Operating experience and development of fluidized-bed denitrators for UNH at Tokai Reprocessing Plant

International Nuclear Information System (INIS)

Sasaki, Minoru; Nakamichi, Hideya; Takeda, Seiichiro; Kubota, Kanya; Katoh, Shuji

1983-01-01

The fluidized bed denitrator for uranyl nitrate hexahydrate (UNH) at Tokai reprocessing plant has been operated since 1976. About 170 tons of spent fuel have been reprocessed, and the denitrator has encountered numerous operational problems during the period. This report deals with these technical problems and the associated countermeasures taken, including the dismantling and reconstruction of equipment and the improvement of operating method. The major problems encountered were as follows: (1) the crystallization of UNH on the UNH feeding line, (2) spray nozzle clogging and candle filter clogging, (3) particle growth, (4) plugging of the drawing-out line by nozzle caking, and (5) slugging in fluidized-bed denitration. The total quantity and quality of UO 3 products obtained so far at the plant are also briefly described together with some future R and D programs such as the improvement of UO 3 reactivity and the automation of denitrators. (Aoki, K.)

Dynamic simulation in the process of pressurized denitration based on oxy-fuel combustion

Science.gov (United States)

Huang, Qiang; Zhou, Dong

2018-02-01

Oxy-fuel combustion is considered as one of the most promising technologies for capturing CO2 from coal-fired power plants. It will greatly reduce the cost of gas purification if we remove NOx in the process of compression, which is the characteristic of oxy-combustion. In this paper, simulation of denitration process of oxy-fuel combustion flue gas was realized by the Aspen Plus software, systematically analyzed the effect of temperature, pressure, initial concentration of O2 and NO in the denitration process. Results show that the increasing of pressure, initial concentration of O2, initial concentration of NO and the decrease of temperature are all beneficial to the denitration process.

Fuel-pellet-fabrication experience using direct-denitration-recycle-PuO2-coprecipitated mixed oxide

International Nuclear Information System (INIS)

Rasmussen, D.E.; Schaus, P.S.

1980-01-01

The fuel pellet fabrication experience described in this paper involved three different feed powders: coprecipitated PuO 2 -UO 2 which was flash calcined in a fluidized bed; co-direct denitrated PuO 2 -UO 2 ; and direct denitrated LWR recycle PuO 2 which was mechanically blended with natural UO 2 . The objectives of this paper are twofold; first, to demonstrate that acceptable quality fuel pellets were fabricated using feed powders manufactured by processes other than the conventional oxalate process; and second, to highlight some pellet fabrication difficulties experienced with the direct denitration LWR recycle PuO 2 feed material, which did not produce acceptable pellets. The direct denitration LWR recycle PuO 2 was available as a by-product and was not specifically produced for use in fuel pellet fabrication. Nevertheless, its characteristics and pellet fabrication behavior serve to re-emphasize the importance of continued process development involving both powder suppliers and fuel fabricators to close the fuel cycle in the future

Study on the mechanism of a manganese-based catalyst for catalytic NOX flue gas denitration

Science.gov (United States)

Zhang, Lei; Wen, Xin; Lei, Zhang; Gao, Long; Sha, Xiangling; Ma, Zhenhua; He, Huibin; Wang, Yusu; Jia, Yang; Li, Yonghui

2018-04-01

Manganese-based bimetallic catalysts were prepared with self-made pyrolysis coke as carrier and its denitration performance of low-temperature SCR (selective catalyst reduction) was studied. The effects of different metal species, calcination temperature, calcination time and the metal load quantity on the denitration performance of the catalyst were studied by orthogonal test. The denitration mechanism of the catalyst was analyzed by XRD (X-ray diffraction), SEM (scanning electron microscope), BET test and transient test. The experiments show that: * The denitration efficiency of Mn-based bimetallic catalysts mainly relates to the metal type, the metal load quantity and the catalyst adjuvant type. * The optimal catalyst preparation conditions are as follows: the load quantity of monometallic MnO2 is 10%, calcined at 300°C for 4h, and then loaded with 8% CeO2, calcined at 350°Cfor 3h. * The denitration mechanism of manganese-based bimetallic oxide catalysts is stated as: NH3 is firstly adsorbed by B acid center Mn-OH which nears Mn4+==O to form NH4+, NH4+ was then attacked by the gas phase NO to form N2, H2O and Mn3+-OH. Finally, Mn3+-OH was oxidized by O2 to regenerate Mn4+.

Study on the mechanism of a manganese-based catalyst for catalytic NOX flue gas denitration

Directory of Open Access Journals (Sweden)

Lei Zhang

2018-04-01

Full Text Available Manganese-based bimetallic catalysts were prepared with self-made pyrolysis coke as carrier and its denitration performance of low-temperature SCR (selective catalyst reduction was studied. The effects of different metal species, calcination temperature, calcination time and the metal load quantity on the denitration performance of the catalyst were studied by orthogonal test. The denitration mechanism of the catalyst was analyzed by XRD (X-ray diffraction, SEM (scanning electron microscope, BET test and transient test. The experiments show that: ① The denitration efficiency of Mn-based bimetallic catalysts mainly relates to the metal type, the metal load quantity and the catalyst adjuvant type. ② The optimal catalyst preparation conditions are as follows: the load quantity of monometallic MnO2 is 10%, calcined at 300°C for 4h, and then loaded with 8% CeO2, calcined at 350°Cfor 3h. ③ The denitration mechanism of manganese-based bimetallic oxide catalysts is stated as: NH3 is firstly adsorbed by B acid center Mn-OH which nears Mn4+==O to form NH4+, NH4+ was then attacked by the gas phase NO to form N2, H2O and Mn3+-OH. Finally, Mn3+-OH was oxidized by O2 to regenerate Mn4+.

Microwave heating denitration device

International Nuclear Information System (INIS)

Sato, Hajime; Morisue, Tetsuo.

1984-01-01

Purpose: To suppress energy consumption due to a reflection of microwaves. Constitution: Microwaves are irradiated to the nitrate solution containing nuclear fuel materials, to cause denitrating reaction under heating and obtain oxides of the nuclear fuel materials. A microwave heating and evaporation can for reserving the nitrate solution is disposed slantwise relative to the horizontal plane and a microwave heating device is connected to the evaporation can, and inert gases for agitation are supplied to the solution within the can. Since the evaporation can is slanted, wasteful energy consumption due to the reflection of the microwaves can be suppressed. (Moriyama, K.)

Preparation of acid deficient solutions of uranyl nitrate and thorium nitrate by steam denitration

International Nuclear Information System (INIS)

Yamagishi, Shigeru; Takahashi, Yoshihisa

1996-01-01

Acid deficient heavy metal (HM) nitrate solutions are often required in the internal gelation processes for nuclear fuel fabrication. The stoichiometric HM-nitrate solutions are needed in a sol-gel process for fuel fabrication. A method for preparing such nitrate solutions with a controlled molar ratio of nitrate/metal by denitration of acid-excess nitrate solutions was developed. The denitration was conducted by bubbling a nitrate solution with a mixture of steam+Ar. It was found that steam was more effective for the denitration than Ar. The acid deficient uranyl nitrate solution with nitrate/U=1.55 was yielded by steam bubbling, while not by only Ar bubbling. As for thorium nitrate, acid deficient solutions of nitrate/Th≥3.1 were obtained by steam bubbling. (author)

Development of the thermal denitration in-storage-can step in the CEUSP process

International Nuclear Information System (INIS)

Vedder, R.J.; Collins, E.D.; Haas, P.A.

1986-01-01

A thermal denitration in-the-storage-can process has been developed for use in the Consolidated Edison Uranium Solidification Program Facility. This process is being used to convert approx.1000 kg of highly fissile and radioactive uranium to a solid form for safe long-term storage. The material being solidified also contains approx.300 kg of cadmium and approx.40 kg of gadolinium which had been combined with the uranium to provide criticality safety. The unique thermal denitration process was found to be extremely susceptible to entrainment of solids by splattering, foaming, or expulsion actions. The process connection nozzle, through which the feed solution and purging air are supplied and the emerging off-gases are discharged, and the off-gas handling system were modified extensively to permit operation without development of nozzle or line pluggage due to accumulation of solid deposits. A process study was made to determine the effects of feed components and process variables on the tendency of the reacting mixture to splatter, foam, or be expelled. Because of the equipment modifications and the selection of appropriate processing conditions, the feed material is being denitrated without significant problems

Restart plan for the prototype vertical denitration calciner

Energy Technology Data Exchange (ETDEWEB)

SUTTER, C.S.

1999-09-01

Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing.

Restart plan for the prototype vertical denitration calciner

International Nuclear Information System (INIS)

SUTTER, C.S.

1999-01-01

Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing

Characteristics of a continuous denitration by formic acid - electrolytic trimming of residual acid with accompanying the precipitation of metal ions

International Nuclear Information System (INIS)

Kim, G. W.; Kim, S. H.; Lim, J. G.; Lee, I. H.

2003-01-01

This work has studied the characteristics of destruction of nitric acid and precipitation of several metal ions in a continuous denitration process combining a denitration by formic acid and a residual acid-electrolytic trimming system. The metal ions of Zr, Mo, Fe, and Nd did not affect the electrodes at the step of electrolytic trimming of the residual acid after denitration by formic acid. The Mo ion in electrolytic solution enhanced the generation of nitrite ion during the electrolytic reaction. The mole ratio of formic acid to nitric acid fed into the continuous denitration reactor using formic acid affected much the final acidity, the precipitation yields of metal ions, the precipitate morphology. At the ratio of 1.65, the process had the lowest final acidity of less than 0.1 M, and the precipitation yields of Zr and Mo reached 95% and 83%, respectively as the highest values

Test Plan for Radioactive Testing of a Vertical Direct Denitration Calciner

International Nuclear Information System (INIS)

COMPTON, J.A.

1999-01-01

A vertical calciner will be used to demonstrate the direct denitration process for converting plutonium-bearing liquors to stable plutonium rich solids. The calciner and some of its associated equipment were previously tested with non-radioactive chemicals to demonstrate operability

Test Plan for Radioactive Testing of a Vertical Direct Denitration Calciner

Energy Technology Data Exchange (ETDEWEB)

COMPTON, J.A.

1999-10-05

A vertical calciner will be used to demonstrate the direct denitration process for converting plutonium-bearing liquors to stable plutonium rich solids. The calciner and some of its associated equipment were previously tested with non-radioactive chemicals to demonstrate operability.

Denitration of simulated high-level liquid wastes and selective removal of cesium with zeolites

Energy Technology Data Exchange (ETDEWEB)

Mimura, Hitoshi; Kanno, Takuji [Tohoku Univ., Sendai (Japan). Research Inst. of Mineral Dressing and Metallurgy; Kimura, Toshiya

1982-03-01

Denitration of high-level liquid wastes (HLW) from nuclear fuel reprocessing has been studied. Selective removal of Cs has been also examined with various types of zeolites. The following zeolites were used in this study; Na-synthetic mordenite (NaSM), Na-natural mordenite (NaNM), Na-natural clinoptilolite (NaCP) and H-synthetic mordenites (HSM). The effective denitration is found in the simulated HLW (15 components, 2N HNO/sub 3/ soln.) containing platinum group elements in the case of the addition of formic acid, and the pH of the solution shows the value of 5.4 when the excess formic acid ((HCOOH)/(HNO/sub 3/) = 2.0) was added. Platinum group elements may react as a catalyst for the decomposition of nitric acid and the excess formic acid. The break-through properties of NaSM column are poor for the simulated HLW, and the selective removal of Cs appears to be difficult. On the other hand, good results are obtained in the denitrated HLW, i.e., break-through capacity, total capacity and column utilization are 59.4 (meq./100 g zeolite), 147 (meq./100 g zeolite) and 40.4 (%), respectively. The break-through properties of NaSM and NaNM are superior to those of HSM. The break-through capacity and column utilization increase with an increase in column temperature.

Denitration of simulated high-level liquid wastes and selective removal of cesium with zeolites

International Nuclear Information System (INIS)

Mimura, Hitoshi; Kanno, Takuji; Kimura, Toshiya.

1982-01-01

Denitration of high-level liquid wastes (HLW) from nuclear fuel reprocessing has been studied. Selective removal of Cs has been also examined with various types of zeolites. The following zeolites were used in this study; Na-synthetic mordenite (NaSM), Na-natural mordenite (NaNM), Na-natural clinoptilolite (NaCP) and H-synthetic mordenites (HSM). The effective denitration is found in the simulated HLW (15 components, 2N HNO 3 soln.) containing platinum group elements in the case of the addition of formic acid, and the pH of the solution shows the value of 5.4 when the excess formic acid ([HCOOH]/[HNO 3 ] = 2.0) was added. Platinum group elements may react as a catalyst for the decomposition of nitric acid and the excess formic acid. The break-through properties of NaSM column are poor for the simulated HLW, and the selective removal of Cs appears to be difficult. On the other hand, good results are obtained in the denitrated HLW, i.e., break-through capacity, total capacity and column utilization are 59.4 (meq./100 g zeolite), 147 (meq./100 g zeolite) and 40.4 (%), respectively. The break-through properties of NaSM and NaNM are superior to those of HSM. The break-through capacity and column utilization increase with an increase in column temperature. (author)

Obtention of Uo3 by means of denitration of uranyl nitrate in a fluidized-bed

International Nuclear Information System (INIS)

Santos, W.R. dos; Costa, P.A.

1990-01-01

A fluidized-bed pilot unit for the production of UO 3 installed at IPEN-CNEN/SP is described. Its capacity is of 20 kg U/h in a continuous process. The main components of this pilot unit are: a system for the concentration of nuclearly pure uranyl nitrate (≅ 100 g U/L), a system for the denitration of the concentrated uranyl nitrate, an absorption system for NO 2 produced during the denitration reaction and, finally, a system for the dissolution of UO 3 that does not meet the specifications. The operational troubles found during the initial runs are presented. The results of the physical and chemical analysis of the UO 3 produced are discussed and a comparison is made for the UO 3 obtained by both fluidized-bed and wet processes. (author) [pt

Progress on flue gas desulfurization and denitration with electron beam irradiation in CAEP

International Nuclear Information System (INIS)

Ren Min; Wang Baojian; Yang Ruizhuang; Huang Wenfeng; He Xiaohai; Mao Benjiang

2005-01-01

The first pilot plant with electron beam irradiation for desulfurization and denitration of flue gas in China and the experimental results based on the pilot plant are briefly introduced in this paper. The FGD (flue gas desulfurization) demonstration installation designed by CAEP (China Academy of Engineering Physics) in Beijing Jingfeng Thermal Powe Co., Ltd. is recommended. (author)

Decomposition of ammonium nitrate in homogeneous and catalytic denitration

International Nuclear Information System (INIS)

Anan'ev, A. V.; Tananaev, I. G.; Shilov, V. P.

2005-01-01

Ammonium nitrate is one of potentially explosive by-products of spent fuel reprocessing. Decomposition of ammonium nitrate in the HNO 3 -HCOOH system was studied in the presence or absence of Pt/SiO 2 catalyst. It was found that decomposition of ammonium nitrate is due to homogeneous noncatalytic oxidation of ammonium ion with nitrous acid generated in the HNO 3 -HCOOH system during denitration. The platinum catalyst initiates the reaction of HNO 3 with HCOOH to form HNO 2 . The regular trends were revealed and the optimal conditions of decomposition of ammonium nitrate in nitric acid solutions were found [ru

Screw calciner mechanical direct denitration process for plutonium nitrate to oxide conversion

International Nuclear Information System (INIS)

Souply, K.R.; Sperry, W.E.

1977-01-01

This report describes a screw calciner direct-denitration process for converting plutonium nitrate to plutonium oxide. The information should be used when making comparisons of alternative plutonium nitrate-to-oxide conversion processes or as a basis for further detailed studies. The report contains process flow sheets with a material balance; a process description; and a discussion of the process including history, advantages and disadvantages, and additional research required

The denitration of simulated fast reactor highly active liquor waste

International Nuclear Information System (INIS)

Saum, C.J.; Ford, L.H.; Blatts, N.

1981-01-01

A short series of tests have been made with simulated HAL containing representative concentrations of palladium and phosphate ion. The information obtained has been confirmed in a small scale continuous denitration plant. These cases of four stirred pot reactors arranged in cascade. One possible advantage of this plant would be the low mean acidity in the first stage compared to the feed material which would limit to some extent the violence of the reaction. This would lead to a lower rate of gas evolution and may permit operation even with liquors where foaming is a problem. (DG)

Effect of Na poisoning catalyst (V2O5-WO3/TiO2) on denitration process and SO3 formation

Science.gov (United States)

Xiao, Haiping; Chen, Yu; Qi, Cong; Ru, Yu

2018-03-01

This paper aims to study the effect of alkali metal sodium (Na) poisoning on the performance of the Selective Catalytic Reduction (SCR) catalyst. The result showed that Na2SO4 poisoning leads to a reduced denitration rate of the SCR catalyst and an increase in the SO3 generation rate. Na2O poisoning leads to a significant reduction in the denitration rate of the SCR catalyst and marginally improves the formation of SO3. The maximum of the SO3 generation rate for a Na2SO4-poisoned catalyst reached 1.35%, whereas it was only 0.85% for the SCR catalyst. When the SO2 was contained in flue gas, the denitration rate for the Na2O-poisoned catalyst clearly increased by more than 28%. However, the effect of SO2 on the Na2SO4-poisoned catalyst was very slight. The denitration rate of the SCR catalyst decreased with an increase in the Na content. The BET and XRD results showed that Na poisoning of the catalyst decreased the number of acid sites, the reducibility of the catalyst, the surface area, and pore volume. The H2-TPR and NH3-TPD results show that Na decreases the number of acid sites and the reducibility of the catalyst. The FT-IR and XPS results showed that Na2O poisoning led to the decrease of V5+dbnd O bonds and the consumptions of oxygen atoms. Na2SO4 poisoning can improve surface adsorbed oxygen, which was beneficial for the SO2-SO3 conversion reaction.

Restart Plan for the Prototype Vertical Denitration Calciner [SD Coversheet has Incorrect Document Number

Energy Technology Data Exchange (ETDEWEB)

SUTTER, C.S.

1999-07-26

Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing.

Application of microwaves in the denitration of nitric solutions of uranium and/or plutonium

International Nuclear Information System (INIS)

Quesada, C.A.; Adelfang, P.

1990-01-01

A method for the conversion of nitric solutions of uranium and/or plutonium that would be an alternative more economic and operatively simpler than the conventional processes is the direct denitration by means of microwaves and vacuum application. This conversion method has the following technical advantages: a) the process is simple, which allows a stable operation; b) neither the addition of chemical reagents nor the dilution of the starting solution are required, thereby the volume of residual liquids is small as compared with other processes; c) one fraction of the evaporation residues is nitric acid which can be reused. The development (on laboratory scale) of this conversion process was initiated. In this first stage, a description of the employed equipment is presented. An example of one of the evaporation and denitration batches and obtained products are fully described. The operative experience leads to deduce that the equipment is satisfactory, due to the following characteristics: 1) it permits an easy manipulation within the glove boxes; 2) the projections, coming out from the reactor, are retained completely; 3) the microwaves oven and the vacuum pump are effectively protected from the corrosive vapors. It is concluded that the employed experimental device is adequate to obtain the necessary materials for the reduction, pressing and sinterability studies. This equipment is adopted for the integral development of sintered pellets fabrication process. (Author) [es

Pilot plant experiments for the denitration and mercury separation from the HEWC solutions

International Nuclear Information System (INIS)

Humblet, L.; Hendrickx, J.P.; Geel, J. van.

1984-06-01

A process development for the elimination of mercury and nitrates from the HEWC (high-enriched waste concentrates) solutions has been achieved. This process is based on the reduction of mercury to metal with formaldehyde. The pilot plant which has enabled to test the developed process is described as well as the experiments. The residual mercury concentration is of 25 mg/1 but the mechanism of the reduction is not yet known. During the denitration the nitrous vapors production calls for an oversized absorption column. The control instruments and the analytical methods are also described. (AF)β

Steam Reforming Technology for Denitration and Immobilization of DOE Tank Wastes

International Nuclear Information System (INIS)

Mason, J. B.; McKibbin, J.; Ryan, K.; Schmoker, D.

2003-01-01

THOR Treatment Technologies, LLC (THOR) is a joint venture formed in June 2002 by Studsvik, Inc. (Studsvik) and Westinghouse Government Environmental Services Company LLC to further develop, market, and deploy Studsvik's patented THORSM non-incineration, steam reforming waste treatment technology. This paper provides an overview of the THORSM steam reforming process as applied to the denitration and conversion of Department of Energy (DOE) tank wastes to an immobilized mineral form. Using the THORSM steam reforming technology to treat nitrate containing tank wastes could significantly benefit the DOE by reducing capital and life-cycle costs, reducing processing and programmatic risks, and positioning the DOE to meet or exceed its stakeholder commitments for tank closure. Specifically, use of the THORSM technology can facilitate processing of up to 75% of tank wastes without the use of vitrification, yielding substantial life-cycle cost savings

Denitration of medium level liquid radioactive wastes by catalytic destruction of nitrogen oxides

International Nuclear Information System (INIS)

Donato, A.; Ricci, G.

1984-01-01

The catalytic abatement by means of NH 3 of the NOsub(x) produced in the radwaste conditioning has been studied. With reference to the gas produced in a bituminization plant, the thermodynamics and the chemistry of the NOsub(x) catalytic reduction to nitrogen and H 2 O have been evaluated. The following operational parameters have been experimentally studied: the catalyst bed temperature; the gas residence time; the vapour concentration; the NOsub(x) concentration; the gas velocity; the catalyst grain size distribution; the catalyst time-life. Abatement yields of the order of 99,5% have been obtained following experimental conditions must be selected. In the case of a bituminization plant, a NOsub(x) catalytic reactor, if installed between the evaporator denitrator and the condenser, could reduce to less than 1/100 the volume of the NaNO 3 secondary wastes produced by the gas scrubbing

Technical evaluation of the direct denitration process to obtain ceramic-grade UO2 powders using microwaves

International Nuclear Information System (INIS)

Lorenzo, Viviana J.; Marchi, Daniel E.; Menghini, Jorge E.

1999-01-01

The direct denitration process to obtain ceramic-grade UO 2 powders using microwaves has been studied and developed at laboratory scale. Conditions were given to obtain powders apt for fuel pellets fabrication within the required specifications, where mechanical treatments before pressing are not necessary. This work describes the equipment used in the process, evaluates the necessary supply and waste generation and describes the characteristics of the product obtained, as well as the conditions for its fabrication. Results show that this method allows to reduce the volume of liquid wastes generated due to their partial re-utilization, simplifying their final disposal treatment, which, in addition to their operational advantages, make this method attractive from the economical point of view. (author)

Study on denitration technology of coal char reduction method

Directory of Open Access Journals (Sweden)

Wenjie FU

2016-06-01

Full Text Available In order to more effectively control NO emissions in coal-fired flue gas, the denitration reaction is carried out with simulated industrial boiler flue gas in a fixed bed reactor. The influence of char types, reaction conditions, the composition of flue gas and other factors on the conversion rate of NO are discussed. The result shows that the industrial semi-coke is the most suitable experimental coal in the three coals studied, and the industrial semi-coke particle size of 0.6 ~ 10 mm is relatively suitable; The conversion rate of NO increases gradually with the increase of temperature, and when the reaction temperature is 700 ℃ and the space velocity is 10 000 h^-1, the conversion rate of NO can reach 99%; the conversion rate of NO decreases gradually as airspeed increases, but the airspeed change has no effect on the conversion rate of NO at 700 ℃; under anaerobic conditions,the change of NO concentration has no effect on the conversion rate of NO; at the same temperature, NO conversion rate is higher at the presence of oxygen compared with that at anaerobic situation, and the conversion rate of NO is the highest when O2 concentration is 4%; under aerobic conditions, the concentration change of SO2 and CO2 has no effect on the conversion rate of NO.

Metal-support interaction: The key factor governing activity of Pd/SnO2 catalyst for denitration of ground water

Directory of Open Access Journals (Sweden)

Bošković Goran C.

2008-01-01

Full Text Available Two mesoporous nanocristalline Pd/SnO2 catalysts were prepared by modified solgel technique differing in the pH conditions (pH = 2 and 9.5 of the synthesis of their supports. Samples achieved different activity and selectivity in water denitration reaction using hydrogen. XPS results of reduced samples indicate a strong interaction between the Pd and the Sn possibly as a result of electron shift from Sn to Pd. The solid solution of Pd2+ and SnO2 is formed by taking O from the surface of the support. In such a way some SnO2-X species may stay onto the surface and be responsible for its pronounced activity.

Test Plan for Radioactive Testing of a Vertical Direct Denitration Calciner

International Nuclear Information System (INIS)

COMPTON, J.A.

1999-01-01

The prototype Vertical Denitration Calciner (VDC) is installed in glovebox 188 in the Plutonium Process Support Laboratory (PPSL). Safety analysis contained in WHC-SD-CP-SAR-021 (FSAR) Rev. 0-L and Addendum to WHC-SD-CP-SAR-021, ''Laboratory Prototype Calciner'' establishes the prototype VDC needs to be shut down if a seismic event of greater than 0.07 g occurs. Shut down is to be automatic upon detection of the seismic event. This requires tie-in of various valves and power for the prototype VDC into the existing Seismic Shutdown System for the Ventilation Supply Fans described in FSAR 5.4.1.2.4. The proposed changes covered by this USQ evaluation include: (1) the physical tie-in modifications, including drawings and Engineering Change Notice (ECN), (2) the work package for accomplishing the modifications, (3) the changes to the System Description Documents, (4) the changes to the Safety Equipment List necessitated by the modifications, and (5) the changes to the failure modes and effects analysis. WHC-SDCP-OSR-010, Plutonium Finishing Plant Operational Safety Requirements Limiting Condition for Operation (LCO) 3.2.3 has been revised to include the requirement for the existing seismic shutdown system to also shut down the laboratory calciner in the event of detection of a greater than 0.07 g seismic event

Test plan for radioactive testing of a vertical direct denitration calciner

Energy Technology Data Exchange (ETDEWEB)

COMPTON, J.A.

1999-08-31

The prototype Vertical Denitration Calciner (VDC) is installed in glovebox 188 in the Plutonium Process Support Laboratory (PPSL). Safety analysis contained in WHC-SD-CP-SAR-021 (FSAR) Rev. 0-L and Addendum to WHC-SD-CP-SAR-021, ''Laboratory Prototype Calciner'' establishes the prototype VDC needs to be shut down if a seismic event of greater than 0.07 g occurs. Shut down is to be automatic upon detection of the seismic event. This requires tie-in of various valves and power for the prototype VDC into the existing Seismic Shutdown System for the Ventilation Supply Fans described in FSAR 5.4.1.2.4. The proposed changes covered by this USQ evaluation include: (1) the physical tie-in modifications, including drawings and Engineering Change Notice (ECN), (2) the work package for accomplishing the modifications, (3) the changes to the System Description Documents, (4) the changes to the Safety Equipment List necessitated by the modifications, and (5) the changes to the failure modes and effects analysis. WHC-SDCP-OSR-010, Plutonium Finishing Plant Operational Safety Requirements Limiting Condition for Operation (LCO) 3.2.3 has been revised to include the requirement for the existing seismic shutdown system to also shut down the laboratory calciner in the event of detection of a greater than 0.07 g seismic event.

Test Plan for Radioactive Testing of a Vertical Direct Denitration Calciner

Energy Technology Data Exchange (ETDEWEB)

COMPTON, J.A.

1999-08-31

The prototype Vertical Denitration Calciner (VDC) is installed in glovebox 188 in the Plutonium Process Support Laboratory (PPSL). Safety analysis contained in WHC-SD-CP-SAR-021 (FSAR) Rev. 0-L and Addendum to WHC-SD-CP-SAR-021, ''Laboratory Prototype Calciner'' establishes the prototype VDC needs to be shut down if a seismic event of greater than 0.07 g occurs. Shut down is to be automatic upon detection of the seismic event. This requires tie-in of various valves and power for the prototype VDC into the existing Seismic Shutdown System for the Ventilation Supply Fans described in FSAR 5.4.1.2.4. The proposed changes covered by this USQ evaluation include: (1) the physical tie-in modifications, including drawings and Engineering Change Notice (ECN), (2) the work package for accomplishing the modifications, (3) the changes to the System Description Documents, (4) the changes to the Safety Equipment List necessitated by the modifications, and (5) the changes to the failure modes and effects analysis. WHC-SDCP-OSR-010, Plutonium Finishing Plant Operational Safety Requirements Limiting Condition for Operation (LCO) 3.2.3 has been revised to include the requirement for the existing seismic shutdown system to also shut down the laboratory calciner in the event of detection of a greater than 0.07 g seismic event.

Test Plan for Radioactive Testing of a Vertical Direct Denitration Calciner

Energy Technology Data Exchange (ETDEWEB)

COMPTON, J.A.

1999-08-13

The prototype Vertical Denitration Calciner (VDC) is installed in glovebox 188 in the Plutonium Process Support Laboratory (PPSL). Safety analysis contained in WHC-SD-CP-SAR-021 (FSAR) Rev. 0-L and Addendum to WHC-SD-CP-SAR-021, ''Laboratory Prototype Calciner'' establishes the prototype VDC needs to be shut down if a seismic event of greater than 0.07 g occurs. Shut down is to be automatic upon detection of the seismic event. This requires tie-in of various valves and power for the prototype VDC into the existing Seismic Shutdown System for the Ventilation Supply Fans described in FSAR 5.4.1.2.4. The proposed changes covered by this USQ evaluation include: (1) the physical tie-in modifications, including drawings and Engineering Change Notice (ECN), (2) the work package for accomplishing the modifications, (3) the changes to the System Description Documents, (4) the changes to the Safety Equipment List necessitated by the modifications, and (5) the changes to the failure modes and effects analysis. WHC-SDCP-OSR-010, Plutonium Finishing Plant Operational Safety Requirements Limiting Condition for Operation (LCO) 3.2.3 has been revised to include the requirement for the existing seismic shutdown system to also shut down the laboratory calciner in the event of detection of a greater than 0.07 g seismic event.

Pilot-scale demonstration of the modified direct denitration process to prepare uranium oxide for fuel fabrication evaluation

International Nuclear Information System (INIS)

Kitts, F.G.

1994-04-01

The Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has the objective of developing a cost-competitive enrichment process that will ultimately replace the gaseous diffusion process used in the United States. Current nuclear fuel fabricators are set up to process only the UF 6 product from gaseous diffusion enrichment. Enriched uranium-iron alloy from the U-AVLIS separator system must be chemically converted into an oxide form acceptable to these fabricators to make fuel pellets that meet American Society for Testing and Materials (ASTM) and utility company specifications. A critical step in this conversion is the modified direct denitration (MDD) that has been selected and presented in the AVLIS Conceptual Design for converting purified uranyl nitrate to UO 3 to be shipped to fabricators for making UO 2 pellets for power reactor fuel. This report describes the MDD process, the equipment used, and the experimental work done to demonstrate the conversion of AVLIS product to ceramic-grade UO 3 suitable for making reactor-grade fuel pellets

Corrosion evaluation of uranyl nitrate solution evaporator and denitrator in Tokai reprocessing plant

International Nuclear Information System (INIS)

Yamanaka, Atsushi; Hashimoto, Kowa; Uchida, Toyomi; Shirato, Yoji; Isozaki, Toshihiko; Nakamura, Yoshinobu

2011-01-01

The Tokai reprocessing plant (TRP) adopted the PUREX method in 1977 and has reprocessed spent nuclear fuel of 1140 tHM (tons of heavy metals) since then. The reprocessing equipment suffers from various corrosion phenomena because of high nitric acidity, solution ion concentrations, such as uranium, plutonium, and fission products, and temperature. Therefore, considering corrosion performance in such a severe environment, stainless steels, titanium steel, and so forth were employed as corrosion resistant materials. The severity of the corrosive environment depends on the nitric acid concentration and the temperature of the solution, and uranium in the solution reportedly does not significantly affect the corrosion of stainless steels and controls the corrosion rates of titanium steel. The TRP equipment that handles uranyl nitrate solution operates at a low nitric acid concentration and has not experienced corrosion problems until now. However, there is a report that corrosion rates of some stainless steels increase in proportion to rising uranium concentrations. The equipment that handles the uranyl nitrate solution in the TRP includes the evaporators, which concentrate uranyl nitrate to a maximum concentration of about 1000 gU/L (grams of uranium per liter), and the denitrator, where uranyl nitrate is converted to UO 3 powder at about 320degC. These equipments are therefore required to grasp the degree of the progress of corrosion to handle high-temperature and high-concentration uranyl nitrate. The evaluation of this equipment on the basis of thickness measurement confirmed only minor corrosion and indicated that the equipment would be fully adequate for future operation. (author)

Safety in connection with the request for approval of the installation alteration in the fuel reprocessing facilities of Power Reactor and Nuclear Fuel Development Corporation (report)

International Nuclear Information System (INIS)

1982-01-01

A report to the Prime Minister by the Nuclear Safety Commission was presented concerning the safety in the installation alteration of the fuel reprocessing facilities, as PNC had requested its approval to the Prime Minister. The safety was confirmed. The items of examination on the safety made by the committee on Examination of Nuclear Fuel Safety of NSC were the aseismic design of liquid waste storage, uranium denitration facility, intermediate gate and radioactive solid waste storage; the criticality safety design of the denitration facility; the radiation shielding design of the liquid waste storage, denitration facility and solid waste storage; the function of radioactive material containment of the liquid waste storage and denitration facility; the radiation control in the liquid waste storage, denitration facility and solid waste storage; the waste management in the liquid waste storage and denitration facility; fire and explosion prevention in the liquid waste storage; exposure dose from the liquid waste storage and denitration facility. (Mori, K.)

Effect of impregnation protocol in the metallic sites of Pt–Ag/activated carbon catalysts for water denitration

Energy Technology Data Exchange (ETDEWEB)

Aristizábal, A. [Departament d’Enginyeria Química, Universitat Rovira i Virgili, Campus Sescelades, Av. Països Catalans 26, 43007 Tarragona (Spain); Contreras, S., E-mail: sandra.contreras@urv.cat [Departament d’Enginyeria Química, Universitat Rovira i Virgili, Campus Sescelades, Av. Països Catalans 26, 43007 Tarragona (Spain); Divins, N.J.; Llorca, J. [Institut de Tècniques Energètiques i Centre de Recerca en Nanoenginyeria, Universitat Politècnica de Catalunya, Diagonal 647, 08028 Barcelona (Spain); Medina, F. [Departament d’Enginyeria Química, Universitat Rovira i Virgili, Campus Sescelades, Av. Països Catalans 26, 43007 Tarragona (Spain)

2014-04-01

Highlights: • Mean particle size is tuned by the Pt precursor. H{sub 2}PtCl{sub 6} leads to smaller size. • H{sub 2}PtCl{sub 6} leads to higher extent of Pt–Ag particles with a composition richer in silver. • Pt(NH{sub 3}){sub 4}(NO{sub 3}){sub 2} leads to Ag{sup 0} particles and some Pt–Ag ensembles in less extent. • Nitrate and nitrite rates are linearly related to mean metal particle size. • Physical mixture of catalysts enhances N{sub 2} selectivities. - Abstract: The influence of the Pt precursor and the impregnation protocol in the catalytic behavior of 3%Pt–1.5%Ag supported on activated carbon for water denitration in a continuous reactor was studied. Pt(NH{sub 3}){sub 4}(NO{sub 3}){sub 2} and H{sub 2}PtCl{sub 6} were selected as Pt precursors. Five protocols were investigated: sequential impregnations (both sequences), co-impregnation, physical mixture of monometallic catalysts, and physical mixture of a bimetallic catalyst with a Pt monometallic catalyst. The samples were characterized by XRD, XPS, TPR, HRTEM and physisorption. It was found that the catalytic activity strongly depends on the synthesis protocol and the Pt precursor, which modify the particle size. Higher nitrate rates are achieved using H{sub 2}PtCl{sub 6} than Pt(NH{sub 3}){sub 4}(NO{sub 3}){sub 2}; this is mainly related to the smaller metal particle size of the former, evidenced by HRTEM. Nitrate consumption rate is directly related with the mean particle size. The physical mixture of monometallic catalysts resulted in the highest nitrogen rate.

Denitration and chemical precipitation of medium level liquid wastes and conditioning of high level wastes from low level liquid wastes by a roll dryer and subsequent vitrification

International Nuclear Information System (INIS)

Halaszovich, S.; Dix, S.; Harms, R.

1987-01-01

Medium level liquid waste (MAW) from the reprocessing need after being fixed in cement an additional shielding to meet required radiation limits for handling and transportation. Normally this shielding consists of concrete and its weight and volume is several times higher than that of the waste product itself. By means of caesium separation using nickel-potassium-hexacyanoferrate and after few years of interim storage waiting for the decay of Ruthenium and Antimony the activities will be reduced below permissible values. (13 MBq/l in waste solution for Cs, 28 MBq/l for Sb and 34 MBq/l for Ru). Below these limits there is no need for additional shielding after cementation in a 400 l drum. Experimental results show, that Caesium can be precipitated and separated effectively not only in laboratory but also in a larger scale under hot cell conditions. The process investigated in this work has been developed from the FIPS process for vitrification of highly radioactive fission product solutions. It consists of: denitration, precipitation, sludge separation, drying and melting

Fiscal 2000 achievement report on the development of energy conservation/environment purification system using cleaning effect of optical irradiation; 2000 nendo hikari clean gijutsu wo mochiita sho energy kankyo joka system no kaihatsu seika hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

The research aims to develop materials and apparatuses for the purification of atmosphere using titanium dioxide that exhibits a powerful oxidizing capability when irradiated with light. A study is conducted to find out an optimum composition for a photocatalytic fluorocarbon polymer sheet suitable for use in a denitration apparatus. A high density fluorocarbon polymer sheet composed of TiO{sub 2} modified with 0.3% of Pd/absorbent zeolite/fluorocarbon polymer PTFE (polytetrafluoroethylene) =48-63/24-36/10-20 is found to show high denitration efficiency, and this achieves the denitration efficiency goal. As for sheet thickness, 0.75mm is found to be enough. The sheet experiences some hardening in an accelerated exposure test, but does not change much in a surface gloss test or a chalking test. Although a slight reduction is observed in denitration efficiency, yet the durability goal is achieved. In the effort to develop an energy conservation type air cleaning apparatus, field tests and experiments are repeatedly conducted. As for photodenitration in the cleaning apparatus, the number of photodenitration stages and the magnitude of equimolar adsorption area necessary for achieving an 80% denitration rate is calculated from the relations of the NOx concentration profile and the denitration rate in the equimolar adsorption module to (gas flow rate/module surface), and the result shows that the initially intended goal is achieved. (NEDO)

Development of FR fuel cycle in Japan (3) - Current state on unified technology of De-nitration conversion and granulation for the simplified pellet fuel production based on microwave heating

International Nuclear Information System (INIS)

Suzuki, M.; Ishii, K.; Yamamoto, T.; Kato, Y.; Kurita, T.; Yoshimoto, K.; Kihara, Y.; Namekawa, T.; Fujii, K. I.

2008-01-01

The major experimental results and current state on unified and simplified de-nitration conversion processing are explained. This technology was enabled by the nitrate solutions mixing method which is our original idea proposed in FS phase II project and transferred to present FaCT project where the mass production is the supreme subject. The major results obtained up to present, which are based on the two originalities, are as follows; (1) The mixing rate of the plutonium nitrate solution (PNS) into the uranium nitrate solution (UNS) was satisfactorily adjusted by the feeding of 0.5 litter step, (2) Excellent size uniformity around 10 μm was obtained in the MOX powders including 30 wt% of PUO 2 products, (3) In the microwave heating, a mild and stable boiling occurred at whole space in the evaporation dish and the size of bubbles gradually became smaller closing to the finishing stage, (4) The Can coefficient of raw MOX powders containing 30 wt% of PUO 2 was ranged 20-40 just before granulation, being rather difficult to pack them in a die, but after the granulation, the packing rate reached 100 %. (authors)

<報文>高レベル放射性廃液の脱硝ならびにCs, Srの分離におけるフッ素イオンの影響

OpenAIRE

三村, 均; 稲野, 昌利; 秋葉, 健一; 菅野, 卓治; Hitoshi, MIMURA; Masatoshi, INANO; Kenichi, AKIBA; Takuji, KANNO; 東北大学選鉱製錬研究所; 東北大学工学部:(現)動力炉・核燃料開発事業団; 東北大学選鉱製錬研究所; 東北大学選鉱製錬研究所

1984-01-01

The effects of HF have been studied on the denitration process of high-level radioactive liquid waste and on the separation of Cs and Sr with zeolite from the denitrated waste solutions. In the denitration process, the addition of HF up to 0.1M gave no effect on the residual nitrate concentration, pH and the residual fluoride concentration, while further addition of HF more than 0.1M resulted in the decrease of the residual nitrate concentration and pH. Zeolite A was stable in HF solution les...

Evaluation report on the development of energy conservation/environment purification system using cleaning effect of optical irradiation; Hikari clean gijutsu wo mochiita sho energy kankyo joka system no kaihatsu hyoka hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

The results achieved in fiscal 1992-1995 under the above-named project are stated. In the development of photocatalytic materials, a photocatalytic fluorocarbon polymer sheet suitable for use in a denitration apparatus is developed. A high density fluorocarbon polymer sheet composed of TiO{sub 2} modified with 0.3% of Pd/absorbent zeolite/fluorocarbon polymer PTFE (polytetrafluoroethylene) =48-63/24-36/10-20 is fabricated, which achieves a level higher than the denitration goal of 70%. Although the sheet in a 500-hour accelerated exposure test undergoes a hardening phenomenon in which elasticity decreases and tensile strength increases, yet degradation is hardly detected. Although a slight reduction is detected in denitration efficiency, yet it does not affect its practical application, and thus the durability goal is achieved. In the development of an energy conservation type air cleaning apparatus usable in underground parking areas or motorway tunnels, an apparatus capable of treating air at a rate of 2,000m{sup 3}/hour is fabricated, and this achieves a denitration level of not less than 80% in a field test (in the absence of rainfall). For denitration in the presence of rainfall, the apparatus is combined with an equimolar adsorption system, and a system capable of 80% denitration is proposed on the basis of data actually measured for each of the two. A conceptual design for a service model comprising a photodenitration and equimolar adsorption systems is evaluated, and it is found that it occupies less space than the existing models. (NEDO)

Basic research on flue gas smoke treatment by electron beam

International Nuclear Information System (INIS)

Namba, Hideki

1995-01-01

Recently, accompanying the increase of the use of fossil fuel, the environment destruction due to the sulfur oxides and nitrogen oxides contained in combustion smoke has become a serious problem. The development of flue gas smoke treatment technology by using electron beam was started in Japan, and attention has been paid worldwide as the promising dry type simultaneous desulfurizing and denitrating process. In this process, by adding ammonia to smoke, and irradiating electron beam on it, ammonium nitrate and ammonium sulfate are formed. As to the reaction mechanism of denitration and desulfurization, radical formation, radical reaction, denitration mechanism, desulfurization mechanism, the particle size distribution of the formed aerosol, the amounts of denitration and desulfurization by electron beam smoke treatment process, the improvement of the denitration efficiency by multi-stage irradiation method and the improvement of the desulfurization rate by low temperature irradiation, and the basic test toward the pilot test are explained. The basic research for putting this system to practical use was carried out jointly by Japan Atomic Energy Research Institute, Chubu Electric Power Co., Inc., and Ebara Seisakusho for standard coal burning smoke in Japan. The verifying test at the pilot plant in Shinnagoya Thermal Power Station was carried out, and it was verified that this process can be used practically for treating coal-burning smoke. (K.I.)

Nuclear fuel conversion and fabrication chemistry

International Nuclear Information System (INIS)

Lerch, R.E.; Norman, R.E.

1984-01-01

Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO 2 material or a mixed PuO 2 -UO 2 fuel material is generated. Several methods are available for preparation of the pure PuO 2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO 2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO 2 . The UO 2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO 2 -PuO 2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO 3 -PuO 2 starting materials is accomplished by calcination-reduction to UO 2 -PuO 2 followed by pellet fabrication. (orig.)

Explosion and fire in the uranium trioxide production facilities at the Savannah River Plant on February 12, 1975. A works technical report

International Nuclear Information System (INIS)

McKibben, J.M.

1976-10-01

On February 12, 1975, an explosion and fire occurred in the denitrator room of the separations A-Line facility, resulting in minor injury to two employees and about $230,000 damage to the building and equipment. The facility, which converts uranyl nitrate solution to UO 3 , had operated for 20 years without major accident. Accidental transfer of tributyl phosphate--uranyl nitrate adduct into a denitrator was followed by rapid decomposition of the organic material. Combustible gases were released into the denitrator room and ignited. No contamination was spread outside the facility. A-Line returned to operation August 11, 1975, after equipment and procedures were modified to lower the probability of similar incidents to a very low level. 18 figures, 9 tables

Effect of phosphate ion on filtration characteristics of solids generated in simulated high level liquid waste

International Nuclear Information System (INIS)

Kondo, Y.

1998-01-01

The effect of phosphate ion on the filtration characteristics of solids generated in a high level liquid waste was experimentally examined. Addition of phosphate ion into the simulated HLLW induced the formation of phosphate such as zirconium phosphate and phosphomolybdic acid. The filtration rate of zirconium phosphate abruptly dropped in the midst of filtration because of a gel-cake formation on the filter surface. The denitration of the simulated HLLW contained zirconium phosphate improved the filterability of this gelatinous solid. The filtration rates of denitrated HLLW decreased with increase of the phosphate ion concentration, since the solids formed by denitration had irregular particle size and configuration in the simulated HLLW with phosphate ion. To increase the filtration rate of denitrated HLLW, a solid suspension filtration tester was designed. The solid-suspension accelerated the filtration rate only in the simulated HLLW with more than 1500 ppm phosphate ion concentration. Under this condition, the simple agitation can easily suspend the constituent solids of filter cake in the solution and a much higher filtration rate can be obtained because the filter cake is continuously swept from the filter surface by rotation of propellers. (authors)

Recovery of americium-241 from aged plutonium metal

International Nuclear Information System (INIS)

Gray, L.W.; Burney, G.A.; Reilly, T.A.; Wilson, T.W.; McKibben, J.M.

1980-01-01

After separation and purification, both actinides were precipitated as oxalates and calcined. A large-scale process was developed using dissolution, separation, purification, precipitation, and calcination. Efforts were made to control corrosion, to avoid product contamination, to keep the volume of process and waste solutions manageable, and to denitrate solutions with formic acid. The Multipurpose Processing Facility (MPPF), designed for recovery of transplutonium isotopes, was used for the first time for the precipitation and calcination of americium. Also, for the first time,, large-scale formic acid denitration was performed in a canyon vessel at SRP

Key enzymes enabling the growth of Arthrobacter sp. strain JBH1 with nitroglycerin as the sole source of carbon and nitrogen.

Science.gov (United States)

Husserl, Johana; Hughes, Joseph B; Spain, Jim C

2012-05-01

Flavoprotein reductases that catalyze the transformation of nitroglycerin (NG) to dinitro- or mononitroglycerols enable bacteria containing such enzymes to use NG as the nitrogen source. The inability to use the resulting mononitroglycerols limits most strains to incomplete denitration of NG. Recently, Arthrobacter strain JBH1 was isolated for the ability to grow on NG as the sole source of carbon and nitrogen, but the enzymes and mechanisms involved were not established. Here, the enzymes that enable the Arthrobacter strain to incorporate NG into a productive pathway were identified. Enzyme assays indicated that the transformation of nitroglycerin to mononitroglycerol is NADPH dependent and that the subsequent transformation of mononitroglycerol is ATP dependent. Cloning and heterologous expression revealed that a flavoprotein catalyzes selective denitration of NG to 1-mononitroglycerol (1-MNG) and that 1-MNG is transformed to 1-nitro-3-phosphoglycerol by a glycerol kinase homolog. Phosphorylation of the nitroester intermediate enables the subsequent denitration of 1-MNG in a productive pathway that supports the growth of the isolate and mineralization of NG.

Result of design and test operation of a coal boiler at Hyogo Refinery

Energy Technology Data Exchange (ETDEWEB)

Miura, Yasuhiko; Sato, Noriyuki

1987-05-01

This boiler is the first coal boiler for the oil refineries in Japan (Installed in Oct., 1986 at Idemitsu Petrochemical Co.) Causes for using coal as a fuel are a conversion to less expensive fuel and offering a technical service to the users of coal through the combustion of coal and learning of a handling technique. The type of boiler is Babcock single barrel radiant type and has 150 t/d capacity with single fuel combustion of coal. Auxiliary equipments are a pulverizer, a transportation and storage unit, a denitration unit, a dust collector, a desulfurization unit, and an ash disposal unit. Main considerations in the design are measures for the security of finely pulverized coal, clogging prevention for coal and ash. A test operation revealed 7 % of combustible loss and 160 - 250 ppm of NOx content at a charge inlet of denitration unit. Actual operation exhibited no clogging at the denitration unit of troubles due to scaling. Design for raw materials is to blend 4 imported coals (from Australia and Canada, etc) and 3 Japanese ones. (7 figs, 2 tabs)

A demonstration test of 4-group partitioning process with real high-level liquid waste

Energy Technology Data Exchange (ETDEWEB)

Morita, Y.; Yamaguchi, I.; Fujiwara, T.; Koizumi, H.; Tachimori, S. [Japan Atomic Energy Research Institute, Tokai-Mura, Ibaraki-Ken (Japan)

2000-07-01

The demonstration test of 4-Group Partitioning Process with concentrated real high-level liquid waste (HLLW) was carried out in the Partitioning Test Facility installed in a hot cell. More than 99.998% of Am and Cm were extracted from the HLLW with the organic solvent containing 0.5 M DIDPA - 0.1 M TBP, and more than 99.98% of Am and Cm were back-extracted with 4 M nitric acid. Np and Pu were extracted simultaneously, and more than 99.93% of Np and more than 99.98% of Pu were back-extracted with oxalic acid. In the denitration step for the separation of Tc and platinum group metals, more than 90% of Rh and more than 97% of Pd were precipitated. About half of Ru were remained in the de-nitrated solution, but the remaining Ru were quantitatively precipitated by neutralization of the de-nitrated solution to pH 6.7. In the adsorption step, both Sr and Cs were separated effectively. Decontamination factors for Cs and Sr were more than 10{sup 6} and 10{sup 4} respectively in all effluent samples. (authors)

Electron-beam flue-gas treatment system

International Nuclear Information System (INIS)

Aoki, Sinji; Suzuki, Ryoji

1994-01-01

The damage of forests in the world due to acid rain has become serious problems, and the development of high efficiency and economical desulfurization and denitration technologies for combustion exhaust gas has been desired. Japan leads the world in exhaust gas treatment technology. The conventional technologies have been the desulfurization by lime gypsum process and the denitration by ammonia catalytic reduction process. The solution by entirely new concept is the electron beam treatment technology for exhaust gas. This technology is a dry process without drain, and does not require catalyst. The byproduct from this technology was approved as a fertilizer. The electron beam treatment technology is called EBA (electron beam with ammonia). The exhaust gas treatment technology by electron beam process is constituted by the cooling of exhaust gas, ammonia addition, electron beam irradiation and the separation of byproduct. The features of the technology are the simultaneous removal of sulfur and nitrogen oxides, dry process, the facilities are simple and the operation is easy, easy following to load variation and the utilization of byproduct. The reaction mechanism of desulfurization and denitration, the course of development, the electron beam generator, and the verifying test are reported. (K.I.)

A low-temperature process for the denitration of Hanford single-shell tank, nitrate-based waste utilizing the nitrate to ammonia and ceramic (NAC) or nitrate to ammonia and glass (NAG) process: Phase 2 report

International Nuclear Information System (INIS)

Mattus, A.J.; Walker, J.F. Jr.; Youngblood, E.L.; Farr, L.L.; Lee, D.D.; Dillow, T.A.; Tiegs, T.N.

1994-12-01

Continuing benchtop studies using Hanford single-shell tank (SST) simulants and actual Oak Ridge National Laboratory (ORNL) low-level waste (LLW), employing a new denitration process for converting nitrate to ammonia and ceramic (NAC), have conclusively shown that between 85 and 99% of the nitrate can be readily converted to gaseous ammonia. In this process, aluminum powders can be used to convert alkaline, nitrate-based supernate to ammonia and an aluminum oxide-sodium aluminate-based solid. The process may be able to use contaminated aluminum scrap metal from DOE sites to effect the conversion. The final, nitrate-free ceramic product can be pressed and sintered like other ceramics or silica and/or fluxing agents can be added to form a glassy ceramic or a flowable glass product. Based upon the starting volumes of 6.2 and 3.1 M sodium nitrate solution, volume reductions of 50 to 70% were obtained for the waste form produced. Sintered pellets produced from supernate from Melton Valley Storage Tanks (MVSTs) have been leached in accordance with the 16.1 leach test for the radioelements 85 Sr and 137 Cs. Despite lengthy counting times, 85 Sr could not be detected in the leachates. 137 Cs was only slightly above background and corresponded to a leach index of 12.2 to 13.7 after 8 months of leaching. Leach testing of unsintered and sintered reactor product spiked with hazardous metals proved that both sintered and unsintered product passed the Toxicity Characteristic Leaching Procedure (TCLP) test. Design of the equipment and flowsheet for a pilot demonstration-scale system to prove the nitrate destruction portion of the NAC process and product formation is under way

Development of partitioning method: confirmation of behavior of technetium in 4-Group Partitioning Process by a small scale experiment

International Nuclear Information System (INIS)

Morita, Yasuji; Yamaguchi, Isoo; Fujiwara, Takeshi; Kubota, Masumitsu; Mizoguchi, Kenichi

1998-08-01

The separation behavior of Tc in the whole of 4-Group Partitioning Process was examined by a flask-scale experiment using simulated high-level liquid waste containing a macro amount of Tc, in order to confirm the reproducibility of the results obtained in previous studies on the Tc behavior at each step of the process. The 4-Group Partitioning Process consists of pre-treatment step, extraction step with diisodecylphosphoric acid (DIDPA), adsorption step with active carbon or precipitation step by denitration for the separation of Tc and platinum group metals (PGM), and adsorption step with inorganic ion exchangers. The present study deals with the behavior of Tc and other elements at all the above steps and additional step for Tc dissolution from the precipitate formed by the denitration. At the pre-treatment step, the ratio of Tc precipitated was very low (about 0.2%) at both operations of heating-denitration and colloid removal. Tc was not extracted with DIDPA and was contained quantitatively in the raffinate from the extraction step. Batch adsorption with active carbon directly from the raffinate showed that distribution coefficient of Tc was more than 100ml/g, which is high enough for the separation. It also revealed much effect of coexisting Mo on the Tc adsorption. At the precipitation step by denitration, 98.2% of Tc were precipitated. At the Tc dissolution from the precipitate with H 2 O 2 , 84.2% of Tc were selectively dissolved in a single operation. Tc was not adsorbed with inorganic ion exchangers. From these results, composition of Tc product from the partitioning process was estimated. The weight ratio of Tc in the Tc product can be increased to about 50% at least. Main contaminating elements are Cr, Ni, Sr, Ba, Mo and Pd. Process optimization to decrease their contamination should be performed in a next study. (J.P.N.)

MOX fuel development: Experience in Argentina

International Nuclear Information System (INIS)

Marchi, D.E.; Adelfang, P.; Menghini, J.E.

1999-01-01

Since 1973, when a laboratory conceived for the safe manipulation of a few hundred grams of plutonium was built, the CNEA (Argentinean Atomic Energy Commission) has been involved in the small-scale development of MOX fuel technology. The plutonium laboratory consists in a glove box facility (α Facility) featuring the necessary equipment to prepare MOX fuel rods for experimental irradiations and to carry out studies on preparative processes development and chemical and physical characterization. The irradiation of the first prototypes of (U,Pu)O 2 fuels fabricated in Argentina began in 1986. These experiments were carried out in the HFR (High Flux Reactor)- Petten , Holland. The rods were prepared and controlled in the CNEA's a Facility. The post-irradiation examinations (PIE) were performed in the KFK (Kernforschungszentrum Karlsruhe), Germany and the JRC (Joint Research Center), Petten. In the period 1991-1995, the development of new laboratory methods of co-conversion of uranium and plutonium were carried out: reverse strike co-precipitation of ADU-Pu(OH) 4 and direct denitration using microwaves. The reverse strike process produced pellets with a high sintered density, excellent micro-homogeneity and good solubility in nitric acid. Liquid wastes showed a very low content of actinides and the process is easy to operate in a glove box environment. The microwave direct denitration was optimized with uranium alone and the conditions to obtain high density pellets, with a good microstructure, without using a milling step, have been developed. At present, new experiments are being carried out to improve the reverse strike co-precipitation process and direct microwave denitration. A new glove box is being installed at the plutonium laboratory, this glove box has process equipment designed to recover scrap from previous fabrication campaigns, and to co-convert mixed U-Pu solutions by direct microwave denitration. (author)

Preparation and Performance of Modified Red Mud-Based Catalysts for Selective Catalytic Reduction of NOx with NH3

Directory of Open Access Journals (Sweden)

Jingkun Wu

2018-01-01

Full Text Available Bayer red mud was selected, and the NH3-SCR activity was tested in a fixed bed in which the typical flue gas atmosphere was simulated. Combined with XRF, XRD, BET, SEM, TG and NH3-Temperature Programmed Desorption (TPD characterization, the denitration characteristics of Ce-doped red mud catalysts were studied on the basis of alkali-removed red mud. The results showed that typical red mud was a feasible material for denitration catalyst. Acid washing and calcining comprised the best treatment process for raw red mud, which reduced the content of alkaline substances, cleared the catalyst pore and optimized the particle morphology with dispersion. In the temperature range of 300–400 °C, the denitrification efficiency of calcined acid washing of red mud catalyst (ARM was more than 70%. The doping of Ce significantly enhanced NH3 adsorption from weak, medium and strong acid sites, reduced the crystallinity of α-Fe2O3 in ARM, optimized the specific surface area and broadened the active temperature window, which increased the NOx conversion rate by an average of nearly 20% points from 250–350 °C. The denitration efficiency of Ce0.3/ARM at 300 °C was as high as 88%. The optimum conditions for the denitration reaction of the Ce0.3/ARM catalyst were controlled as follows: Gas Hourly Space Velocity (GHSV of 30,000 h−1, O2 volume fraction of 3.5–4% and the NH3/NO molar ratio ([NH3/NO] of 1.0. The presence of SO2 in the feed had an irreversible negative effect on the activity of the Ce0.3/ARM catalyst.

Cause analysis and suggestion of urea consumption in denitrification system of coal-fired power plant

Science.gov (United States)

Zhang, Xueying; Dong, Ruifeng; Guo, Yang; Wang, Fangfang; Yang, Shuo

2018-02-01

In the daily operation of many power plants, the urea consumption of denitration system is much more than normal. Therefore, the process of site testing and laboratory analysis are carried out. Several suggestions are given out. (1) The position of sampling hole on the exit flue of denitrification system should be redesigned. (2) The denitrification optimization and adjustment should be carried out based on the technical specifications for the operation system. (3) The flue gas CEMS system for single point sampling should be transformed into two or three point sampling mode. (4) When the coal - fired unit is shutting down, examine the ammonia injection and nozzle branch, in order to improve the operation reliability of denitration system.

Rapid formation of phase-clean 110 K (Bi-2223) powders derived via freeze-drying process

Science.gov (United States)

Balachandran, U.

1996-06-04

A process for the preparation of amorphous precursor powders for Pb-doped Bi{sub 2}Sr{sub 2} Ca{sub 2}Cu{sub 3}O{sub x} (2223) includes a freeze-drying process incorporating a splat-freezing step. The process generally includes splat freezing a nitrate solution of Bi, Pb, Sr, Ca, and Cu to form flakes of the solution without any phase separation; grinding the frozen flakes to form a powder; freeze-drying the frozen powder; heating the dried powder to form a dry green precursor powders; denitrating the green-powders; heating the denitrated powders to form phase-clean Bi-2223 powders. The grain boundaries of the 2223 grains appear to be clean, leading to good intergrain contact between 2223 grains. 11 figs.

Heterogeneous-catalytic redox reactions in nitrate - formate systems

International Nuclear Information System (INIS)

Ananiev, A.V.; Shilov, V.P.; Tananaev, I.G.; Brossard, Ph.; Broudic, J.Ch.

2000-01-01

It was found that an intensive destruction of various organic and mineral substances - usual components of aqueous waste solutions (oxalic acid, complexones, urea, hydrazine, ammonium nitrate, etc.) takes place under the conditions of catalytic denitration. Kinetics and mechanisms of urea and ammonium nitrate decomposition in the system HNO 3 - HCOOH - Pt/SiO 2 are comprehensively investigated. The behaviour of uranium, neptunium and plutonium under the conditions of catalytic denitration is studied. It is shown, that under the certain conditions the formic acid is an effective reducer of the uranium (VI), neptunium (VI, V) and plutonium (VI, IV) ions. Kinetics of heterogeneous-catalytic red-ox reactions of uranium (VI), neptunium (VI, V) and plutonium (VI, IV) with formic acid are investigated. The mechanisms of the appropriate reactions are evaluated. (authors)

In-Depth Chemistry in Plasma-Exposed M30 and JA2 Gun Propellants

National Research Council Canada - National Science Library

Pesce-Rodriguez, Rose

2001-01-01

...) method to detect low levels of NO in the propellant. It appears that for M30, profiles for radiation-induced denitration of nitrate esters are consistent with Beer's law, and that effects occur as deep...

Computer software configuration management plan for the Honeywell modular automation system

International Nuclear Information System (INIS)

Cunningham, L.T.

1997-01-01

This document provides a Computer Software management plan for a new Honeywell Modular Automation System (MAS) being installed in the Plutonium Finishing Plant (PFP). This type of system will be used to control new thermal stabilization furnaces, a vertical denitrator calciner, and a pyrolysis furnace

Honeywell Modular Automation System Computer Software Documentation

International Nuclear Information System (INIS)

CUNNINGHAM, L.T.

1999-01-01

This document provides a Computer Software Documentation for a new Honeywell Modular Automation System (MAS) being installed in the Plutonium Finishing Plant (PFP). This system will be used to control new thermal stabilization furnaces in HA-211 and vertical denitration calciner in HC-230C-2

40 CFR 421.323 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Science.gov (United States)

2010-07-01

....885 Nickel 3.503 2.357 Fluoride 222.900 126.700 (d) Digestion wet air pollution control. BAT... denitration wet air pollution control. BAT Limitations for the Secondary Uranium Subcategory Pollutant or... representing the degree of effluent reduction attainable by the application of the best available technology...

Development of the process for production of UO2 powder by atomization of uranyl nitrate

International Nuclear Information System (INIS)

Oliveira Lainetti, P.E. de.

1991-01-01

A method of direct conversion of uranyl nitrate hexahydrate (UNH) solution to ceramic grade uranium dioxide powders by thermal denitration in a furnace that combines atomization nozzle and a gas stirred bed is described. The main purpose of this work is to show that this alternative process is technically viable, specially if the recovery of the scrap generated in the nuclear fuel pellet production is required, without further generation of new liquid wastes. The steps for the development of the denitration unit as well as the characteristics of the final powders are described. Powder production experiments have been carried out for different atomization gas pressures and furnace upper section temperatures. Determination of impurity content, specific surface area, particle size and pore size distribution, density, U content, and O/U rate of uranium dioxide powders have been done; phase identification and morphology studies have also been performed. Sintered pellets have been studied by hydrostatic density determination and microstructure analyses. (author)

Feature of flue gas treatment by electron-beam irradiation and details of its development

International Nuclear Information System (INIS)

Tokunaga, Okihiro; Suzuki, Nobutake.

1986-01-01

The method of flue gas treatment with an electron beam, developed jointly by Japan Atomic Energy Research Institute and Ebara Corporation, is promising as a simple, dry process, not using a catalyst, of the desulfurization and denitration. In the procedure, flue gas is irradiated with an electron beam in the presence of ammonia, so that sulfurous acid gas and nitrogen oxide are converted to ammonium sulfate and ammonium nitrate particles, which are then removed. The method is already demonstrated in the flue gas treatment of an iron ore sintering furnace as pilot test. And further, the pilot tests in coal combustion flue gas treatment are proceeding in the United States and West Germany. For the flue gas treatment method using an electron beam, the mechanisms of desulfurization and denitration, the course taken in its development and the present state of development are described, and also the future outlook and problems. (Mori, K.)

Conversion of highly active waste to solids

International Nuclear Information System (INIS)

Scheffler, K.

Borosilicate glasses were selected as matrix material for solidification of highly radioactive wastes. Current laboratory work on the VERA process is described. Goals were met by a five-component glass VG-38 and a glass-ceramic VC-15. The VERA process is described: flowsheet, denitration, calcinator, fusion facility

Development of a process for co-conversion of Pu-U nitrate mixed solutions to mixed oxide powder using microwave heating method

International Nuclear Information System (INIS)

Koizumi, Masumichi; Ohtsuka, Katsuyuki; Ohshima, Hirofumi; Isagawa, Hiroto; Akiyama, Hideo; Todokoro, Akio; Naruki, Kaoru

1983-01-01

For the complete nuclear fuel cycle, the development of a process for the co-conversion of Pu-U nitrate mixed solutions to mixed oxide powder has been performed along the line of non-proliferation policy of nuclear materials. A new co-conversion process using a microwave heating method has been developed and successfully demonstrated with good results using the test unit with a capacity of 2 kg MOX/d. Through the experiments and engineering test operations, several important data have been obtained concerning the feasibility of the test unit, powder characteristics and homogeneity of the product, and impurity pickups during denitration process. The results of these experimental operations show that the co-conversion process using a microwave heating method has many excellent advantages, such as good powder characteristics of the product, good homogeneity of Pu-U oxide, simplicity of the process, minimum liquid waste, no possibility of changing the Pu/U ratio and stable operability of the plant. Since August 1979, plutonium nitrate solution transported from the Tokai Reprocessing Plant has been converted to mixed oxide powder which has the Pu/U ratio = 1. The products have been processed to the ATR ''FUGEN'' reloading fuel. Based on the successful development of the co-conversion process, the microwave heating direct denitration facility with a 10 kg MOX/d capacity has been constructed adjacent to the reprocessing plant. This facility will come into hot operation by the fall of this year. For future development of the microwave heating method, a continuous direct denitration, a vitrification of high active liquid waste and a solidification of the plutonium-contaminated waste are investigated in Power Reactor and Nuclear Fuel Development Corp. (author)

Denitration of High Nitrate Salts Using Reductants

Energy Technology Data Exchange (ETDEWEB)

HD Smith; EO Jones; AJ Schmidt; AH Zacher; MD Brown; MR Elmore; SR Gano

1999-05-03

This report describes work conducted by Pacific Northwest National Laboratory (PNNL), in conjunction with Idaho National Engineering and Environmental Laboratory (INEEL), to remove nitrates in simulated low-activity waste (LAW). The major objective of this work was to provide data for identifying and demonstrating a technically viable and cost-effective approach to condition LAW for immobilization (grout).

Electronic Denitration Savannah River Site Radioactive Waste

International Nuclear Information System (INIS)

Hobbs, D.T.

1995-01-01

Electrochemical destruction of nitrate in radioactive Savannah River Site Waste has been demonstrated in a bench-scale flow cell reactor. Greater than 99% of the nitrate can be destroyed in either an undivided or a divided cell reactor. The rate of destruction and the overall power consumption is dependent on the cell configuration and electrode materials. The fastest rate was observed using an undivided cell equipped with a nickel cathode and nickel anode. The use of platinized titanium anode increased the energy requirement and costs compared to a nickel anode in both the undivided and divided cell configurations

Novel finishing concepts within BNFL'S advanced reprocessing programme

Energy Technology Data Exchange (ETDEWEB)

Hobbs, J.W.; Booth, R.; Lawson, S.; Parkes, P. [BNFL British Nuclear Fuels, Sellafield, Seascale, Cumbria, Research and Technology (United Kingdom)

2000-07-01

New methods of converting actinide nitrate solutions to oxide and fabricating the products from the reprocessing of high burn up and MOX fuels are necessary for the next generation of fuel cycle facilities in order to meet dose criteria and cost reduction targets. Options to support this include never fully separating the plutonium from the uranium and reducing decontamination factors. The product stream from such a reprocessing plant will require a finishing route capable of significant levels of automation and dose minimization. Casting of molten uranyl nitrate into pellets followed by de-nitration under vacuum has been investigated as a novel way of manufacturing oxide fuel pellets. Pellets were successfully cast over a range of temperatures and de-nitrated. Incorporation of uranium oxide into the melt was investigated to increase the density of the cast pellet. It has been demonstrated that it is possible to produce extrudates from powder and molten magnesium nitrate mixtures. Results of a preliminary study of the flow behaviour during extrusion of magnesium nitrate simulant loaded with alumina powder are also discussed. (authors)

Continuous solvent extraction feed adjustment for HTGR fuel reprocessing. Interim development report

International Nuclear Information System (INIS)

Olguin, L.J.

1978-06-01

The two-cycle Acid-Thorex solvent extraction process requires that the feed stream to each thorium cycle be processed to reduce its nitric acid concentration (feed adjustment). This interim development report presents the results of bench-scale and pilot-plant-scale feed adjustment experiments using a continuous mode of operation. An examination of formic acid denitration and fluoride ion volatilization is also included

Cold and semi-hot tests of 4-group partitioning process at NUCEF

Energy Technology Data Exchange (ETDEWEB)

Morita, Yasuji; Yamaguchi, Isoo; Fujiwara, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Mizoguchi, Kenichi [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan); Kubota, Masumitsu [Research Organization for Information Science and Technology, Tokai, Ibaraki (Japan)

2000-06-01

The 4-Group Partitioning Process was tested in the Partitioning Test Facility installed in a hot cell at NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) using simulated high-level liquid waste (HLLW) or the simulated HLLW added with a small amount of real HLLW and Tc. Behavior of each element was examined in a series of the following separation steps: pretreatment for HLLW to prepare the feed solution to extraction step, extraction with diisodecylphosphoric acid for the separation of transuranium elements, precipitation by denitration and adsorption step with active carbon for the separation of Tc and platinum group metals, and adsorption with inorganic ion exchangers for the separation of Sr and Cs. It was confined that each element behaved as expected. More than 99.99% of Am were extracted with DIDPA and 99.92% of Am were back-extracted with 4 M nitric acid. In the precipitation step by denitration, ratio of Tc precipitated was 96.2%. The present tests confined the expected performance of each equipment in the Partitioning Test Facility for the separation of elements and gave useful data for the comparison of element behavior with a result of a partitioning test using real HLLW. (author)

Demonstration of pyropartitioning process by using genuine high-level liquid waste. Reductive-extraction of actinide elements from chlorination product

International Nuclear Information System (INIS)

Uozumi, Koichi; Iizuka, Masatoshi; Kurata, Masaki; Ougier, Michel; Malmbeck, Rikard; Winckel, Stefaan van

2009-01-01

The pyropartitioning process separates the minor actinide elements (MAs) together with uranium and plutonium from the high-level liquid waste generated at the Purex reprocessing of spent LWR fuel and introduces them to metallic fuel cycle. For the demonstration of this technology, a series experiment using 520g of genuine high-level liquid waste was started and the conversion of actinide elements to their chlorides was already demonstrated by denitration and chlorination. In the present study, a reductive extraction experiment in molten salt/liquid cadmium system to recover actinide elements from the chlorination product of the genuine high-level liquid waste was performed. The results of the experiment are as following; 1) By the addition of the cadmium-lithium alloy reductant, almost all of plutonium and MAs in the initial high-level liquid waste were recovered in the cadmium phase. It means no mass loss during denitration, chlorination, and reductive-extraction. 2) The separation factor values of plutonium, MAs, and rare-earth fission product elements versus uranium agreed with the literature values. Therefore, actinide elements will be separated from fission product elements in the actual system. Hence, the pyropartitioning process was successfully demonstrated. (author)

Engineering report (conceptual design) PFP solution stabilization

Energy Technology Data Exchange (ETDEWEB)

Witt, J.B.

1997-07-17

This Engineering Report (Conceptual Design) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage.

Engineering report (conceptual design) PFP solution stabilization

International Nuclear Information System (INIS)

Witt, J.B.

1997-01-01

This Engineering Report (Conceptual Design) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage

PFP solution stabilization

International Nuclear Information System (INIS)

Aftanas, B.L.

1996-01-01

This Functional Design Criteria (FDC) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage

Determination of aristolochic acids by high-performance liquid chromatography with fluorescence detection.

Science.gov (United States)

Wang, Yinan; Chan, Wan

2014-06-25

Nephrotoxic and carcinogenic aristolochic acids (AAs) are naturally occurring nitrophenanthrene carboxylic acids in the herbal genus Aristolochia. The misuse of AA-containing herbs in preparing slimming drugs has caused hundred of cases of kidney disease in Belgium women in a slimming regime in the early 1990s. Accumulating evidence also suggested that prolong dietary intake of AA-contaminated food is one of the major causes to the Balkan endemic nephropathy that was first observed in the late 1950s. Therefore, analytical methods of high sensitivity are extremely important for safeguarding human exposure to AA-containing herbal medicines, herbal remedies, and food composites. In this paper, we describe the development of a new high-performance liquid chromatography coupled fluorescence detector (HPLC-FLD) method for the sensitive determination of AAs. The method makes use of a novel cysteine-induced denitration reaction that "turns on" the fluorescence of AAs for fluorometric detections. Our results showed that the combination of cysteine-induced denitration and HPLC-FLD analysis allows for sensitive quantification of AA-I and AA-II at detection limits of 27.1 and 25.4 ng/g, respectively. The method was validated and has been successfully applied in quantifying AAs in Chinese herbal medicines.

Electron beam treatment technology for exhaust gas for preventing acid rain

International Nuclear Information System (INIS)

Aoki, Shinji

1990-01-01

Recently, accompanying the increase of the use of fossil fuel, the damage due to acid rain such as withering of trees and extinction of fishes and shells has occurred worldwide, and it has become a serious problem. The sulfur oxides and nitrogen oxides contained in exhaust gas are oxidized by the action of sunbeam to become sulfuric acid and nitric acid mists, which fall in the form of rain. Acid rain is closely related to the use of the coal containing high sulfur, and it hinders the use of coal which is rich energy source. In order to simplify the processing system for boiler exhaust gas and to reduce waste water and wastes, Ebara Corp. developed the dry simultaneous desulfurizing and denitrating technology utilizing electron beam in cooperation with Japan Atomic Energy Research Institute. The flow chart of the system applied to the exhaust gas treatment in a coal-fired thermal power station is shown. The mechanism of desulfurization and denitration, and the features of this system are described. The demonstration plant was constructed in a coal-fired thermal power station in Indianapolis, Indiana, USA, and the trial operation was completed in July, 1987. The test results are reported. (K.I.)

Behavior of tributyl phosphate in A-line processes

International Nuclear Information System (INIS)

Harmon, H.D.; Hyder, M.L.; Tiffany, B.; Gray, L.W.; Soltys, P.A.

1976-08-01

The chemical and physical properties of uranyl nitrate--tributyl phosphate adduct (UO 2 ) (NO 3 ) 2 (TBP) 2 were studied to define optimum, safe operating conditions for denitration of uranyl nitrate (UN) solutions containing low concentrations of adduct. The distribution of TBP between aqueous UN solutions and organic phases (TBP pure or diluted in n-paraffin) was measured. Specific gravity measurements confirmed published data for aqueous UN solutions, but disagreed with literature data for 100 percent TBP. Intersection of the UN specific gravity curve and the 100 percent TBP data of this work indicated that phase inversion cannot occur when the aqueous phase contains greater than or equal to 330 g U/1. Thermal decomposition of adduct occurs in one gradual weight loss step below 170 0 C and in two successive steps above 170 0 C. Rate constants for the first reaction were calculated for 130 to 210 0 C. The major flammable decomposition product was l-butene; inorganic gases formed were N 2 , NO, N 2 O, CO, and CO 2 . Adduct decomposition during denitration of UN was characterized by foaming and by gas evolution (mostly nonflammable) at 135 to 185 0 C. Maximum flammable gas evolution and potential self-heating by TBP oxidation were calculated

Fabrication of ceramic grade UO2 by direct conversion of uranyl nitrate hexahydrate

International Nuclear Information System (INIS)

Lainetti, P.E.O.; Riella, H.G.

1992-01-01

A method of direct conversion of uranyl nitrate hexahydrate (UNH) solution to ceramic grade uranium dioxide powders by thermal denitration in a furnace that combines atomization nozzle and a gas stirred bed is described. The main purpose of this work is to show that this alternative process is technically viable, specially if the recovery of the scrap generated in the nuclear fuel pellet production is required, without further generation of new liquid wastes. (author)

Review of experience gained in fabricating nuclear grade uranium and thorium compounds and their analytical quality control at the Instituto de Energia Atomica, Sao Paulo, Brazil

International Nuclear Information System (INIS)

Abrao, A.; Franca Junior, J.M.; Ikuta, A.

1977-01-01

The main activities developed at 'Instituto de Energia Atomica' Sao Paulo, Brazil, on the recovery of uranium from ores, the purification of uranium and thorium raw concentrates and their transformation in nuclear grade compounds, are reviewed. The design and assemble of pilot facilities for ammonium diuranate (ADV) uranium tetrafluoride, uranium trioxide, uranium oxide microspheres, uranyl nitrate denitration, uranim hexafluoride and thorium compounds are discussed. The establishment of analytical procedures are emphasized [pt

Thermal denitration of high concentration nitrate salts waste water

International Nuclear Information System (INIS)

Hwang, D. S.; Oh, J. H.; Choi, Y. D.; Hwang, S. T.; Park, J. H.; Latge, C.

2003-01-01

This study investigated the thermodynamic and the thermal decomposition properties of high concentration nitrate salts waste water for the lagoon sludge treatment. The thermodynamic property was carried out by COACH and GEMINI II based on the composition of nitrate salts waste water. The thermal decomposition property was carried out by TG-DTA and XRD. Ammonium nitrate and sodium nitrate were decomposed at 250 .deg. C and 730 . deg. C, respectively. Sodium nitrate could be decomposed at 450 .deg. C in the case of adding alumina for converting unstable Na 2 O into stable Na 2 O.Al 2 O 3 . The flow sheet for nitrate salts waste water treatment was proposed based on the these properties data. These will be used by the basic data of the process simulation

Gamma radiolysis of Cu(II) complex of metronidazole

International Nuclear Information System (INIS)

Mandal, P.C.; Bardhan, D.K.; Bhattacharyya, S.N.

1990-01-01

Aqueous solutions of Cu(II)Metronidazole (Cu(II)M) at neutral pH were irradiated with 60 Co γ-rays under different conditions of radiolysis. The radiolytic formation of HNO 2 and Cu(I) was followed. The radiolytic yields of chromophore loss of Cu(II)M were also determined under different conditions. The OH radicals attack the metal complex to give the OH adducts of the ligand at C 2 , C 4 , and C 5 either directly or through the formation of Cu(III) species. The C 5 -OH adduct, however, undergoes oxidative denitration and as a result the metal complex is decomposed. The OH adducts also undergo electron transfer to Cu(II) ion to give reduced complex. No denitration was observed due to the reaction of e eq - with the metal complex. On the other hand, the nitro group of the ligand in the complex undergoes successive 4-electron reduction to give hydroxylamino derivative. From the competition kinetics using t-butyl alcohol as the scavenger of OH in N 2 O saturated solution of the metal complex, the rate constant for the reaction of OH with complex was evaluated to be ca. 2.1x10 9 dm 3 mol -1 s -1 which is of the same order as that observed in the case of free metronidazole. (author)

Trends in desulphurisation and denitration techniques in the cement industry

Energy Technology Data Exchange (ETDEWEB)

Kupper, D

1991-03-01

A wide ranging article which examines various topics including primary SOx and NOx control measures, staged combustion, preheaters, precalciners, and comments on prospects for the future. As emission limits become more stringent, regulations for kiln systems will be adapted to those applicable to general combustion systems.

Structural investigations of zirconia powders obtained by hf-plasmochemical denitration

International Nuclear Information System (INIS)

Dedov, N.V.; Ivanov, Yu.F.; Dorda, F.A.; Paul', A.V.; Zav'yalov, A.V.; Koneva, N.A.; Korobtsev, V.P.; Kutyavin, E.M.; Mazin, V.I.; Matyukha, V.A.

1992-01-01

Results are presented of structural and physicochemical investigations of unstabilized and stabilized (using yttria, magnesia, calcium oxide and alumina) zirconia, obtained from nitrate solutions on pilot HF-plasmochemical equipment with an electric rating of 63 kW. The phase composition of the ultradispersed powder is shown. Morphological and grain-size analyses were carried out on the EM-125K electron diffraction microscope. specimens were prepared by applying powder to a carbon film obtained in a VUP-4 vacuum cell. The phase composition was studied by x-ray analysis on the DRON-3 diffractometer. These studies established that the main morphological constituents of the ZrO 2 powder are polycrystalline hollow spheres and fragments of films. The average sizes (diameter) of the spheres in 0.77 μ (mean square deviation σ n = 0.57μ) and for the grains in them 31 nm (σ n = 9.5 nm). There exists a certain correlation between the dimensions of the polycrystalline spheres and their grain structure - the coarser the powders, the larger the grains observed

Method and equipment to prepare aqueous radioactive waste for non-polluting and safe handling, transport, and final storage

Energy Technology Data Exchange (ETDEWEB)

Drobmik, S; Hild, W; Kaufmann, F; Koschorke, H

1977-09-15

A system is proposed which allows the joint execution of the process steps of denitration, drying and calcination necessary for the treatment of radioactive waste and the subsequent injection of superheated water vapor into the reaction chamber so as to also clean the offgases produced and discharge them from the facility through a filter system without polluting the environment. Several design variants are described which mainly relate to the arrangement of the spray nozzle and its environment. The vitrified radioactive materials are then put into repositories.

Pilot test of flue gas treatment by electron beam

International Nuclear Information System (INIS)

Tokunaga, Okihiro

1995-01-01

The development of the technology of the desulfurization and denitration for flue gas by using electron beam was started in Japan in 1970s, and since then, the development research for putting it to practical use and the basic research on the subjects which must be resolved for promoting the practical use have been advanced. Based on these results, the verifying test using a pilot scale plant was carried out from 1991 to 1994 for the treatment of coal-burning flue gas, municipal waste-burning flue gas and highway tunnel exhaust gas. The operation of the pilot plant was already finished, and the conceptual design of a practical scale plant based on the results and the assessment of the economical efficiency were performed. As for the coal-burning flue gas treatment by using electron beam, the basic test, the pilot test and the conceptual design of a practical scale plant and the assessment of the economical efficiency are reported. As for the municipal waste-burning flue gas treatment by using electron beam, the basic test and the pilot test are reported. Also the pilot test on the denitration of exhaust gas in highway tunnels in reported. In Poland, the pilot test on the treatment of flue gas in coal-burning thermal power stations is carried out. In Germany, the technical development for cleaning the air contaminated by volatile organic compounds by electron beam irradiation is advanced. (K.I.)

Trace Gas Evolution in the Present and Past Atmosphere

DEFF Research Database (Denmark)

Winther, Malte Nordmann

.Continuous incubation experiments are presented with nitrifying bacteria Nitrosomonasmobilis revealing strong indications of N2O production from different chemical reactions.The measurements revealed a three step site preference pattern in the range of nitricationand denitrication and we therefore suggest...... performed 1) on the Arctictundra and 2) on an inclined temperate slope. 1) Previous studies has shown that largeamounts of N2O is being emitted after thawing of permafrost. We investigated a downslopesite covering a moisture gradient area in the arctic tundra. Moss-covered sites revealed highnitrication...

Improvements to a uranium solidification process by in-plant testing

International Nuclear Information System (INIS)

Rindfleisch, J.A.

1984-01-01

When a process is having operational or equipment problems, often there is not enough time or money available for an extensive pilot plant program. This is when in-plant testing becomes imperative. One such process at the Idaho Chemical Processing Plant (ICPP) to undergo such an in-plant testing program was the uranium product solidification (denitrator) system. The testing program took approximately six months of in-plant testing that would have required at least two years of pilot plant preparation and operation to obtain the same information. This paper describes the results of the testing program, and the equipment and procedural changes

Pilot plant experience on high-level waste solidification and design of the engineering prototype VERA

Energy Technology Data Exchange (ETDEWEB)

Guber, W; Diefenbacher, W; Hild, W; Krause, H; Schneider, E; Schubert, G

1972-11-01

In the present paper the solidification process for highly active waste solutions as developed in the Karlsruhe Nuclear Research Center is presented. Its principal steps are: denitration, calcination in a spray calciner operated with superheated steam, melting of the calcine with appropriate additives to borosilicate glass in an induction-heated melting furnace. The operational experiences gained so far in the inactive 1:1 pilot plant are reported. Furthermore, a description is given of the projected multi-purpose experimental facility VERA 2 which is provided for processing the highly active waste solutions from the first German reprocessing plant WAK.

Alpha process with biological elimination of nitrogen. Application of mathematical models; Proceso alpha con eliminacion biologica de nitrogeno. Aplicacion de modelos matematicos

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, J. C.; Lopez-Carrasco, M. D.; Cortacans, J. A.; Larrea, L.; Larrea, A.

1999-07-01

This article illustrates the advantages of a step feed process for the biological elimination of nitrogen by presenting the experiments carried out by INFILCO at a pilot plant in San Sebastian. This arrangement, also known as the alpha (alternative phase step feed) process, reduces the volume of the biological reactor, eliminates the need for internal recycling and optimised the consumption of the organic matter used for denitrication. This article also demonstrates the possibility of employing a mathematical model as a tool in assessing, designing and operating full scale treatment plants for typically urban sewage. (Author) 6 refs.

Integrated removal of NO and mercury from coal combustion flue gas using manganese oxides supported on TiO2.

Science.gov (United States)

Zhang, Shibo; Zhao, Yongchun; Wang, Zonghua; Zhang, Junying; Wang, Lulu; Zheng, Chuguang

2017-03-01

A catalyst composed of manganese oxides supported on titania (MnO x /TiO 2 ) synthesized by a sol-gel method was selected to remove nitric oxide and mercury jointly at a relatively low temperature in simulated flue gas from coal-fired power plants. The physico-chemical characteristics of catalysts were investigated by X-ray fluorescence (XRF), X-ray diffraction (XRD), and X-ray photoelectron spectroscopy (XPS) analyses, etc. The effects of Mn loading, reaction temperature and individual flue gas components on denitration and Hg 0 removal were examined. The results indicated that the optimal Mn/Ti molar ratio was 0.8 and the best working temperature was 240°C for NO conversion. O 2 and a proper ratio of [NH 3 ]/[NO] are essential for the denitration reaction. Both NO conversion and Hg 0 removal efficiency could reach more than 80% when NO and Hg 0 were removed simultaneously using Mn0.8Ti at 240°C. Hg 0 removal efficiency slightly declined as the Mn content increased in the catalysts. The reaction temperature had no significant effect on Hg 0 removal efficiency. O 2 and HCl had a promotional effect on Hg 0 removal. SO 2 and NH 3 were observed to weaken Hg 0 removal because of competitive adsorption. NO first facilitated Hg 0 removal and then had an inhibiting effect as NO concentration increased without O 2 , and it exhibited weak inhibition of Hg 0 removal efficiency in the presence of O 2 . The oxidation of Hg 0 on MnO x /TiO 2 follows the Mars-Maessen and Langmuir-Hinshelwood mechanisms. Copyright © 2016. Published by Elsevier B.V.

A new method for the determination of the nitrogen content of nitrocellulose based on the molar ratio of nitrite-to-nitrate ions released after alkaline hydrolysis

Energy Technology Data Exchange (ETDEWEB)

Alinat, Elodie, E-mail: elodie.alinat@chimie-paristech.fr [PSL Research University, Chimie ParisTech, Laboratory of Physicochemistry of Electrolytes, Colloids and Analytical Sciences (PECSA), 11 rue Pierre et Marie Curie, 75005 Paris (France); Central Laboratory of Police Prefecture (LCPP), 39 bis rue de Dantzig, 75015 Paris (France); CNRS, UMR 7195 PECSA, 11 rue Pierre et Marie Curie, 75005 Paris (France); Sorbonne Universités, UPMC Univ Paris 06, LBM, 4 place Jussieu, F-75005 Paris (France); Delaunay, Nathalie, E-mail: nathalie.delaunay@espci.fr [PSL Research University, Chimie ParisTech, Laboratory of Physicochemistry of Electrolytes, Colloids and Analytical Sciences (PECSA), 11 rue Pierre et Marie Curie, 75005 Paris (France); CNRS, UMR 7195 PECSA, 11 rue Pierre et Marie Curie, 75005 Paris (France); Sorbonne Universités, UPMC Univ Paris 06, LBM, 4 place Jussieu, F-75005 Paris (France); Archer, Xavier, E-mail: xavier.archer@interieur.gouv.fr [Central Laboratory of Police Prefecture (LCPP), 39 bis rue de Dantzig, 75015 Paris (France); Mallet, Jean-Maurice, E-mail: jean-maurice.mallet@es.fr [École Normale Supérieure-PSL Research University, Département de Chimie, 24 rue Lhomond, 75005 Paris (France); Sorbonne Universités, UPMC Univ Paris 06, LBM, 4 place Jussieu, F-75005 Paris (France); CNRS, UMR 7203 LBM, F-75005 Paris (France); Gareil, Pierre, E-mail: pierre.gareil@chimie-paristech.fr [PSL Research University, Chimie ParisTech, Laboratory of Physicochemistry of Electrolytes, Colloids and Analytical Sciences (PECSA), 11 rue Pierre et Marie Curie, 75005 Paris (France); CNRS, UMR 7195 PECSA, 11 rue Pierre et Marie Curie, 75005 Paris (France); Sorbonne Universités, UPMC Univ Paris 06, LBM, 4 place Jussieu, F-75005 Paris (France)

2015-04-09

Highlights: • New insights into the nitrocellulose alkaline denitration mechanism. • Linear correlation for molar ratio of nitrite-to-nitrate ions and nitrogen content. • Capillary electrophoresis monitoring of nitrite and nitrate ions. • Applications to explosive and non-explosive nitrocellulose-containing samples. • Improved performances (including safety) over classical methods. - Abstract: A new method was proposed to determine the nitrogen content of nitrocelluloses (NCs). It is based on the finding of a linear relationship between the nitrogen content and the molar ratio of nitrite-to-nitrate ions released after alkaline hydrolysis. Capillary electrophoresis was used to monitor the concentration of nitrite and nitrate ions. The influences of hydrolysis time and molar mass of NC on the molar ratio of nitrite-to-nitrate ions were investigated, and new insights into the understanding of the alkaline denitration mechanism of NCs, underlying this analytical strategy is provided. The method was then tested successfully with various explosive and non-explosive NC-containing samples such as various daily products and smokeless gunpowders. Inherently to its principle exploiting a concentration ratio, this method shows very good repeatability in the determination of nitrogen content in real samples with relative standard deviation (n = 3) inferior to 1.5%, and also provides very significant advantages with respect to sample extraction, analysis time (1 h for alkaline hydrolysis, 3 min for electrophoretic separation), which was about 5 times shorter than for the classical Devarda's method, currently used in industry, and safety conditions (no need for preliminary drying NC samples, mild hydrolysis conditions with 1 M sodium hydroxide for 1 h at 60 °C)

A new method for the determination of the nitrogen content of nitrocellulose based on the molar ratio of nitrite-to-nitrate ions released after alkaline hydrolysis

International Nuclear Information System (INIS)

Alinat, Elodie; Delaunay, Nathalie; Archer, Xavier; Mallet, Jean-Maurice; Gareil, Pierre

2015-01-01

Highlights: • New insights into the nitrocellulose alkaline denitration mechanism. • Linear correlation for molar ratio of nitrite-to-nitrate ions and nitrogen content. • Capillary electrophoresis monitoring of nitrite and nitrate ions. • Applications to explosive and non-explosive nitrocellulose-containing samples. • Improved performances (including safety) over classical methods. - Abstract: A new method was proposed to determine the nitrogen content of nitrocelluloses (NCs). It is based on the finding of a linear relationship between the nitrogen content and the molar ratio of nitrite-to-nitrate ions released after alkaline hydrolysis. Capillary electrophoresis was used to monitor the concentration of nitrite and nitrate ions. The influences of hydrolysis time and molar mass of NC on the molar ratio of nitrite-to-nitrate ions were investigated, and new insights into the understanding of the alkaline denitration mechanism of NCs, underlying this analytical strategy is provided. The method was then tested successfully with various explosive and non-explosive NC-containing samples such as various daily products and smokeless gunpowders. Inherently to its principle exploiting a concentration ratio, this method shows very good repeatability in the determination of nitrogen content in real samples with relative standard deviation (n = 3) inferior to 1.5%, and also provides very significant advantages with respect to sample extraction, analysis time (1 h for alkaline hydrolysis, 3 min for electrophoretic separation), which was about 5 times shorter than for the classical Devarda's method, currently used in industry, and safety conditions (no need for preliminary drying NC samples, mild hydrolysis conditions with 1 M sodium hydroxide for 1 h at 60 °C)

Optimization of an Sbr process for nitrogen removal from concentrated wastewater via nitrite

International Nuclear Information System (INIS)

Longhi, L.; Basilico, D.; Meloni, A.; Canziani, R.

2009-01-01

The results of an experimentation carried out on a pilot-scale Sbr for nitrogen removal via nitridation-denitration are reported. The experimentation was carried out in the period July October 2007 and was aimed at achieving design data for the upgrade of a full scale wastewater treatment plant (WWTP), following the new regulations issued by Lombardy Regional Authority for the discharge of effluents into sensitive areas. One aspect that has been considered in the upgrade is nitrogen removal from the supernatant coming from anaerobic sludge digestion. The experimental results provided sound design data based on real biological activity measurements and operational process parameters such as oxygen and organic carbon requirements. [it

FIPS: a process for the solidification of fission product solutions using a drum drier. [HTGR fuel reprocessing

Energy Technology Data Exchange (ETDEWEB)

Halaszovich, St.; Laser, M.; Merz, E.; Thiele, D.

1976-08-01

A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the crucible by rising level in-pot-melting, and off-gas treatment and recovery of nitric acid is under development. A small plant with a capacity of 1 kg glass per hour has been tested in hot cells with fission product solutions from LWR fuel element reprocessing since December 1974. The equipment is very simple to operate and to control. No serious problems arose during operation.

Criticality safety evaluation in Tokai Reprocessing Plant

International Nuclear Information System (INIS)

Shirai, Nobutoshi; Nakajima, Masayoshi; Takaya, Akikazu; Ohnuma, Hideyuki; Shirouzu, Hidetomo; Hayashi, Shinichiro; Yoshikawa, Koji; Suto, Toshiyuki

2000-04-01

Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 'Criticality safety of single unit' in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units. (author)

Application of non-reductive partial partitioning in FBR fuel reprocessing: a simulation study

International Nuclear Information System (INIS)

Shekhar Kumar; Koganti, S.B.

2002-06-01

The observed performance of conventional partitioning contactor in the Purex process in the Purex process is seldom satisfactory due to over-consumption of reductant and poor U-Pu decontamination factors. Contemporary trends indicate gradually move-over to MOX fuels for FBRs. In this scenario, it is not necessary to separate uranium and plutonium completely. By controlling the acid concentration and flow rates, it is possible to selectively strip essentially all the plutonium and part of the uranium in the aqueous stream. Therefore, a mixed product enriched in plutonium is obtained which can be precipitated and denitrated. Alternatively direct denitration by microwave heating can be used. The idea is particularly attractive for the flowsheet meant for the first core of FBTR where Pu/(U+Pu) ratio of 0.7 (in the discharged fuel) is diluted by addition of uranium to 0.3. By partial partitioning in the 2B contactor, a product enriched in plutonium (Pu/(U+Pu) ratio ∼0.6) can be obtained. After a minor uranium addition, this product will be suitable for the direct fabrication of II core of FBTR where a Pu/(U+Pu) ratio of 0.55 will be required. In a similar fashion, the enriched product can be used for multiple core zones of proposed PFBR by selective additions of uranium for each zone. To explore the feasibility of such partial partitioning step, an exhaustive simulation study was made using the in-house developed computer code SIMPSEX. 1300 simulations runs were completed for different combinations of parameters and results were analyzed. In this report, the results of this study and a possible flowsheet step have been discussed. Simultaneous variation in the flow rates has been considered and safe operating limits for the partial partitioning step have been established. (author)

Design of an ion exchange column for plutonium recovery

International Nuclear Information System (INIS)

Araujo, J.A. de; Matsuda, H.T.; Santos Tome Lobao, A. dos; Quesada, A.C.

1994-01-01

An ion exchange column design for plutonium recovering from scraps of the MOX fuel elements fabrication is presented. The proposed column is constructed in 304 stainless steel and borosilicate glass provided of heating-jacket and temperature control and pressure relief devices. Safety aspects required for alpha emitters handling have been also considered. The design and construction were performed totally at Brazilian Institute for Energetic and Nuclear Research. The equipment will be used in the plutonium separation step as a part of an installation named Facilidad Alfa at the Centro Atomico de Constituyentes-CNEA/Buenos Aires, where other processes, including dissolution denitration by microwaves and final steps of MOX pellets re-fabrication will be performed. (author). 4 refs, 3 figs

Role of thermal analysis in uranium oxide fuel fabrication process

International Nuclear Information System (INIS)

Balaji Rao, Y.; Yadav, R.B.

2006-01-01

The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)

Process for denitrating waste solutions containing nitric acid actinides simultaneously separating the actinides

International Nuclear Information System (INIS)

Gompper, K.

1984-01-01

The invention should reduce the acid and nitrate content of waste solutions containing nitric acid as much as possible, should reduce the total salt content of the waste solution, remove the actinides contained in it by precipitation and reduce the α radio-activity in the remaining solution, without having to worry about strong reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig.) [de

Powder technological vitrification of simulated high-level waste

International Nuclear Information System (INIS)

Gahlert, S.

1988-03-01

High-level waste simulate from the reprocessing of light water reactor and fast breeder fuel was vitrified by powder technology. After denitration with formaldehyde, the simulated HLW is mixed with glass frit and simultaneously dried in an oil-heated mixer. After 'in-can calcination' for at least 24 hours at 850 or 950 K (depending on the type of waste and glass), the mixture is hot-pressed in-can for several hours at 920 or 1020 K respectively, at pressures between 0.4 and 1.0 MPa. The technology has been demonstrated inactively up to diameters of 30 cm. Leach resistance is significantly enhanced when compared to common borosilicate glasses by the utilization of glasses with higher silicon and aluminium content and lower sodium content. (orig.) [de

Development of a pyro-partitioning process for long-lived radioactive nuclides. Process test for pretreatment of simulated high-level waste containing uranium

International Nuclear Information System (INIS)

Kurata, Masateru; Hijikata, Takatoshi; Kinoshita, Kensuke; Inoue, Tadashi

2000-01-01

A pyro-partitioning process developed at CRIEPI requires a pre-treatment process to convert high-level liquid waste to chloride. A combination process of denitration and chlorination has been developed for this purpose. Continuous process tests using simulated high-level waste were performed to certify the applicability of the process. Test results indicated a successful material balance sufficient for satisfying pyro-partitioning process criteria. In the present study, process tests using simulated high-level waste containing uranium were also carried out to prove that the pre-treatment process is feasible for uranium. The results indicated that uranium can be converted to chloride appropriate for the pyro-partitioning process. The material balance obtained from the tests is to be used to revise the process flow diagram. (author)

Synthesis and Characterization of Oxide Feedstock Powders for the Fuel Cycle R and D Program

International Nuclear Information System (INIS)

Voit, Stewart L.; Vedder, Raymond James; Johnson, Jared A.

2010-01-01

Nuclear fuel feedstock properties, such as physical, chemical, and isotopic characteristics, have a significant impact on the fuel fabrication process and, by extension, the in-reactor fuel performance. This has been demonstrated through studies with UO 2 spanning greater than 50 years. The Fuel Cycle R and D Program with The Department of Energy Office of Nuclear Energy has initiated an effort to develop a better understanding of the relationships between oxide feedstock, fresh fuel properties, and in-reactor fuel performance for advanced mixed oxide compositions. Powder conditioning studies to enable the use of less than ideal powders for ceramic fuel pellet processing are ongoing at Los Alamos National Laboratory (LANL) and an understanding of methods to increase the green density and homogeneity of pressed pellets has been gained for certain powders. Furthermore, Oak Ridge National Laboratory (ORNL) is developing methods for the co-conversion of mixed oxides along with techniques to analyze the degree of mixing. Experience with the fabrication of fuel pellets using co-synthesized multi-constituent materials is limited. In instances where atomically mixed solid solutions of two or more species are needed, traditional ceramic processing methods have been employed. Solution-based processes may be considered viable synthesis options, including co-precipitation (AUPuC), direct precipitation, direct-conversion (Modified Direct Denitration or MDD) and internal/external gelation (sol-gel). Each of these techniques has various advantages and disadvantages. The Fiscal Year 2010 feedstock development work at ORNL focused on the synthesis and characterization of one batch of UO x and one batch of U 80 Ce 20 O x . Oxide material synthesized at ORNL is being shipped to LANL for fuel fabrication process development studies. The feedstock preparation was performed using the MDD process which utilizes a rotary kiln to continuously thermally denitrate double salts of ammonium

Development of a partitioning method for the management of high-level liquid waste

International Nuclear Information System (INIS)

Kubota, M.; Dojiri, S.; Yamaguchi, I.; Morita, Y.; Yamagishi, I.; Kobayashi, T.; Tani, S.

1989-01-01

Fundamental studies especially focused on the separation of neptunium and technetium have been carried out to construct the advanced partitioning process of fractioning elements in a high-level liquid waste into four groups: transuranium elements, technetium-noble metals, strontium-cesium, and other elements. For the separation of neptunium by solvent extraction, DIDPA proved excellent for extracting Np(V), and its extraction rate was accelerated by hydrogen peroxide. Np(V) was found to be also separated quantitatively as precipitate with oxalic acid. For the separation of technetium, the denitration with formic acid was effective in precipitating it along with noble metals, and the adsorption with activated carbon was also effective for quantitative separation. Through these fundamental studies, the advanced partitioning process is presented as the candidate to be examined with an actual high-level liquid waste

Pressure effects on nanostructured manganites

International Nuclear Information System (INIS)

Acha, C.; Garbarino, G.; Leyva, A.G.

2007-01-01

We have measured the pressure sensitivity of magnetic properties on La 5/8-y Pr y Ca 3/8 MnO 3 (y=0.3) nanostructured powders. Samples were synthesized following a microwave assisted denitration process and a final heat treatment at different temperatures to control the grain size of the samples. A span in grain diameters from 40 nm to ∼1000 nm was obtained. Magnetization curves as a function of temperature were measured following different thermomagnetic histories. AC susceptibility as a function of temperature was also measured at different hydrostatic pressures (up to 10 kbar) and for different frequencies. Our results indicate that the nanostructuration plays a role of an internal pressure, producing a structural deformation with similar effects to those obtained under an external hydrostatic pressure

Formation and filtration characteristics of solids generated in a high level liquid waste treatment process. Solids formation behavior from simulated high level liquid waste

International Nuclear Information System (INIS)

Kondo, Y.; Kubota, M.

1997-01-01

The solids formation behavior in a simulated high level liquid waste (HLLW) was experimentally examined, when the simulated HLLW was treated in the ordinary way of actual HLLW treatment process. Solids formation conditions and mechanism were closely discussed. The solids formation during a concentration step can be explained by considering the formation of zirconium phosphate, phosphomolybdic acid and precipitation of strontium and barium nitrates and their solubilities. For the solids formation during the denitration step, at least four courses were observed; formation of an undissolved material by a chemical reaction with each other of solute elements (zirconium, molybdenum, tellurium) precipitation by reduction (platinum group metals) formation of hydroxide or carbonate compounds (chromium, neodymium, iron, nickel, strontium, barium) and a physical adsorption to stable solid such as zirconium molybdate (nickel, strontium, barium). (author)

Recovery of Am-Cm from high-activity waste concentrate by in-canyon-tank precipitation as oxalates

International Nuclear Information System (INIS)

Gray, L.W.; Burney, G.A.; Wilson, T.W.; McKibben, J.M.

1980-01-01

Savannah River Laboratory and Savannah River Plant have been separating actinides for more than 25 years. Work continues to upgrade processes and to initiate new processes. This report summarizes work on a precipitation process to separate kg amounts of Am and Cm from hundreds of kilograms of NaNO 3 and Al(NO 3 ) 3 . The developed process includes formic acid denitration of the Am-Cm bearing streams for acid adjustment; oxalate precipitation of the Am-Cm; and Mn +2 catalyzed oxidation of oxalate in both the decanted supernate and the precipitated actinides. The new process generates one fourth the radioactive waste as the solvent extraction process which it replaced, and produces a cleaner feed solution for downstream processing to separate the Am and Cm before conversion to their respective oxides

Evaluation and development plan of NRTA measurement methods for the Rokkasho Reprocessing Plant

International Nuclear Information System (INIS)

Li, T.K.; Hakkila, E.A.; Flosterbuer, S.F.

1995-01-01

Near-real-time accounting (NRTA) has been proposed as a safeguards method at the Rokkasho Reprocessing Plant (RRP), a large-scale commercial boiling water and pressurized water reactors spent-fuel reprocessing facility. NRTA for RRP requires material balance closures every month. To develop a more effective and practical NRTA system for RRP, we have evaluated NRTA measurement techniques and systems that might be implemented in both the main process and the co-denitration process areas at RRP to analyze the concentrations of plutonium in solutions and mixed oxide powder. Based on the comparative evaluation, including performance, reliability, design criteria, operation methods, maintenance requirements, and estimated costs for each possible measurement method, recommendations for development were formulated. This paper discusses the evaluations and reports on the recommendation of the NRTA development plan for potential implementation at RRP

Application of plutonium inventory measurement system (PIMS) and temporary canister verification system (TCVS) at RRP

International Nuclear Information System (INIS)

Noguchi, Yoshihiko; Nakamura, Hironobu; Adachi, Hideto; Iwamoto, Tomonori

2004-01-01

In U-Pu co-denitration area at Rokkasho Reprocessing Plant (RRP), Plutonium Inventory Measurement System (PIMS) and Temporary Canister Verification System (TCVS) are installed to provide efficient and effective safeguards. PIMS measures Pu quantity inside pipes and vessels installed in glove boxes by total neutron counting method. PIMS consists of total 142 neutron detector attached on the wall and top of glove boxes and neutron count rates of each detectors are related to each other to calculate Pu quantity of each process areas. In this moment, inactive calibration using Cf-source was completed. On the other hand, TCVS measures Pu quantity of canisters inside temporary storage by coincidence counting method and it will be installed before the active test. These systems have monitoring function as additional measures. This paper describes specification, performance and measurement principles of PIMS and TCVS. (author)

Am/Cm Vitrification Process: Pretreatment Material Balance Calculations

International Nuclear Information System (INIS)

Smith, F.G.

2001-01-01

This report documents material balance calculations for the pretreatment steps required to prepare the Americium/Curium solution currently stored in Tank 17.1 in the F-Canyon for vitrification. The material balance uses the latest analysis of the tank contents to provide a best estimate calculation of the expected plant operations during the pretreatment process. The material balance calculations primarily follow the material that directly leads to melter feed. Except for vapor products of the denitration reactions and treatment of supernate from precipitation and precipitate washing, the flowsheet does not include side streams such as acid washes of the empty tanks that would go directly to waste. The calculation also neglects tank heels. This report consolidates previously reported results, corrects some errors found in the spreadsheet and provides a more detailed discussion of the calculation basis

Combined removal of sulfur compounds and nitrate by autotrophic denitrication in bioaugmented activated sludge system

NARCIS (Netherlands)

Manconi, I.; Carucci, A.; Lens, P.N.L.

2007-01-01

An autotrophic denitrification process using reduced sulfur compounds (thiosulfate and sulfide) as electron donor in an activated sludge system is proposed as an efficient and cost effective alternative to conventional heterotrophic denitrification for inorganic (or with low C/N ratio) wastewaters

Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

International Nuclear Information System (INIS)

Gompper, K.

1986-01-01

The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

Applications of Fuzzy adaptive PID control in the thermal power plant denitration liquid ammonia evaporation

Directory of Open Access Journals (Sweden)

Li Jing

2016-01-01

Full Text Available For the control of the liquid level of liquid ammonia in thermal power plant’s ammonia vaporization room, traditional PID controller parameter tuning is difficult to adapt to complex control systems, the setting of the traditional PID controller parameters is difficult to adapt to the complex control system. For the disadvantage of bad parameter setting, poor performance and so on the fuzzy adaptive PID control is proposed. Fuzzy adaptive PID control combines the advantages of traditional PID technology and fuzzy control. By using the fuzzy controller to intelligent control the object, the performance of the PID controller is further improved, and the control precision of the system is improved[1]. The simulation results show that the fuzzy adaptive PID controller not only has the advantages of high accuracy of PID controller, but also has the characteristics of fast and strong adaptability of fuzzy controller. It realizes the optimization of PID parameters which are in the optimal state, and the maximum increase production efficiency, so that are more suitable for nonlinear dynamic system.

Separation of Am-Cm from NaNO3 waste solutions by in-canyon-tank precipitation as oxalates

International Nuclear Information System (INIS)

Gray, L.W.; Burney, G.A.; Wilson, T.W.; McKibben, J.M.

1981-09-01

A process for the purification of Am-Cm residues was developed specifically for application in Savannah River Plant (SRP) canyon tanks. These Am-Cm residues were collected from several campaigns to produce plutonium containing high isotopic concentrations of 240 Pu. An initial purification from Al 3+ had already been accomplished by a solvent extraction process. The product of this process was contaminated with NaNO 3 as a result of entrainment of the solvent extraction NaNO 3 scrub solution. To produce an acceptable feed solution for a later pressurized cation exchange chromatography separation and purification step, the bulk of the NaNO 3 must be removed. This purification process includes formic acid denitration, adjustment of contaminating cations by evaporation and water dilution, and oxalate precipitation of the actinides and lanthanides. After washing, the precipitate was dissolved in 8M nitric acid, and the oxalate was destroyed by nitric acid oxidation

Properties and thermal decomposition of the double salts of uranyl nitrate-ammonium nitrate

International Nuclear Information System (INIS)

Notz, K.J.; Haas, R.A.

1989-01-01

The formation of ammonium nitrate-uranyl nitrate double salts has important effects on the thermal denitration process for the preparation of UO 3 and on the physical properties of the resulting product. Analyses were performed, and properties and decomposition behavior were determined for three double salts: NH 4 UO 2 (NO 3 ) 3 , (NH 4 ) 2 UO 2 (NO 3 ) 4 , and (NH 4 ) 2 UO 2 (NO 3 ) 4 ·2H 2 O. The tinitrate salt decomposes without melting at 270-300 C to give a γ-UO 3 powder of ∼3-μm average size, with good ceramic properties for fabrication into UO 2 nuclear fuel pellets. The tetranitrate dihydrate melts at 48 C; it also dehydrates to the anhydrous salt. The anhydrous tetranitrate decomposes exothermically, without melting, at 170-270 C by losing one mole of ammonium nitrate to form the trinitrate salt

Effect of UV on De-NOx performance and microbial community of a hybrid catalytic membrane biofilm reactor

Science.gov (United States)

Chen, Zhouyang; Huang, Zhensha; He, Yiming; Xiao, Xiaoliang; Wei, Zaishan

2018-02-01

The hybrid membrane catalytic biofilm reactor provides a new way of flue gas denitration. However, the effects of UV on denitrification performance, microbial community and microbial nitrogen metabolism are still unknown. In this study, the effects of UV on deNO x performance, nitrification and denitrification, microbial community and microbial nitrogen metabolism of a bench scale N-TiO2/PSF hybrid catalytic membrane biofilm reactor (HCMBR) were evaluated. The change from nature light to UV in the HCMBR leads to the fall of NO removal efficiency of HCMBR from 92.8% to 81.8%. UV affected the microbial community structure, but did not change microbial nitrogen metabolism, as shown by metagenomics sequencing method. Some dominant phyla, such as Gammaproteobacteria, Bacteroidetes, Firmicutes, Actinobacteria, and Alphaproteobacteria, increased in abundance, whereas others, such as Proteobacteria and Betaproteobacteria, decreased. There were nitrification, denitrification, nitrogen fixation, and organic nitrogen metabolism in the HCMBR.

The new development of radiation processing in China

International Nuclear Information System (INIS)

Chen Dianhua

1998-12-01

China Isotope and Radiation Association (CIRA) investigated the status of radiation processing in China's mainland. There are 45 accelerators each with beam power over 5 kW, the total beam power is 2005 kW. There are 48 γ facilities each with designed capacity over 1.11 x 10 4 TBq (0.3 million curies) and other 75 units with designed capacity less than 1.11 x 10 4 TBq. The total loaded capacity is 4.63 x 10 5 TBq, more than one third of the designed capacity. radiation processing is mainly used in producing chemical-industrial products (as heat-shrinkable products and radiation cross-linked wire and cable) in China. Some enterprises with annual output value over a hundred million RMB have emerged. Radiation processing is also used in preservation and disinfection of food. In 1977, six kinds of hygienic standards for irradiated food were authorized. Radiation sterilization of disposable medical products, radiation desulphurization and denitration are also being developed in China

Studies on the reaction of nitric acid and sugar

International Nuclear Information System (INIS)

MacDougall, C.S.; Bayne, C.K.; Roberson, R.B.

1982-01-01

The design of vessels and off-gas systems for denitrating acidic radioactive process solutions by reacting nitric acid with sugar requires a fairly accurate determination of the rate of the controlling step. Therefore, the reaction of sugar with concentrated nitric acid was closely examined at temperatures of 100 and 110 0 C and in the presence of low levels of iron )0 to 0.2 M Fe(III)). Efficiencies of the sugar destruction by nitric acid ranged from 2.56 to 2.93 mol of acid consumed per mole of carbon added. Product off-gases were examined throughout the reaction. Release of CO was fairly constant throughout the reaction, but amounts of CO 2 increased as the nitric acid began to attack the terminal carboxylic acids produced from the consumption of sucrose. Voluminous quantities of NO 2 were released at the beginning of the reaction, but larger relative concentrations of NO were observed toward the end

Defense by-products production and utilization program: noble metal recovery screening experiments

International Nuclear Information System (INIS)

Hazelton, R.F.; Jensen, G.A.; Raney, P.J.

1986-03-01

Isotopes of the platinum metals (rutheium, rhodium, and palladium) are produced during uranium fuel fission in nuclear reactors. The strategic values of these noble metals warrant considering their recovery from spent fuel should the spent fuel be processed after reactor discharge. A program to evaluate methods for ruthenium, rhodium, and palladium recovery from spent fuel reprocessing liquids was conducted at Pacific Northwest Laboratory (PNL). The purpose of the work reported in this docuent was to evaluate several recovery processes revealed in the patent and technical literature. Beaker-scale screening tests were initiated for three potential recovery processes: precipitation during sugar denitration of nitric acid reprocessing solutions after plutonium-uranium solvent extraction, adsorption using nobe metal selective chelates on active carbon, and reduction forming solid noble metal deposits on an amine-borane reductive resin. Simulated reprocessing plant solutions representing typical nitric acid liquids from defense (PUREX) or commercial fuel reprocessing facilities were formulated and used for evaluation of the three processes. 9 refs., 3 figs., 9 tabs

Product Conversion: The Link between Separations and Fuel Fabrication

International Nuclear Information System (INIS)

Felker, L.K.; Vedder, R.J.; Walker, E.A.; Collins, E.D.

2008-01-01

Several chemical processing flowsheets are under development for the separation and isolation of the actinide, lanthanide, and fission product streams in spent nuclear fuel. The conversion of these product streams to solid forms, typically oxides, is desired for waste disposition and recycle of product fractions back into transmutation fuels or targets. The modified direct denitration (MDD) process developed at Oak Ridge National Laboratory (ORNL) in the 1980's offers significant advantages for the conversion of the spent fuel products to powder form suitable for direct fabrication into recycle fuels. A glove-box-contained MDD system and a fume-hood-contained system have been assembled at ORNL for the purposes of testing the co-conversion of uranium and mixed-actinide products. The current activities are focused on the conversion of the first products from the processing of spent nuclear fuel in the Coupled End-to-End Demonstration currently being conducted at ORNL. (authors)

TG/DTA and X ray Diffraction Studies on Ammonium Uranyl Nitrate

International Nuclear Information System (INIS)

Kim, Byung Ho; Lee, Young Bum; Jeong, Ji Young; Choi, Jong Hyun; Kim, Tae Joon; Nam, Ho Yun; Kim, Jong Man

2011-01-01

Ammonium uranyl nitrate (AUN) is an important intermediate product during conversion of a uranyl nitrate[UO 2 (NO 3 ) 2 ] solution to UO 2 powder for the fabrication of nuclear fuels, the so-called modified direct denitration (MDD) process. The MDD process involves the thermal decomposition of AUN double salts, which are prepared from a mixture consisting of a UO 2 (NO 3 ) 2 solution and NH 4 NO 3 . The physical and chemical properties of an oxide powder depend upon its thermal treatment. Three double salts are known for the UO 2 (NO 3 ) 2 - NH 4 NO 3 -H 2 O system, but there have been only a few studies done on thermal decomposition of these salts. Therefore, the objective of this study is to investigate the reaction pathways during a thermal decomposition and reduction of AUN to achieve a better knowledge of the influence of an AUN preparation process and thermal decomposition procedures on uranium oxides under a nitrogen, air, or hydrogen atmosphere

The chemistry of plutonium in sol-gel processes

International Nuclear Information System (INIS)

Lloyd, M.H.; Haire, R.G.

1978-01-01

Studies of plutonium chemical behavior conducted in conjunction with plutonia sol-gel process development at ORNL are described. The colloidal solutions produced consist of 'Pu(IV) polymer,' and this is therefore the study of polymeric plutonium behavior. Spectrophotometric, electron diffraction, and electron microscopy studies, in addition to specific studies that were concerned with the colloidal behavior of Pu(IV) polymer, indicate several characteristics of polymer that are not generally recognized. The particle nature of Pu(IV) polymer indicated by electron microscopy, the amorphous-crystalline characteristics of primary polymer particles demonstrated by electron diffraction, and the reversible and irreversible aggregation of the primary particles shown by spectrophotometric techniques present a useful view of the nature of Pu(IV) polymer that has been helpful in solving or understanding various types of processing problems involving plutonium hydrolytic behavior. The colloidal characteristics of Pu(IV) polymer and crystallite growth of primary polymer particles by thermal denitration are also described. (orig.) [de

Bio-desulfurization technology in Japan; Wagakuni ni okeru baio datsuryu gijutsu

Energy Technology Data Exchange (ETDEWEB)

Maruhashi, K. [Petroleum Energy Center, Tokyo (Japan)

2000-05-01

A bio-reaction of microbes (catalytic reaction by an enzyme) is characterized in that the reaction is carried out at a normal temperature and under a normal pressure and has particularly high specificity with respect to substrate (reactant). It is considered that a low loading process of environment harmony type can be constructed by applying the bio-reaction in petroleum refinery process. CO{sub 2} exhaust and energy consumption in the bio-desulfurization (BDS) is estimated to be 70 to 80% lower than those in hydrodesulfurization (HDS). The bio-technologies that can be applied to the petroleum refinery process include, for example, desulfurization, demetallation, dewaxing, denitration, cracking and so on. In this paper, the present state of bio-desulphurization technology is introduced. Particularly, as the research results in Japan, acquirement of mesophile R.erythropolis KA2-5-1 strain, thermophile Paenibacillus sp. A11-2 strain whose optimum temperature is 50 degrees C, BT degradation fungus Rhodococcus sp. T09 and the like are introduced. (NEDO)

Development of long-lived radionuclide partitioning technology

Energy Technology Data Exchange (ETDEWEB)

Yoo, Jae Hyung; Lee, Eil Hee; Kim, Kwang Wook; Yang, Han Beom; Chung, Dong Yong; Lim, Jae Kwan; Shin, Young Jun; Kim, Heung Ho; Kown, Sun Gil; Kim, Young Hwan; Hwang, Doo Seung

1996-07-01

This study has been focused on the development of unit processes for partitioning in the 1st stage, and experimentally carried out to examine the separation characteristics and operation conditions on the following unit processes. (1) Removal of a small amount of uranium by extraction with TBP, (2) Removal of Zr and Mo and destruction of nitric acid by uranium by denitration with formic acid, (3) Co-precipitation of Am, Np and RE oxalic acid, (4) Dissolution and destruction of oxalate by hydrogen peroxide, (5) Co-extraction of Am, Np and RE by nitric acid, (8) Back-extraction of Np by oxalic acid, (9) Adsorption and elution of Cs and Sr by zeolite, and (10) Advanced separation of radionuclide by electrochemical REDOX method. The results obtained from each unit process will be use as the basic materials for the establishment of optimal partitioning and design of process equipment. (author). 46 refs., 54 tabs., 222 figs.

Development of partitioning method

International Nuclear Information System (INIS)

Kobayashi, Tsutomu; Shirahashi, Koichi; Kubota, Masumitsu

1989-11-01

Precipitation behavior of elements in a high-level liquid waste (HLW) was studied by using the simulated liquid waste, when the transuranic elements group was precipitated and separated as oxalate from HLW generated from the reprocessing of spent nuclear fuel. The results showed that over 90 % of strontium and barium were precipitated when oxalic acid was directly added to HLW to precipitate the transuranic elements group, and the percentages of these elements precipitated were affected by molybdenum and or zirconium. Therefore, a method of adding oxalic acid into the filtrate was studied after removing previously molybdenum and zirconium as precipitate by denitrating HLW, and it was found that precipitated fractions of strontium and barium could be suppressed about 10 %. Adding oxalic acid under the co-existance of ascorbic acid is effective for quantitative precipitation of neptunium in HLW. In this case, it was found that adding ascorbic acid had little influence on precipitation behavior of the other elements except palladium. (author)

Report of third regular inspection of Tokai reprocessing facilities, Power Reactor and Nuclear Fuel Development Corp

International Nuclear Information System (INIS)

Anon.

1987-01-01

The reprocessing facilities passed the inspection before use on December 25, 1980, and started the full operation. Since then, this is the third regular inspection. It was begun on April 1, 1986, and finished on August 18, 1986, with the inspection of the rate of recovery of products. The reprocessing facilities which became the object of inspection were the facilities for accepting and storing spent fuel, the reprocessing facilities proper (the facilities of shearing, dissolution, separation, refining, denitration and recovery of acid and solvent), the facilities for storing products, measurement and control system, radioactive waste facilities, radiation control facilities and attached facilities (power, water, steam and testing). The main works carried out during the period of this regular inspection were the repair of an enriched uranium dissolution tank by welding, the renewal of a piping for a low activity waste liquid storing tank, and the removal of a washing tank. The total exposure dose in the first half of fiscal year 1986 was about 30.81 man-rem. (Kako, I.)

Fiscal 1995 achievement report. Development of entrained bed coal gasification power plant (Part 4 - Pilot plant operation); 1995 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 4. Pilot plant unten sosa hen

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-03-01

The 200 tons/day entrained bed coal gasification pilot plant constructed for establishing the technology of integrated coal gasification combined cycle was subjected to operational tests, and the fiscal 1995 results are compiled. In fiscal 1995, 1328 hours and 3 minutes (8 gasification operations) was recorded with gasification furnace facility, 899 hours and 53 minutes with the gas clean-up facility, 831 hours and 27 minutes with the gas turbine facility (11 startups for the generation of 6657 MWh), and 1958 hours and 2 minutes with the treatment furnace and 1331 hours and 10 minutes with the denitration unit of the safety/environment-related facility. The details of starts and stops were described in graphs which covered Runs D13, D14-1, D14-2, E1, D15, and A14. Operating procedures were studied and compiled for the plant start/stop schedule, general guidelines, gasification furnace facility, gas clean-up facility (dry type desulfurization facility), gas clean-up facility (dry type dedusting facility), gas turbine facility, real-pressure natural-size combustor test facility, and the safety/environment-related facility. (NEDO)

Fiscal 2000 report on the development of high-efficiency refuse-fueled power generation technology; Kokoritsu haikibutsu hatsuden gijutsu kaihatsu 2000 nendo hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

Efforts were made to develop a refuse gasification/fusion power generation technology to contribute to the effective utilization of unexploited energy and to reduction in greenhouse gas emissions. Developed in the technology of elevating steam temperature were the evaluation of high-temperature corrosion of SH materials and a high temperature dust removing system, dechlorination technology for the thermolysis process, and a ceramic-made high-temperature air heater. For the avoidance of exhaust gas reheating, development was carried out for a low-temperature denitration unit, stable refuse feeding system for reduction in the self-heat melting critical calorific value, waste plastic injection technology for reduction in the amount of external fuel injection, and so forth. The effect of the developed element technologies were evaluated and a detailed feasibility study was conducted for a refuse gas conversion power generation system using gas engine power generation for minor-scale general waste treatment facilities. In the survey of the trend of refuse-fueled power generation technologies, trend in Japan and advanced refuse-fueled power generation systems and their introduction in Europe and America were investigated. (NEDO)

Achievement report for fiscal 2000 on international research cooperation project. Research on technologies to decompose heavy hydrocarbon resources; 2000 nendo jushitsu tanka suiso shigen no bunkai gijutsu ni kansuru kenkyu seika hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

Researches have been made on structuring a hydrocracking process for vacuum distillation residue of Marlim crude oil produced in Brazil (Marlim-VR) being one of the heavy hydrocarbon resources. This paper summarizes the achievements in fiscal 2000. In discussing the hydrocracking conditions, discussions were given on the effects of different iron-based catalysts, reaction pressures, reaction temperatures, effects of solvent addition, and the catalyst made of natural limonite produced in Brazil. As a result, the bottom yield was reduced successfully to 5.5% VR, and conversely oil yield was enhanced up to 84.8% VR by using the limonite produced in Brazil as a catalyst, and by performing bottom recycles under the conditions of 450 degrees C and 10 MPa. In discussing the up-grading of hydrocracking produced oil, a result of 99% by weight or higher was obtained in both of denitration and desulfurization rates. In the fuel evaluation test, a product exceeding 45 being the JIS standard for light oil was derived, which was found a clean product having lower sulfur concentration than the existing heavy oil. (NEDO)

U.S. Department of Energy radioactive nitric acid shipping campaign

International Nuclear Information System (INIS)

Penn, H.R.

1996-01-01

This report is about the disposal of a large quantity of chemicals previously used in the Plutonium/Uranium Extraction Plant (PUREX). Several alternatives were considered for disposal of the over 700,000 liters of this radiologically contaminated nitric acid. These alternatives included sugar denitration, biodenitrification, calcination, chemical conversion to solid sodium nitrate or to ammonium nitrate, or decontamination and re-use. Another alternative was to solicit interest from others that might be able to utilize this material in its current condition. British Nuclear Fuels Inc., located in the United Kingdom, expressed interest in this alternative. DOE Headquarters requested Westinghouse Hanford Company (WHC) Transportation and Packaging group to investigate the feasibility of transferring the radiologically contaminated nitric acid to the United Kingdom. Shipments began in May 1995, and were monitored with DOE's satellite tracking system TRANSCOM. This shipping campaign was successfully completed, with no incidents, and savings realized for cleanup of the PUREX facility in excess of $37 million. This process will be duplicated at the Savannah River Site, with cooperation between SRS and Hanford personnel sharing lessons learned

Development of partitioning method : cold experiment with partitioning test facility in NUCEF (I)

International Nuclear Information System (INIS)

Yamaguchi, Isoo; Morita, Yasuji; Kondo, Yasuo

1996-03-01

A test facility in which about 1.85 x 10 14 Bq of high-level liquid waste can be treated has been completed in 1994 at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) for research and development of Partitioning Method. The outline of the partitioning test facility and support equipments for it which were design terms, constructions, arrangements, functions and inspections were given in JAERI-Tech 94-030. The present report describes the results of the water transfer test and partitioning tests, which are methods of precipitation by denitration, oxalate precipitation, solvent extraction, and adsorption with inorganic ion exchanger, using nitric acid to master operation method of the test facility. As often as issues related to equipments occurred during the tests, they were improved. As to issues related to processes such as being stopped up of columns, their measures of solution were found by testing in laboratories. They were reflected in operation of the Partitioning Test Facility. Their particulars and improving points were described in this report. (author)

Separation of Am-Cm from Al(NO3)3 waste solutions by in-canyon-tank precipitation as oxalates

International Nuclear Information System (INIS)

Gray, L.W.; Burney, G.A.; Wilson, T.W.; McKibben, J.M.; Bibler, N.E.; Holtzscheiter, E.W.; Campbell, T.G.

1982-04-01

A process for recovery of Am-Cm residues from high-activity waste concentrates has been developed specifically for application in Savannah River Plant (SRP) canyon tanks. The Am-Cm residues were collected from a campaign to produce plutonium containing high isotopic concentrations of 242 Pu. The separation of Am-Cm from the high-activity waste stream, containing about 2M Al(NO 3 ) 3 , is necessary to produce an acceptable feed solution for a later pressurized cation exchange chromatography separation and purification step. The new process includes formic acid denitration, adjustment of contaminating cations by evaporation and water dilution, and oxalate precipitation of the actinides and lanthanides. After washing, the precipitate was dissolved in 8M nitric acid and the oxalate was destroyed by nitric acid oxidation that was catalyzed by manganous ions. This new process generates about one-fourth the waste of the californium solvent extraction process, which it replaced. The new process also produces a cleaner feed solution for the pressurized cation exchange chromatography separation and purification step

NO-donating aspirin inhibits the growth of leukemic Jurkat cells and modulates β-catenin expression

International Nuclear Information System (INIS)

Nath, Niharika; Labaze, Georges; Rigas, Basil; Kashfi, Khosrow

2004-01-01

β-Catenin has been implicated in leukemic cell proliferation. We compared the effects of aspirin (ASA) and the ortho, meta, and para positional isomers of NO-donating aspirin (NO-ASA) on cell growth and β-catenin expression in human Jurkat T leukemic cells. Cell growth inhibition was strong: IC 50 for p-, o-, and m- were 20 ± 1.6 (mean ± SEM), 15 ± 1.5, and 200 ± 12 μM, respectively, in contrast to that of ASA (3200 ± 375 μM). The para isomer of NO-ASA degraded β-catenin in a dose- and time-dependent manner coinciding with increasing expression of activated caspase-3. The caspase inhibitor ZVAD blocked β-catenin cleavage by p-NO-ASA and partially reversed cell growth inhibition by p-NO-ASA but not that by ASA. A denitrated analog of p-NO-ASA did not degrade β-catenin indicating the importance of the NO-donating moiety. Our findings suggest that NO-ASA merits further study as an agent against leukemia

The uranium fuel cycle at IPEN - Energy and Nuclear Research Institute, SP, Brazil

International Nuclear Information System (INIS)

Abrao, Alcidio

1994-09-01

This paper summarizes the progress of research concerning the uranium fuel cycle set up at the IPEN, Sao Paulo, from the raw yellow-cake to the uranium hexafluoride. It covers the reconversion of the hexafluoride to ammonium uranyl tricarbonate and the manufacturing of the fuel elements for the swimming pool IEA-R1 reactor. This review extends the coverage of two pilot plants for uranium purification based upon ion exchange, one demonstration unity for the purification of uranyl nitrate by solvent extraction in pulsed columns, the unity of uranium tetrafluoride into moving bed reactors and a second one based upon the wet chemistry via uranium dioxide and aqueous hydrogen fluoride. The paper mentions the pilot plant for the preparation of uranium trioxide by the thermal decomposition of ammonium diuranate and a second unity by the thermal denitration of uranyl nitrate. The paper outlines the fluorine plant and the unity for the hexafluoride preparation, the unity for the conversion of the hexa to the ammonium uranyl tricarbonate and the fabrication of fuel elements for the IEA-R1 reactor. (author)

Operation databook of the fuel treatment system of the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY). JFY 2004 to JFY 2008

International Nuclear Information System (INIS)

Kokusen, Junya; Sumiya, Masato; Seki, Masakazu; Kobayashi, Fuyumi; Ishii, Junichi; Umeda, Miki

2013-02-01

Uranyl nitrate solution fuel used in the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY) is adjusted in the Fuel Treatment System, in which such parameters are varied as concentration of uranium, free nitric acid, soluble neutron poison, and so on. Operations for concentration and denitration of the solution fuel were carried out with an evaporator from JFY 2004 to JFY 2008 in order to adjust the fuel to the experimental condition of the STACY and the TRACY. In parallel, the solution fuel in which some kinds of soluble neutron poison were doped was also adjusted in JFY 2005 and JFY 2006 for the purpose of the STACY experiments to determine neutron absorption effects brought by fission products, etc. After these experiments in the STACY, a part of the solution fuel including the soluble neutron poison was purified by the solvent extraction method with mixer-settlers in JFY 2006 and JFY 2007. This report summarizes operation data of the Fuel Treatment System from JFY 2004 to JFY 2008. (author)

Application of radiation to environmental protection

International Nuclear Information System (INIS)

Tokunaga, Okihiro; Machi, Sueo

1987-01-01

A survey was made to investigate the current applications of radiations to environmental protection. Electron irradiation is used for desulfurization and denitration of flue gas. Flue gas generated from combusted fossil fuel such as coal and oil is irradiated with electrons to oxidize sulfur dioxide and nitrogen oxides, which are then removed to clean the flue gas. Ammonia or lime may be added before irradiation to convert sulfer dioxide and nitrogen oxides into ammonium sulfate and ammonium nitrate or convert sulfur dioxide into calcium sulfite and calcium sulfate. Radiations are helpful for treatment of waste water. When water is irradiated with radiations, most of the energy is absorbed by water to cause decomposition of water molecules, resulting in the production of highly reactive species such as OH radical and H atom. Contaminants then react with these species and are oxidized into carbon dioxide and water. Other methods are also available for applying radiations to water treatment. Sludge contains a large amount of organic substances that serve as fertilizer components. Thus, radiations can be used to sterilize and compost sludge. (Nogami, K.)

Chemistry of the 8-Nitroguanine DNA Lesion: Reactivity, Labelling and Repair.

Science.gov (United States)

Alexander, Katie J; McConville, Matthew; Williams, Kathryn R; Luzyanin, Konstantin V; O'Neil, Ian A; Cosstick, Richard

2018-02-26

The 8-nitroguanine lesion in DNA is increasingly associated with inflammation-related carcinogenesis, whereas the same modification on guanosine 3',5'-cyclic monophosphate generates a second messenger in NO-mediated signal transduction. Very little is known about the chemistry of 8-nitroguanine nucleotides, despite the fact that their biological effects are closely linked to their chemical properties. To this end, a selection of chemical reactions have been performed on 8-nitroguanine nucleosides and oligodeoxynucleotides. Reactions with alkylating reagents reveal how the 8-nitro substituent affects the reactivity of the purine ring, by significantly decreasing the reactivity of the N2 position, whilst the relative reactivity at N1 appears to be enhanced. Interestingly, the displacement of the nitro group with thiols results in an efficient and specific method of labelling this lesion and is demonstrated in oligodeoxynucleotides. Additionally, the repair of this lesion is also shown to be a chemically feasible reaction through a reductive denitration with a hydride source. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Protein Tyrosine Nitration : Selectivity, Physicochemical and Biological Consequences, Denitration, and Proteomics Methods for the Identification of Tyrosine-Nitrated Proteins

NARCIS (Netherlands)

Abello, Nicolas; Kerstjens, Huib A. M.; Postma, Dirkje S.; Bischoff, Rainer

Protein tyrosine nitration (PTN) is a post-translational modification occurring under the action of a nitrating agent. Tyrosine is modified in the 3-position of the phenolic ring through the addition of a nitro group (NO(2)). In the present article, we review the main nitration reactions and

Fiscal 1998 achievement report on regional consortium research and development project. Venture business fostering regional consortium in its 2nd year--Creation of key industries (Development of novel manufacturing technology capable of dealing with multiple types of environmental preservation oriented fine ceramic porous structures); 1998 nendo kankyoyo fine ceramics takotai no tahinshu taiogata shinseizo gijutsu no kaihatsu seika hokokusho. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-03-01

This effort aims to develop a technology to manufacture various types of high-temperature dust collecting porous ceramic bodies. In the development of a molding technology, guidelines regarding foam containing slurry adjustment by use of surfactant are established and, under the guidelines, an alumina body with its average porosity exceeding 80% is fabricated, with the pores structured homogenous, gradient on the surface, formed in multiple layers, and composite. As for coating, a surface reforming method is developed by which a coating that is a few tens of nanometers thick is uniformly formed on an alumina panel surface, on the exterior of a porous body, and inside a model alumina porous body. It is found that the coating enhances the anti-corrosion capability of alumina. When a titanium oxide coating with 2% silica added thereto is formed on a porous body surface by the said surface reforming method, it is found that there is a catalytic activity achieving an 80% denitrating rate at 700 degrees C. Thanks to a newly developed dust collecting performance evaluating unit, it is proved that the ceramic filter meet the purpose of a dust collector sufficiently. (NEDO)

Experimental investigation of the IFMIF target mock-up

International Nuclear Information System (INIS)

Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Yu.; Arnol'dov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.; Nakamura, H.

2009-01-01

The international fusion materials irradiation facility (IFMIF) lithium neutron target mock-ups have been constructed and tested at water and lithium test facilities in the IPPE of Russia. Jet velocity in both mock-ups was up to 20 m/s. Calculations and experiments showed lithium flow instability at conjunction point of straight and concave sections of the mock-up back wall. Water velocity profile across the mock-up width, jet thickness, and wave height were measured. The significant increase of thickness of both water and lithium jets near the mock-up sidewalls was observed. The influence of shape of the nozzle outlet part on jet stability was investigated. Lithium evaporation from the jet free surface was investigated as well as lithium deposition on vacuum pipe walls of the target mock-up. It was shown that these phenomena are not very critical for the target efficiency. The possibility of lithium denitration down to 2 ppm (at 10 ppm requested) by means of aluminium getter was shown. Two types of cold traps and plug indicators of impurities were tested. The results are presented in the paper.

Demonstration plant of smoke treatment by electron beam irradiation

International Nuclear Information System (INIS)

Kawamura, Keita

1989-01-01

The acid rain caused by sulfur oxides and nitrogen oxides has become the large social problem as it damages forests, lakes and agricultural crops and also buildings in Europe and America. In such circumstances, concern has been expressed in various countries on the smoke treatment technology, EBA process, which removes the sulfur oxides and nitrogen oxides contained in smoke simultaneously by irradiating electron beam on the smoke which is exhausted from power station boilers and industrial boilers and mainly causes acid rain. The research and development of this technology were begun in 1971 based on the original idea of Ebara Corp., and from 1972, those were advanced as the joint research with Japan Atomic Energy Research Institute. Thereafter, by the joint research with the technical research association on prevention of nitrogen oxides in iron and steel industry, by ammonia addition and irradiation process, the desulfurization and denitration performance was heightened, and the byproduct was successfully captured as powder, in this way, the continuous dry treatment process was established. The demonstration test plant was constructed in a coal-firing power station in Indiana, USA, and the trial operation was carried out from 1985 for two years. (K.I.)

The flashcal process for the fabrication of fuel-metal oxides using the whiteshell roto-spray calciner

International Nuclear Information System (INIS)

Sridhar, T.S.

1988-01-01

A one-step, continuous, thermochemical calcination process, called the FLASHCAL (Flash Calcination) process has been developed for the production of single- and mixed-oxide powders of fuel metals (uranium, thorium and plutonium) from the respective nitrate solutions using the Whiteshell Roto-Spray Calciner (RSC). The metal-nitrate feed solution, either by itself or mixed with a suitable chemical reactant or additive, is converted to its oxide powder in the RSC at temperatures between 300 and 600 0 C. Rapid denitration takes place in the calciner, yielding the metal-oxide powders while simultaneously destroying any excess chemical additive and reaction by-products. In the production of precursor oxide powders suitable for fuel fabrication, the FLASHCAL process has advantages over batch calcination and other processes that involve precipitation and filtration steps because fewer processing and handling operations are needed. Results obtained with thorium nitrate and uranium nitrate-thorium nitrate mixtures indicate that some measure of control over the size distribution and morphology of the oxide product powders is possible in this process with the proper selection of chemical additive, as well as the operating parameters of the calciner

Development of long-lived radionuclide partitioning technology

Energy Technology Data Exchange (ETDEWEB)

Yoo, Jae Hyung; Lee, Eil Hee; Kwon, Sun Gil; Kim, Heung Ho; Yang, Han Beom; Kim, Kwang Wook; Chung, Dong Yong; Lim, Jae Kwan; Kim, Young Hwan; Hwang, Doo Seong; Lee, Kue Il; Park, Won Seok; Gu, Je Hue

1997-09-01

This study has been performed with focus on the modification of the process, advancement of separation efficiency, adaptability check for another separation technology and optimization of separation conditions. The works are summarized as follows : (1) Kinetics of denitration with formic acid by mathematical modeling, (2) Evaluation of a mutual separation process for MA and RE using DEHPA, (3) Preparation of Zr-DEHPA, and its extraction and selective stripping behaviors to separate MA form RE, (4) Adsorption and elution behaviors of RE by anion-exchange chromatography. (5) Selection of reducing agent for selective separation of Pd, and its separation condition, (6) Development of a liquid-liquid extraction device composed of a highly packed fiber bundle, and its extraction behavior by experimental and theoretical analysis. In addition, characteristics of in-situ stripping and separation of metal ions by electrolysis system of glassy carbon fiber, (7) Behavior of reductive precipitation of Eu by photochemical reaction. The results will be use as basic materials for the design and installationof the engineering test process which is scheduled to conduct in phase II. (author). 13 refs., 31 tabs., 92 figs.

Fiscal 1994 achievement report. Development of entrained bed coal gasification power plant (Part 4 - Pilot plant operation); 1994 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 4. Pilot plant unten sosa hen

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

The 200 tons/day entrained bed coal gasification pilot plant constructed for the establishment of the technology of integrated coal gasification combined cycle power generation was operated for testing, and the results are put together. Operating hours recorded were 1347 hours and 7 minutes for the gasification furnace facility (7 gasification operations), 752 hours and 22 minutes for the gas clean-up facilities, 425 hours and 20 minutes for the gas turbine facility (6 startups for generating 2616.1 MWh), and 1852 hours for the treatment furnace and 1304 hours and 32 minutes for the denitration system in the safety/environment-related facility. Detailed graphs were drawn for the description of starts and stops in Run D8, Run D9 (1-3), Run D10, Run D11, and in Run D12. Operating procedures were studied and then compiled for the plant start-stop schedule, general guidelines, gasification furnace facility, gas clean-up facility (dry type desulfurization facility), gas clean-up facility (dry type dedusting facility), gas turbine facility, real-pressure natural-size combustor test facility, and for the safety/environment related facility. (NEDO)

Exhaust gas treatment by electron beam irradiation

International Nuclear Information System (INIS)

Shibamura, Yokichi; Suda, Shoichi; Kobayashi, Toshiki

1991-01-01

Among global environmental problems, atmospheric pollution has been discussed since relatively old days, and various countermeasures have been taken, but recently in connection with acid rain, the efficient and economical treatment technology is demanded. As the denitration and desulfurization technology for the exhaust gas from the combustion of fossil fuel, the incineration of city trash and internal combustion engines, three is the treatment method by electron beam irradiation. By irradiating electron beam to exhaust gas, nitrogen oxides and sulfur oxides are oxidized to nitric acid and sulfuric acid, and by promoting the neutralization of these acids with injected alkali, harmless salts are recovered. This method has the merit that nitrogen oxides and surfur oxides can be removed efficiently with a single system. In this report, as for the exhaust gas treatment by electron beam irradiation, its principle, features, and the present status of research and development are described, and in particular, the research on the recent exhaust gas treatment in city trash incineration is introduced. This treatment method is a dry process, accordingly, waste water disposal is unnecessary. The reaction products are utilized as fertilizer, and waste is not produced. (K.I.)

Method of processing plutonium and uranium solution

International Nuclear Information System (INIS)

Otsuka, Katsuyuki; Kondo, Isao; Suzuki, Toru.

1989-01-01

Solutions of plutonium nitrate solutions and uranyl nitrate recovered in the solvent extraction step in reprocessing plants and nuclear fuel production plants are applied with low temperature treatment by means of freeze-drying under vacuum into residues containing nitrates, which are denitrated under heating and calcined under reduction into powders. That is, since complicate processes of heating, concentration and dinitration conducted so far for the plutonium solution and uranyl solution are replaced with one step of freeze-drying under vacuum, the process can be simplified significantly. In addition, since the treatment is applied at low temperature, occurrence of corrosion for the material of evaporation, etc. can be prevented. Further, the number of operators can be saved by dividing the operations into recovery of solidification products, supply and sintering of the solutions and vacuum sublimation. Further, since nitrates processed at a low temperature are powderized by heating dinitration, the powderization step can be simplified. The specific surface area and the grain size distribution of the powder is made appropriate and it is possible to obtain oxide powders of physical property easily to be prepared into pellets. (N.H.)

Recovery of americium-241 from aged plutonium metal

International Nuclear Information System (INIS)

Gray, L.W.; Burney, G.A.; Reilly, T.A.; Wilson, T.W.; McKibben, J.M.

1980-12-01

About 5 kg of ingrown 241 Am was recovered from 850 kg of aged plutonium using a process developed specifically for Savannah River Plant application. The aged plutonium metal was first dissolved in sulfamic acid. Sodium nitrite was added to oxidize the plutonium to Pu(IV) and the residual sulfamate ion was oxidized to nitrogen gas and sulfate. The plutonium and americium were separated by one cycle of solvent extraction. The recovered products were subsequently purified by cation exchange chromatography, precipitated as oxalates, and calcined to the oxides. Plutonium processng was routine. Before cation exchange purification, the aqueous americium solution from solvent extraction was concentrated and stripped of nitric acid. More than 98% of the 241 Am was then recovered from the cation exchange column where it was effectively decontaminated from all major impurities except nickel and chromium. This partially purified product solution was concentrated further by evaporation and then denitrated by reaction with formic acid. Individual batches of americium oxalate were then precipitated, filtered, washed, and calcined. About 98.5% of the americium was recovered. The final product purity averaged 98% 241 AmO 2 ; residual impurities were primarily lead and nickel

Development of high-efficiency wastes-burning electric power generating technology. Volume 2. Report for fiscal 1999; Kokoritsu haikibutsu hatsuden gijutsu kaihatsu 1999 nendo hokokusho. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-03-01

In high-efficiency power generation using general wastes and combustible industrial wastes as fuel, development has been performed on a wastes gasifying and melting power generation technology. This technology is capable of suppressing generation of dioxines, recovering slag that can be utilized effectively, and reducing ash volume, by thermally decomposing the wastes and melting combustion ash at elevated temperatures by using thermally decomposed gases. With regard to the evaluation on high temperature corrosiveness of SH materials and the development of a high temperature dust removing system, a steam heater was designed, fabricated, and installed in a model plant, wherein the operation test has been performed for about 1,620 hours. For the technology of dechlorination during a thermal decomposition process, dechlorination rate of 90% was confirmed at 425 degrees C or higher in a demonstration plant. In addition, developments were made on a low temperature denitration device to avoid re-heating of waste gases, a stable wastes supply system to reduce quantity of self-heated melt limiting heat generation, and a waste plastics blowing technology to reduce external fuel charge quantity. Furthermore, a survey was carried out on the trends in wastes electric power generation technologies. (NEDO)

The use dynamic avalanching and fractal analysis to characterise uranium oxide powders

International Nuclear Information System (INIS)

Hobbs, J.W.; Rhodes, D.

2000-01-01

Direct thermal denitration is an attractive method of co-converting mixed-metal nitrate solutions of plutonium and uranium into oxide because of its apparent simplicity. Such benefits are often marred by the relatively poor powder quality and handling characteristics, which can be overcome by modifications to the process chemistry. To ensure that powder synthesis routes under assessment require the minimal further processing it is necessary to be able to characterise the powder fully in term of the key fundamental properties. This paper will demonstrate the use of a dynamic avalanching technique, fractal analysis and morphology to assess processing behaviour. The use of dynamic avalanching to uniquely characterise the chaotic flow properties of urania powders has proved successful and results have shown that this technique is capable of detecting small differences in processing behaviour due changes in morphologies and particle size distribution. This technique has promise for being able to provide nearly instantaneous feedback to the powder generation process being monitored (e.g. calcination, milling, mixing). The use of fractals to describe powders is an interesting characterisation tool when combined with morphological shape factors and the flow index. (authors)

Solidification of TRU wastes in a ceramic matrix

International Nuclear Information System (INIS)

Loida, A.; Schubert, G.

1991-01-01

Aluminumsilicate based ceramic materials have been evaluated as an alternative waste form for the incorporation of TRU wastes. These waste forms are free of water and - cannot generate hydrogen radiolyticly, - they show good compatibility between the compounds of the waste and the matrix, - they are resistent against aqueous solutions, heat and radiation. R and D-work has been performed to demonstrate the suitability of this waste form for the immobilization of TRU-wastes. Four kinds of original TRU-waste streams and a mixture of all of them have been immobilized by ceramization, using glove box and remote operation technique as well. Clay minerals, (kaolinite, bentonite) and reactive corundum were selected as ceramic raw materials (KAB 78) in an appropriate ratio yielding 78 wt% Al 2 O 3 and 22 wt%SiO 2 . The main process steps are (i) pretreatment of the liquid waste (concentration, denitration, neutralization, solid- liquid separation), (ii) mixing with ceramic raw materials and forming, (iii) heat treatment with T max. of 1300 0 C for 15 minutes. The waste load of the ceramic matrix has been increased gradually from 20 to 50, in some cases to 60 wt.%

Laboratory-scale vitrification and leaching of Hanford high-level waste for the purpose of simulant and glass property models validation

International Nuclear Information System (INIS)

Morrey, E.V.; Elliott, M.L.; Tingey, J.M.

1993-02-01

The Hanford Waste Vitrification Plant (HWVP) is being built to process the high-level and TRU waste into canistered glass logs for disposal in a national repository. Testing programs have been established within the Project to verify process technology using simulated waste. A parallel testing program with actual radioactive waste is being performed to confirm the validity of using simulates and glass property models for waste form qualification and process testing. The first feed type to be processed by and the first to be tested on a laboratory-scale is pretreated neutralized current acid waste (NCAW). The NCAW is a neutralized high-level waste stream generated from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant at Hanford. As part of the fuel reprocessing, the high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). Sodium hydroxide and sodium nitrite were added to the CAW to minimize corrosion in the tanks, thus yielding neutralized CAW. The NCAW contains small amounts of plutonium, fission products from the irradiated fuel, stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. This paper will discuss the results and status of the laboratory-scale radioactive testing

Chemical engineering problems of radioactive waste fixation by vitrification

International Nuclear Information System (INIS)

Taylor, R.F.

1985-01-01

Basic features are reviewed of the chemical engineering problems faced in the vitrification of the high-level radioactive liquid wastes resulting from the reprocessing of nuclear fuel. After an outline of glass solution properties and formation kinetics the constituent elements of the vitrification route are examined in turn: waste feed evaporation and denitration, calcination, offgas treatment, and finally melting and product quality. Plant and experimental data for each stage are discussed with comparison between process routes and with reference to the underlying principles. Attention is drawn to the future need for higher trapping efficiencies and for dealing with a wider range of species in offgas treatments as higher burnup fuels are processed after shorter cooling times from reactor. Two areas of present study where deeper insight into underlying process mechanics is needed are, firstly, the association of waste material with glass formers in the wet or sinter stages and secondly their incorporation and mixing reaction in the melt. Fuller understanding here would bring direct benefit to process performance and handling. The problems discussed are not of a nature to jeopardize the vitrification routes but if product quality does come to rely heavily on process control then demonstrable confidence in the behaviour of the central physico-chemical interactions is indispensable. (author)

Synthesis, processing behavior, and characterization of bismuth superconductors using freeze dried nitrate precursors

International Nuclear Information System (INIS)

Coppa, N.V.; Hults, W.L.; Smith, J.L.; Brynestad, J.

1994-01-01

The synthesis of Bi 2-x Pb x Sr 2 Ca 2 Cu 3 O 10 (Bi2223) powders from a freeze dried nitrate precursor is reported here. We examine the composition and morphology of the precursor material, describe the chemistry and kinetics of product formation, and evaluate the phase composition and superconducting properties of the products. A nitrate solution containing the appropriate ratio of cations was rapidly frozen and then freeze dried at low temperatures to form an atomic mixture of the component salts. The thermal processing of the freeze dried material consisted of three steps: (i) dehydration, (ii) denitration, and (iii) solid state reaction to form the Bi2223 superconducting product. Calcium substituted bismuthates and strontium-substituted calcium cuprate, not Bi2201, are the intermediates between the nitrates and the superconducting products. These highly disordered phases rapidly transform into Bi2212 or Bi2223 at higher temperatures (>790 degree C). The kinetics of product formation was studied using XRD analysis and magnetic susceptibility. The kinetics were shown to follow the nucleation and growth mechanism. Bi2223 formed after only 30 min at a few degrees below the melting point, and after 37 h Bi2223 products exhibited excellent phase composition and magnetic susceptibility characteristics

Fiscal 1995 achievement report. Development of entrained bed coal gasification power plant (Part 3 - Pilot plant operational test - 2/2); 1995 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 3. Pilot plant unten shiken hen (2/2)

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-03-01

The 200 tons/day entrained bed coal gasification pilot plant constructed for establishing the technology of integrated coal gasification combined cycle was subjected to operational tests, and the fiscal 1995 results are detailed. During Runs D13, D14, E1, D15, and A14 in the operational test of the gas clean-up facility (dry type dedusting facility), 10 troubles occurred, including damage of the separator screen, leak in the seal valve, and leak of the expansion gas, and measures were taken to deal with each of the troubles. The results of the gas turbine facility operational test were satisfactory, without any trouble worth discussion. In the operational test of the safety/environment-related facility, it was found that the produced gas was stably incinerated and that denitration performance during gas turbine operation roughly achieved the intended level. In the operational test of electric and control facilities, an overall test was conducted, inspection was made of the indoor switching facility, etc., and 13 improvements were made, which included the alteration of the high ANN setting in the water tank for slag, the alteration of the mill exit temperature setting for enabling the use of Taiheiyo coal, and proper methods for carrying out high-load operation. (NEDO)

Studies on the separation and purification of strontium from the highly radioactive waste flow of fuel element reprocessing

International Nuclear Information System (INIS)

Tuerker, A.

1989-10-01

The quantity of spent fuel elements is increasing due to the extended peaceful uses of nuclear energy. Of the numerous fission products, strontium has an interesting application potential as a radiation or energy source for utilization in various industrial fields. A necessary condition for its use is its isolation in high radiochemical purity from the highly radioactive waste flow of the Purex process. In the present study, precipitation and coprecipitation reactions, and in particular ion exchange reactions, were chosen from among the various possible chemical and physical separation methods and examined with respect to their suitability for a selective separation of strontium from the other fission products. In selecting separation materials, particularly with respect to radiation resistance, thermal stability and selectivity, polyantimonic acid proved to be the best absorbent (even in a very acid medium) for strontium. Furthermore, the behaviour of the most important radionuclides was studied with respect to the denitration reaction from a 5 molar nitric acid solution. On the basis of the high demands made on the purity of the product, a method was developed by combining lead sulphate carrier precipitation with an ion exchange reaction on polyantimonic acid and is shown in a flow chart. (orig.) [de

Density of simulated americium/curium melter feed solution

International Nuclear Information System (INIS)

Rudisill, T.S.

1997-01-01

Vitrification will be used to stabilize an americium/curium (Am/Cm) solution presently stored in F-Canyon for eventual transport to Oak Ridge National Laboratory and use in heavy isotope production programs. Prior to vitrification, a series of in-tank oxalate precipitation and nitric/oxalic acid washes will be used to separate these elements and lanthanide fission products from the bulk of the uranium and metal impurities present in the solution. Following nitric acid dissolution and oxalate destruction, the solution will be denitrated and evaporated to a dissolved solids concentration of approximately 100 g/l (on an oxide basis). During the Am/Cm vitrification, an airlift will be used to supply the concentrated feed solution to a constant head tank which drains through a filter and an in-line orifice to the melter. Since the delivery system is sensitive to the physical properties of the feed, a simulated solution was prepared and used to measure the density as a function of temperature between 20 to 70 degrees C. The measured density decreased linearly at a rate of 0.0007 g/cm3/degree C from an average value of 1.2326 g/cm 3 at 20 degrees C to an average value of 1.1973g/cm 3 at 70 degrees C

Report of short term research group on environment safety in nuclear fuel cycle, 1983

International Nuclear Information System (INIS)

1984-01-01

The research group on environment safety in nuclear fuel cycle was organized in fiscal 1979 as the research group in the range of the common utilization of Yayoi, and this is the third year since it developed into the short term research group in the Nuclear Engineering Research Laboratory. The results obtained so far were summarized in three reports, UTNL-R110, 134 and 147. In this fiscal year, ''The chemistry of reprocessing'' is the subtheme, and this short term research is to be carried out. The meeting is held on March 23 and 24, 1984, in this Laboratory, and the following reports are presented. The conference on institutional stability and the disposal of nuclear and chemically toxic wastes held at MIT, the social scientific analysis of nuclear power development, the present status of reprocessing research in foreign countries, the problems based on the operation experience of actual plants, the chemistry of fuel dissolution, the chemistry of solvent extraction, reprocessing offgas treatment and problems, the chemistry of fixing Kr and I in zeolite, waste treatment in the Tokai Reprocessing Plant of Power Reactor and Nuclear Fuel Development Corp., the chemistry of actinoids, denitration process and the chemistry of MOX production, and future reprocessing research. (Kako, I.)

Density of simulated americium/curium melter feed solution

Energy Technology Data Exchange (ETDEWEB)

Rudisill, T.S.

1997-09-22

Vitrification will be used to stabilize an americium/curium (Am/Cm) solution presently stored in F-Canyon for eventual transport to Oak Ridge National Laboratory and use in heavy isotope production programs. Prior to vitrification, a series of in-tank oxalate precipitation and nitric/oxalic acid washes will be used to separate these elements and lanthanide fission products from the bulk of the uranium and metal impurities present in the solution. Following nitric acid dissolution and oxalate destruction, the solution will be denitrated and evaporated to a dissolved solids concentration of approximately 100 g/l (on an oxide basis). During the Am/Cm vitrification, an airlift will be used to supply the concentrated feed solution to a constant head tank which drains through a filter and an in-line orifice to the melter. Since the delivery system is sensitive to the physical properties of the feed, a simulated solution was prepared and used to measure the density as a function of temperature between 20 to 70{degrees} C. The measured density decreased linearly at a rate of 0.0007 g/cm3/{degree} C from an average value of 1.2326 g/cm{sup 3} at 20{degrees} C to an average value of 1.1973g/cm{sup 3} at 70{degrees} C.

Analysis of lagoon sludge characteristics for choice of treatment process

Energy Technology Data Exchange (ETDEWEB)

Park, J. H.; Hwang, D. S.; Choi, Y. D.; Lee, K. I.; Hwang, S. T.; Jung, K. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

The Korea Atomic Energy Research Institute has launched a decommissioning program of uranium conversion plant. One of the important tasks in the decommissioning program is the treatment of the sludge, which was generated during operation and stored in the two ponds of the lagoon. The treatment requires the volume reduction of lagoon sludges for the low cost of the program and the conversion of the chemical forms, including uranium, for the acceptance at the final disposal site. The physical properties, such as densities, were measured and chemical compositions and radiological properties were analyzed. The denitration was a candidate process which would satisfy the requirements for sludge treatment, and the characteristics of thermal decomposition and dissolution with water were analyzed. The main compounds of the sludge were ammonium and sodium nitrate from conversion plant and calcium nitrate, calcium carbonate from Ca precipitation and impurities of the yellow cake. The content of uranium, thorium and Ra-226 was high in pond-1 and low in pond-2 because those were removed during Ca precipitation. On the base of the characteristics of the sludge and available technologies, reviewed in this study and being developed in Korea Atomic Energy Research Institute, two processes were proposed and evaluated in points of the expected technological difficulties. And the cost for treatment of sludges are estimated for both processes. 79 refs., 44 figs., 37 tabs. (Author)

Treatment of off-gas evolved from thermal decomposition of sludge waste

International Nuclear Information System (INIS)

Doo-Seong Hwang; Yun-Dong Choi; Gyeong-Hwan Jeong; Jei-Kwon Moon

2013-01-01

Korea Atomic Energy Research Institute (KAERI) started a decommissioning program of a uranium conversion plant. The treatment of the sludge waste, which was generated during the operation of the plant, is one of the most important tasks in the decommissioning program of the plant. The major compounds of sludge waste are nitrate salts and uranium. The sludge waste is denitrated by thermal decomposition. The treatment of off-gas evolved from the thermal decomposition of nitrate salts in the sludge waste is investigated. The nitrate salts in the sludge were decomposed in two steps: the first decomposition is due to the ammonium nitrate, and the second is due to the sodium and calcium nitrate and calcium carbonate. The components of off-gas from the decomposition of ammonium nitrate at low temperature are NH 3 , N 2 O, NO 2 , and NO. In addition, the components from the decomposition of sodium and calcium nitrate at high temperature are NO 2 and NO. Off-gas from the thermal decomposition is treated by the catalytic oxidation of ammonia and selective catalytic reduction (SCR). Ammonia is converted into nitrogen oxides through the oxidation catalyst and all nitrogen oxides are removed by SCR treatment besides nitrous oxide, which is greenhouse gas. An additional process is needed to remove nitrous oxide, and the feeding rate of ammonia in SCR should be controlled properly for evolved nitrogen oxides. (author)

Preliminary investigation of topical nitroglycerin formulations containing natural wound healing agent in diabetes-induced foot ulcer.

Science.gov (United States)

Hotkar, Mukesh S; Avachat, Amelia M; Bhosale, Sagar S; Oswal, Yogesh M

2015-04-01

Nitroglycerin (NTG) is an organic nitrate rapidly denitrated by enzymes to release free radical nitric oxide and shows improved wound healing and tissue protection from oxidative damage. The purpose of this study was to evaluate whether topical application of NTG in the form of gel/ointment along with a natural wound healing agent, aloe vera, would bring about wound healing by using diabetes-induced foot ulcer model and rat excision wound model. All these formulations were evaluated for pH, viscosity, drug content and ex vivo diffusion studies using rat skin. Based on ex vivo permeation studies, the formulation consisting of carbopol 974p as a gelling agent and aloe vera was found to be suitable. The in vivo study used streptozotocin-induced diabetic foot ulcer and rat excision wound models to analyse wound healing activity. The wound size in animals of all treated groups was significantly reduced compared with that of the diabetic control and marketed treated animals. This study showed that the gel formed with carbopol 974p (1%) and aloe vera promotes significant wound healing and closure in diabetic rats compared with the commercial product and provides a promising product to be used in diabetes-induced foot ulcer. © 2013 The Authors. International Wound Journal © 2013 Medicalhelplines.com Inc and John Wiley & Sons Ltd.

Preliminary engineering and economic analysis of the fixation of high-level radioactive wastes in concrete

International Nuclear Information System (INIS)

Weeren, H.O.; Perona, J.J.

1979-07-01

This feasibility study was based on a waste fixation facility that would serve a reprocessing plant with a capacity of 5 metric tons of uranium per day (MTU/day). Postirradiation cooling times of 3 to 10 years prior to waste solidification were assumed. The waste solution would be concentrated, denitrated, mixed with cement, and cast under pressure in cylindrical canisters similar to those envisioned for a glass facility. The solidified waste grout would be vented, to allow the free water to escape, and then sealed. The filled canisters would be shipped to a geologic repository for permanent storage. Recent work with concretes formed under elevated temperatures and pressures (FUETAP) indicates that they are highly leach resistant. The operating costs were estimated for a waste fixation facility under several conditions. Operating costs for a glass fixation facility were also estimated and compared with the operating costs for a concrete fixation facility. The principal conclusion is that concrete could be an alternative to glass as a matrix for fixation of wastes with high heat-generation rates. The operating costs of an optimized concrete fixation process would probably not be greatly higher than the operating costs of a glass plant, and the capital costs would almost surely be lower. In addition, the concrete process is not a high-temperature process and would not have the consequent operating problems

Features of hydrotreating catalytic cracking feed and heavy slow coking gas oils

Energy Technology Data Exchange (ETDEWEB)

Yefremov, N.I.; Kushnarev, D.F.; Frolov, P.A.; Chagovets, A.N.; Kalabin, G.A.

1993-12-31

A possible means of more extensive processing of crude oil is the use, in catalytic cracking, of heavy coking gas oils (HCGOs), a feature of which is a higher content of polycyclic aromatic compounds and resins by comparison with straight-run vacuum distillates. The presence of these compounds in catalytic cracking feed causes a reduction in the product yield and increased coke formation. Therefore, one of the problems of hydrotreating feedstock of this kind is the hydrogenation of polycyclic arenes. Processes of extensive desulphurization and denitration occur in parallel, since the sulphur and nitrogen compounds of HCGO are chiefly condensed benzoderivatives of thiophene, pyridine and carbazole, and largely concentrated in heavy aromatic and resinous fractions. The composition of the saturated part of the cracking feed plays a large role in achieving the optimum yields of gaseous and gasoline fractions. Thus an increase in the proportion of cyclanes in the feed raises the gasoline yield. In this way, an investigation of the hydrocarbon conversions during the hydrotreatment of cracking feed is of great importance. The present paper sets out the results for studying the change in the group-structural characteristics of the hydrogenation products of a mixture containing 30% HCGOs according to data of {sup 1}H and {sup 13}C NMR spectroscopy. 7 refs., 7 figs., 1 tab.

Development of recycling processes for clean rejected MOX fuel pellets

International Nuclear Information System (INIS)

Khot, P.M.; Singh, G.; Shelke, B.K.; Surendra, B.; Yadav, M.K.; Mishra, A.K.; Afzal, Mohd.; Panakkal, J.P.

2014-01-01

Highlights: • Dry and wet (MWDD) methods were developed for 100% recycling of CRO (0.4–44% PuO 2 ). • Dry method showed higher productivity and comparable powder/product characteristics. • MWDD batches demonstrated improved powder/product characteristics to that of virgin. • Second/multiple recycling is possible with MWDD with better powder/product characteristics. • MWDD batches prepared by little milling showed better macroscopic homogeneity to that of virgin. - Abstract: The dry and wet recycling processes have been developed for 100% recycling of Clean Reject Oxide (CRO) generated during the fabrication of MOX fuel, as CRO contains significant amount of plutonium. Plutonium being strategic material need to be circumvented from its proliferation issues related to its storage for long period. It was difficult to recycle CRO containing higher Pu content even with multiple oxidation and reduction steps. The mechanical recycling comprising of jaw crushing and sieving has been coupled with thermal pulverization for recycling CRO with higher Pu content in dry recycling technique. In wet recycling, MicroWave Direct Denitration (MWDD) technique has been developed for 100% recycling of CRO. The powder prepared by dry and wet (MWDD) recycling techniques was characterized by XRD and BET techniques and their effects on the pellets were evaluated. (U,21%Pu)O 2 pellets fabricated from virgin powder and MWDD were characterized using optical microscopy and α-autoradiography and the results obtained were compared

Investigation into rationalization of low decontamination pellet fuel fabrication plant configuration

International Nuclear Information System (INIS)

Maekawa, Kazuhiko; Yoshimura, Tadahiro; Hoshino, Yasushi; Munekata, Hideki; Tamaki, Yoshihisa

2005-02-01

In feasibility studies on commercialized FBR cycle system, a comprehensive system investigation and properties evaluation for candidate FBR cycle systems has been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation into rationalization of low decontamination pellet fuel fabrication plant configuration was carried out. Until last fiscal year, conceptual design studies of the fuel fabrication plant in 200t-HM/y scale were conducted, and system properties data concerning economics and environmental burden reduction of fuel fabrication plant was acquired. In addition to this, 50t-HM/y scale plant was also schematically studied. In this fiscal year, a rationalization study on conceptual design of 50t-HM/y scale plant was conducted with main aim of economic improvement, and the 200t-HM/y scale plant design was revised based on the recent R and D progress. The system properties data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In both case of the 50t-HM/y and 200t-HM/y scale plant, it was suggested that the equipment costs were reduced in several percentages because of reduction of maintenance equipments and cut in line number at the pellet fabrication process although granulation process fro denitration converted powder and O/M control process for pellets were added. System properties data for comparative evaluation of candidate fuel fabrication systems was also prepared. (author)

Alternate fuel cycle technologies, nitrate-to-oxide conversion project. Progress report, January--June 1977

International Nuclear Information System (INIS)

Lehmkuhl, G.D.

1977-01-01

Work is being done at the Rocky Flats Plant (RFP) under contract from the Savannah River Operations Office (SROO) of the U.S. Energy Research and Development Administration (ERDA) to critically analyze and evaluate existing technology for converting plutonium nitrate to plutonium oxide, and to recommend flow sheets and equipment for this process. Seven such processes were compared using an expanded process-comparison scheme. The results of the comparison differed somewhat from the initial comparison made in September, 1976. The direct calcination methods, headed by the screw calciner process, received the highest ratings when operating experience was considered with a small weighting factor. These methods are much simpler than the others. The oxalate precipitation methods, headed by the plutonium(IV) oxalate precipitation and calcination process, received highest ratings when operating experience was strongly considered. Thus, in the long term, the screw calciner or other direct-conversion methods should be developed. For a plant to be built in the short term, however, an oxalate precipitation method should be used since a larger amount of experience exists with these processes. The block flow diagrams, material balances, and equipment flow sheets for each of the seven processes compared are included in this report. A process-design criterion is being prepared for a mechanical (screw calciner) direct-denitration process, and includes process flow sheets, a material balance, a process description, equipment performance specifications, the control philosophy and specifications, the operating philosophy, and a general process layout

Review of experience gained in fabricating nuclear grade uranium and thorium compounds and their analytical quality control at the Instituto de Energia Atomica, Sao Paulo, Brazil

International Nuclear Information System (INIS)

Abrao, A.; Franca, J.M. Jr.; Ikuta, A.; Pueschel, C.R.; Federgruen, L.; Lordello, A.R.; Tomida, E.K.; Moraes, S.; Brito, J. de; Gomes, R.P.; Araujo, J.A.; Floh, B.; Matsuda, H.T.

1977-01-01

This paper summarizes the main activities dealing with the fabrication of nuclear grade uranium and thorium compounds at the Instituto de Energia Atomica, Sao Paulo. Identification of problems and their resolutions, the experience gained in plant operation, the performance characteristics of an ion-exchange facility and a solvent extraction unit (a demonstration plant based on pulsed columns for purification of uranium and production of ammonium diuranate) are described. A moving-bed facility for UF 4 preparation and its operation is discussed. A pilot plant for uranium and thorium oxide microsphere preparation based on internal gelation for HTGR fuel type is also described. A solvent extraction pilot plant for thorium purification based on a compound extraction-scrubbing column and a mixer-settler battery and the involved technology for thorium purification are commented. The main products, namely ammonium diuranate, uranyl amonium tricarbonate, uranium trioxide, uranium tetrafluoride, thorium nitrate and thorium oxalate and their quality are commented. The development of necessary analytical procedures for the quality control of the mentioned nuclear grade products is summarized. A great majority of such procedures was particularly suitable for analyzing traces impurities. Designed for installation are the units for denitration of uranyl nitrate solutions and pilot plants for elemental fluorine and UF 6 . The installation of a laboratory-scale plant designed for reprocessing irradiated uranium and an experimental unit for the recovery of protactinium from irradiated thorium is in progress

A low-temperature process for the denitration of Hanford single-shell tank, nitrate-based waste utilizing the nitrate to ammonia and ceramic (NAC) process

International Nuclear Information System (INIS)

Mattus, A.J.; Lee, D.D.; Dillow, T.A.; Farr, L.L.; Loghry, S.L.; Pitt, W.W.; Gibson, M.R.

1994-12-01

Bench-top feasibility studies with Hanford single-shell tank (SST) simulants, using a new, low-temperature (50 to 60C) process for converting nitrate to ammonia and ceramic (NAC), have conclusively shown that between 85 to 99% of the nitrate can be readily converted. In this process, aluminum powders or shot can be used to convert alkaline, nitrate-based supernate to ammonia and an aluminum oxide-sodium aluminate-based solid which might function as its own waste form. The process may actually be able to utilize already contaminated aluminum scrap metal from various DOE sites to effect the conversion. The final, nearly nitrate-free ceramic-like product can be pressed and sintered like other ceramics. Based upon the starting volumes of 6.2 and 3.1 M sodium nitrate solution, volume reductions of 50 to 55% were obtained for the waste form produced, compared to an expected 35 to 50% volume increase if the Hanford supernate were grouted. Engineering data extracted from bench-top studies indicate that the process will be very economical to operate, and data were used to cost a batch, 1,200-kg NO 3 /h plant for working off Hanford SST waste over 20 years. Their total process cost analysis presented in the appendix, indicates that between $2.01 to 2.66 per kilogram of nitrate converted will be required. Additionally, data on the fate of select radioelements present in solution are presented in this report as well as kinetic, operational, and control data for a number of experiments. Additionally, if the ceramic product functions as its own waste form, it too will offer other cost savings associated with having a smaller volume of waste form as well as eliminating other process steps such as grouting

Usefulness of myocardial scintigraphy using cigarette smoking and isosorbide dinitrate in patients with angina pectoris

International Nuclear Information System (INIS)

Igarashi, Takeki; Horimoto, Masashi; Funayama, Naoki

1989-01-01

Cigarette smoking is not only one of the most important risk factors for coronary artery disease, but also anginal attacks often occur during smoking. Coronary vasoconstriction is considered to be one of the mechanisms which cause anginal episodes. Thirty patients with angina pectoris, 27 men and 3 women, were investigated. Group I comprised 22 patients with rest angina and group II consisted of 8 patients with effort angina. Five minutes after smoking one cigarette, a first study of 20l Tl myocardial SPECT was performed. Consequently, after sublingual administration of isosorbide denitrate, a second SPECT was done. Most cases improved and it was recognized as positive when changes of localized perfusion defects were observed between both SPECTs. In the group I, 19 out of 22 patients (86%) were positive, but in the group II, 2 out of 8 (25%) were positive. The differences were statistically significant (p<0.05). By the examination of coronarty angiography in 20 patients, the result of myocardial scintigraphy using cigarette smoking and isosorbide dinitrate seemed inversely proportional to the severity of narrowing of the coronary artery. In addition, background factors including risk factors were also compared for both groups. Because the result of this myocardial scintigraphy was positive in the rest angina group and in the group with minimal coronary arteriogrpahic abnormalities, it seggests that myocardial scintigraphy using cigarette smoking and isosorbide dinitrate is useful as a non-invasive auxiliary diagnostic method to detect coronary spasm or coronary vasoconstriction-prone angina pectoris including silent myocardial ischemia. (author)

Development of holdup monitor system (HMOS) during facility maintenance

International Nuclear Information System (INIS)

Nakamura, Hironobu; Hosoma, Takashi; Tanaka, Izumi

1999-01-01

Holdup MOnitor System (HMOS) was developed for the purpose of verifying the constant holdup during facility maintenance in Plutonium Conversion Development Facility (PCDF). The glove box assay system (GBAS; big slab) has been used by inspectors, measures the holdup periodically (i.e. IIV) using coincidence counting. The GBAS couldn't be used for inspection during maintenance period. Because many glove boxes (GB) set in process area had been occupied by large vinyl green-houses due to maintenance. We aimed that the holdup except for the maintenance GB should be constant during maintenance period, the HMOSs were set to 3 GBs. The system had been used from June '98 to July '99 for verification. The HMOS detector is located top/bottom of the GB, counts total neutron variation in the GB continuously. Detector efficiencies are 1.2%(top) and 0.12%(bottom), respectively. The measurement variation is observed up to 1.5%(3σ). The HMOS has high sensitivity 8 to 90g Pu (3σ; In case of 1kg Pu holdup, the sensitivity depends on position in the GB). The movement of equipment or nuclear material from/in the GB can be detected effectively. Though the HMOS observes measurement variation related to humidity in the GB, hygroscopic effect of denitration MOX powder, material/equipment movement and mainly 241 Pu nuclear decay, this system can verify that the holdup is constant qualitatively. As a result, in PCDF, safeguard related to the inventory verification during maintenance period (more than 1 year) were successfully performed using holdup monitor system. (author)

Simultaneous estimation of a binary mixture of a weak acid and a strong acid by volumetric titration and pH measurement

International Nuclear Information System (INIS)

Karmakar, Sanat; Mallika, C.; Kamachi Mudali, U.

2012-01-01

High level liquid waste (HLLW) generated in the aqueous reprocessing of spent nuclear fuels for the separation of uranium and plutonium by PUREX process, comprises the fission and corrosion products in 4 M nitric acid. Reduction in waste volume is accomplished by destroying the acidity of the waste solution from 4 to less than 2 M by treating it with formaldehyde and subsequent concentration by evaporation. In the denitration by HCHO, nitric acid in the waste solution is reduced to NOx and water via nitrous acid as the intermediate product: whereas formaldehyde is oxidized to formic acid which is converted to CO 2 and H 2 O subsequently. The reaction is highly exothermic and the release of all gaseous products may lead to uncontrollable process conditions. Hence, for the safe operation, it is desirable to estimate the concentration of residual formic acid as well as nitric acid in the product stream as a function of time. The acidity in the feed solution is 4 M and the concentration of HNO 3 in the product solution is in the range 1- 4 M. Since the formic acid generated during the reaction will be consumed immediately, the concentration of residual acid will be in the range 0.05-0.5 M. A simultaneous titration method based on pH measurement and volumetric analysis has been developed in the present work for the quantitative determination of the weak acid (HCOOH)with known pKa value and the strong acid (HNO 3 ) in the binary mixture

Physical, Chemical and Structural Evolution of Zeolite - Containing Waste Forms Produced from Metakaolinite and Calcined HLW

International Nuclear Information System (INIS)

Grutzeck, Michael

2005-01-01

During the seventh year of the current grant (DE-FG02-05ER63966) we completed an exhaustive study of cold calcination and began work on the development of tank fill materials to fill empty tanks and control residuals. Cold calcination of low and high NOx low activity waste (LAW) SRS Tank 44 and Hanford AN-107 simulants, respectively with metallic Al + Si powders was evaluated. It was found that a combination of Al and Si powders could be used as reducing agents to reduce the nitrate and nitrite content of both low and high NOx LAW to low enough levels to allow the LAW to be solidified directly by mixing it with metakaolin and allowing it to cure at 90 C. During room temperature reactions, NOx was reduced and nitrogen was emitted as N2 or NH3. This was an important finding because now one can pretreat LAW at ambient temperatures which provides a low-temperature alternative to thermal calcination. The significant advantage of using Al and Si metals for denitration/denitrition of the LAW is the fact that the supernate could potentially be treated in situ in the waste tanks themselves. Tank fill materials based upon a hydroceramic binder have been formulated from mixtures of metakaolinite, Class F fly ash and Class C flue gas desulphurization (FGD) ash mixed with various concentrations of NaOH solution. These harden over a period of hours or days depending on composition. A systematic study of properties of the tank fill materials (leachability) and ability to adsorb and hold residuals is under way

Nitroreductase catalyzed biotransformation of CL-20

International Nuclear Information System (INIS)

Bhushan, Bharat; Halasz, Annamaria; Hawari, Jalal

2004-01-01

Previously, we reported that a salicylate 1-monooxygenase from Pseudomonas sp. ATCC 29352 biotransformed CL-20 (2,4,6,8,10,12-hexanitro-2,4,6,8,10,12-hexaaza-isowurtzitane) (C 6 H 6 N 12 O 12 ) and produced a key metabolite with mol. wt. 346Da corresponding to an empirical formula of C 6 H 6 N 10 O 8 which spontaneously decomposed in aqueous medium to produce N 2 O, NH4+, and HCOOH [Appl. Environ. Microbiol. (2004)]. In the present study, we found that nitroreductase from Escherichia coli catalyzed a one-electron transfer to CL-20 to form a radical anion (CL-20 - ) which upon initial N-denitration also produced metabolite C 6 H 6 N 10 O 8 . The latter was tentatively identified as 1,4,5,8-tetranitro-1,3a,4,4a,5,7a,8,8a-octahydro-diimidazo[4,5-b:4',5'-e] pyrazine [IUPAC] which decomposed spontaneously in water to produce glyoxal (OHCCHO) and formic acid (HCOOH). The rates of CL-20 biotransformation under anaerobic and aerobic conditions were 3.4+/-0.2 and 0.25+/-0.01nmolmin -1 mg of protein -1 , respectively. The product stoichiometry showed that each reacted CL-20 molecule produced about 1.8 nitrite ions, 3.3 molecules of nitrous oxide, 1.6 molecules of formic acid, 1.0 molecule of glyoxal, and 1.3 ammonium ions. Carbon and nitrogen products gave mass-balances of 60% and 81%, respectively. A comparative study between native-, deflavo-, and reconstituted-nitroreductase showed that FMN-site was possibly involved in the biotransformation of CL-20

Outline of facility for studying high level radioactive materials (CPF) and study programmes

International Nuclear Information System (INIS)

Sakamoto, Motoi

1983-01-01

The Chemical Processing Facility for studying high level radioactive materials in Tokai Works of Power Reactor and Nuclear Fuel Development Corp. is a facility for fundamental studies centering around hot cells, necessary for the development of fuel recycle techniques for fast breeder reactors, an important point of nuclear fuel cycle, and of the techniques for processing and disposing high level radioactive liquid wastes. The operation of the facility was started in 1982, for both the system A (the test of fuel recycle for fast breeder reactors) and the system B (the test of vitrification of high level liquid wastes). In this report, the outline of the facility, the contents of testings and the reflection of the results are described. For the fuel recycle test, the hot test of the spent fuel pins of JOYO MK-1 core was started, and now the uranium and plutonium extraction test is underway. The scheduled tests are fuel solubility, the confirmation of residual properties in fuel melting, the confirmation of extracting conditions, the electrolytic reduction of plutonium, off-gas behaviour and the test of material reliability. For the test of vitrification of high level liquid wastes, the fundamental test on the solidifying techniques for the actual high level wastes eluted from the Tokai reprocessing plant has been started, and the following tests are programmed: Assessment of the properties of actual liquid wastes, denitration and concentration test, vitrification test, off-gas treatment test, the test of evaluating solidified wastes, and the test of storing solidified wastes. These test results are programmed to be reflected to the safety deliberation and the demonstration operation of a vitrification pilot plant. (Wakatsuki, Y.)

Nuclear fuel reprocessing deactivation plan for the Idaho Chemical Processing Plant, Revision 1

International Nuclear Information System (INIS)

Patterson, M.W.

1994-10-01

The decision was announced on April 28, 1992 to cease all United States Department of Energy (DOE) reprocessing of nuclear fuels. This decision leads to the deactivation of all fuels dissolution, solvent extraction, krypton gas recovery operations, and product denitration at the Idaho Chemical Processing Plant (ICPP). The reprocessing facilities will be converted to a safe and stable shutdown condition awaiting future alternate uses or decontamination and decommissioning (D ampersand D). This ICPP Deactivation Plan includes the scope of work, schedule, costs, and associated staffing levels necessary to achieve a safe and orderly deactivation of reprocessing activities and the Waste Calcining Facility (WCF). Deactivation activities primarily involve shutdown of operating systems and buildings, fissile and hazardous material removal, and related activities. A minimum required level of continued surveillance and maintenance is planned for each facility/process system to ensure necessary environmental, health, and safety margins are maintained and to support ongoing operations for ICPP facilities that are not being deactivated. Management of the ICPP was transferred from Westinghouse Idaho Nuclear Company, Inc. (WINCO) to Lockheed Idaho Technologies Company (LITCO) on October 1, 1994 as part of the INEL consolidated contract. This revision of the deactivation plan (formerly the Nuclear Fuel Reprocessing Phaseout Plan for the ICPP) is being published during the consolidation of the INEL site-wide contract and the information presented here is current as of October 31, 1994. LITCO has adopted the existing plans for the deactivation of ICPP reprocessing facilities and the plans developed under WINCO are still being actively pursued, although the change in management may result in changes which have not yet been identified. Accordingly, the contents of this plan are subject to revision

Direct and irreversible inhibition of cyclooxygenase-1 by nitroaspirin (NCX 4016).

Science.gov (United States)

Corazzi, Teresa; Leone, Mario; Maucci, Raffaella; Corazzi, Lanfranco; Gresele, Paolo

2005-12-01

Benzoic acid, 2-(acetyl-oxy)-3-[(nitrooxy)methyl]phenyl ester (NCX 4016), a new drug made by an aspirin molecule linked, through a spacer, to a nitric oxide (NO)-donating moiety, is now under clinical testing for the treatment of atherothrombotic conditions. Aspirin exerts its antithrombotic activity by irreversibly inactivating platelet cyclooxygenase (COX)-1. NCX 4016 in vivo undergoes metabolism into deacetylated and/or denitrated metabolites, and it is not known whether NCX 4016 needs to liberate aspirin to inhibit COX-1, or whether it can block it as a whole molecule. The aim of our study was to evaluate the effects of NCX 4016 and its analog or metabolites on platelet COX-1 and whole blood COX-2 and on purified ovine COX (oCOX)-1 and oCOX-2. In particular, we have compared the mechanism by which NCX 4016 inhibits purified oCOX enzymes with that of aspirin using a spectrophotometric assay. All the NCX 4016 derivatives containing acetylsalicylic acid inhibited the activity of oCOX-1 and oCOX-2, whereas the deacetylated metabolites and the nitric oxide-donating moiety were inactive. Dialysis experiments showed that oCOX-1 inhibition by NCX 4016, similar to aspirin, is irreversible. Reversible COX inhibitors (indomethacin) or salicylic acid incubated with the enzyme before NCX 4016 prevent the irreversible inhibition of oCOX-1 by NCX 4016 as well as by aspirin. In conclusion, our data show that NCX 4016 acts as a direct and irreversible inhibitor of COX-1 and that the presence of a spacer and NO-donating moiety in the molecule slows the kinetics of COX-1 inhibition by NCX 4016, compared with aspirin.

Evaluation of high-level waste vitrification feed preparation chemistry for an NCAW simulant, FY 1994: Alternate flowsheets (DRAFT)

International Nuclear Information System (INIS)

Smith, H.D.; Merz, M.D.; Wiemers, K.D.; Smith, G.L.

1996-02-01

High-level radioactive waste stored in tanks at the U.S. Department of Energy's (DOE's) Hanford Site will be pretreated to concentrate radioactive constituents and fed to the vitrification plant A flowsheet for feed preparation within the vitrification plant (based on the Hanford Waste Vitrification Plant (HWVP) design) called for HCOOH addition during the feed preparation step to adjust rheology and glass redox conditions. However, the potential for generating H 2 and NH 3 during treatment of high-level waste (HLW) with HCOOH was identified at Pacific Northwest Laboratory (PNL). Studies at the University of Georgia, under contract with Savannah River Technology Center (SRTC) and PNL, have verified the catalytic role of noble metals (Pd, Rh, Ru), present in the waste, in the generation of H 2 and NH 3 . Both laboratory-scale and pilot-scale studies at SRTC have documented the H 2 and NH 3 generation phenomenal Because H 2 and NH 3 may create hazardous conditions in the vessel vapor space and offgas system of a vitrification plant, reducing the H 2 generation rate and the NH 3 generation to the lowest possible levels consistent with desired melter feed characteristics is important. The Fiscal Year 1993 and 1994 studies were conducted with simulated (non-radioactive), pre-treated neutralized current acid waste (NCAW). Neutralized current acid waste is a high-level waste originating from the plutonium/uranium extraction (PUREX) plant that has been partially denitrated with sugar, neutralized with NaOH, and is presently stored in double-shell tanks. The non-radioactive simulant used for the present study includes all of the trace components found in the waste, or substitutes a chemically similar element for radioactive or very toxic species. The composition and simulant preparation steps were chosen to best simulate the chemical processing characteristics of the actual waste

Experimental Investigation of the IFMIF Target Mock-up

International Nuclear Information System (INIS)

Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Y.; Arnoldov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.M.; Nakamura, H.

2007-01-01

Full text of publication follows: The IFMIF lithium neutron target mock-ups have been constructed and tested at the water and lithium test facilities. Description of the mock-ups and test facilities is presented in the paper, as well as the main results obtained. Reference geometry was used but the mockup flow cross-section was decreased. Velocity of water and lithium was up to reference value of 20 m/s. Features of lithium and water hydrodynamics were observed. The calculations and experiments showed that conjunction point of back wall straight and concave sections generated instability of lithium flow because of centrifugal force sudden change at this place. Therefore, it was proposed to use parabolic shape of the target back wall. Generation of wakes at the corners of cross-section of the Shima nozzle outlet was observed, and, as a result, surface waves appeared on the lithium jet. Observations of lithium and water jets and measurements of water jet thickness showed significant increasing the thickness near sidewalls of the mock-up concave section. It is because of absence of the centrifugal force at these places. Very large instability of the water jet surface was observed when outlet part of the Shima nozzle was divergent slightly (about 1 deg.), and vice versa very smooth jet surface occurred in confusing case (of about 0.5 deg.). So, nozzle outlet shape is very critical. Evaporation of lithium from the jet surface was investigated as well as deposition of vapor on vacuum pipe wall. It turned out to be not so critical. Significant part of the work concerned purification of lithium and monitoring impurities. The possibility of denitration of lithium down to 2 ppm by means of aluminum soluble getter was showed. Two types of both cold traps and plug indicators of impurities were tested. The results are presented in the paper. (authors)

Laboratory characterization and vitrification of Hanford radioactive high-level waste

International Nuclear Information System (INIS)

Tingey, J.M.; Elliott, M.L.; Larson, D.E.; Morrey, E.V.

1991-05-01

Radioactive high-level wastes generated at the Department of Energy's Hanford Site are stored in underground carbon steel tanks. Two double-shell tanks contain neutralized current acid waste (NCAW) from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant. The tanks were sampled for characterization and waste immobilization process/product development. The high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). The CAW was ''neutralized'' to a pH of approximately 14 by adding sodium hydroxide to reduce corrosion of the tanks. This ''neutralized'' waste is called Neutralized Current Acid Waste. Both precipitated solids and liquids are stored in the NCAW waste tanks. The NCAW contains small amounts of plutonium and most of the fission products and americium from the irradiated fuel. NCAW also contains stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. The NCAW will be retrieved, pretreated, and immobilized prior to final disposal. Pretreatment consists of water washing the precipitated NCAW solids for sulfate and soluble salts removal as a waste reduction step prior to vitrification. This waste is expected to be the first waste type to be retrieved and vitrified in the Hanford Waste Vitrification Plant (HWVP). A characterization plan was developed that details the processing of the small-volume NCAW samples through retrieval, pretreatment and vitrification process steps. Physical, rheological, chemical, and radiochemical properties were measured throughout these process steps. The results of nonradioactive simulant tests were used to develop appropriate pretreatment and vitrification process steps. The processing and characterization of simulants and actual NCAW tank samples are used to evaluate the operation of these processes. 3 refs., 1 fig., 4 tabs

Nitrate to ammonia and ceramic (NAC) process during batch and continuous operation

International Nuclear Information System (INIS)

Muguercia, I.; Solomon, S.; Ebadian, M.A.

1996-01-01

The nitrate to ammonia and ceramic (NAC) process is an innovative technology for the denitration of radioactive sodium nitrate-based liquid waste found throughout Department of Energy (DOE) facilities in the United States. In the present investigation, two reaction systems were studied. The first utilized only sodium nitrate as the substrate for the aluminum. The second consisted of the multication composition of waste forms located at the Hanford facility. Studies were carried out on the batch reaction at three different starting nitrate ion concentrations, each at three different temperatures. For each of these conditions, the rate of nitrate depletion was determined, and rate constants were calculated. The reaction did not demonstrate simple kinetics; rather, it appeared to involve two zero order reactions. Certain generalities were obtained in both the batch reaction and in the continuous process, nonetheless. It was found that the conversion of nitrate to ammonia seemed to be most efficient at the lowest temperature studied, 50 degrees C. This behavior was more obvious in the case of the unadulterated nitrate solution than with the Hanford simulant. To elaborate a practical, marketable product, it was necessary to develop a process that could be carried out in a continuous matter, whereby reactants were continuously fed into a reactor while the products of the reaction were simultaneously removed. Thus, the objective has been to develop the prototype procedures for carrying out this continuous reaction. As a corollary of this research, it was first necessary to define the characteristics of the reaction with respect to rate, conversion efficiency, and safety. To achieve this end, reactions were run under various batch conditions, and an attempt was made to measure the rates of the depletion of nitrate and the production of ammonia and hydrogen as well as pH and temperature changes

Development of radioactive waste management at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Miyanaga, I.; Sakata, S.; Ito, A.; Amano, H.

1977-01-01

For low- and medium-level waste treatment, main efforts have been put on the reduction of waste volume. For high-level wastes, studies are being carried out on the solidification and partitioning techniques in preparation for completion of the fuel cycle in Japan. For sea disposal of low-level wastes planned by the JAEC, significant information has been obtained regarding integrity and leaching behavior of cement solidified wastes. This paper describes the present status of development of the techniques in the following sections; 1. Treatment of Low- and Medium-Level Wastes; an incinerator with two stage ceramic filters has been tested, and the decontamination factor was found to be 10 4 for various nuclides; reverse osmosis method with a cellulose acetate membrane has been tested for laundry liquid waste, and 60 Co was removed more than 99% together with detergents; and solidification products of spent ion-exchange resin with polyethylene have been proved to be superior in mechanical properties, water resistance and volume reduction to asphalt products. 2. Safety Evaluation of Cement Solidified Wastes for Sea Disposal; homogeneous cement-solidified wastes in 200 l sealed drums did not show any cracks or defects under high hydrostatic pressure; the leaching ratio of 137 Cs for the first one year was estimated to be lower than 0.3%. 3. Treatment of High-Level Wastes; vitrification using natural zeolite has been developed and properties of the products were proved to be excellent; and a partitioning procedure consisting mainly of solvent extraction and ion-exchange method has been studied; reduction of the amount of alkaline agent by introducing a denitration technique, and reduction of resin volume by adopting a porous type resin were achieved

Technological advances in (U,Pu)O2 CRO recycling using microwave heating

International Nuclear Information System (INIS)

Das, D.K.; Singh, G.; Khot, P.M; Kumar, S.; Mishra, A.K.; Behere, P.G.; Afzal, Mohd; Kumar, Arun

2014-01-01

A batch type wet recycling process viz. microwave direct de-nitration and calcination technique (MWDDC) has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, India. The process was developed for complete and multiple recycling of PFBR clean rejected (U,Pu)O 2 MOX fuel pellets (CRO) up to 30(wt%) of PuO 2 . The complete recycling of CRO containing higher Pu content with conventional dry recycling was difficult to achieve and certain amount of virgin powder is always needed to obtain the required product characteristics. The conditioned co-de-nitrated powder via MWDDC process have more or less similar characteristics to that of virgin powder with respect to particle size, apparent and tap density, surface area. This paper presents an insight into MWDDC process details and recent advancements made for improvement of powder and product characteristics. Low temperature microwave calcination (LTMC) was incorporated to improve the quality of co-de-nitrated powder with regard to volatile impurities and nitrate content. MWDDC powder and pellets were subjected to extensive chemical and physical characterization as per PFBR specification document. MOX pellets were fabricated from virgin and MWDDC powder via powder oxide pelletizing route and characterized. The homogeneity in the MOX pellets fabricated from MWDDC powder was found as good as that of virgin. Industrial microwave heating systems are indigenously developed and have advanced applicator and wave transmission designs to achieve high throughput, precise control of microwave power hence the temperature during the course of the process. It was demonstrated that MWDDC is a novel technique for (U,Pu)O 2 MOX rejects recycling in view of complete and multiple recycling. Key words: (U,Pu)O 2 MOX, CRO, Recycling, MWDDC. (authors)

Idaho Chemical Processing Plant low-level waste grout stabilization development program FY-96 status report

International Nuclear Information System (INIS)

Herbst, A.K.

1996-09-01

The general purpose of the Grout Stabilization Development Program is to solidify and stabilize the liquid low-level wastes (LLW) generated at the Idaho Chemical Processing Plant (ICPP). It is anticipated that LLW will be produced from the following: (1) chemical separation of the tank farm high-activity sodium-bearing waste; (2) retrieval, dissolution, and chemical separation of the aluminum, zirconium, and sodium calcines; (3) facility decontamination processes; and (4) process equipment waste. The main tasks completed this fiscal year as part of the program were chromium stabilization study for sodium-bearing waste and stabilization and solidification of LLW from aluminum and zirconium calcines. The projected LLW will be highly acidic and contain high amounts of nitrates. Both of these are detrimental to Portland cement chemistry; thus, methods to precondition the LLW and to cure the grout were explored. A thermal calcination process, called denitration, was developed to solidify the waste and destroy the nitrates. A three-way blend of Portland cement, blast furnace slag, and fly ash was successfully tested. Grout cubes were prepared at various waste loadings to maximize loading while meeting compressive strength and leach resistance requirements. For the sodium LLW, a 25% waste loading achieves a volume reduction of 3.5 and a compressive strength of 2,500 pounds per square inch while meeting leach, mix, and flow requirements. It was found that the sulfur in the slag reduces the chromium leach rate below regulatory limits. For the aluminum LLW, a 15% waste loading achieves a volume reduction of 8.5 and a compressive strength of 4,350 pounds per square inch while meeting leach requirements. Likewise for zirconium LLW, a 30% waste loading achieves a volume reduction of 8.3 and a compressive strength of 3,570 pounds per square inch

Research program on development of advanced treatment technology for americium-containing aqueous waste in NUCEF

Energy Technology Data Exchange (ETDEWEB)

Mineo, Hideaki; Matsumura, Tatsuro; Tsubata, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1996-10-01

A research program was prepared on the development of an advanced treatment process for the americium-containing concentrated aqueous waste in NUCEF, than allows americium recovery for the reuse and the reduction of TRU waste generation. A preliminary analysis was conducted on the separation requirements based on the components estimated for the waste. An R and D strategy was proposed from the view to reduce TRU waste generated in the processing that the highest priority is given on the control of TRU leakage such as americium into the effluent stream after americium recovery and the minimization of salt used in the separation over the decontamination of impurities from americium. The extraction chromatographic method was selected as a candidate technology for americium separation under the principle to use reagents that are functional in acidic conditions such as bidentate extractants of DHEDECMP, CMPO or diamides, considering the larger flexibilities in process modification and possible multi-component separation with compact equipment and the past achievements on the recovery of kg quantities of americium. Major R and D items extracted are screening and evaluation of extractants for americium and plutonium, optimization of separation conditions, selection of denitration method, equipment developments and development of solidification methods of discarded americium after reuse and of various kinds of separation residues. In order to cope these items, four steps of R and D program were proposed, i.e., fundamental experiment in beaker-scale on screening and evaluation of extractants, flowsheet study in bench-scale using simulated and small amount of americium aqueous waste solution to evaluate candidate process, americium recovery test in iron-shielded cell to be installed in NUCEF. It is objected to make recovery of 100g orders of americium used for research on fundamental TRU fuel properties. (J.P.N.)

Proposed Atomic Energy of Canada Ltd. 99Mo waste calcination process

International Nuclear Information System (INIS)

Ramey, D.W.; Haas, P.A.; Malkemus, D.W.; McGinnis, C.P.; Meyers, E.S.; Patton, B.D.; Birdwell, J.F.; Jubin, R.T.; Coltharp, K.A.

1994-10-01

Atomic Energy of Canada Limited (AECL), at its Chalk River Laboratory, generates from 3000 to 5000 L/year of high-level fissile waste solution from the production of 99 Mo. In this Mo process, highly enriched uranium (93 wt % 235 U, total uranium basis) contained in uranium-aluminum alloy target rods is irradiated to produce the 99 Mo product. The targets are removed from the reactor and dissolved in a mercury nitrate-catalyzed reaction with nitric acid. The 99 Mo product is then recovered by passing the solution through an alumina (Al 2 O 3 ) column. During discussions with personnel from the Oak Ridge National Laboratory (ORNL) on September 10, 1992, the ORNL-developed technology formerly applied to the solidification of aqueous uranium waste (Consolidated Edison Uranium Solidification Program or CEUSP) was judged potentially applicable to the AECL 99 Mo waste. Under a Work-for-Others contract (no. ERD-92-1132), which began May 24, 1993, ORNL was tasked to determine the feasibility of applying the CEUSP (or a similar) calcination process to solidify AECL's 99 Mo waste for > 30 years of safe dry storage. This study was to provide sufficient detailed information on the applicability of a CEUSP-type waste solidification process to allow AECL to select the process which best suited its needs. As with the CEUSP process, evaporation of the waste and a simultaneously partial destruction of acid by reaction with formaldehyde followed by in situ waste can thermal denitration waste was chosen as the best means of solidification. Unlike the CEUSP material, the 99 Mo waste has a considerable number of problem volatile and semivolatile constituents which must be recovered in the off-gas treatment system. Mercury removal before calcination was seen as the best option

Volatilities of ruthenium, iodine, and technetium on calcining fission product nitrate wastes

International Nuclear Information System (INIS)

Rimshaw, S.J.; Case, F.N.

1980-01-01

Various high-level nitrate wastes were subjected to formic acid denitration. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases according to the following equation: 4 HCOOH + 2 HNO 3 → N 2 O + 4 CO 2 + 5 H 2 O. These gases can be scrubbed free of 106 Ru, 131 I, and 99 Tc radioactivities prior to elimination from the plant by passage through HEPA filters. The formation of deleterious NO/sub x/ is avoided. Moreover, formic acid reduces ruthenium to a lower valence state with a sharp reduction in RuO 4 volatility during subsequent calcination of the pretreated waste. It is shown that a minimum of 3% of RuO 4 in an off-gas stream reacts with Davison silica gel (Grade 40) to give a fine RuO 2 aerosol having a particle size of 0.5 μ. This RuO 2 aerosol passes through water or weak acid scrub solutions but is trapped by a caustic scrub solution. Iodine volatilizes almost completely on calcining an acidic waste, and the iodine volatility increases with increasing calcination temperature. On calcining an alkaline sodium nitrate waste the iodine volatility is about an order of magnitude lower, with a relatively low iodine volatility of 0.39% at a calcination temperature of 250 0 C and a moderate volatility of 9.5% at 600 0 C. Volatilities of 99 Tc were generally 0 C. Data are presented to indicate that 99 Tc concentrates in the alkaline sodium nitrate supernatant waste, with approx. 10 mg 99 Tc being associated with each curie of 137 Cs present in the waste. It is shown that lutidine (2,4 dimethyl-pyridine) extracts Tc(VII) quantitatively from alkaline supernatant wastes. The distribution coefficient (K/sub D/) for Tc(VII) going into the organic phase in the above system is 102 for a simulated West Valley waste and 191 for a simulated Savannah River Plant (SRP) waste

Fluidized Bed Steam Reforming of Hanford LAW Using THORsm Mineralizing Technology

International Nuclear Information System (INIS)

Olson, Arlin L.; Nicholas R Soelberg; Douglas W. Marshall; Gary L. Anderson

2004-01-01

The U.S. Department of Energy (DOE) documented, in 2002, a plan for accelerating cleanup of the Hanford Site, located in southeastern Washington State, by at least 35 years. A key element of the plan was acceleration of the tank waste program and completion of tank waste treatment by 2028 by increasing the capacity of the planned Waste Treatment Plant (WTP) and using supplemental technologies for waste treatment and immobilization. The plan identified steam reforming technology as a candidate for supplemental treatment of as much as 70% of the low-activity waste (LAW). Mineralizing steam reforming technology, offered by THOR Treatment Technologies, LLC would produce a denitrated, granular mineral waste form using a high-temperature fluidized bed process. A pilot scale demonstration of the technology was completed in a 15-cm-diameter reactor vessel. The pilot scale facility was equipped with a highly efficient cyclone separator and heated sintered metal filters for particulate removal, a thermal oxidizer for reduced gas species and NOx destruction, and a packed activated carbon bed for residual volatile species capture. The pilot scale equipment is owned by the DOE, but located at the Science and Technology Applications Research (STAR) Center in Idaho Falls, ID. Pilot scale testing was performed August 2-5, 2004. Flowsheet chemistry and operational parameters were defined through a collaborative effort involving Idaho National Engineering and Environmental Laboratory, Savannah River National Laboratory (SRNL), and THOR Treatment Technologies personnel. Science Application International Corporation, owners of the STAR Center, personnel performed actual pilot scale operation. The pilot scale test achieved a total of 68.7 hrs of cumulative/continuous processing operation before termination in response to a bed de-fluidization condition. 178 kg of LAW surrogate were processed that resulted in 148 kg of solid product, a mass reduction of about 17%. The process achieved

Comparative evaluation of glasses reprocessing and reversible conditioning of calcinates; Evaluation comparative de la reprise des verres et du conditionnement reversible des calcinats

Energy Technology Data Exchange (ETDEWEB)

Boen, R

2000-01-15

Fission products and minor actinides separated during the spent fuel reprocessing treatment are industrially vitrified on-line and thus confined inside a glass matrix with admittedly durability properties. In the framework of the feasibility of a reversible conditioning, this document examines first the possible alternative ways of conditioning and storage of calcinates before vitrification, which may simplify the reversibility aspect. Such a conditioning must be compatible with the storage process, with a possible extraction of actinides and long-lived fission products, and with the vitrification process if no extraction is performed. Calcinates are pulverulent and comprise an important soluble fraction, a proportion of nitrates of about 30%, and release a high thermal power (17 kW/m{sup 3}) combined to a low thermal conductivity (0.1 to 0.15 W.m{sup -1} k{sup -1}). Among the different foreseeable solutions (denitration, mixing with another material, with or without compacting, dissolution inside another material..), the dissolution inside a borate seems to be the most acceptable with respect to the safety, feasibility and vitrification aspects. The thermal aspect of the storage remains complex as a specific container is necessary. In a second part, this report analyzes the possibility to re-extract back the long-lived radionuclides from vitrified wastes. The different possible ways to destroy the glass structure and to transfer the fission products and minor actinides in an aqueous solution compatible with an hydrometallurgical separation process are explored. Two processes are foreseeable: a low temperature dissolution process which requires a preliminary crushing and the handling of huge amounts of acids, and a both high and low temperature process which comprises the following steps: melting, fractionation by water tempering, addition of Na{sub 2}O or sodium tetraborate to make it sensible to hot leaching, separation of fission products and minor actinides

Fluidized Bed Steam Reforming of Hanford LAW Using THORsm Mineralizing Technology

Energy Technology Data Exchange (ETDEWEB)

Olson, Arlin L.; Nicholas R Soelberg; Douglas W. Marshall; Gary L. Anderson

2004-11-01

The U.S. Department of Energy (DOE) documented, in 2002, a plan for accelerating cleanup of the Hanford Site, located in southeastern Washington State, by at least 35 years. A key element of the plan was acceleration of the tank waste program and completion of ''tank waste treatment by 2028 by increasing the capacity of the planned Waste Treatment Plant (WTP) and using supplemental technologies for waste treatment and immobilization.'' The plan identified steam reforming technology as a candidate for supplemental treatment of as much as 70% of the low-activity waste (LAW). Mineralizing steam reforming technology, offered by THOR Treatment Technologies, LLC would produce a denitrated, granular mineral waste form using a high-temperature fluidized bed process. A pilot scale demonstration of the technology was completed in a 15-cm-diameter reactor vessel. The pilot scale facility was equipped with a highly efficient cyclone separator and heated sintered metal filters for particulate removal, a thermal oxidizer for reduced gas species and NOx destruction, and a packed activated carbon bed for residual volatile species capture. The pilot scale equipment is owned by the DOE, but located at the Science and Technology Applications Research (STAR) Center in Idaho Falls, ID. Pilot scale testing was performed August 2–5, 2004. Flowsheet chemistry and operational parameters were defined through a collaborative effort involving Idaho National Engineering and Environmental Laboratory, Savannah River National Laboratory (SRNL), and THOR Treatment Technologies personnel. Science Application International Corporation, owners of the STAR Center, personnel performed actual pilot scale operation. The pilot scale test achieved a total of 68.7 hrs of cumulative/continuous processing operation before termination in response to a bed de-fluidization condition. 178 kg of LAW surrogate were processed that resulted in 148 kg of solid product, a mass reduction of about 17%. The process

Comparative evaluation of glasses reprocessing and reversible conditioning of calcinates

International Nuclear Information System (INIS)

Boen, R.

2000-01-01

Fission products and minor actinides separated during the spent fuel reprocessing treatment are industrially vitrified on-line and thus confined inside a glass matrix with admittedly durability properties. In the framework of the feasibility of a reversible conditioning, this document examines first the possible alternative ways of conditioning and storage of calcinates before vitrification, which may simplify the reversibility aspect. Such a conditioning must be compatible with the storage process, with a possible extraction of actinides and long-lived fission products, and with the vitrification process if no extraction is performed. Calcinates are pulverulent and comprise an important soluble fraction, a proportion of nitrates of about 30%, and release a high thermal power (17 kW/m 3 ) combined to a low thermal conductivity (0.1 to 0.15 W.m -1 k -1 ). Among the different foreseeable solutions (denitration, mixing with another material, with or without compacting, dissolution inside another material..), the dissolution inside a borate seems to be the most acceptable with respect to the safety, feasibility and vitrification aspects. The thermal aspect of the storage remains complex as a specific container is necessary. In a second part, this report analyzes the possibility to re-extract back the long-lived radionuclides from vitrified wastes. The different possible ways to destroy the glass structure and to transfer the fission products and minor actinides in an aqueous solution compatible with an hydrometallurgical separation process are explored. Two processes are foreseeable: a low temperature dissolution process which requires a preliminary crushing and the handling of huge amounts of acids, and a both high and low temperature process which comprises the following steps: melting, fractionation by water tempering, addition of Na 2 O or sodium tetraborate to make it sensible to hot leaching, separation of fission products and minor actinides, recycling of

Fluidized Bed Steam Reforming of INEEL SBW Using THORsm Mineralizing Technology

Energy Technology Data Exchange (ETDEWEB)

Arlin L. Olson; Nicholas R. Soelberg; Douglas W. Marshall; Gary L. Anderson

2004-12-01

Sodium bearing waste (SBW) disposition is one of the U.S. Department of Energy (DOE) Idaho Operation Office’s (NE-ID) and State of Idaho’s top priorities at the Idaho National Engineering and Environmental Laboratory (INEEL). Many studies have resulted in the identification of five treatment alternatives that form a short list of perhaps the most appropriate technologies for the DOE to select from. The alternatives are (a) calcination with maximum achievable control technology (MACT) upgrade, (b) steam reforming, (c) cesium ion exchange (CsIX) with immobilization, (d) direct evaporation, and (e) vitrification. Each alternative has undergone some degree of applied technical development and preliminary process design over the past four years. DOE desired further experimental data, with regard to steam reforming technology, to make informed decisions concerning selection of treatment technology for SBW. Mineralizing steam reforming technology, offered by THOR Treatment Technologies, LLC would produce a denitrated, granular mineral waste form using a high-temperature fluidized bed process. A pilot scale demonstration of the technology was performed in a 15-cm-diameter reactor vessel September 27 through October 1, 2004. The pilot scale equipment is owned by the DOE, and located at the Science and Technology Applications Research (STAR) Center in Idaho Falls, ID. Flowsheet chemistry and operational parameters were defined through a collaborative effort involving Idaho National Engineering and Environmental Laboratory, Savannah River National Laboratory (SRNL), and THOR Treatment Technologies personnel. Personnel from Science Applications International Corporation, owners of the STAR Center, operated the pilot plant. The pilot scale test was terminated as planned after achieving a total of 100 hrs of cumulative/continuous processing operation. About 230 kg of SBW surrogate were processed that resulted in about 88 kg of solid product, a mass reduction of about 62

MINERALIZING, STEAM REFORMING TREATMENT OF HANFORD LOW-ACTIVITY WASTE (a.k.a. INEEL/EXT-05-02526)

International Nuclear Information System (INIS)

A. L. Olson; N. R. Soelberg; D. W. Marshall; G. L. Anderson

2005-01-01

The U.S. Department of Energy (DOE) documented, in 2002, a plan for accelerating cleanup of the Hanford Site, located in southeastern Washington State, by at least 35 years. A key element of the plan was acceleration of the tank waste program and completion of ''tank waste treatment by 2028 by increasing the capacity of the planned Waste Treatment Plant (WTP) and using supplemental technologies for waste treatment and immobilization''. The plan identified steam reforming technology as a candidate for supplemental treatment of as much as 70% of the low-activity waste (LAW). Mineralizing steam reforming technology, offered by THOR Treatment Technologies, LLC would produce a denitrated, granular mineral waste form using a high-temperature fluidized bed process. A pilot scale demonstration of the technology was completed in a 15-cm-diameter reactor vessel. The pilot scale facility was equipped with a cyclone separator and heated sintered metal filters for particulate removal, a thermal oxidizer for reduced gas species and NOx destruction, and a packed activated carbon bed for residual volatile species capture. The pilot scale equipment is owned by the DOE, but located at the Science and Technology Applications Research (STAR) Center in Idaho Falls, ID. Pilot scale testing was performed August 2-5, 2004. Flowsheet chemistry and operational parameters were defined through a collaborative effort involving Idaho National Engineering and Environmental Laboratory (INEEL), Savannah River National Laboratory (SRNL), and THOR Treatment Technologies personnel. Science Application International Corporation, owners of the STAR Center, personnel performed actual pilot scale operation. The pilot scale test achieved a total of 68.4 hours of cumulative/continuous processing operation before termination in response to a bed de-fluidization condition. 178 kg of LAW surrogate were processed that resulted in 148 kg of solid product, a mass reduction of about 17%. The process achieved

Volatility of ruthenium-106, technetium-99, and iodine-129, and the evolution of nitrogen oxide compounds during the calcination of high-level, radioactive nitric acid waste

International Nuclear Information System (INIS)

Rimshaw, S.J.; Case, F.N.; Tompkins, J.A.

1980-02-01

The nitrate anion is the predominant constituent in all high-level nuclear wastes. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases (N 2 or N 2 O), which can be scrubbed free of 106 Ru, 129 I, and 99 Tc radioactivities prior to elimination from the plant by passing through HEPA filters. Treatment of a high-level authentic radioactive waste with two moles of formic acid per mole of nitrate anion leads to a low RuO 4 volatility of about 0.1%, which can be reduced to an even lower level of 0.007% on adding a 15% excess of formic acid. Without pretreatment of the nitrate waste with formic acid, a high RuO 4 volatility of approx. 35% is observed on calcining a 4.0 N HNO 3 solution in quartz equipment at 350 0 C. The RuO 4 volatility falls to approx. 1.0% on decreasing the initial HNO 3 concentration to 1.0 N or lower. It is postulated that thermal denitration of a highly nitrated ruthenium complex leads to the formation of volatile RuO 4 , while decarboxylation of a ruthenium-formate complex leads to the formation of nonvolatile RuO 2 . Wet scrubbing with water is used to remove RuO 4 from the off-gas stream. In all glass equipment, small amounts of particulate RuO 2 are formed in the gas phase by decomposition of RuO 4 . The 99 Tc volatility was found to vary from 0.2 to 1.4% on calcining HNO 3 and HCOOH (formic acid) solutions over the temperature range of 250 to 600 0 C. These unexpectedly low volatilities of 99 Tc are correlated to the high thermal stability limits of various metal pertechnetates and technetates. Iodine volatilities were high, varying from a low of 30% at 350 0 C to a high of 97% at 650 0 C. It is concluded that with a proper selection of pretreatment and operating conditions the 106 Ru and 99 Tc activities can be retained in the calcined solid with recycle of the wet scrubbing solution

Development of Concentration and Calcination Technology For High Level Liquid Waste

International Nuclear Information System (INIS)

Pande, D.P.

2006-01-01

The concentrated medium and high-level liquid radio chemicals effluents contain nitric acid, water along with the dissolved chemicals including the nitrates of the radio nuclides. High level liquid waste contain mainly nitrates of cesium, strontium, cerium, zirconium, chromium, barium, calcium, cobalt, copper, pickle, iron etc. and other fission products. This concentrated solution requires further evaporation, dehydration, drying and decomposition in temperature range of 150 to 700 deg. C. The addition of the calcined solids in vitrification pot, instead of liquid feed, helps to avoid low temperature zone because the vaporization of the liquid and decomposition of nitrates do not take place inside the melter. In our work Differential and thermo gravimetric studies has been carried out in the various stages of thermal treatment including drying, dehydration and conversion to oxide forms. Experimental studies were done to characterize the chemicals present in high-level radioactive waste. A Rotary Ball Kiln Calciner was used for development of the process because this is amenable for continuous operation and moderately good heat transfer can be achieved inside the kiln. This also has minimum secondary waste and off gases generation. The Rotary Ball Kiln Calciner Demonstration facility system was designed and installed for the demonstration of calcination process. The Rotary Ball Kiln Calciner is a slowly rotating slightly inclined horizontal tube that is externally heated by means of electric resistance heating. The liquid feed is sprayed onto the moving bed of metal balls in a slowly rotating calciner by a peristaltic type-metering pump. The vaporization of the liquid occurs in the pre-calcination zone due to counter current flow of hot gases. The dehydration and denitration of the solids occurs in the calcination zone, which is externally heated by electrical furnace. The calcined powder is cooled in the post calcination portion. It has been demonstrated that the

Nitrous oxide distribution and its origin in the central and eastern South Pacific Subtropical Gyre

Directory of Open Access Journals (Sweden)

J. Charpentier

2007-09-01

Full Text Available The mechanisms of microbial nitrous oxide (N₂O production in the ocean have been the subject of many discussions in recent years. New isotopomeric tools can further refine our knowledge of N₂O sources in natural environments. This study compares hydrographic, N₂O concentration, and N₂O isotopic and isotopomeric data from three stations along a coast-perpendicular transect in the South Pacific Ocean, extending from the center (Sts. GYR and EGY of the subtropical oligotrophic gyre (~26° S; 114° W to the upwelling zone (St. UPX off the central Chilean coast (~34° S. Although AOU/N₂O and NO₃⁻ trends support the idea that most of the N₂O (mainly from intermediate water (200–600 m comes from nitrification, N₂O isotopomeric composition (intramolecular distribution of ¹⁵N isotopes expressed as SP (site preference of ¹⁵N shows low values (10 to 12permil that could be attributed to the production through of microbial nitrifier denitrification (reduction of nitrite to N₂O mediated by ammonium oxidizers. The coincidence of this SP signal with high – stability layer, where sinking organic particles can accumulate, suggests that N₂O could be produced by nitrifier denitrification inside particles. It is postulated that deceleration of particles in the pycnocline can modify the advection - diffusion balance inside particles, allowing the accumulation of nitrite and O₂ depletion suitable for nitrifier denitrication. As lateral advection seems to be relatively insignificant in the gyre, in situ nitrifier denitrification could account for 40–50% of the N₂O produced in this layer. In contrast, coastal upwelling system is characterized by O₂ deficient condition and some N deficit in a eutrophic system. Here, N₂O accumulates up to 480% saturation, and isotopic and

Iron-Dependent Enzyme Catalyzes the Initial Step in Biodegradation of N-Nitroglycine by Variovorax sp. Strain JS1663.

Science.gov (United States)

Mahan, Kristina M; Zheng, Hangping; Fida, Tekle T; Parry, Ronald J; Graham, David E; Spain, Jim C

2017-08-01

Nitramines are key constituents of most of the explosives currently in use and consequently contaminate soil and groundwater at many military facilities around the world. Toxicity from nitramine contamination poses a health risk to plants and animals. Thus, understanding how nitramines are biodegraded is critical to environmental remediation. The biodegradation of synthetic nitramine compounds such as hexahydro-1,3,5-trinitro-1,3,5-triazine (RDX) has been studied for decades, but little is known about the catabolism of naturally produced nitramine compounds. In this study, we report the isolation of a soil bacterium, Variovorax sp. strain JS1663, that degrades N -nitroglycine (NNG), a naturally produced nitramine, and the key enzyme involved in its catabolism. Variovorax sp. JS1663 is a Gram-negative, non-spore-forming motile bacterium isolated from activated sludge based on its ability to use NNG as a sole growth substrate under aerobic conditions. A single gene ( nnlA ) encodes an iron-dependent enzyme that releases nitrite from NNG through a proposed β-elimination reaction. Bioinformatics analysis of the amino acid sequence of NNG lyase identified a PAS (Per-Arnt-Sim) domain. PAS domains can be associated with heme cofactors and function as signal sensors in signaling proteins. This is the first instance of a PAS domain present in a denitration enzyme. The NNG biodegradation pathway should provide the basis for the identification of other enzymes that cleave the N-N bond and facilitate the development of enzymes to cleave similar bonds in RDX, nitroguanidine, and other nitramine explosives. IMPORTANCE The production of antibiotics and other allelopathic chemicals is a major aspect of chemical ecology. The biodegradation of such chemicals can play an important ecological role in mitigating or eliminating the effects of such compounds. N -Nitroglycine (NNG) is produced by the Gram-positive filamentous soil bacterium Streptomyces noursei This study reports the

Spatial and temporal evaluation of long term trend (2005-2014) of OMI retrieved NO2 and SO2 concentrations in Henan Province, China

Science.gov (United States)

Zhang, Leishi; Lee, Chih Sheng; Zhang, Ruiqin; Chen, Liangfu

2017-04-01

Tropospheric NO2 and SO2 concentrations are of great importance with regard to air quality, atmospheric chemistry, and climate change. Due to lack of surface monitoring stations, this study analyzes long term trend of NO2 and SO2 levels (2005-2014), retrieved from Ozone Monitoring Instrument (OMI) board on the NASA's Aura satellite, in an important region of China - Henan Province. Henan Province, located in North China Plain, has encountered serious air pollution problems including extremely high PM2.5 concentrations and as one of the most polluted region in China. The satellite spatial images clearly show that high levels of both NO2 and SO2 are concentrated in north and northeastern regions with much lower levels observed in other parts of Henan. Both pollutants exhibit the highest levels in winter with the least in summer/spring. The temporal trend analysis based on moving average of deseasonalized and decyclic data indicates that for NO2, there is a continuous increasing pattern from 2005 to 2011 at 6.4% per year, thereafter, it shows a decreasing trend (10.6% per year). As for SO2, the increasing trend is about 16% per year from 2005 to 2007 with decreasing rate 7% per year from 2007 to 2014. The economic development with incredible annual 11% GDP growth in Henan is responsible for increasing levels of NO2 and SO2. The observed decreasing SO2 level starting in 2007 is due to reduced SO2 emission, utilization of flue gas desulfurization (FGD) devices and to some extent, in preparation of Beijing 2008 Olympic Games. On the other hand, increasing vehicle numbers (155% from 2006 to 2012) and coal consumption (37% during the same span), along with the lack of denitration process for removing flue/exhaust gas NOx are responsible for increasing NO2 trend until 2011. The ratio of SO2/NO2 started decreasing in 2007 and dropped significantly from 2011 to 2013 indicating good performance of FGD and ever increasing NOx contribution from mobile sources. Unlike those

MDD Status Letter Report (AFCI CETE Milestone)

International Nuclear Information System (INIS)

Vedder, Raymond James; Jubin, Robert Thomas

2009-01-01

Current flow sheets for processing used nuclear fuels do not produce separated streams of all of the actinides. These aqueous processing streams must be converted into solid forms suitable for recycle (fuel/target fabrication), storage, or disposal, necessitating co-conversion. A process developed at ORNL in the 1980s to make UO 3 suitable as fuel feedstock was studied for preparation of mixed actinide oxides with similarly favorable ceramic properties. The process, Modified Direct Denitration (MDD), uses ammonium nitrate to alter the thermal decomposition behavior of metal nitrates and improve the ceramic properties of the resulting solid oxide. Since plutonium (IV) and neptunium(IV) form compounds similar to uranium with the ammonium ion [(NH 4 ) 2 Pu(NO 3 ) 6 , (NH 4 ) 2 Np(NO 3 ) 6 ], MDD-conversion of these metals was considered to be applicable. Co-conversion has advantages for making mixed oxides over individual element conversions that are followed by dry mixing of the oxide powders. Issues associated with preparing a mixture from individual oxides include use of additional equipment, dusting associated with feeding and milling, time requirements for milling, blending to obtain a uniform mixture, and inhomogenity at higher plutonium concentrations. These issues can be partially or wholly avoided by using MDD coconversion in which the mixing of the individual metals occurs in liquid solution; thus, adjusting relative metal concentrations is simpler and the resulting mixed oxide is more uniform than that produced by blending the individual oxides. Utilizing MDD also eliminates the need for mechanical treatment of the powder to obtain the desired ceramic properties, such as surface area and particle size distribution, since these characteristics are acceptable as-produced. The original MDD development work established that uranium oxide with good ceramic properties could be made. Following the discovery, a more fundamental understanding of the chemistry of the

MDD Status Letter Report (AFCI CETE Milestone)

Energy Technology Data Exchange (ETDEWEB)

Vedder, Raymond James [ORNL; Jubin, Robert Thomas [ORNL

2009-09-01

Current flow sheets for processing used nuclear fuels do not produce separated streams of all of the actinides. These aqueous processing streams must be converted into solid forms suitable for recycle (fuel/target fabrication), storage, or disposal, necessitating co-conversion. A process developed at ORNL in the 1980s to make UO{sub 3} suitable as fuel feedstock was studied for preparation of mixed actinide oxides with similarly favorable ceramic properties. The process, Modified Direct Denitration (MDD), uses ammonium nitrate to alter the thermal decomposition behavior of metal nitrates and improve the ceramic properties of the resulting solid oxide. Since plutonium (IV) and neptunium(IV) form compounds similar to uranium with the ammonium ion [(NH{sub 4}){sub 2}Pu(NO{sub 3}){sub 6}, (NH{sub 4}){sub 2}Np(NO{sub 3}){sub 6}], MDD-conversion of these metals was considered to be applicable. Co-conversion has advantages for making mixed oxides over individual element conversions that are followed by dry mixing of the oxide powders. Issues associated with preparing a mixture from individual oxides include use of additional equipment, dusting associated with feeding and milling, time requirements for milling, blending to obtain a uniform mixture, and inhomogenity at higher plutonium concentrations. These issues can be partially or wholly avoided by using MDD coconversion in which the mixing of the individual metals occurs in liquid solution; thus, adjusting relative metal concentrations is simpler and the resulting mixed oxide is more uniform than that produced by blending the individual oxides. Utilizing MDD also eliminates the need for mechanical treatment of the powder to obtain the desired ceramic properties, such as surface area and particle size distribution, since these characteristics are acceptable as-produced. The original MDD development work established that uranium oxide with good ceramic properties could be made. Following the discovery, a more fundamental

BENCH-SCALE STEAM REFORMING OF ACTUAL TANK 48H WASTE

International Nuclear Information System (INIS)

Burket, P; Gene Daniel, G; Charles Nash, C; Carol Jantzen, C; Michael Williams, M

2008-01-01

Fluidized Bed Steam Reforming (FBSR) has been demonstrated to be a viable technology to remove >99% of the organics from Tank 48H simulant, to remove >99% of the nitrate/nitrite from Tank 48H simulant, and to form a solid product that is primarily carbonate based. The technology was demonstrated in October of 2006 in the Engineering Scale Test Demonstration Fluidized Bed Steam Reformer1 (ESTD FBSR) at the Hazen Research Inc. (HRI) facility in Golden, CO. The purpose of the Bench-scale Steam Reformer (BSR) testing was to demonstrate that the same reactions occur and the same product is formed when steam reforming actual radioactive Tank 48H waste. The approach used in the current study was to test the BSR with the same Tank 48H simulant and same Erwin coal as was used at the ESTD FBSR under the same operating conditions. This comparison would allow verification that the same chemical reactions occur in both the BSR and ESTD FBSR. Then, actual radioactive Tank 48H material would be steam reformed in the BSR to verify that the actual tank 48H sample reacts the same way chemically as the simulant Tank 48H material. The conclusions from the BSR study and comparison to the ESTD FBSR are the following: (1) A Bench-scale Steam Reforming (BSR) unit was successfully designed and built that: (a) Emulated the chemistry of the ESTD FBSR Denitration Mineralization Reformer (DMR) and Carbon Reduction Reformer (CRR) known collectively as the dual reformer flowsheet. (b) Measured and controlled the off-gas stream. (c) Processed real (radioactive) Tank 48H waste. (d) Met the standards and specifications for radiological testing in the Savannah River National Laboratory (SRNL) Shielded Cells Facility (SCF). (2) Three runs with radioactive Tank 48H material were performed. (3) The Tetraphenylborate (TPB) was destroyed to > 99% for all radioactive Bench-scale tests. (4) The feed nitrate/nitrite was destroyed to >99% for all radioactive BSR tests the same as the ESTD FBSR. (5) The

BENCH-SCALE STEAM REFORMING OF ACTUAL TANK 48H WASTE

Energy Technology Data Exchange (ETDEWEB)

Burket, P; Gene Daniel, G; Charles Nash, C; Carol Jantzen, C; Michael Williams, M

2008-09-25

Fluidized Bed Steam Reforming (FBSR) has been demonstrated to be a viable technology to remove >99% of the organics from Tank 48H simulant, to remove >99% of the nitrate/nitrite from Tank 48H simulant, and to form a solid product that is primarily carbonate based. The technology was demonstrated in October of 2006 in the Engineering Scale Test Demonstration Fluidized Bed Steam Reformer1 (ESTD FBSR) at the Hazen Research Inc. (HRI) facility in Golden, CO. The purpose of the Bench-scale Steam Reformer (BSR) testing was to demonstrate that the same reactions occur and the same product is formed when steam reforming actual radioactive Tank 48H waste. The approach used in the current study was to test the BSR with the same Tank 48H simulant and same Erwin coal as was used at the ESTD FBSR under the same operating conditions. This comparison would allow verification that the same chemical reactions occur in both the BSR and ESTD FBSR. Then, actual radioactive Tank 48H material would be steam reformed in the BSR to verify that the actual tank 48H sample reacts the same way chemically as the simulant Tank 48H material. The conclusions from the BSR study and comparison to the ESTD FBSR are the following: (1) A Bench-scale Steam Reforming (BSR) unit was successfully designed and built that: (a) Emulated the chemistry of the ESTD FBSR Denitration Mineralization Reformer (DMR) and Carbon Reduction Reformer (CRR) known collectively as the dual reformer flowsheet. (b) Measured and controlled the off-gas stream. (c) Processed real (radioactive) Tank 48H waste. (d) Met the standards and specifications for radiological testing in the Savannah River National Laboratory (SRNL) Shielded Cells Facility (SCF). (2) Three runs with radioactive Tank 48H material were performed. (3) The Tetraphenylborate (TPB) was destroyed to > 99% for all radioactive Bench-scale tests. (4) The feed nitrate/nitrite was destroyed to >99% for all radioactive BSR tests the same as the ESTD FBSR. (5) The

Integrated Waste Treatment Unit (IWTU) Input Coal Analyses and Off-Gass Filter (OGF) Content Analyses

Energy Technology Data Exchange (ETDEWEB)

Jantzen, Carol M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, David M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Guenther, Chris P. [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Shekhawat, Dushyant [National Energy Technology Lab. (NETL), Morgantown, WV (United States); VanEssendelft, Dirk T. [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Means, Nicholas C. [AECOM Technology Corp., Oak Ridge, TN (United States)

2015-04-23

A full engineering scale Fluidized Bed Steam Reformer (FBSR) system is being used at the Idaho Nuclear Technology and Engineering Center (INTEC) to stabilize acidic Low Activity Waste (LAW) known as Sodium Bearing Waste (SBW). The INTEC facility, known as the Integrated Waste Treatment Unit (IWTU), underwent an Operational Readiness Review (ORR) and a Technology Readiness Assessment (TRA) in March 2014. The IWTU began non-radioactive simulant processing in late 2014 and by January, 2015 ; the IWTU had processed 62,000 gallons of simulant. The facility is currently in a planned outage for inspection of the equipment and will resume processing simulated waste feed before commencing to process 900,000 gallons of radioactive SBW. The SBW acidic waste will be made into a granular FBSR product (carbonate based) for disposal in the Waste Isolation Pilot Plant (WIPP). In the FBSR process calcined coal is used to create a CO2 fugacity to force the waste species to convert to carbonate species. The quality of the coal, which is a feed input, is important because the reactivity, moisture, and volatiles (C,H,N,O, and S) in the coal impact the reactions and control of the mineralizing process in the primary steam reforming vessel, the Denitration and Mineralizing Reformer (DMR). Too much moisture in the coal can require that additional coal be used. However since moisture in the coal is only a small fraction of the moisture from the fluidizing steam this can be self-correcting. If the coal reactivity or heating value is too low then the coal feedrate needs to be adjusted to achieve the desired heat generation. Too little coal and autothermal heat generation in the DMR cannot be sustained and/or the carbon dioxide fugacity will be too low to create the desired carbonate mineral species. Too much coal and excess S and hydroxide species can form. Excess sulfur from coal that (1) is too rich in sulfur or (2) from overfeeding coal can promote wall scale and contribute to corrosion

Nuclear Materials Management in a Recovery Facility for Unirradiated Enriched Uranium; Gestion des Matieres Nucleaires dans une Installation de Recuperation d'Uranium Enrichi Non Irradie; Administrativno-khozyajstvennyj uchet yadernykh materialov na ustanovke po regeneratsii neobluchennogo obogashchennogo urana; Administracion de Materiales Nucleares en una Planta de Recuperacion para Uranio Enriquecido No Irradiado

Energy Technology Data Exchange (ETDEWEB)

Jasny, G. R. [Union Carbide Corporation, Oak Ridge, TN (United States)

1966-02-15

The United States Atomic Energy Commission's Y-12 Plant, in Oak Ridge, Tennessee, has been processing and recovering various forms of unirradiated enriched uranium for over twenty years. Today, the Y-12 recovery facilities consist of a semi-continuous train of unit operations including dissolution, combustion, evaporation, extraction, denitration, and hydrofluorination. The processing and storage equipment is of restricted geometry and has a combined capacity of several hundred kilograms of enriched Uranium per month. Feed to the recovery operation comes both from the Plant and from other USAEC installations in the United States. This feed has included, at one time or another, practically every type of unirradiated enriched uranium scrap. Physical control is maintained successfully by the following techniques: 1. Careful design of equipment to permit cleaning and sampling and to minimize material trapping; 2. Continuous monitoring of all discards and waste streams including sewer and stack effluents and contaminated discards; 3. Periodic physical inventories; 4. Careful sampling and analysis of all external feed streams and product streams; 5. Stringent control of the quality of analytical measurements and of sampling; 6. Assignment of the responsibility for material control to operating personnel rather than to accounting personnel; 7. Careful indoctrination of operating personnel. Information input to the nuclear material accounting system is achieved by means of a paper flow closely paralleling the physical flow of material, i.e. batch identity is maintained by means of individual batch cards and batch disposition is signalled by transfer of the batch card to the nuclear material accounting department. Most accounting operations are performed on electronic data-processing equipment. In addition to the records and calculations required for material balances, historical records of hold-up and concentration are kept to detect anomalies in input-output or inventory