inform the international safeguards society transparently but confidentially about the nuclear fuel cycle related activities and trade, and to confirm that there are no undeclared activities related to the nuclear fuel cycle in the states. Only strong SSAC with enhanced capabilities, activities and rights can meet the demand. The proliferation of nuclear weapons is a threat that also urges cooperation at all levels. International, regional and state systems must work closely together, and there is need and space for ...

Energy Technology Data Exchange (ETDEWEB)

Ammonium sulfate is produced from spent sulfuric acid from the production of alcohols. Certain mineral and organic impurities exist in the sulfuric acid that impair the quality of the ammonium sulfate. A detailed method for the chemical analysis for the impurities is presented.

Science.gov (United States)

... is disengaged from the spent rocket motor by a ... of the outbound and incoming modu- ... meteorological sensor and reference resistor is accomplished ...

Science.gov (United States)

The report gives results of a study of sintering and leaching mechanisms of fly ash/spent sodium sorbent mixtures from a dry injection flue gas desulfurization (FGD) process. It includes an estimate of the economics of pelletizing and sintering to handle the fly ash and spent sor...

Energy Technology Data Exchange (ETDEWEB)

The synovial and bone uptake of tracer in the knees of patients with rheumatoid arthritis (RA) was quantified using {sup 99}Tc{sup m}-hexamethyl propylene amine oxime-labelled leucocytes and {sup 99}Tc{sup m}-methylene diphosphonate (MDP), respectively. Significant neutrophil migration and MDP uptake occurred in the knees of patients with RA irrespective of the disease duration. In all but one patient neutrophil migration was reduced after intra-articular steroid injection. The change in MDP uptake after steroid injection was variable. There was a significant correlation between the percentage reduction in neutrophil migration and pain score, while the latter correlated poorly with the change in MDP uptake. The quantification of the neutrophil component of the inflammatory process is a sensitive index for monitoring RA activity and response to pharmacological interventions, while quantitative bone scintigraphy should not be ...

UK PubMed Central (United Kingdom)

There is ample evidence suggesting that superantigens may act as a triggering factor in the pathogenesis of rheumatoid arthritis (RA). We investigated whether superantigen could activate T cells in...Full Text Available

International Nuclear Information System (INIS)

This paper describes the evaluation of antibody coated polystyrene beads using "1"2"5I-labelled T_3 ("1"2"5I-T_3). (author).

International Nuclear Information System (INIS)

A conceptual framework is proposed within which the notion of risk as normally used in risk assessment (RA) could be enlarged in line with the real substance of social issues of technology policy, to help avoid RA's threatened irrelevance to social decision making. It is argued that the frequent organizational incoherence and thus the unviability of modern technology arises from 'social alienation' between the innovation-commitment phase and the implementation of the technology in society. The roles of technical elites and of particular concepts of technology in this alienation are emphasized. One of the case studies deals with 'Nuclear power - myths of scientific and organizational realism' and discusses the UK nuclear 'programme' and the Three Mile Island accident. (author).

International Nuclear Information System (INIS)

Twenty rheumatoid arthritis (RA) patients with involvement of the cranio-cervical junction (CCJ) were given CT scan. In most cases the myelo-CT technique with the lumbar injection of small quantities of non-ionic contrast medium was adopted; CT easily diagnosed the horizontal and vertical dislocations on the CCJ, the bone strcture lesions and the inflammatory masses. A high frequency of medullary compressions was also found in these patients. Myelo-CT appears to be the technique of choice for the precise definition of the various factors responsible for the neurological complications found in RA of the CCJ.

Energy Technology Data Exchange (ETDEWEB)

Promoting the exchange of information related to implementation of the As Low as Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report was prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health. It contains the fifth in a series of bibliographies on dose reduction at DOE facilities. The BNL ALARA Center was originally established in 1983 under the sponsorship of the Nuclear Regulatory Commission to monitor dose-reduction research and ALARA activities at nuclear power plants. This effort was expanded in 1988 by the DOE`s Office of Environment, Safety and Health, to include DOE nuclear facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, ...

International Nuclear Information System (INIS)

In Japan, the development of next-generation Light Water Reactor has been launched since April 2008. The development program will be completed in 2015. The purpose of development is to cope with the replacement for existing nuclear power plants after 2030 in Japan and the expanding demand for nuclear power in the world; 'Nuclear Renaissance.' The reactor also aims to be global standard at around 2030. The requirements for global standard and domestic users have been investigated through the feasibility study of past 2 years, 2006-2007, and six innovative features or 'Core-Concepts' were established as follows. A) Reactor core system with uranium enrichment above 5% for significant decrease of spent fuel discharge and prominent higher availability B) Long-life materials and innovative water chemistry technologies for 80 years plant lifetime and significant reduction of occupational dose C) Seismic isolation technologies to standardize plant ...

Energy Technology Data Exchange (ETDEWEB)

Fabrication processes for FFTF fuel pins are described.

International Nuclear Information System (INIS)

Full text: Dissolution test (DT) is one of the important specifications of PFBR MOX fuel. It is a macro homogeneity test used to ascertain that the dissolution of spent fuel will be easier during reprocessing. The DT specification limit for PFBR MOX fuel is ? 1.0 wt.% residue of the whole pellet when it is dissolved in HNO3 under reflux conditions. For this test, sintered MOX pellet is dissolved in reflux condition in 60 ml of 12M HNO3 for 10 hours. This test is carried out for each batch and during complete MOX campaign a lot of MOX waste in liquid form is being generated. This liquid waste is in clean form and U and Pu is required to be recovered from the clean waste. Different methods like direct denitration, precipitation and solvent extraction are available but either the product is not of good quality for pelletization or secondary liquid wastes are generated. Microwave denitration is a faster and ...

Science.gov (United States)

...identified as oil when the shipment document accurately describes the material as: aviation fuel, diesel fuel, fuel oil, gasoline, jet fuel, kerosene, motor fuel, or...

Science.gov (United States)

To assess the influence of a live attenuated oral vaccine against porcine post-weaning colibacillosis (PWC) induced by enterotoxigenic Escherichia coli (ETEC) on mucosal lymphoid cell CD45 isoforms expression, experimental group of weaned pigs (n=6) was immunized orally with F4ac+ non-ETEC strain (day 0) and challenged with F4ac+ ETEC strain 7 days latter. Non-immunized ETEC-infected pigs (n=6) served as control. All pigs were killed on post-challenge day 7. The small intestine was excised for isolation of jejunal lamina propria (JLP) and ileal Peyer's patch (IPP) lymphocytes and immunohistochemical studies. The results obtained by immunophenotyping of isolated cells show that the proportion of CD45RA+ and CD45RC+ JLP, but not IPP, cells were higher in the non-ETEC-immunized ETEC-infected pigs versus non-immunized infected. Additionally, while CD45RA+ JLP cells increased only slightly, the expression of CD45RC isoform on the JLP cells was ...

International Nuclear Information System (INIS)

The author describes feasibility study on the scheme of disposing spent guide tubes at Daya Bay nuclear power plant, including scheme selection, packing cask design, disposal operation and technological process, budget estimation and safety evaluation. The research results show that the new scheme is feasible on technology, and reasonable on budget. If the scheme is adopted to dispose the spent guide tubes at Daya Bay nuclear power station, it can not only save large funds, but also accumulate experience on disposing non-compressible low and intermediate level solid radioactive wastes produced at nuclear power station for the country

International Nuclear Information System (INIS)

As part of the national survey to evaluate natural radioactivity in the environment, concentration levels of natural uranium and "2"2"6Ra have been analyzed in over 300 drinking water samples taken from different locations in Argentina. "2"2"6Ra was determined by "2"2"2Rn emanation and liquid scintillation counting, and natural uranium by a fluorimetric procedure. Values ranging from 0.03 to 24 #mu#g.l"-"1 of natural uranium and from 0.06 to 50 #mu#g.l"-"1, were measured on drinking water samples taken from tap water systems and private wells, respectively. Concentrations up to 15 mBq.l"-"1 and to 22 mBq.l"-"1 of "2"2"6Ra were found in drinking water samples taken from tap water systems and private wells, respectively. These values are compared with the reference values accepted for drinking water. Based on the water intake rate, the age distribution and the measured concentrations, an annual collective effective dose of ...

Energy Technology Data Exchange (ETDEWEB)

Following a referendum in Italy in 1987, the four Nuclear Power Plants (NPPs) owned and operated by the state utility ENEL were closed. After closing the NPPs, ENEL selected a ''safestore'' decommissioning strategy; anticipating a safestore period of some 40-50 years. This approach was consistent with the funds collected during plant operation, and was reinforced by the lack of both a waste repository and a set of national free release limits for contaminated materials in Italy. During 1999, twin decisions were made to privatize ENEL and to transform the nuclear division into a separate subsidiary of the ENEL group. This group was renamed Sogin and during the following year, ownership of the company was transferred to the Italian Treasury. On formation, Sogin was asked by the Italian government to review the national decommissioning strategy. The objective of the review was to move from a safestore strategy to a prompt decommissioning strategy, with ...

International Nuclear Information System (INIS)

The recovery of noble metals from the quaternary Mo-Ru-Rh-Pd alloy in the simulated insoluble residue was studied by means of Pb extraction and the recovery yield was found to be more than 90% for Ru, Rh and Pd. The recovery yield of Mo was very small due to the oxidative vaporization in air. The decontamination factor for Ce as a stand-in of Pu was determined to be around 200 for one stage. These results on the recovery yields of noble metals from the quaternary alloy and the decontamination factor for Ce are in good agreement with those for the ternary Mo-Ru-Pd alloy previously reported by the present authors. The mutual separation of noble metal elements recovered in Pb button was examined by means of re-extraction with molten Zn and the dissolution in boiling 3M or 6M nitric acid solution. The dissolution fractions of Pd and Rh recovered in the Pb button in boiling 3M and 6M nitric acid solution for 12 h were about 90 and 30%, respectively, and that of Ru was extremely low (#approx ...

International Nuclear Information System (INIS)

Borosilicate based glass formulations have been found suitable for fixing the HLW (high level radioactive liquid waste) generated after reprocessing of the spent nuclear fuel. As the glass experiences continuous irradiation by #alpha#, #beta#a, and #gamma# radiations from the radioactive components of HLW, alteration in the glass structure may occur. Understanding of these structural evolutions of the nuclear waste glasses under irradiation is crucial to secure long term disposal and predict their behavior. In the present work, alkali based barium borosilicate glasses, having composition similar to that of Trombay Research Reactor waste glass were irradiated with high energy "1"2C beam and the radiation induced changes were monitored by micro Raman experiment. Since a "1"2C atom can be considered as a cluster of alphas, this beam was chosen to yield linear energy losses (LET) comparable to that in case of a particles. The irradiations were ...

International Nuclear Information System (INIS)

Active Process Water (APW) system is provided as a unitized system in TAPP-3 and 4. Maintenance on APW system requires shutdown of this system. As shut down heat exchangers are fed by APW system; during APW system shutdown cold shutdown state cannot be maintained. Therefore safety analysis is done to optimize the duration of reactor shutdown (which means low decay heat) after which APW shutdown can be taken with minimum water supply to the shutdown heat exchangers. Based on this analysis, it is proposed in technical specification that APW system shutdown can be taken after 7 days of reactor shutdown with shutdown heat exchangers supplied with about 20 % of normal APW flow. With this configuration, PHTS, moderator, end shield, calandria vault water temperature can be maintained within limits. A design provision is made at TAPP-3 and 4 to interconnect APW system of both units in such a way that one unit in addition to its own APW requirement can provide limited supply to shutdown heat ...

International Nuclear Information System (INIS)

In the Japan Research Reactor-II (JRR-2), the main source of _4_1Ar generation is the exhaust air from the horizontal experimental holes and the pneumatic tubes. For the horizontal experimental holes, the flow of exhaust air through the holes was decreased by improving the airtightness, and a decay duct of capacity 2.4 m_3 was installed in the middle of the exhaust line. In consequence, the release rate of _4_1Ar was reduced by 6-8%. For the pneumatic tubes, a mechanical shutter was installed in the tube. The shutter stops the exhaust air flow, except when the pneumatic tube is used. Prior to the use, the activated air in the tube is led to a decay tank. As a result, the _4_1Ar release rate was reduced by 10-20%. By the above means, the yearly exposure at the site boundary was reduced to 0.36 mR from 2.6 mR. In Hot Laboratory for metallurgical examination of spent fuel, the exhaust filtration system consists of filters in the cave, i.e. frame ...

Energy Technology Data Exchange (ETDEWEB)

During the years 1995-1999, an international effort was going on under the title DECOVALEX II, in order to study the coupled thermo-hydro-mechanical (THM) effects in the water containing fractured rock mass caused by the heat generation of spent fuel canisters in a repository. The project was a continuation project of DECOVALEX (1991-1996). The name comes from the acronym 'an international co-operative project for the DEvelopment of COupled models and their VALidation against EXperiments in nuclear waste isolation'. Radiation and Nuclear Safety Authority (STUK) was one of the eleven Funding Organizations in the international project. STUK's Research Team was the Technical Research Centre of Finland, VTT Communities and Infrastructure. To support and coordinate the national research work, STUK formed a National DECOVALEX II Group (NDG). The group had representatives from six research organisations and rock engineering ...

British Library Electronic Table of Contents (United Kingdom)

The features of present deposits that form in the vicinity of hot springs can provide clues to the parameters of paleowaters in places of past hydrothermal activity marked by remnant carbonate and/or siliceous sinter. We investigated a large carbonate body at the Garga hot spring developing in the Baikal zone of nitric hydrotherms in the Barguzin Rift Zone valley. The main focus was on the structure of the carbonate mound, as well as on the partitioning of radioactive elements between the cyanobacterial mat and the inorganic component of the body (the issue that has never been explored before). The cyanobacterial community of the Garga spring is an active biosorbent of 226Ra, 228Ra, 210Pb. The radionuclides accumulated by biosorption become preserved in minerals that form within the bacter...

Energy Technology Data Exchange (ETDEWEB)

The quality of fuel oils is discussed. The problems that the fuel oil marketer must deal with that relate to the quality of the fuel oil are described.

Science.gov (United States)

fuel obligation under the RFS2 program for the production or importation of conventional jet fuel, RINs can be generated for renewable jet fuel. Is that right? A: As described in...

Science.gov (United States)

of different transportation fuels (e.g., gasoline, diesel fuel, aviation gasoline, jet fuel, residual fuel oil). Subsequent calculations are performed to estimate the share...

UK PubMed Central (United Kingdom)

In t(15;17) acute promyelocytic leukemia, all-trans retinoic acid (RA) induces leukemic cell maturation in vitro and remission in acute promyelocytic leukemia patients, but in vivo treatments invariably...Full Text Available

Science.gov (United States)

Apr 16, 2007 ... with a 1-sigma error radius of about 0.5 arc sec. ... It was consistent with a single point source at: RA (J2000) = 03h 13m 57.6s ... At 13000 sec. after the trigger, the magnitude in the UVOT B filter was ~19.9 with a ...

International Nuclear Information System (INIS)

Concentrations of 226Ra, 232Th, and 40K measured in bauxite waste, local building materials, and soils are presented and used in model equations to estimate the effective gamma dose-equivalent increments over background in the center of a standard-sized room in a prototype house. Calculated and measured values compare reasonably well.

Energy Technology Data Exchange (ETDEWEB)

This report summarizes the activities and major accomplishments for the Idaho National Laboratory Advanced Radiotherapy Research Program for calendar year 2004. Topics covered include boron analysis in biological samples, computational dosimetry and treatment planning software development, medical neutron source development and characterization, and collaborative dosimetry studies at the RA-1 facility in Buenos Aires, Argentina.

International Nuclear Information System (INIS)

The leaching of phosphorus (P), nitrogen (N), and radionuclides "2"3"2Th, "2"2"6Ra, "2"2"8Ra, and "4"0K from Joel sands amended with red mud/gypsum (RMG) at 9 rates (0, 2, 4, 8, 16, 32, 64, 128, and 256 t/ha) was measured using columns. Intense leaching conditions (34 mm/day for 12 days) and a high rate of applied P (320 kg/ha as superphosphate) and N (680 kg/ha as ammonium nitrate) were used to simulate extremes of irrigated vegetable production on the Swan Coastal Plain. Addition of the highest rate of RMG (256 t/ha) reduced leaching of fertiliser P and ammonium-nitrogen (NH4-N) by 85% and 50%, respectively, compared with 0 t/ha after 12 days. At 64 t RMG/ha P leaching was reduced 50% compared with 0 t/ha. Nitrate-nitrogen (NO3-N) leaching was not affected by addition of RMG. Reduced leaching of NH4-N was attributed to an increase in cation exchange capacity of the soil with the addition of RMG. Bicarbonate-extractable P in the soil increased ...

Energy Technology Data Exchange (ETDEWEB)

Recycling of the spent rinse water discharged from the wet benches commonly used in semiconductor processing is one tactic for responding to the targets for water usage published in the 1997 National Technology Roadmap for Semiconductors (NTRS). Not only does the NTRS list a target that dramatically reduces total water usage/unit area of silicon manufactured by the industry in the future but for the years 2003 and beyond, the NTRS actually touts goals which would have semiconductor manufacturers drawing less water from a regional water supply per unit area of silicon manufactured than the quantity of ultrapure water (UPW) used in the production of that same silicon. Achieving this latter NTRS target strongly implies more widespread recycling of spent rinse waters at semiconductor manufacturing sites. In spite of the fact that, by most metrics, spent rinse waters are of much higher purity than incoming municipal waters, ...

International Nuclear Information System (INIS)

... NUCLEAR FUEL CYCLE AND FUEL MATERIALS borophosphate glass

Science.gov (United States)

No Abstract Available

International Nuclear Information System (INIS)

... electric utilities electrochemistry energy storage fuel cells organizational models

International Nuclear Information System (INIS)

... FUEL CYCLE AND FUEL MATERIALS MATERIALS SCIENCE chromium-nickel

UK PubMed Central (United Kingdom)

1. The mechanical power spent to accelerate the limbs relative to the trunk in level walking and running, Ẇ_int, has been measured at various `constant' speeds (3-33...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires ...

Energy Technology Data Exchange (ETDEWEB)

Numerical solutions given by a vorticity-velocity method are presented for combined free and forced laminar convection in the thermal entrance region of a horizontal rectangular channel without the assumptions of large Prandtl number and small Grashof number. The channel wall is heated with a uniform wall heat flux. Typical developments of temperature profile, secondary flow, and axial velocity at various axial positions in the entrance region are presented. Local friction factor and Nusselt number variations are shown for Rayleigh numbers Ra = 10{sup 4}, 3 {times} 10{sup 4}, 6 {times} 10{sup 4}, and 10{sup 5} with the Prandtl number as a parameter. The solution for the limiting case of large Prandtl number and small Grashof number obtained from the present study confirms the data of existing literature. It is observed that the large Prandtl number assumption is valid for Pr = 10 when Ra {le} 3 {times} 10{sup 4} but for a larger Prandtl number ...

Energy Technology Data Exchange (ETDEWEB)

Conoco`s Crosswell Reservoir Characterization (CRC) team is using Decision and Risk Analysis (D and RA) to evaluate the potential economic impact of the authors` projects. D and RA can be a useful tool for refining the direction of a project and communicating the value of information in economic terms. Initial studies show that CRC has high potential economic value. Three steps were used to analyze the value of the CRC project. A high-level, industry-wide model was developed to look at the technical, application, and commercialization success of the technology. The team considered an infill drilling decision in shallow shelf carbonates as a detailed case. For that case, fully integrating crosswell information with supporting data (well logs, cores, production information, etc.) leads to a three-fold increase in expected value. The third step, a portfolio analysis of all crosswell applications, was initiated, but not completed. Given the ...