Dynamic process model of a plutonium oxalate precipitator. Final report

Energy Technology Data Exchange (ETDEWEB)

Miller, C.L.; Hammelman, J.E.; Borgonovi, G.M.

1977-11-01

In support of LLL material safeguards program, a dynamic process model was developed which simulates the performance of a plutonium (IV) oxalate precipitator. The plutonium oxalate precipitator is a component in the plutonium oxalate process for making plutonium oxide powder from plutonium nitrate. The model is based on state-of-the-art crystallization descriptive equations, the parameters of which are quantified through the use of batch experimental data. The dynamic model predicts performance very similar to general Hanford oxalate process experience. The utilization of such a process model in an actual plant operation could promote both process control and material safeguards control by serving as a baseline predictor which could give early warning of process upsets or material diversion. The model has been incorporated into a FORTRAN computer program and is also compatible with the DYNSYS 2 computer code which is being used at LLL for process modeling efforts.

Dynamic process model of a plutonium oxalate precipitator. Final report

International Nuclear Information System (INIS)

Miller, C.L.; Hammelman, J.E.; Borgonovi, G.M.

1977-11-01

In support of LLL material safeguards program, a dynamic process model was developed which simulates the performance of a plutonium (IV) oxalate precipitator. The plutonium oxalate precipitator is a component in the plutonium oxalate process for making plutonium oxide powder from plutonium nitrate. The model is based on state-of-the-art crystallization descriptive equations, the parameters of which are quantified through the use of batch experimental data. The dynamic model predicts performance very similar to general Hanford oxalate process experience. The utilization of such a process model in an actual plant operation could promote both process control and material safeguards control by serving as a baseline predictor which could give early warning of process upsets or material diversion. The model has been incorporated into a FORTRAN computer program and is also compatible with the DYNSYS 2 computer code which is being used at LLL for process modeling efforts

Electronic spectra of plutonium ions in nitric acid and in lithium nitrate solutions

International Nuclear Information System (INIS)

Mekhail, F.M.

1987-01-01

The absorption spectra of plutonium ions in nitric acid have been described. There is a characteristic change in the absorption spectra of Pu v in lithium nitrate solutions. In 2 M-lithium nitrate a new peak at 969 nm and high absorption at 1200 nm are noticed. A decrease in the absorption by about 20% and the appearance of a new shoulder at 1120 nm in 6 M-lithium nitrate are found. There is no change in the spectrum in 4 M-lithium nitrate. The absorption spectra of plutonium ions in the spectral range 200 - 400 nm are interesting. All plutonium ions have an intense band in the region 250 - 260 nm as well as a less intense and rather diffuse band at 320 - 330 nm in lithium nitrate solutions the sharp band at 250 - 260 nm has disappeared. This suggests that this band is very sensitive to the environmental field. The band is probably produced by 5 F q → 5 f q-1 6 d transition as well as electron transfer. It is believed that the spectrum of Pu V at pH 6.5 represents the hydrolysis product Pu O 2 (O H). 9 fig., 4 tab

Plutonium Chemistry in the UREX+ Separation Processes

Energy Technology Data Exchange (ETDEWEB)

ALena Paulenova; George F. Vandegrift, III; Kenneth R. Czerwinski

2009-10-01

The project "Plutonium Chemistry in the UREX+ Separation Processes” is led by Dr. Alena Paulenova of Oregon State University under collaboration with Dr. George Vandegrift of ANL and Dr. Ken Czerwinski of the University of Nevada at Las Vegas. The objective of the project is to examine the chemical speciation of plutonium in UREX+ (uranium/tributylphosphate) extraction processes for advanced fuel technology. Researchers will analyze the change in speciation using existing thermodynamics and kinetic computer codes to examine the speciation of plutonium in aqueous and organic phases. They will examine the different oxidation states of plutonium to find the relative distribution between the aqueous and organic phases under various conditions such as different concentrations of nitric acid, total nitrates, or actinide ions. They will also utilize techniques such as X-ray absorbance spectroscopy and small-angle neutron scattering for determining plutonium and uranium speciation in all separation stages. The project started in April 2005 and is scheduled for completion in March 2008.

Reclamation of plutonium from pyrochemical processing residues

International Nuclear Information System (INIS)

Gray, L.W.; Gray, J.H.; Holcomb, H.P.; Chostner, D.F.

1987-04-01

Savannah River Laboratory (SRL), Savannah River Plant (SRP), and Rocky Flats Plant (RFP) have jointly developed a process to recover plutonium from molten salt extraction residues. These NaCl, KCL, and MgCl 2 residues, which are generated in the pyrochemical extraction of 241 Am from aged plutonium metal, contain up to 25 wt % dissolved plutonium and up to 2 wt % americium. The overall objective was to develop a process to convert these residues to a pure plutonium metal product and discardable waste. To meet this objective a combination of pyrochemical and aqueous unit operations was used. The first step was to scrub the salt residue with a molten metal (aluminum and magnesium) to form a heterogeneous ''scrub alloy'' containing nominally 25 wt % plutonium. This unit operation, performed at RFP, effectively separated the actinides from the bulk of the chloride salts. After packaging in aluminum cans, the ''scrub alloy'' was then dissolved in a nitric acid - hydrofluoric acid - mercuric nitrate solution at SRP. Residual chloride was separated from the dissolver solution by precipitation with Hg 2 (NO 3 ) 2 followed by centrifuging. Plutonium was then separated from the aluminum, americium and magnesium using the Purex solvent extraction system. The 241 Am was diverted to the waste tank farm, but could be recovered if desired

Evaluation of the Magnesium Hydroxide Treatment Process for Stabilizing PFP Plutonium/Nitric Acid Solutions

Energy Technology Data Exchange (ETDEWEB)

Gerber, Mark A.; Schmidt, Andrew J.; Delegard, Calvin H.; Silvers, Kurt L.; Baker, Aaron B.; Gano, Susan R.; Thornton, Brenda M.

2000-09-28

This document summarizes an evaluation of the magnesium hydroxide [Mg(OH)2] process to be used at the Hanford Plutonium Finishing Plant (PFP) for stabilizing plutonium/nitric acid solutions to meet the goal of stabilizing the plutonium in an oxide form suitable for storage under DOE-STD-3013-99. During the treatment process, nitric acid solutions bearing plutonium nitrate are neutralized with Mg(OH)2 in an air sparge reactor. The resulting slurry, containing plutonium hydroxide, is filtered and calcined. The process evaluation included a literature review and extensive laboratory- and bench-scale testing. The testing was conducted using cerium as a surrogate for plutonium to identify and quantify the effects of key processing variables on processing time (primarily neutralization and filtration time) and calcined product properties.

SEPARATION OF PLUTONIUM

Science.gov (United States)

Maddock, A.G.; Smith, F.

1959-08-25

A method is described for separating plutonium from uranium and fission products by treating a nitrate solution of fission products, uranium, and hexavalent plutonium with a relatively water-insoluble fluoride to adsorb fission products on the fluoride, treating the residual solution with a reducing agent for plutonium to reduce its valence to four and less, treating the reduced plutonium solution with a relatively insoluble fluoride to adsorb the plutonium on the fluoride, removing the solution, and subsequently treating the fluoride with its adsorbed plutonium with a concentrated aqueous solution of at least one of a group consisting of aluminum nitrate, ferric nitrate, and manganous nitrate to remove the plutonium from the fluoride.

Process for plutonium rextraction in aqueous solution from an organic solvent, especially for uranium plutonium partition

International Nuclear Information System (INIS)

Germain, M.; Gillet, B.; Pasquiou, J.Y.

1989-01-01

The organic solvent containing plutonium is contacted with an aqueous solution of a uranous salt, for instance uranous nitrate, and a hydroxylamine salt, for instance the nitrate. In these conditions uranous nitrate is a reducing agent of Pu III and hydroxylamine nitrate stabilizes Pu III and U IV in the aqueous phase. Performances are similar to these of the U IV-hydrazine nitrate without interference of hydrazine nitrate degradation products [fr

Plutonium scrap waste processing based on aqueous nitrate and chloride media

International Nuclear Information System (INIS)

Navratil, J.D.

1985-01-01

A brief review of plutonium scrap aqueous waste processing technology at Rocky Flats is given. Nitric acid unit operations include dissolution and leaching, anion exchange purification and precipitation. Chloride waste processing consists of cation exchange and carbonate precipitation. Ferrite and carrier precipitation waste treatment processes are also described. 3 figs

Radiolytic and thermal stability of selected plutonium salts containing nitrate groups

International Nuclear Information System (INIS)

Bryan, G.H.

1976-04-01

(Pu(NO 3 ) 4 . XH 2 O, K 2 Pu(NO 3 ) 6 , and (NH 4 ) 2 Pu(NO 3 ) 6 ) were studied to evaluate their ability to serve as shipping forms that meet criteria. The radiolytic gas evolution study eliminated (NH 4 ) 2 Pu(NO 3 ) 6 from further consideration. None of the compounds produced H 2 or O 2 in sufficient quantity to produce a flammable mixture, except Pu(NO 3 ) 4 . XH 2 O which produced O 2 and H 2 in a ratio that is above explosive limits after long storage time. The ammonium salt decomposition appears to be about the same as that observed upon heating of NH 4 NO 3 to produce N 2 , H 2 O, and nitrous oxides. Plutonium nitrate contains hydration water. This water is of some concern due to the production of hydrogen by alpha-radiolysis. Two waters of hydration appear to be the lower limit to which Pu(NO 3 ) 4 . XH 2 O may be taken before decomposition of the solid begins. TGA results indicate the simple nitrate (Pu(NO 3 ) 4 . XH 2 O) is somewhat less thermally stable than either the ammonium or potassium hexanitrato plutonate which detracts somewhat from its suitability as a shipping form. Maintaining large quantities of this compound with a high 238 Pu content (less than 1 percent 238 Pu) may require specially designed and larger containers to prevent thermal degradation and gas pressure buildup problems. The informaion was presented to plutonium processors; the final consensus of this group was that in spite of some thermal instability of Pu(NO 3 ) 4 . XH 2 O at fairly low temperatures, it was preferable to K 2 Pu(NO 3 ) 6 due to the additional waste disposal problems the potassium would present. (Pu(NO 3 ) 4 . XH 2 O also has several other advantages. A possible problem that could arise due to the variable weight of plutonium nitrate could be in plutonium accountability, but this would be prevented if the plutonium content of the solution prior to evaporation to the solid is known

Reduction stripping of plutonium (IV) with hydroxylamine nitrate in the centrifugal contactor

International Nuclear Information System (INIS)

Zhang Zefu; Ding Dachun; Fei Hongcheng

1989-01-01

Reduction stripping of Pu(IV) with hydroxylamine nitrate under conditions of Purex 3B contactor is carried out with 16-stage mixer-centrifugal settler in laboratory scale. Results show that the apparent recovery of Pu is more than 99.9%. But there is a great difference in the concentration of plutonium of every stage in extraction equipment at different temperature. For example, plutonium concentrations in organic phase of several stages near the organic effluent at 45-50 deg C are lower than that at 20-24 deg C. Thus, recovery efficiency of plutonium can be assured only for the former. However, plutonium will probably be lost, if operating conditions are fluctuated. In addition, plutonium accumulated in the extraction equipment increased appreciably for the case of 20-24 deg C. Therefore, in order to ensure recovery efficiency of Pu, reduction stripping of Pu should be made at higher temperature

The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

International Nuclear Information System (INIS)

Barney, G.S.; Cooper, T.D.

1994-01-01

Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125 degrees C and 250 degrees C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110 degrees C to 270 degrees C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252 degrees C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant

The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

Energy Technology Data Exchange (ETDEWEB)

Barney, G.S.; Cooper, T.D.

1994-06-01

Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125{degrees}C and 250{degrees}C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110{degrees}C to 270{degrees}C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252{degrees}C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant.

Risk of transporting plutonium dioxide and liquid plutonium nitrate by truck and rail

International Nuclear Information System (INIS)

Williams, L.D.; Hall, R.J.

1976-01-01

This paper presents an analysis method developed to assess the risk in the transportation of radioactive materials and demonstrates its application by summarizing the results of risk assessments of the shipment of plutonium dioxide powder and liquid plutonium nitrate by truck and rail in the United States. In the risk assessment, postulated release modes (series of events that could result in a release of radioactive material) are identified using the fault tree analysis method. Two categories of events that could contribute to a release during transportation are considered: accidents and substandard packaging conditions. The likelihood of basic events in these categories are determined from general accident experience, radioactive material shipping experience, engineering analysis, etc. Using the laws of combining probabilities, the likelihood of radioactive material release is then evaluated. Since accidents vary greatly in severity, the releases from these postulated events also vary in magnitude from the no release condition up through release of the entire package contents. The bases for the release quantities assigned to the postulated accidents sequences are discussed. The consequences of these postulated releases are evaluated based on the amount of plutonium estimated to be released to the environs, the probable weather conditions at the time of the accident, and population density downwind from the accident scene. The likelihood and consequence of postulated releases are coupled and expressed as a risk spectra. (author)

CSER 00-003: Criticality Safety Evaluation report for PFP Magnesium Hydroxide Precipitation Process for Plutonium Stabilization Glovebox 3

International Nuclear Information System (INIS)

LAN, J.S.

2000-01-01

This Criticality Safety Evaluation Report analyzes the stabilization of plutonium/uranium solutions in Glovebox 3 using the magnesium hydroxide precipitation process at PFP. The process covered are the receipt of diluted plutonium solutions into three precipitation tanks, the precipitation of plutonium from the solution, the filtering of the plutonium precipitate from the solution, the scraping of the precipitate from the filter into boats, and the initial drying of the precipitated slurry on a hot plate. A batch (up to 2.5 kg) is brought into the glovebox as plutonium nitrate, processed, and is then removed in boats for further processing. This CSER establishes limits for the magnesium hydroxide precipitation process in Glovebox 3 to maintain criticality safety while handling fissionable material

Chloride removal from plutonium alloy

International Nuclear Information System (INIS)

Holcomb, H.P.

1983-01-01

SRP is evaluating a program to recover plutonium from a metallic alloy that will contain chloride salt impurities. Removal of chloride to sufficiently low levels to prevent damaging corrosion to canyon equipment is feasible as a head-end step following dissolution. Silver nitrate and mercurous nitrate were each successfully used in laboratory tests to remove chloride from simulated alloy dissolver solution containing plutonium. Levels less than 10 ppM chloride were achieved in the supernates over the precipitated and centrifuged insoluble salts. Also, less than 0.05% loss of plutonium in the +3, +4, or +6 oxidation states was incurred via precipitate carrying. These results provide impetus for further study and development of a plant-scale process to recover plutonium from metal alloy at SRP

Literature review for oxalate oxidation processes and plutonium oxalate solubility

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-01

A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign.

LITERATURE REVIEW FOR OXALATE OXIDATION PROCESSES AND PLUTONIUM OXALATE SOLUBILITY

Energy Technology Data Exchange (ETDEWEB)

Nash, C.

2012-02-03

A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign. H Canyon plans to commence conversion of plutonium metal to low-fired plutonium oxide in 2012 for eventual use in the Mixed Oxide Fuel (MOX) Facility. The flowsheet includes sequential operations of metal dissolution, ion exchange, elution, oxalate precipitation, filtration, and calcination. All processes beyond dissolution will occur in HB-Line. The filtration step produces an aqueous filtrate that may have as much as 4 M nitric acid and 0.15 M oxalate. The oxalate needs to be removed from the stream to prevent possible downstream precipitation of residual plutonium when the solution is processed in H Canyon. In addition, sending the oxalate to the waste tank farm is undesirable. This report addresses the processing options for destroying the oxalate in existing H Canyon equipment.

An improved, computer-based, on-line gamma monitor for plutonium anion exchange process control

International Nuclear Information System (INIS)

Pope, N.G.; Marsh, S.F.

1987-06-01

An improved, low-cost, computer-based system has replaced a previously developed on-line gamma monitor. Both instruments continuously profile uranium, plutonium, and americium in the nitrate anion exchange process used to recover and purify plutonium at the Los Alamos Plutonium Facility. The latest system incorporates a personal computer that provides full-feature multichannel analyzer (MCA) capabilities by means of a single-slot, plug-in integrated circuit board. In addition to controlling all MCA functions, the computer program continuously corrects for gain shift and performs all other data processing functions. This Plutonium Recovery Operations Gamma Ray Energy Spectrometer System (PROGRESS) provides on-line process operational data essential for efficient operation. By identifying abnormal conditions in real time, it allows operators to take corrective actions promptly. The decision-making capability of the computer will be of increasing value as we implement automated process-control functions in the future. 4 refs., 6 figs

Contribution to the study of the process of purification of plutonium by extraction with trilaurylamine

International Nuclear Information System (INIS)

Saey, Jean-Claude

1966-01-01

This work addresses the process of plutonium purification which uses trilaurylamine nitrate. In order to use this nitrate in its solid state and at ordinary temperature, a secondary solvent must be added which must have some properties: low volume mass and viscosity, high boiling and ignition temperatures, rather low miscibility with water, high stability in front of joint actions of nitric acid and radiations, and no reaction with the alkylammonium nitrate and the complex. Thus, the author addresses phenomena of immiscibility and identifies some important molecular characteristics which could lead to the selection of another secondary solvent than dodecane. The decalin seem interesting and its behaviour is studied. A mixing of dodecane and decalin is used as extraction mixing. The obtained results are discussed. Finally, the author notices that using this mixing in the plutonium purification process results in a large increase of metal concentrations and a decrease of risks of crystallisation, without any major drawback, in a continuously operating micro-industrial installation

Precipitation of plutonium (III) oxalate and calcination to plutonium oxide

International Nuclear Information System (INIS)

Esteban, A.; Orosco, E.H.; Cassaniti, P.; Greco, L.; Adelfang, P.

1989-01-01

The plutonium based fuel fabrication requires the conversion of the plutonium nitrate solution from nuclear fuel reprocessing into pure PuO2. The conversion method based on the precipitation of plutonium (III) oxalate and subsequent calcination has been studied in detail. In this procedure, plutonium (III) oxalate is precipitated, at room temperature, by the slow addition of 1M oxalic acid to the feed solution, containing from 5-100 g/l of plutonium in 1M nitric acid. Before precipitation, the plutonium is adjusted to trivalent state by addition of 1M ascorbic acid in the presence of an oxidation inhibitor such as hydrazine. Finally, the precipitate is calcinated at 700 deg C to obtain PuO2. A flowsheet is proposed in this paper including: a) A study about the conditions to adjust the plutonium valence. b) Solubility data of plutonium (III) oxalate and measurements of plutonium losses to the filtrate and wash solution. c) Characterization of the obtained products. Plutonium (III) oxalate has several potential advantages over similar conversion processes. These include: 1) Formation of small particle sizes powder with good pellets fabrication characteristics. 2) The process is rather insensitive to most process variables, except nitric acid concentration. 3) Ambient temperature operations. 4) The losses of plutonium to the filtrate are less than in other conversion processes. (Author) [es

Vapor-liquid equilibria for nitric acid-water and plutonium nitrate-nitric acid-water solutions

International Nuclear Information System (INIS)

Maimoni, A.

1980-01-01

The liquid-vapor equilibrium data for nitric acid and nitric acid-plutnonium nitrate-water solutions were examined to develop correlations covering the range of conditions encountered in nuclear fuel reprocessing. The scanty available data for plutonium nitrate solutions are of poor quality but allow an order of magnitude estimate to be made. A formal thermodynamic analysis was attempted initially but was not successful due to the poor quality of the data as well as the complex chemical equilibria involved in the nitric acid and in the plutonium nitrate solutions. Thus, while there was no difficulty in correlating activity coefficients for nitric acid solutions over relatively narrow temperature ranges, attempts to extend the correlations over the range 25 0 C to the boiling point were not successful. The available data were then analyzed using empirical correlations from which normal boiling points and relative volatilities can be obtained over the concentration ranges 0 to 700 g/l Pu, 0 to 13 M nitric acid. Activity coefficients are required, however, if estimates of individual component vapor pressures are needed. The required ternary activity coefficients can be approximated from the correlations

SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

Science.gov (United States)

Spence, R.; Lister, M.W.

1958-12-16

Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

Coordinated safeguards for materials management in a uranium--plutonium nitrate-to-oxide coconversion facility: Coprecal

International Nuclear Information System (INIS)

Dayem, H.A.; Cobb, D.D.; Dietz, R.J.; Hakkila, E.A.; Kern, E.A.; Schelonka, E.P.; Shipley, J.P.; Smith, D.B.

1979-02-01

This report describes the conceptual design of an advanced materials-management system for safeguarding special nuclear materials in a uranium--plutonium nitrate-to-oxide coconversion facility based on the Coprecal process. Design concepts are presented for near real-time (dynamic) accountability by forming dynamic materials balances from information provided by chemical and nondestructive analyses and from process-control instrumentation. Modeling and simulation techniques are used to compare the sensitivities of proposed dynamic materials accounting strategies to both abrupt and protracted diversion. The safeguards implications of coconversion as well as some unique features of the reference process are discussed and design criteria are identified to improve the safeguardability of the Coprecal coconversion process

The research of technology and equipment for a microwave denitration process of the uranyl nitrate solution

International Nuclear Information System (INIS)

Bao Weimin; Wang Xuejun; Ma Xuquan; Shi Miaoyi; Zhang Zhicheng; Bao Zhu Tian.

1991-01-01

In order to improve the present process of converting the plutonium nitrate into oxide powder in the nuclear fuel cycle, a new conversion process for the direct denitration using microwave heating has been developed. Microwave denitration is based on intramolecular polarization of a material in electric field and has no need of a process of heat transfer during microwave heating, so that the whole material can be heated quickly and uniformly. The thermal decomposition reactions of Pu, U, Th and RE nitrate have been analyzed and compared. The uranyl nitrate solution was chosen as imitative plutonium nitrate solution. The performance parameters ε r tanδ of U, Th and RE nitrate and oxide in microwave field were measured. The data obtained show that all of them could absorb microwave energy well and cause heating decomposition reactions. The microwave denitration test unit was designed and made. Denitration tests for rare-earths nitrate and uranyl nitrate solutions were performed. It could be completed in one step that the uranyl nitrate solution was evaporated, dryed and denitrated in a vessel. The denitrated products are a porous lump and easy to scrape off from the denitration vessel. The main forms of the products UO 3 ·0.8H 2 O and U 3 O 8 which have excellent powder properties. The capacity of the denitration unit is 1.3 kg UO 3 /h. According to the experimental results the simplicity, feasibility and good repeatability of the process have been fully proved. The unit operates easily and is adaptable to conversion of nitrate in nuclear fuel cycle. (author)

Continuous plutonium(IV) oxalate precipitation, filtration, and calcination process. [From product streams from Redox, Purex, or Recuplex solvent extraction plants

Energy Technology Data Exchange (ETDEWEB)

Beede, R L

1956-09-27

A continuous plutonium (IV) oxalate precipitation, filtration, and calcination process has been developed. Continuous and batch decomposition of the oxalate in the filtrates has been demonstrated. The processes have been demonstrated in prototype equipment. Plutonium (IV) oxalate was precipitated continuously at room temperature by the concurrent addition of plutonium (IV) nitrate feed and oxalic acid into the pan of a modified rotary drum filter. The plutonium (IV) oxalate was calcined to plutonium dioxide, which could be readily hydrofluorinated. Continuous decomposition of the oxalate in synthetic plutonium (IV) oxalate filtrates containing plutonium (IV) oxalate solids was demonstrated using co-current flow in a U-shaped reactor. Feeds containing from 10 to 100 g/1 Pu, as plutonium (IV) nitrate, and 1.0 to 6.5 M HNO/sub 3/, respectively, can be processed. One molar oxalic acid is used as the precipitant. Temperatures of 20 to 35/sup 0/C for the precipitation and filtration are satisfactory. Plutonium (IV) oxalate can be calcined at 300 to 400/sup 0/C in a screw-type drier-calciner to plutonium dioxide and hydrofluorinated at 450 to 550/sup 0/C. Plutonium dioxide exceeding purity requirements has been produced in the prototype equipment. Advantages of continuous precipitation and filtration are: uniform plutonium (IV) oxalate, improved filtration characteristics, elimination of heating and cooling facilities, and higher capacities through a single unit. Advantages of the screw-type drier-calciner are the continuous production of an oxide satisfactory for feed for the proposed plant vibrating tube hydrofluorinator, and ease of coupling continuous precipitation and filtration to this proposed hydrofluorinator. Continuous decomposition of oxalate in filtrates offers advantages in decreasing filtrate storage requirements when coupled to a filtrate concentrator. (JGB)

History and stabilization of the Plutonium Finishing Plant (PFP) complex, Hanford Site

Energy Technology Data Exchange (ETDEWEB)

Gerber, M.S., Fluor Daniel Hanford

1997-02-18

The 231-Z Isolation Building or Plutonium Metallurgy Building is located in the Hanford Site`s 200 West Area, approximately 300 yards north of the Plutonium Finishing Plant (PFP) (234-5 Building). When the Hanford Engineer Works (HEW) built it in 1944 to contain the final step for processing plutonium, it was called the Isolation Building. At that time, HEW used a bismuth phosphate radiochemical separations process to make `AT solution,` which was then dried and shipped to Los Alamos, New Mexico. (AT solution is a code name used during World War II for the final HEW product.) The process was carried out first in T Plant and the 224-T Bulk Reduction Building and B Plant and the 224-B Bulk Reduction Building. The 224-T and -B processes produced a concentrated plutonium nitrate stream, which then was sent in 8-gallon batches to the 231-Z Building for final purification. In the 231-Z Building, the plutonium nitrate solution underwent peroxide `strikes` (additions of hydrogen peroxide to further separate the plutonium from its carrier solutions), to form the AT solution. The AT solution was dried and shipped to the Los Alamos Site, where it was made into metallic plutonium and then into weapons hemispheres.` The 231-Z Building began `hot` operations (operations using radioactive materials) with regular runs of plutonium nitrate on January 16, 1945.

Comparison of predicted with observed biokinetics of inhaled plutonium nitrate and gadolinium oxide in humans

International Nuclear Information System (INIS)

Hodgson, A.; Shutt, A.L.; Etherington, G.; Hodgson, S.A.; Rance, E.; Stradling, G.N.; Youngman, M.J.; Ziesenis, A.; Kreyling, W.G.

2003-01-01

The absorption kinetics to blood of plutonium and gadolinium after inhalation as nitrate and oxide in humans and animals has been studied. For each material, values describing the time dependence of absorption were derived from the studies in animals and used with the ICRP human respiratory tract model to predict lung retention and cumulative amounts to blood for the volunteers inhaling the same materials. Comparison with the observed behaviour in the volunteers suggests that absorption of plutonium and gadolinium is reasonably species independent, and that data obtained from animal studies can be used to assess their biokinetic behaviour in humans. (author)

An investigation to compare the performance of methods for the determination of free acid in highly concentrated solutions of plutonium and uranium nitrate

International Nuclear Information System (INIS)

Crossley, D.

1980-08-01

An investigation has been carried out to compare the performance of the direct titration method and the indirect mass balance method, for the determination of free acid in highly concentrated solutions of uranium nitrate and plutonium nitrate. The direct titration of free acid with alkali is carried out in a fluoride medium to avoid interference from the hydrolysis of uranium or plutonium, while free acid concentration by the mass balance method is obtained by calculation from the metal concentration, metal valency state, and total nitrate concentration in a sample. The Gran plot end-point prediction technique has been used extensively in the investigation to gain information concerning the hydrolysis of uranium and plutonium in fluoride media and in other complexing media. The use of the Gran plot technique has improved the detection of the end-point of the free acid titration which gives an improvement in the precision of the determination. The experimental results obtained show that there is good agreement between the two methods for the determination of free acidity, and that the precision of the direct titration method in a fluoride medium using the Gran plot technique to detect the end-point is 0.75% (coefficient of variation), for a typical separation plant plutonium nitrate solution. The performance of alternative complexing agents in the direct titration method has been studied and is discussed. (author)

Minimum critical values of uranyl and plutonium nitrate solutions calculated by various routes of the french criticality codes system CRISTAL using the new isopiestic nitrate density law

International Nuclear Information System (INIS)

Anno, Jacques; Rouyer, Veronique; Leclaire, Nicolas

2003-01-01

This paper provides for various cases of 235 U enrichment or Pu isotopic vectors, and different reflectors, new minimum critical values of uranyl nitrate and plutonium nitrate solutions (H + =0) obtained by the standard IRSN calculation route and the new isopiestic density laws. Comparisons are also made with other more accurate routes showing that the standard one's results are most often conservative and usable for criticality safety assessments. (author)

Role of ion transfer membrane in the production of uranous nitrate

International Nuclear Information System (INIS)

Nair, M.K.T.; Singh, R.K.; Bajpai, D.D.; Venugopalan, A.K.; Singh, R.R.; Gurba, P.B.; Thomas, Mathew

1992-01-01

In Purex process, plutonium and uranium are co-extracted into organic phase and these are partitioned by reducing Pu(IV) to Pu(III) using hydrazine stabilized uranous nitrate solution. Usually, uranous nitrate is added in much higher quantity than the stoichiometric requirement to effect complete reduction of plutonium. In conventional electrolytic cells only 60 to 70% of uranyl to uranous conversion is achieved. Use of this solution results in dilution of plutonium product. In addition to this, each externally fed uranous nitrate batch increases uranium processing load and affects the plant throughput. In order to keep the additional uranium processing load to a minimum, it is necessary to increase the uranous content to near cent percent level in the externally fed uranous nitrate solution. The studies carried out at PREFRE (Power Reactor Fuel Reprocessing) laboratory have shown that it is possible to produce concentrated uranous nitrate solution, nearly free from uranyl nitrate, by using a cation exchange membrane. This paper describes the development work carried out at PREFRE plant, Tarapur for production of cent percent uranous nitrate solution. Development of electrolytic cells for uranous production, from laboratory scale to pilot plant scale, has been explained. (author). 24 refs., 8 figs., 8 tabs

Recovery of plutonium by pyroredox processing

International Nuclear Information System (INIS)

McNeese, J.A.; Bowersox, D.F.; Christensen, D.C.

1985-09-01

Using pyrochemical oxidation and reduction, we have developed a process to recover the plutonium in impure scrap with less than 95% plutonium. This plutonium metal was further purified by pyrochemical electrorefining. During development of the procedures, depleted electrorefining anodes were processed, and over 80% of the plutonium was recovered as high-purity metal in one electrorefining cycle. Over 40 kg of plutonium has been recovered from 55 kg of impure anodes with our procedures. 6 refs., 7 figs., 4 tabs

Recovery of plutonium by pyroredox processing

International Nuclear Information System (INIS)

McNeese, J.A.; Bowersox, D.F.; Christensen, D.C.

1985-01-01

Using pyrochemical oxidation and reduction, we have developed a process to recover the plutonium in impure scrap with less than 95% plutonium. This plutonium metal was further purified by pyrochemical electrorefining. During development of the procedures, depleted electrorefining anodes were processed, and over 80% of the plutonium was recovered as high-purity metal in one electrorefining cycle. Over 40 kg of plutonium has been recovered from 55 kg of impure anodes with our procedures. 6 refs., 2 figs., 5 tabs

Effects of inhaled plutonium nitrate on bone and liver in dogs

International Nuclear Information System (INIS)

Dagle, G.E.; Weller, R.E.; Watson, C.R.; Buschbom, R.L.

1994-04-01

The life-span biological effects of inhaled soluble, alpha-emitting radionuclides deposited in the skeleton and liver were studied in 5 groups of 20 beagles exposed to initial lung depositions ranging from 0.48 to 518 Bq/g of lung. Average plutonium amounts in the lungs decreased to approximately 1% of the final body deposition in dogs surviving 5 years or more; more than 90% of the final depositions accumulated in the liver and skeleton. The liver-to-skeletal ratio of deposited plutonium was 0.83. The incidence of bone tumors, primarily osteogenic sarcomas causing early mortality, at final group average skeletal depositions of 15.8, 2.1, and 0.5 Bq/g was, respectively, 85%, 50%, and 5%; there were no bone tumors in exposure groups with mean average depositions lower than 0.5 Bq/g. Elevated serum liver enzyme levels were observed in exposure groups down to 1.3 Bq/g. The incidence of liver tumors at final group average liver depositions of 6.9, 1.3, 0.2, and 0.1 Bq/g, was, respectively, 25%, 15%, 15%, and 15%; one hepatoma occurred among 40 control dogs. The risk of the liver cancer produced by inhaled plutonium nitrate was difficult to assess due to the competing risks of life shortening from lung and bone tumors

Uranous nitrate production for purex process applications using PtO2 catalyst and H2/H2-gas mixtures

International Nuclear Information System (INIS)

Sreenivasa Rao, K.; Shyamali, R.; Narayan, C.V.; Patil, A.R.; Jambunathan, U.; Ramanujam, A.; Kansara, V.P.

2003-04-01

In the Purex process of spent fuel reprocessing. the twin objectives- decontamination and partitioning are achieved by extracting uranium (VI) and plutonium (IV) together in the solvent 30% TBP-dodecane and then selectively reducing Pu (IV) to Pu (III) in which valency it is least extractable in the solvent. Uranous nitrate stabilized with hydrazine nitrate is the widely employed partitioning agent. The conventional method of producing U(IV) is by the electrolytic reduction of uranyl nitrate with hydrazine nitrate as uranous ion stabilizer. Tre percentage conversion of U(VI) to U(IV) obtained in this method is 50 -60 %. The use of this solution as partitioning agent leads not only to the dilution of the plutonium product but also to increase in uranium processing load by each externally fed uranous nitrate batch. Also the oxide coating of the anode, TSIA (Titanium Substrate Insoluble Anode) wears out after a certain period of operation. This necessitates recoating which is quite cumbersome considering the amount of the decontamination involved. An alternative to the conventional electrolytic method of reduction of uranyl nitrate to uranous nitrate was explored at FRD laboratory .The studies have revealed that near 100% uranous nitrate can be produced by reducing uranyl nitrate with H 2 gas or H 2 (8%)- Ar/N 2 gas mixture in presence of PtO 2 catalyst. This report describes the laboratory scale studies carried out to optimize the various parameters. Based on these studies reduction of uranyl nitrate on a pilot plant scale was carried out. The design and operation of the reductor column and also the various studies carried out in the pilot plant studies are discussed. Near 100% conversion of uranyl nitrate to uranous nitrate and also the redundancy of supply of electrical energy make this process a viable alternative to the existing electrolytic method. (author)

Fused salt processing of impure plutonium dioxide to high-purity plutonium metal

International Nuclear Information System (INIS)

Mullins, L.J.; Christensen, D.C.; Babcock, B.R.

1982-01-01

A process for converting impure plutonium dioxide (approx. 96% pure) to high-purity plutonium metal (>99.9%) was developed. The process consists of reducing the oxide to an impure plutonium metal intermediate with calcium metal in molten calcium chloride. The impure intermediate metal is cast into an anode and electrorefined to produce high-purity plutonium metal. The oxide reduction step is being done now on a 0.6-kg scale with the resulting yield being >99.5%. The electrorefining is being done on a 4.0-kg scale with the resulting yield being 80 to 85%. The purity of the product, which averages 99.98%, is essentially insensitive to the purity of the feed metal. The yield, however, is directly dependent on the chemical composition of the feed. To date, approximately 250 kg of impure oxide has been converted to pure metal by this processing sequence. The availability of impure plutonium dioxide, together with the need for pure plutonium metal, makes this sequence a valuable plutonium processing tool

Potentiometric determination of free nitric-acid in trilaurylamine solutions containing plutonium nitrate; Dosage potentiometrique de l'acidite nitrique libre dans les solutions organiques de trilaurylamine

Energy Technology Data Exchange (ETDEWEB)

Perez, J J; Saey, J C [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1965-07-01

A potentiometric method of determination of the free nitric acid in trilaurylamine solutions containing plutonium or thorium nitrates is described. The potentiometric titration is carried out in a mixture of benzene and 1,2-dichloro ethane with a standard solution of trilaurylamine as the titrant. When thorium nitrate is present the metal complex is not dissociated then the titration has a single end-point. In the case of plutonium nitrate the partial dissociation of the plutonium complex corresponds to a second point. The experimental error in duplicate analyses of 50 samples is about 1 per cent for free acid concentrations in the range of 0,03 to 0,1 N and plutonium concentrations between 1 to 5 g/l. (authors) [French] Une methode potentiometrique de dosage de l'acidite nitrique libre dans les solutions de trilaurylamine contenant un complexe de plutonium ou de thorium est decrite. La potentiometrie est effectuee en prenant comme base titrante la trilaurylamine et comme milieu de dilution un melange de benzene et de 1,2 dichloroethane. Dans le cas du thorium, le complexe organometallique n'est pas deplace et la courbe de titrage presente un seul point d'inflexion. Dans le cas du plutonium le complexe est partiellement dissocie ce qui correspond a un second saut de potentiel. La moyenne des erreurs experimentales sur 50 echantillons doses a ete d'environ {+-} 1 pour cent sur l'acide libre. Les solutions experimentees contenaient de 0,03 a 0,1 N en acide et de 1 a 5 g/l en plutonium. (auteurs)

Standard test method for plutonium assay by plutonium (III) diode array spectrophotometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method describes the determination of total plutonium as plutonium(III) in nitrate and chloride solutions. The technique is applicable to solutions of plutonium dioxide powders and pellets (Test Methods C 697), nuclear grade mixed oxides (Test Methods C 698), plutonium metal (Test Methods C 758), and plutonium nitrate solutions (Test Methods C 759). Solid samples are dissolved using the appropriate dissolution techniques described in Practice C 1168. The use of this technique for other plutonium-bearing materials has been reported (1-5), but final determination of applicability must be made by the user. The applicable concentration range for plutonium sample solutions is 10–200 g Pu/L. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropria...

Addressing mixed waste in plutonium processing

International Nuclear Information System (INIS)

Christensen, D.C.; Sohn, C.L.; Reid, R.A.

1991-01-01

The overall goal is the minimization of all waste generated in actinide processing facilities. Current emphasis is directed toward reducing and managing mixed waste in plutonium processing facilities. More specifically, the focus is on prioritizing plutonium processing technologies for development that will address major problems in mixed waste management. A five step methodological approach to identify, analyze, solve, and initiate corrective action for mixed waste problems in plutonium processing facilities has been developed

Selecting a plutonium vitrification process

Energy Technology Data Exchange (ETDEWEB)

Jouan, A. [Centre d`Etudes de la Vallee du Rhone, Bagnols sur Ceze (France)

1996-05-01

Vitrification of plutonium is one means of mitigating its potential danger. This option is technically feasible, even if it is not the solution advocated in France. Two situations are possible, depending on whether or not the glass matrix also contains fission products; concentrations of up to 15% should be achievable for plutonium alone, whereas the upper limit is 3% in the presence of fission products. The French continuous vitrification process appears to be particularly suitable for plutonium vitrification: its capacity is compatible with the required throughout, and the compact dimensions of the process equipment prevent a criticality hazard. Preprocessing of plutonium metal, to convert it to PuO{sub 2} or to a nitric acid solution, may prove advantageous or even necessary depending on whether a dry or wet process is adopted. The process may involve a single step (vitrification of Pu or PuO{sub 2} mixed with glass frit) or may include a prior calcination step - notably if the plutonium is to be incorporated into a fission product glass. It is important to weigh the advantages and drawbacks of all the possible options in terms of feasibility, safety and cost-effectiveness.

Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 These test methods cover procedures for the chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Plutonium by Controlled-Potential Coulometry Plutonium by Amperometric Titration with Iron(II) Plutonium by Diode Array Spectrophotometry Free Acid by Titration in an Oxalate Solution 8 to 15 Free Acid by Iodate Precipitation-Potentiometric Titration Test Method 16 to 22 Uranium by Arsenazo I Spectrophotometric Test Method 23 to 33 Thorium by Thorin Spectrophotometric Test Method 34 to 42 Iron by 1,10-Phenanthroline Spectrophotometric Test Method 43 to 50 Impurities by ICP-AES Chloride by Thiocyanate Spectrophotometric Test Method 51 to 58 Fluoride by Distillation-Spectrophotometric Test Method 59 to 66 Sulfate by Barium Sulfate Turbidimetric Test Method 67 to 74 Isotopic Composition by Mass Spectrom...

Plutonium Oxide Process Capability Work Plan

Energy Technology Data Exchange (ETDEWEB)

Meier, David E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2014-02-28

Pacific Northwest National Laboratory (PNNL) has been tasked to develop a Pilot-scale Plutonium-oxide Processing Unit (P3U) providing a flexible capability to produce 200g (Pu basis) samples of plutonium oxide using different chemical processes for use in identifying and validating nuclear forensics signatures associated with plutonium production. Materials produced can also be used as exercise and reference materials.

Chromium in aqueous nitrate plutonium process streams: Corrosion of 316 stainless steel and chromium speciation

International Nuclear Information System (INIS)

Smith, W.H.; Purdy, G.M.

1995-01-01

This study was undertaken to determine if chromium(+6) could exist in plutonium process solutions under normal operating conditions. Four individual reactions were studied: the rate of dissolution of stainless steel, which is the principal source of chromium in process solutions; the rate of oxidation of chromium(+3) to chromium(+6) by nitric acid; and the reduction of chromium(+6) back to chromium(+3) by reaction with stainless steel and with oxalic acid. The stainless steel corrosion rate was found to increase with increasing nitric acid concentration, increasing hydrofluoric acid concentration, and increasing temperature. Oxidation of chromium(+3) to chromium(+6) was negligible at room temperature and only became significant in hot concentrated nitric acid. The rate of reduction of chromium(+6) back to chromium(+3) by reaction with stainless steel or oxalic acid was found to be much greater than the rate of the reverse oxidation reaction. Based on these findings and taking into account normal operating conditions, it was determined that although there would be considerable chromium in plutonium process streams it would rarely be found in the (+6) oxidation state and would not exist in the (+6) state in the final process waste solutions

Chromium in aqueous nitrate plutonium process streams: Corrosion of 316 stainless steel and chromium speciation

International Nuclear Information System (INIS)

Smith, W.H.; Purdy, G.

1995-01-01

This study was undertaken to determine if chromium +6 could exist in plutonium process solutions under normal operating conditions. Four individual reactions were studied: the rate of dissolution of stainless steel, which is the principal source of chromium in process solutions; the rate of oxidation of chromium +3 to chromium +6 by nitric. acid; and the reduction of chromium +6 back to chromium +3 by reaction with stainless steel and with oxalic acid. The stainless steel corrosion rate was found to increase with increasing nitric acid concentration, increasing hydrofluoric acid concentration, and increasing temperature. Oxidation of chromium +3 to chromium +6 was negligible at room temperature and only became significant in hot concentrated nitric acid. The rate of reduction of chromium +6 back to chromium +3 by reaction with stainless steel or oxalic acid was found to be much greater than the rate of the reverse oxidation reaction. Based on these findings and taking into account normal operating conditions, it was determined that although there would be considerable chromium in plutonium process streams it would rarely be found in the +6 oxidation state and would not exist in the +6 state in the final process waste solutions

Method to manufacture a nuclear fuel from uranium-plutonium monocarbide or uranium-plutonium mononitride

International Nuclear Information System (INIS)

Krauth, A.; Mueller, N.

1977-01-01

Pure uranium carbide or nitride is converted with plutonium oxide and carbon (all in powder form) to uranium-plutonium monocarbide or mononitride by cold pressing and sintering at about 1600 0 C. Pure uranium carbide or uranium nitride powder is firstly prepared without extensive safety measures. The pure uranium carbide or nitride powder can also be inactivated by using chemical substances (e.g. stearic acid) and be handled in air. The sinterable uranium carbide or nitride powder (or also granulate) is then introduced into the plutonium line and mixed with a nonstoichiometrically adjusted, prereacted mixture of plutonium oxide and carbon, pressed to pellets and reaction sintered. The surface of the uranium-plutonium carbide (higher metal content) can be nitrated towards the end of the sinter process in a stream of nitrogen. The protective layer stabilizes the carbide against the water and oxygen content in air. (IHOE) [de

Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

Energy Technology Data Exchange (ETDEWEB)

ULLAH, M K

2001-02-26

The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

Study of the solubility of plutonyl nitrate with a view to its extraction by a solvent; Etude de la solubilite du nitrate de plutonyle en vue de son extraction par solvant

Energy Technology Data Exchange (ETDEWEB)

Vergnaud, G [Commissariat a l' Energie Atomique, Fontenay aux Roses (France). Centre d' Etudes Nucleaires

1965-09-01

The research covers the determination of the partition coefficient of plutonyl nitrate and their application to the industrial treatment of plutonium. The solvent used is a tri-n-butyl phosphate solution diluted to 30 per cent in dodecane. The parameters which have been studied and which can affect the extraction of the plutonyl nitrate are: the plutonyl nitrate concentration; the nitric acid concentration; the uranyl nitrate concentration; the tetravalent plutonium nitrate concentration. The industrial application has been studied using a battery of mixer-separators. The operational conditions are defined in the case of a conventional plutonium separation plant. The results obtained, for a higher plutonyl nitrate concentration than that actually planned, are satisfactory and encouraging as far as the operation is concerned. Only the presence of polymerized tetravalent plutonium can adversely affect the satisfactory extraction of the plutonium. (author) [French] L'etude porte sur les coefficients de partage du nitrate de plutonyle et leur application a l'elaboration industrielle du plutonium. Le solvant utilise est une solution de phosphate de tri-n-butyle, dilue a 30 pour cent dans le dodecane. Les parametres etudies et pouvant influencer l'extraction du nitrate de plutonyle, par ce solvant, sont: concentration du nitrate de plutonyle; concentration de l'acide nitrique; concentration du nitrate d'uranyle; concentration au nitrate de plutonium tetravalent. L'application industrielle est etudiee au moyen d'une batterie de melangeurs-decanteurs. Les conditions operatoires sont definies d'apres un schema classique d'usine de separation de plutonium. Les resultats obtenus, pour une concentration du nitrate de plutonyle superieure a celle reellement prevue, sont satisfaisants et encourageants pour l'exploitation. Seule, la presence de plutonium tetraralent polymerise, peut nuire a une bonne extraction du plutonium. (auteur)

Study of the solubility of plutonyl nitrate with a view to its extraction by a solvent; Etude de la solubilite du nitrate de plutonyle en vue de son extraction par solvant

Energy Technology Data Exchange (ETDEWEB)

Vergnaud, G. [Commissariat a l' Energie Atomique, Fontenay aux Roses (France). Centre d' Etudes Nucleaires

1965-09-01

The research covers the determination of the partition coefficient of plutonyl nitrate and their application to the industrial treatment of plutonium. The solvent used is a tri-n-butyl phosphate solution diluted to 30 per cent in dodecane. The parameters which have been studied and which can affect the extraction of the plutonyl nitrate are: the plutonyl nitrate concentration; the nitric acid concentration; the uranyl nitrate concentration; the tetravalent plutonium nitrate concentration. The industrial application has been studied using a battery of mixer-separators. The operational conditions are defined in the case of a conventional plutonium separation plant. The results obtained, for a higher plutonyl nitrate concentration than that actually planned, are satisfactory and encouraging as far as the operation is concerned. Only the presence of polymerized tetravalent plutonium can adversely affect the satisfactory extraction of the plutonium. (author) [French] L'etude porte sur les coefficients de partage du nitrate de plutonyle et leur application a l'elaboration industrielle du plutonium. Le solvant utilise est une solution de phosphate de tri-n-butyle, dilue a 30 pour cent dans le dodecane. Les parametres etudies et pouvant influencer l'extraction du nitrate de plutonyle, par ce solvant, sont: concentration du nitrate de plutonyle; concentration de l'acide nitrique; concentration du nitrate d'uranyle; concentration au nitrate de plutonium tetravalent. L'application industrielle est etudiee au moyen d'une batterie de melangeurs-decanteurs. Les conditions operatoires sont definies d'apres un schema classique d'usine de separation de plutonium. Les resultats obtenus, pour une concentration du nitrate de plutonyle superieure a celle reellement prevue, sont satisfaisants et encourageants pour l'exploitation. Seule, la presence de plutonium tetraralent polymerise, peut nuire a une bonne extraction du plutonium

Cigarette smoke and plutonium

International Nuclear Information System (INIS)

Filipy, R.E.

1985-01-01

Autoradiographic techniques with liquid photographic emulsion and cellulose nitrate track-etch film are being used to investigate the spatial distribution of inhaled plutonium in the lungs of beagle dogs exposed to cigarette smoke or to the plutonium aerosol only. More plutonium than expected was detected on the inner surfaces of bronchi, and particles were observed beneath the bronchial mucosa. 2 figures, 2 tables

Plutonium recovery from spent glass fiber paper fine air filter

International Nuclear Information System (INIS)

Rovnyj, S.I.; Guzhavin, V.I.; Pyatin, N.P.; Evlanov, D.S.

2002-01-01

Investigations into the realizing technology of plutonium recovery from waste glass paper filters of fine purification were conducted. Two process schemes involving the nitro-fluoro-acid treatment of glass paper in the mixture of nitric and hydrofluoric acids and the previous alkali treatment of glass paper with the following nitro-fluoro-acid leaching of plutonium from pulp by the mixture of nitric and hydrofluoric acids were developed. Alkali, nitrate solutions and insoluble precipitants were analyzed for plutonium content [ru

Continuous precipitation process of plutonium salts

International Nuclear Information System (INIS)

Richard, P.

1967-03-01

This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [fr

COGEMA Experience in Uranous Nitrate Preparation

International Nuclear Information System (INIS)

Tison, E.; Bretault, Ph.

2006-01-01

Separation and purification of plutonium by PUREX process is based on a sequence of extraction and back extraction which requires reducing plutonium Pu IV (extractable form) into Pu III (inextractable form) Different reducers can be used to reduce Pu IV into Pu III. Early plants such as that for Magnox fuel at Sellafield used ferrous sulfamate while UP 1 at Marcoule used uranous sulfamate. These reducers are efficient and easy to prepare but generates ferric and/or sulphate ions and so complicates management of the wastes from the plutonium purification cycle. Recent plants such as UP3 and UP2 800 at La Hague, THORP at Sellafield, and RRP at Rokkasho Mura (currently under tests) use uranous nitrate (U IV) stabilized by hydrazinium nitrate (N 2 H 5 NO 3 ) and hydroxyl ammonium nitrate (HAN). In the French plants, uranous nitrate is used in U-Pu separation and alpha barrier and HAN is used in Pu purification. Compared to sulfamate, U IV does not generate extraneous chemical species and uranyl nitrate (U VI) generated by reducing Pu IV follows the main uranium stream. More over uranous nitrate is prepared from reprocessed purified uranyl nitrate taken at the outlet of the reprocessing plant. Hydrazine and HAN offer the advantage to be salt-free reagents. Uranous nitrate can be generated either by electrolysis or by catalytic hydrogenation process. Electrolytic process has been implemented in early plant UP 1 at Marcoule (when changing reducer from uranous sulfamate to uranous nitrate) and was used again in UP2 plant at La Hague. However, the electrolytic process presented several disadvantages such as a low conversion rate and problems associated with the use of mercury. Electrolysis cells with no mercury were developed for the Eurochemic plant in Belgium and then implemented in the first Japanese reprocessing plant in Tokai-Mura. But finally, in 1975, the electrolytic process was abandoned in favor of the catalytic hydrogenation process developed at La Hague. The

Electrolytic production of uranous nitrate

International Nuclear Information System (INIS)

Orebaugh, E.G.; Propst, R.C.

1980-04-01

Efficient production of uranous nitrate is important in nuclear fuel reprocessing because U(IV) acts as a plutonium reductant in solvent extraction and can be coprecipitated with plutonium and/or throium as oxalates during fuel reprocessing. Experimental conditions are described for the efficient electrolytic production of uranous nitrate for use as a reductant in the SRP Purex process. The bench-scale, continuous-flow, electrolysis cell exhibits a current efficiency approaching 100% in combination with high conversion rates of U(VI) to U(IV) in simulated and actual SRP Purex solutions. High current efficiency is achieved with a voltage-controlled mercury-plated platinum electrode and the use of hydrazine as a nitrite scavenger. Conversion of U(VI) to U(IV) proceeds at 100% efficiency. Cathodic gas generation is minimal. The low rate of gas generation permits a long residence time within the cathode, a necessary condition for high conversions on a continuous basis. Design proposals are given for a plant-scale, continuous-flow unit to meet SRP production requirements. Results from the bench-scale tests indicate that an 8-kW unit can supply sufficient uranous nitrate reductant to meet the needs of the Purex process at SRP

Preventing pollution from plutonium processing

International Nuclear Information System (INIS)

Pillay, K.K.S.

1993-01-01

The plutonium processing facility at Los Alamos has adopted the strategic goal of becoming a facility that processes plutonium in a way that produces only environmentally benign waste streams. Pollution prevention through source reduction and environmentally sound recycling are being pursued. General approaches to waste reductions are administrative controls, modification of process technologies, and additional waste polishing. Recycling of waste materials, such as spent acids and salts, are technical possibilities and are being pursued to accomplish additional waste reduction. Liquid waste stream polishing to remove final traces of plutonium and hazardous chemical constituents is accomplished through (a) process modifications, (b) use of alternative chemicals and sorbents for residue removal, (c) acid recycling, and (d) judicious use of a variety of waste polishing technologies. Technologies that show promise in waste minimization and pollution prevention are identified. Working toward this goal of pollution prevention is a worthwhile endeavor, not only for Los Alamos, but for the Nuclear Complex of the future

Preventing pollution from plutonium processing

International Nuclear Information System (INIS)

Pillay, K.K.S.

1995-01-01

The plutonium processing facility at Los Alamos has adopted the strategic goal of becoming a facility that processes plutonium in a way that produces only environmentally benign waste streams. Pollution prevention through source reduction and environmentally sound recycling are being pursued. General approaches to waste reductions are administrative controls, modification of process technologies, and additional waste polishing. Recycling of waste materials, such as spent acids and salts, are technical possibilities and are being pursued to accomplish additional waste reduction. Liquid waste stream polishing to remove final traces of plutonium and hazardous chemical constituents is accomplished through process modifications, use of alternative chemicals and sorbents for residue removal, acid recycling, and judicious use of a variety of waste polishing technologies. Technologies that show promise in waste minimization and pollution prevention are identified. Working toward this goal of pollution prevention is a worthwhile endeavor , not only for Los Alamos, but for the Nuclear Complex of the future. (author) 12 refs.; 2 figs

Modified titrimetric determination of plutonium using photometric end-point detection

International Nuclear Information System (INIS)

Baughman, W.J.; Dahlby, J.W.

1980-04-01

A method used at LASL for the accurate and precise assay of plutonium metal was modified for the measurement of plutonium in plutonium oxides, nitrate solutions, and in other samples containing large quantities of plutonium in oxidized states higher than +3. In this modified method, the plutonium oxide or other sample is dissolved using the sealed-reflux dissolution method or other appropriate methods. Weighed aliquots, containing approximately 100 mg of plutonium, of the dissolved sample or plutonium nitrate solution are fumed to dryness with an HC1O 4 -H 2 SO 4 mixture. The dried residue is dissolved in dilute H 2 SO 4 , and the plutonium is reduced to plutonium (III) with zinc metal. The excess zinc metal is dissolved with HCl, and the solution is passed through a lead reductor column to ensure complete reduction of the plutonium to plutonium (III). The solution, with added ferroin indicator, is then titrated immediately with standardized ceric solution to a photometric end point. For the analysis of plutonium metal solutions, plutonium oxides, and nitrate solutions, the relative standard deviation are 0.06, 0.08, and 0.14%, respectively. Of the elements most likely to be found with the plutonium, only iron, neptunium, and uranium interfere. Small amounts of uranium and iron, which titrate quantitatively in the method, are determined by separate analytical methods, and suitable corrections are applied to the plutonium value. 4 tables, 4 figures

Base case Pu-nitrate to Pu-oxide conversion plant

International Nuclear Information System (INIS)

1978-10-01

This paper explains that the plutonium recovered in the course of reprocessing spent fuel is obtained in the form of plutonium nitrate, whereas PuO 2 is required for the fabrication of fuel rods. The reference conversion plant described in the paper converts plutonium nitrate into plutonium dioxide powder by precipitation with oxalic acid followed by calcination. The paper also describes the main features of the safety, maintenance and safeguards philosophy used in its design

Transport of plutonium nitrate

International Nuclear Information System (INIS)

1982-02-01

This leaflet discusses the subject under the headings: why do we need plutonium; why must we transport it; what action is carried out; how is it moved; what are the risks. The transport of the material in specially designed containers, from Dounreay in Caithness by road and sea to Sellafield in Cumbria, is described. (U.K.)

Skeletal lesions from inhaled plutonium in beagles

International Nuclear Information System (INIS)

Dagle, G.E.; Park, J.F.; Weller, R.E.; Ragan, H.A.; McClanahan, B.J.; Fisher, D.R.

1984-10-01

The report briefly reviews the skeletal effects observed in ongoing lifespan studies in beagle dogs at 13, 10, and 7 years, respectively, after inhalation exposure to 239 Pu oxide and nitrate or 238 Pu oxide. Plutonium nitrate was chosen to represent soluble material more readily translocated to bone and other tissues than the oxide. Bone lesions related to plutonium exposure were observed only in dogs exposed to 238 Pu oxide and 239 Pu nitrate. The skeleton accumulated approximately 2% ( 239 Pu oxide), 45% ( 238 Pu oxide) or 50% ( 239 Pu nitrate) of the final body burdens at 13, 10, and 7 years, respectively, after exposure. 11 references, 2 figures

Separation of neptunium from uranium and plutonium in the Purex process

International Nuclear Information System (INIS)

Kolarik, Z.; Schuler, R.

1984-01-01

The possibility of removing neptunium from the Purex process in the first extraction cycle was investigated. Butyraldehyde was found to reduce Np(VI) to Np(V), but not Pu(IV) to Pu(III). Up to 99.7% Np can be separated from uranium and plutonium in the 1A extractor or, much more favourably, in an additional partitioning extractor. Hydroxylamine nitrate can be used for reducing Np(VI) to Np(V) in a uranium purification cycle at a high U concentration in the feed solution. Here the decontamination factor for Np can be as high as 2300 and is lowered if iron is present in the feed. (author)

Chromium in aqueous nitrate plutonium process streams: Corrosion of 316 stainless steel and chromium speciation

International Nuclear Information System (INIS)

Smith, W.H.; Purdy, G.

1994-01-01

According to the measurements made in this study, the only situation in which chromium (+6) could exist in a plutonium process solution is one in which a feed containing chromium is dissolved in a glass pot dissolver in high nitric acid concentration and at high temperature. But when the resulting feed is prepared for ion exchange, the chemical treatment reduces chromium to the +3 state. Any solution being processed through the evaporator will only contain chromium in the +3 state and any chromium salts remaining in the evaporator bottoms will be chromium +3 salts

The influence of plutonium concentration and solution flow rate on the effective capacity of macroporous anion exchange resin

International Nuclear Information System (INIS)

Marsh, S.F.; Gallegos, T.D.

1987-07-01

The principal aqueous process used to recover and purify plutonium at the Los Alamos Plutonium Facility is anion exchange in nitric acid. Previous studies with gel-type anion exchange resin have shown an inverse relationship between plutonium concentration in the feed solution and the optimum flow rate for this process. Because gel-type resin has been replaced with macroporous resin at Los Alamos, the relationship between plutonium concentration and solution flow rate was reexamined with the selected Lewatit MP-500-FK resin using solutions of plutonium in nitric acid and in nitric acid with high levels of added nitrate salts. Our results with this resin differ significantly from previous data obtained with gel-type resin. Flow-rate variation from 10 to 80 liters per hour had essentially no effect on the measured quantities of plutonium sorbed by the macroporous resin. However, the effect of plutonium concentration in the feed solutions was pronounced, as feed solutions that contained the highest concentrations of plutonium also produced the highest resin loadings. The most notable effect of high concentrations of dissolved nitrate salts in these solutions was an increased resin capacity for plutonium at low flow rates. 16 refs., 7 figs., 2 tabs

A vision for environmentally conscious plutonium processing

International Nuclear Information System (INIS)

Avens, L.R.; Eller, P.G.; Christensen, D.C.; Miller, W.L.

1998-01-01

Regardless of individual technical and political opinions about the uses of plutonium, it is virtually certain that plutonium processing will continue on a significant global scale for many decades for the purposes of national defense, nuclear power and remediation. An unavoidable aspect of plutonium processing is that radioactive contaminated gas, liquid, and solid streams are generated. These streams need to be handled in a manner that is not only in full compliance with today's laws,but also will be considered environmentally and economically responsible now and in the future. In this regard, it is indeed ironic that the multibillion dollar and multidecade radioactive cleanup mortgage that the US Department of Energy (and its Russian counterpart) now owns resulted from waste management practices that were at the time in full legal compliance. The theme of this paper is that recent dramatic advances in actinide science and technology now make it possible to drastically minimize or even eliminate the problematic waste streams of traditional plutonium processing operations. Advanced technology thereby provides the means to avoid passing on to our children and grandchildren significant environmental and economic legacies that traditional processing inevitably produces. This paper will describe such a vision for plutonium processing that could be implemented fully within five years at a facility such as the Los Alamos Plutonium Facility (TA55). As a significant bonus, even on this short time scale, the initial technology investment is handsomely returned in avoided waste management costs

Benchmark Evaluation of Plutonium Nitrate Solution Arrays

International Nuclear Information System (INIS)

Marshall, M.A.; Bess, J.D.

2011-01-01

In October and November of 1981 thirteen approach-to-critical experiments were performed on a remote split table machine (RSTM) in the Critical Mass Laboratory of Pacific Northwest Laboratory (PNL) in Richland, Washington, using planar arrays of polyethylene bottles filled with plutonium (Pu) nitrate solution. Arrays of up to sixteen bottles were used to measure the critical number of bottles and critical array spacing with a tight fitting Plexiglas(reg s ign) reflector on all sides of the arrays except the top. Some experiments used Plexiglas shells fitted around each bottles to determine the effect of moderation on criticality. Each bottle contained approximately 2.4 L of Pu(NO3)4 solution with a Pu content of 105 g Pu/L and a free acid molarity H+ of 5.1. The plutonium was of low 240Pu (2.9 wt.%) content. These experiments were performed to fill a gap in experimental data regarding criticality limits for storing and handling arrays of Pu solution in reprocessing facilities. Of the thirteen approach-to-critical experiments eleven resulted in extrapolations to critical configurations. Four of the approaches were extrapolated to the critical number of bottles; these were not evaluated further due to the large uncertainty associated with the modeling of a fraction of a bottle. The remaining seven approaches were extrapolated to critical array spacing of 3-4 and 4-4 arrays; these seven critical configurations were evaluation for inclusion as acceptable benchmark experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. Detailed and simple models of these configurations were created and the associated bias of these simplifications was determined to range from 0.00116 and 0.00162 ± 0.00006 ?keff. Monte Carlo analysis of all models was completed using MCNP5 with ENDF/BVII.0 neutron cross section libraries. A thorough uncertainty analysis of all critical, geometric, and material parameters was performed using parameter

The transport of plutonium nitrate liquor in the UK

International Nuclear Information System (INIS)

Barrett, T.R.; Wilson, D.; Bancroft, M.; Higson, J.

1994-01-01

The PFR fuel cycle reasons for plutonium nitrate liquor transport are discussed, identifying the pre 1980 'one litre' bottle associated with nuclear R and D activities and then the 250 litre PUNIT transport container to support the PFR fuel cycle. The PUNIT vessel design is discussed, addressing, in particular, the need for clean filling and criticality design features together with the identification of the IAEA transport regulations criteria and the late 1970s testing programme -the full scale drop tests and the model fire tests. Radiolysis is discussed together with the need for inert gas padding. The filling facility and operation are then outlined together with the fissile material accountancy requirements. The shipments and emergency arrangements are then described drawing attention to time and weather constraints. The risk assessment is discussed, together with reference to the Competent Authority (DoT) 'authorisation' and any constraints. A history of the amount of material transferred is then given - relating that to the reprocessing plant operations since 1980. The most satisfactory performance over 12 years is noted. (author)

A vision for environmentally conscious plutonium processing

International Nuclear Information System (INIS)

Avens, L.R.; Eller, P.G.; Christensen, D.C.; Miller, W.L.

1998-01-01

Regardless of individual technical and political opinions about the uses of plutonium, it is virtually certain that plutonium processing will continue on a significant global scale for many decades for the purposes of national defense, nuclear power, and remediation. An unavoidable aspect of plutonium processing is that radioactively contaminated gas, liquid, and solid waste streams are generated. These streams need to be handled in a manner that not only is in full compliance with today's laws but also will be considered environmentally and economically responsible now and in the future. In this regard, it is indeed ironic that the multibillion dollar and multidecade radioactive cleanup mortgage that the US Department of Energy (and its Russian counterpart) now owns resulted from waste management practices that were at the time in full legal compliance. It is now abundantly evident that in the long run, these practices have proven to be neither environmentally nor economically sound. Recent dramatic advances in actinide science and technology now make it possible to drastically minimize or even eliminate the problematic waste streams of traditional plutonium processing operations. Advanced technology thereby provides the means to avoid passing on to children and grandchildren significant environmental and economic legacies that traditional processing inevitably produces. The authors describe such a vision for plutonium processing that could be implemented fully within 5 yr at a facility such as the Los Alamos National Laboratory Plutonium Facility (TA55). As a significant bonus, even on this short timescale, the initial technology investment is handsomely returned in avoided waste management costs

Waste minimization at a plutonium processing facility

International Nuclear Information System (INIS)

Pillay, K.K.S.

1995-01-01

As part of Los Alamos National Laboratory's (LANL) mission to reduce the nuclear danger throughout the world, the plutonium processing facility at LANL maintains expertise and skills in nuclear weapons technologies as well as leadership in all peaceful applications of plutonium technologies, including fuel fabrication for terrestrial and space reactors and heat sources and thermoelectric generators for space missions. Another near-term challenge resulted from two safety assessments performed by the Defense Nuclear Facilities Safety Board and the U.S. Department of Energy during the past two years. These assessments have necessitated the processing and stabilization of plutonium contained in tons of residues so that they can be stored safely for an indefinite period. This report describes waste streams and approaches to waste reduction of plutonium management

IMPROVED PROCESS OF PLUTONIUM CARRIER PRECIPITATION

Science.gov (United States)

Faris, B.F.

1959-06-30

This patent relates to an improvement in the bismuth phosphate process for separating and recovering plutonium from neutron irradiated uranium, resulting in improved decontamination even without the use of scavenging precipitates in the by-product precipitation step and subsequently more complete recovery of the plutonium in the product precipitation step. This improvement is achieved by addition of fluomolybdic acid, or a water soluble fluomolybdate, such as the ammonium, sodium, or potassium salt thereof, to the aqueous nitric acid solution containing tetravalent plutonium ions and contaminating fission products, so as to establish a fluomolybdate ion concentration of about 0.05 M. The solution is then treated to form the bismuth phosphate plutonium carrying precipitate.

Magnetic separation as a plutonium residue enrichment process

International Nuclear Information System (INIS)

Avens, L.R.; Gallegos, U.F.; McFarlan, J.T.

1990-01-01

Several plutonium contaminated residues have been subjected to Open Gradient Magnetic Separation (OGMS) on an experimental scale. OGMS experiments on graphite and bomb reduction residues resulted in a plutonium rich fraction and a plutonium lean fraction. Values for the bulk quantity rejected to the lean fraction varied between about 20% to 85% of the feed bulk. The plutonium content of the lean fraction can be reduced from about 2% in the feed to the 0.1% to 0.5% range dependent on the portion of the feed rejected to this lean fraction. These values are low enough in plutonium to meet economic discard limits and be considered for direct discard. Magnetic separation of pyrochemical salts gave less favorable results. While a fraction very rich in plutonium could be obtained, the lean fraction plutonium content was too high for direct discard. This may still have chemical processing applications. OGMS experiments at low magnetic field strength on incinerator ash did give two fractions but the plutonium content of each fraction was essentially identical. Thus, no chemical processing advantage was identified for magnetic separation of this residue. 6 refs., 1 fig., 9 tabs

Optimisation of parameters for co-precipitation of uranium and plutonium - results of simulation studies

International Nuclear Information System (INIS)

Pandey, N.K.; Velvandan, P.V.; Murugesan, S.; Ahmed, M.K.; Koganti, S.B.

1999-01-01

Preparation of plutonium oxide from plutonium nitrate solution generally proceeds via oxalate precipitation route. In a nuclear fuel reprocessing scheme this step succeeds the partitioning step (separation of uranium and plutonium). Results of present studies confirm that it is possible to avoid partitioning step and recover plutonium and uranium as co-precipitated product. This also helps in minimising the risk of proliferation of fissile material. In this procedure, the solubility of uranium oxalate in nitric acid is effectively used. Co-precipitation parameters are optimised with simulated solutions of uranium nitrate and thorium nitrate (in place of plutonium). On the basis of obtained results a reconversion flow-sheet is designed and reported here. (author)

Continuous precipitation process of plutonium salts; Procede continu de precipitation des sels de plutonium

Energy Technology Data Exchange (ETDEWEB)

Richard, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-03-01

This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [French] Cette etude porte sur la precipitation continue de l'oxalate de plutonium. L'etude de la solubilite des differentes valences du plutonium dans des milieux acides nitrique-oxalique, puis nitrique-sulfurique-oxalique conduit a choisir la precipitation de l'oxalate de plutonium tetravalent. L'etude porte ensuite sur la sedimentation et la granulometrie de l'oxalate de Pu{sup 4+} obtenue en faisant varier differents parametres de la precipitation : la temperature, l'acidite, l'exces oxalique et le temps de murissement. La derniere partie traite des resultats obtenus avec plusieurs types d'appareils continus essayes au laboratoire. En conclusion sont donnees les conditions de marche de l'appareil tubulaire adopte, ainsi qu'une extrapolation a l'echelle industrielle sous forme d'un flow-sheet. (auteur)

Extraction and purification of plutonium by a tertiary amine; Extraction et purification du plutonium par une amine tertiaire

Energy Technology Data Exchange (ETDEWEB)

Trentinian, M de; Chesne, A [Commissariat a l' Energie Atomique, Fontenay aux Roses, Section de Chimie des Actimides (France).Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

Trilaurylamine diluted with a paraffinic solvent (dodecane) was studied as part of the research dealing with the separation and purification of plutonium. The physical properties (solubility of nitrates in the amine as a function of temperature) and the resistance to radiations of this substance were examined. The extraction characteristics of nitric solutions of plutonium, uranium and certain fission products are given as a function of the following factors: concentration of the various ions in solution, valency states. A method of plutonium purification based on these results is presented. (author) [French] La trilaurylamine diluee par un solvant paraffinique (dodecane) a ete etudiee dans le cadre des recherches concernant la separation et la purification du plutonium. Une etude des caracteres physiques (solubilite des nitrates dans l'amine en fonction de la temperature) s'ajoute a celle de la tenue aux radiations de ce corps. Les caracteristiques d'extraction de solutions nitriques de plutonium, uranium, et certains produits de fission, sont donnes en fonction des facteurs suivants: concentration des differents ions en solution, etats de valence. On presente une methode de purification du plutonium basee sur ces resultats. (auteur)

Contribution to the characterization of the ideality deviation of concentrated solutions of electrolytes: application to the case plutonium and uranium (IV) nitrates

International Nuclear Information System (INIS)

Charrin, N.

1999-01-01

The purpose of this work is to establish a base of binary data referring to the plutonium and uranium nitrates (IV) activity coefficients, which will permit to take account the medium effects in the process of liquid-liquid extraction set in action during the reprocessing of irradiated combustibles in a more scrupulous way. The first chapter sticks to establish the problematic of acquisition of actinides binary data at an oxidation state (IV) linked to two characteristics of this type of electrolyte its radioactive properties and its chemical properties. Its chemical properties bring us to define the fictitious binary data and to use an approach based on the thermodynamic concept of simple solutions, on the measurements of water activity of ternary or quaternary mixtures of the actinide, in nitric acid medium and on the binary data of nitric acid. The second chapter intended to propose reliable binary data concerning nitric acid. The validation of acquisition of fictitious binary data method suggested is undertaken. The electrolyte test is the thorium nitrate (IV). The very encouraging results has determined the carrying out of this work of research in that way. The third chapter is based on the experimental acquisition of uranium and plutonium nitrates (IV) binary data. It emphasises the importance given to the preparation of the studied mixtures which characteristics, very high actinide concentrations and low acidities, make them atypical solutions and without any referenced equivalents. The last chapter describes the exploitation which was made of the established binary data. The characteristic parameters of Pu(NO 3 ) 4 and U(NO 3 ) 4 of Pitzer model and of the specific interaction theory has been appraised. Then the application of' the concept of simple solutions to the calculation of the density or quaternary mixtures like Pu(NO 3 ) 4 / UO 2 (NO 3 ) 2 /HNO 3 / H 2 O was proposed. (author)

Process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide

International Nuclear Information System (INIS)

Heremanns, R.H.; Vandersteene, J.J.

1983-01-01

The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined

Study of accurate volume measurement system for plutonium nitrate solution

Energy Technology Data Exchange (ETDEWEB)

Hosoma, T. [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

1998-12-01

It is important for effective safeguarding of nuclear materials to establish a technique for accurate volume measurement of plutonium nitrate solution in accountancy tank. The volume of the solution can be estimated by two differential pressures between three dip-tubes, in which the air is purged by an compressor. One of the differential pressure corresponds to the density of the solution, and another corresponds to the surface level of the solution in the tank. The measurement of the differential pressure contains many uncertain errors, such as precision of pressure transducer, fluctuation of back-pressure, generation of bubbles at the front of the dip-tubes, non-uniformity of temperature and density of the solution, pressure drop in the dip-tube, and so on. The various excess pressures at the volume measurement are discussed and corrected by a reasonable method. High precision-differential pressure measurement system is developed with a quartz oscillation type transducer which converts a differential pressure to a digital signal. The developed system is used for inspection by the government and IAEA. (M. Suetake)

Laboratory-scale evaluations of alternative plutonium precipitation methods

International Nuclear Information System (INIS)

Martella, L.L.; Saba, M.T.; Campbell, G.K.

1984-01-01

Plutonium(III), (IV), and (VI) carbonate; plutonium(III) fluoride; plutonium(III) and (IV) oxalate; and plutonium(IV) and (VI) hydroxide precipitation methods were evaluated for conversion of plutonium nitrate anion-exchange eluate to a solid, and compared with the current plutonium peroxide precipitation method used at Rocky Flats. Plutonium(III) and (IV) oxalate, plutonium(III) fluoride, and plutonium(IV) hydroxide precipitations were the most effective of the alternative conversion methods tested because of the larger particle-size formation, faster filtration rates, and the low plutonium loss to the filtrate. These were found to be as efficient as, and in some cases more efficient than, the peroxide method. 18 references, 14 figures, 3 tables

233-S Plutonium Concentration Facility data quality objectives

International Nuclear Information System (INIS)

Encke, D.B.

1996-08-01

This document is a summary of the decision-making associated with the Data Quality Objective process that pertains to the characterization activities in the 233-S Plutonium Concentration Facility at the Hanford Site in Richland, Washington. The 233-S Plutonium Concentration Facility is located adjacent to, and north of, the REDOX Plant. The facility was used to concentrate the plutonium nitrate product solution from the REDOX facility. The 233-S Pipe Gallery, Control Room, SWP Change Room, Toilet, Equipment Room and the Electrical Cubicle are currently scheduled for decontamination and cleanout to support future demolition (D and D). Identification of the radiological contamination and presence of hazardous materials is needed to allow for disposal of the D and D debris

Dry sample storage system for an analytical laboratory supporting plutonium processing

International Nuclear Information System (INIS)

Treibs, H.A.; Hartenstein, S.D.; Griebenow, B.L.; Wade, M.A.

1990-01-01

The Special Isotope Separation (SIS) plant is designed to provide removal of undesirable isotopes in fuel grade plutonium by the atomic vapor laser isotope separation (AVLIS) process. The AVLIS process involves evaporation of plutonium metal, and passage of an intense beam of light from a laser through the plutonium vapor. The laser beam consists of several discrete wavelengths, tuned to the precise wavelength required to ionize the undesired isotopes. These ions are attracted to charged plates, leaving the bulk of the plutonium vapor enriched in the desired isotopes to be collected on a cold plate. Major portions of the process consist of pyrochemical processes, including direct reduction of the plutonium oxide feed material with calcium metal, and aqueous processes for purification of plutonium in residues. The analytical laboratory for the plant is called the Material and Process Control Laboratory (MPCL), and provides for the analysis of solid and liquid process samples

Design of plutonium processing facilities

International Nuclear Information System (INIS)

Derbyshire, W.; Sills, R.J.

1982-01-01

Five considerations for the design of plutonium processing facilities are identified. These are: Toxicity, Radiation, Criticality, Containment and Remote Operation. They are examined with reference to reprocessing spent nuclear fuel and application is detailed both for liquid and dry processes. (author)

The extraction of plutonium with triethylene glycol dichloride

International Nuclear Information System (INIS)

Aikin, A.M.; Moss, M.; Bruce, T.

1951-03-01

The extraction of plutonium by triethylene glycol dichloride (trigly) has been investigated briefly. The effect of (1) the valence state of the plutonium, (2) the concentration of nitric acid, (3) the concentration of ammonium nitrate and (4) the conditioning of the trigly was measured. The solubility of plutonium IV in trigly was found to be 70 mgms/ml. Solutions of plutonium in trigly and in concentrated nitric acid solutions have been examined spectrophotometrically. (author)

The extraction of plutonium with triethylene glycol dichloride

Energy Technology Data Exchange (ETDEWEB)

Aikin, A M; Moss, M; Bruce, T

1951-03-15

The extraction of plutonium by triethylene glycol dichloride (trigly) has been investigated briefly. The effect of (1) the valence state of the plutonium, (2) the concentration of nitric acid, (3) the concentration of ammonium nitrate and (4) the conditioning of the trigly was measured. The solubility of plutonium IV in trigly was found to be 70 mgms/ml. Solutions of plutonium in trigly and in concentrated nitric acid solutions have been examined spectrophotometrically. (author)

PROCESS OF ELIMINATING HYDROGEN PEROXIDE IN SOLUTIONS CONTAINING PLUTONIUM VALUES

Science.gov (United States)

Barrick, J.G.; Fries, B.A.

1960-09-27

A procedure is given for peroxide precipitation processes for separating and recovering plutonium values contained in an aqueous solution. When plutonium peroxide is precipitated from an aqueous solution, the supernatant contains appreciable quantities of plutonium and peroxide. It is desirable to process this solution further to recover plutonium contained therein, but the presence of the peroxide introduces difficulties; residual hydrogen peroxide contained in the supernatant solution is eliminated by adding a nitrite or a sulfite to this solution.

Effect of compositional variation in plutonium on process shielding design

Energy Technology Data Exchange (ETDEWEB)

Brown, T.H.

1997-11-01

Radiation dose rate from plutonium with high {sup 239}Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations.

Effect of compositional variation in plutonium on process shielding design

International Nuclear Information System (INIS)

Brown, T.H.

1997-11-01

Radiation dose rate from plutonium with high 239 Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations

PLUTONIUM LOADING CAPACITY OF REILLEX HPQ ANION EXCHANGE COLUMN - AFS-2 PLUTONIUM FLOWSHEET FOR MOX

Energy Technology Data Exchange (ETDEWEB)

Kyser, E.; King, W.; O' Rourke, P.

2012-07-26

Radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the dependence of column loading performance on the feed composition in the H-Canyon dissolution process for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). These loading experiments show that a representative feed solution containing {approx}5 g Pu/L can be loaded onto Reillex{trademark} HPQ resin from solutions containing 8 M total nitrate and 0.1 M KF provided that the F is complexed with Al to an [Al]/[F] molar ratio range of 1.5-2.0. Lower concentrations of total nitrate and [Al]/[F] molar ratios may still have acceptable performance but were not tested in this study. Loading and washing Pu losses should be relatively low (<1%) for resin loading of up to 60 g Pu/L. Loading above 60 g Pu/L resin is possible, but Pu wash losses will increase such that 10-20% of the additional Pu fed may not be retained by the resin as the resin loading approaches 80 g Pu/L resin.

A solvent proceed for the extraction of the irradiate uranium and plutonium in the reactor core; Un procede par solvant pour l'extraction du plutonium de l'uranium irradie dans les piles

Energy Technology Data Exchange (ETDEWEB)

Goldschmidt, B; Regnaut, P; Prevot, I [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

Description of the conditions of plutonium, fission products and of uranium separation by selective extraction of the nitrates by organic solvent, containing a simultaneous extraction of plutonium and uranium, followed by a plutonium re-extraction after reduction, and an uranium re-extraction. The rates of decontamination being insufficient in this first stage, we also describes the processes of decontamination permitting separately to get the rates wanted for uranium and plutonium. Finally, we describes the beginning of the operation that consists in a nitric dissolution of the active uranium while capturing the products of gaseous fission, as well as the final concentration of the products of fission in a concentrated solution. (authors) [French] Description des conditions de separation du plutonium, des produits de fission et de l'uranium au moyen d'une extraction selective des nitrates par solvant organique, comprenant une extraction simultanee du plutonium et de l'uranium, suivie d'une reextraction du plutonium apres reduction, et d'une reextraction de l'uranium. Les taux de decontamination etant insuffisants dans ce premier stade, on decrit egalement les processus de decontamination permettant separement d'obtenir les taux desires pour l'uranium et le plutonium. Enfin, on decrit aussi le debut de l'operation qui consiste en une dissolution nitrique de l'uranium actif en captant les produits de fission gazeux, ainsi que la concentration finale des produits de fission sous forme de solution concentree. (auteurs)

On line spectrophotometry with optical fibers. Application to uranium-plutonium separation in a spent fuel reprocessing plant

International Nuclear Information System (INIS)

Boisde, G.; Mus, G.; Tachon, M.

1985-06-01

Optimization of mixer-settler operation for uranium-plutonium separation in the Purex process can be obtained by remote spectrophotometry with optical fibers. Data acquisition on uranium VI, uranium IV and plutonium III is examined in function of acidity and nitrate content of the solution. Principles for on line multicomponent monitoring and mathematical modelization of the measurements are described [fr

Study of the influence of radiolysis on the stability of plutonium III. Application to a heterogeneous medium formed by a nitric solution of ferrous ions and an organic solution of trilauryl-ammonium nitrate

International Nuclear Information System (INIS)

Fourmaux, J.M.

1980-01-01

The objective of this research thesis is to study the behaviour of plutonium 238 in media which are commonly used to isolate it from other elements such as neptunium and fission products created during the neutron irradiation of the neptunium 237 isotope. As plutonium 238 purification processes are all based on redox reaction, it is essential to know the influence of radiolysis on the redox behaviour, and on the distribution coefficients of this isotope in solutions used during its separation from the neptunium 237 isotope. Therefore, it is necessary to study the influence of radiolysis on the stability of plutonium with an oxidation III level. As this extraction is performed by an organic solvent (trilauryl-ammonium nitrate), this study addresses the behaviour of plutonium in an emulsion formed by this solvent and the nitric aqueous solution previously adjusted in terms of Fe 2+ ions. After a brief recall of bibliographical generalities related to radiolysis, the author presents and comments the Nernst law in the case of a two-phase system (emulsion), and reports the use of this law to obtain the plutonium potential-distribution coefficient relationship. The last part reports experimental data

Study of plutonium sorption in aluminia column in the system HNO3-HF

International Nuclear Information System (INIS)

Araujo, J.A. de.

1977-01-01

The column chromatographic method using alumina has been applied successfully to study the sorption-desorption behavior of plutonium traces in HNO 3 -HF and HNO 3 -HF-UO 2 (NO 3 ) 2 systems, aiming to elaborate a process for recovering plutonium traces from reprocessing wastes, mainly in existing solutions where uranium is presented in macro quantities. Basically, the method consists in the sorption of plutonium by percolating a solution containing HNO 3 (0,1 to 0,8M) or uranyl nitrate (1-50 gU/l) and HF(0,1 to 0,3M) through an Al 2 O 3 collumn. The plutonium is fixed on Al 2 O 3 whereas the uranyl ions is collected in the efluent. The adsorption of Pu-III, Pu-IV and Pu-VI in the presence of HF was determined and Pu-IV can be almost completely sorbed. The Pu-IV is eluted by reduction to Pu-III in the column using 3 M HNO 3 -0,005M FeSO 4 at 50 0 C as elutrient. This method is very simple and can be applied for separation and purification of plutonium (traces) from uranyl nitrate or others coming solutions from wet chemistry of irradiated fuels [pt

A rapid method of dosing plutonium in radioactive effluents; Methode de dosage rapide du plutonium dans les effluents radioactifs (1961)

Energy Technology Data Exchange (ETDEWEB)

Scheidhauer, J; Messainguiral, L [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1961-07-01

The plutonium is first separated by a lanthanum fluoride precipitation. The precipitated fluorides are dissolved in normal nitric acid solution in the presence of aluminium nitrate. The plutonium transformed to the tetravalent state is then extracted with thenoyltrifluoroacetone and returned to the aqueous phase with 10 N nitric acid. After evaporation on a watch glass the residue is calcined on a Meker burner and counted using a counting system fitted with a zinc sulphide scintillator. When necessary, the calcium is eliminated at the beginning of the dosage by a fluoride precipitation, the plutonium being oxidised to the valency IV. (authors) [French] Le plutonium est d'abord separe par entrainement au fluorure de lanthane. Le precipite des fluorures est remis en solution en milieu acide nitrique normal, en presence de nitrate d'aluminium. Le plutonium amene a la valence IV est alors extrait par la thenoyltrifluoroacetone et remis en phase aqueuse dans l'acide nitrique 10 N. Apres evaporation sur verre de montre, le residu est calcine sur bec Meker et compte sur un ensemble de comptage equipe d'un scintillateur au sulfure de zinc. Lorsque cela est necessaire, le calcium est elimine, au debut du dosage, par precipitation du fluorure, le plutonium etant oxyde a la valence VI. (auteurs)

Accountability methods for plutonium and uranium: the NRC manuals

Energy Technology Data Exchange (ETDEWEB)

Gutmacher, R.G.; Stephens, F.B.

1977-09-28

Four manuals containing methods for the accountability of plutonium nitrate solutions, plutonium dioxide, uranium dioxide and mixed uranium-plutonium oxide have been prepared by us and issued by the U.S. Nuclear Regulatory Commission. A similar manual on methods for the accountability of uranium and plutonium in reprocessing plant dissolver solutions is now in preparation. In the present paper, we discuss the contents of the previously issued manuals and give a preview of the manual now being prepared.

Accountability methods for plutonium and uranium: the NRC manuals

International Nuclear Information System (INIS)

Gutmacher, R.G.; Stephens, F.B.

1977-01-01

Four manuals containing methods for the accountability of plutonium nitrate solutions, plutonium dioxide, uranium dioxide and mixed uranium-plutonium oxide have been prepared by us and issued by the U.S. Nuclear Regulatory Commission. A similar manual on methods for the accountability of uranium and plutonium in reprocessing plant dissolver solutions is now in preparation. In the present paper, we discuss the contents of the previously issued manuals and give a preview of the manual now being prepared

PLUTONIUM PURIFICATION PROCESS EMPLOYING THORIUM PYROPHOSPHATE CARRIER

Science.gov (United States)

King, E.L.

1959-04-28

The separation and purification of plutonium from the radioactive elements of lower atomic weight is described. The process of this invention comprises forming a 0.5 to 2 M aqueous acidffc solution containing plutonium fons in the tetravalent state and elements with which it is normally contaminated in neutron irradiated uranium, treating the solution with a double thorium compound and a soluble pyrophosphate compound (Na/sub 4/P/sub 2/O/sub 7/) whereby a carrier precipitate of thorium A method is presented of reducing neptunium and - trite is advantageous since it destroys any hydrazine f so that they can be removed from solutions in which they are contained is described. In the carrier precipitation process for the separation of plutonium from uranium and fission products including zirconium and columbium, the precipitated blsmuth phosphate carries some zirconium, columbium, and uranium impurities. According to the invention such impurities can be complexed and removed by dissolving the contaminated carrier precipitate in 10M nitric acid, followed by addition of fluosilicic acid to about 1M, diluting the solution to about 1M in nitric acid, and then adding phosphoric acid to re-precipitate bismuth phosphate carrying plutonium.

Technical report for generic site add-on facility for plutonium polishing

International Nuclear Information System (INIS)

1998-06-01

The purpose of this report is to provide environmental data and reference process information associated with incorporating plutonium polishing steps (dissolution, impurity removal, and conversion to oxide powder) into the genetic-site Mixed-Oxide Fuel Fabrication Facility (MOXFF). The incorporation of the plutonium polishing steps will enable the removal of undesirable impurities, such as gallium and americium, known to be associated with the plutonium. Moreover, unanticipated impurities can be removed, including those that may be contained in (1) poorly characterized feed materials, (2) corrosion products added from processing equipment, and (3) miscellaneous materials contained in scrap recycle streams. These impurities will be removed to the extent necessary to meet plutonium product purity specifications for MOX fuels. Incorporation of the plutonium polishing steps will mean that the Pit Disassembly and Conversion Facility (PDCF) will need to produce a plutonium product that can be dissolved at the MOXFF in nitric acid at a suitable rate (sufficient to meet overall production requirements) with the minimal usage of hydrofluoric acid, and its complexing agent, aluminum nitrate. This function will require that if the PDCF product is plutonium oxide powder, that powder must be produced, stored, and shipped without exceeding a temperature of 600 C

Conceptual Design for the Pilot-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

Energy Technology Data Exchange (ETDEWEB)

Lumetta, Gregg J.; Meier, David E.; Tingey, Joel M.; Casella, Amanda J.; Delegard, Calvin H.; Edwards, Matthew K.; Jones, Susan A.; Rapko, Brian M.

2014-08-05

This report describes a conceptual design for a pilot-scale capability to produce plutonium oxide for use as exercise and reference materials, and for use in identifying and validating nuclear forensics signatures associated with plutonium production. This capability is referred to as the Pilot-scale Plutonium oxide Processing Unit (P3U), and it will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including plutonium dioxide (PuO2) dissolution, purification of the Pu by ion exchange, precipitation, and conversion to oxide by calcination.

Simulation of uranium and plutonium oxides compounds obtained in plasma

Science.gov (United States)

Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

2018-03-01

The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

Design safety features of containments used for handling plutonium in Reprocessing Plants

International Nuclear Information System (INIS)

Aherwal, P.; Achuthan, P.V.

2016-01-01

The plutonium present in spent fuel is separated from the associated uranium and fission products using solvent extraction cycles in process cells. Product plutonium nitrate solution containing trace concentrations of uranium and fission products is treated in the reconversion facility through a precipitation-calcination route and converted to sinterable grade plutonium oxide (PuO 2 ). All chemical operations involving materials with high plutonium content, both in solid and solution forms are carried out in glove boxes. Glove box provides an effective isolation from radioactive materials handled and acts as a barrier between the operator and the source of radiation. These glove boxes are interconnected for sequential operations and the interconnected glove box trains are installed within secondary enclosures called double skin which provides double barrier protection to operators

Studies in biological excretion of inhaled plutonium in the case of a few occupational workers in a fuel reprocessing plant

International Nuclear Information System (INIS)

Hedge, A.G.; Chandramouli, S.; Iyer, R.S.; Bhat, I.S.

1992-01-01

A power reactor fuel reprocessing plant is in operation at Tarapur. The various processes involved in the plant are: fuel rod cutting, dissolution in nitric acid, separation of plutonium, and handling of separated plutonium. The chemical form of plutonium could be nitrate, TBP complex, or oxide depending upon the nature of the process involved. Possible internal exposure to plant personnel occurs mainly by inhalation and occasionally through a contaminated wound. Occupational workers are regularly monitored for internal contamination by urinary excretion analysis as well as by in-vivo lung counting. This paper presents a follow-up study of plutonium elimination in four inhalation exposure cases. (author) 8 refs.; 6 figs

Plutonium production story at the Hanford site: processes and facilities history

Energy Technology Data Exchange (ETDEWEB)

Gerber, M.S., Westinghouse Hanford

1996-06-20

This document tells the history of the actual plutonium production process at the Hanford Site. It contains five major sections: Fuel Fabrication Processes, Irradiation of Nuclear Fuel, Spent Fuel Handling, Radiochemical Reprocessing of Irradiated Fuel, and Plutonium Finishing Operations. Within each section the story of the earliest operations is told, along with changes over time until the end of operations. Chemical and physical processes are described, along with the facilities where these processes were carried out. This document is a processes and facilities history. It does not deal with the waste products of plutonium production.

Study on the process variables in the anion exchange plutonium separation process

Energy Technology Data Exchange (ETDEWEB)

Nishimura, D T

1957-11-15

This report discusses the study of the process variables in the Anion Exchange Process Pilot Plant for the separation of plutonium from irradiated uranium. Variables associated with the feed, wash and elution cycles were studied with the aim of improving the quality of the final plutonium product, reduce cycling time and reagent requirements, and also to obtain data for prediction of resin column behaviour under various feed conditions. A cation resin column and a silica gel column were installed in the system and these were studied for plutonium recovery and product quality. The product obtained from the plant was acceptable in all the impurities except the associated gamma activity which was too high for easy product handling. (author)

A rapid and specific titrimetric method for the precise determination of plutonium using redox indicator

International Nuclear Information System (INIS)

Chitnis, R.T.; Dubey, S.C.

1976-01-01

A simple and rapid method for the determination of plutonium in plutonium nitrate solution and its application to the purex process solutions is discussed. The method involves the oxidation of plutonium to Pu(VI) with the help of argentic oxide followed by the destruction of the excess argentic oxide by means of sulphamic acid. The determination of plutonium is completed by adding ferrous ammonium sulphate solution which reduces Pu(VI) to Pu(IV) and titrating the excess ferrous with standard potassium dichromate solution using sodium diphenylamine sulphonate as the internal indicator. The effect of the various reagents add during the oxidation and reduction of plutonium, on the final titration has been investigated. The method works satisfactorily for the analysis of plutonium in the range of 0.5 to 5 mg. The precision of the method is found to be within 0.1%. (author)

Liquid waste processing from plutonium (III) oxalate precipitation

International Nuclear Information System (INIS)

Esteban, A.; Cassaniti, P.; Orosco, E.H.

1990-01-01

Plutonium (III) oxalate filtrates contain about 0.2M oxalic acid, 0.09M ascorbic acid, 0.05M hydrazine, 1M nitric acid and 20-100 mg/l of plutonium. The developed treatment of liquid wastes consist in two main steps: a) Distillation to reduce up to 10% of the initial volume and refluxing to destroy organic material. Then, the treated solution is suitable to adjust the plutonium at the tetravalent state by addition of hydrogen peroxide and the nitric molarity up to 8.6M. b) Recovery and purification of plutonium by anion exchange using two columns in series containing Dowex 1-X4 resin. With the proposed process, it is possible to transform 38 litres of filtrates with 40mg/l of Pu into 0.1 l of purified solution with 15-20g/l of Pu. This solution is suitable to be recycled in the Pu (III) oxalate precipitation process. This process has several potential advantages over similar liquid waste treatments. These include: 1) It does not increase the liquid volume. 2) It consumes only few reagents. 3) The operations involved are simple, requiring limited handling and they are feasible to automatization. 4) The Pu recovery factor is about 99%. (Author) [es

The study of reductive reextraction of plutonium in the Purex process

International Nuclear Information System (INIS)

Poczynajlo, A.

1985-01-01

The methods of separation of U and Pu in the Purex process and the thermodynamic and kinetic properties of Pu(4) reductants are discussed. The kinetic equation of the process of reductive reextraction of plutonium for the first order reaction with respect to Pu(4) is derived. The kinetics of plutonium reextraction with the use of uranium (4), ascorbic acid and other reductants has been studied. The necessity of application of the stoichiometric excess of reductant has been explained by simultaneously occured reoxidation process of plutonium. The method of calculation of the steady- state plutonium concentration profiles has been elaborated for counter-current separation of U and Pu in multistage contactor. 90 refs., 20 tabs., 29 figs. (author)

Reaction of alkali nitrates with PuO2

International Nuclear Information System (INIS)

Yamashita, T.; Ohuchi, K.; Takahashi, K.; Fujino, T.

1990-01-01

Improvement of solubility of plutonium dioxide (PuO 2 ) in acid solution is important to establish the nuclear fuel reprocessing technique for uranium-plutonium mixed oxide fuels. If insoluble PuO 2 can be converted into any soluble plutonium compounds, problems arising from the fuel dissolution process will be reduced to a great extent. Alkali metal plutonates and alkaline-earth plutonates are known to have enhanced solubility in mineral acids. However, the reaction conditions to form such plutonates and characterization thereof are not well elucidated. Then the reactivity and reaction conditions to form lithium and sodium plutonates from their nitrates and PuO 2 were studied at temperatures between 500 and 900 degree C and alkali metal to plutonium atom ratios between 0.5 and 6 by means of thermogravimetry as well as X-ray diffraction technique. The reaction behavior of alkali plutonates will be discussed in comparison with corresponding alkali uranates

Plutonium Finishing Plant. Interim plutonium stabilization engineering study

Energy Technology Data Exchange (ETDEWEB)

Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J. [Pacific Northwest Lab., Richland, WA (United States); Nass, R. [Nuclear Fuel Services, Inc. (United States)

1995-08-01

This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage.

Plutonium Finishing Plant. Interim plutonium stabilization engineering study

International Nuclear Information System (INIS)

Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J.; Nass, R.

1995-08-01

This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage

Aqueous recovery of plutonium from pyrochemical processing residues

International Nuclear Information System (INIS)

Gray, L.W.; Gray, J.H.

1984-01-01

Pyrochemical processes provide rapid methods to reclaim plutonium from scrap residues. Frequently, however, these processes yield an impure plutonium product and waste residues that are contaminated with actinides and are therefore nondiscardable. The Savannah River Laboratory and Plant and the Rocky Flats Plant are jointly developing new processes using both pyrochemistry and aqueous chemistry to generate pure product and discardable waste. An example of residue being treated is that from the molten salt extraction (MSE), a mixture of NaCl, KCl, MgCl 2 , PuCl 3 , AmCl 3 , PuO 2 , and Pu 0 . This mixture is scrubbed with molten aluminum containing a small amount of magnesium to produce a nonhomogeneous Al-Pu-Am-Mg alloy. This process, which rejects most of the NaCl-KCl-MgCl 2 salts, results in a product easily dissolved in 6M HNO 3 -0.1M HF. Any residual chloride in the product is removed by precipitation with Hg(I) followed by centrifuging. Plutonium and americium are then separated by the standard Purex process. The americium, initially diverted to the solvent extraction waste stream, can either be recovered or sent to waste

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)

Complexes of pentavalent plutonium in lithium nitrate solutions

International Nuclear Information System (INIS)

Mekhail, F.M.; Zaki, M.R.

1977-01-01

Pu 0 2 ion can form nitrate complexes in concentrated solution of lithium nitrate of PH 3.5. Spectrophotometric and ion exchange studies revealed the existence of two complexes, presumably the mono-and the dinitro. The rate of adsorption of the dinitrato complex, formed in 4 to 6 M-lithium nitrate solutions, on De-Acidite FF has been investigated and suggested to be diffusion controlled. The adsorption isotherm found to obey satisfactorily Freundlich equation

R.4. Innovative concept for plutonium finishing facility

International Nuclear Information System (INIS)

Bertolotti, G.; Laguerie, I.V. de; Richter, R.; Gillet, B.

1998-01-01

After complete shutdown of the units of the previous UP2 plant, the new R4 facility will ensure the purification of Plutonium of the UP2-800 plant for the whole range of fuel to be reprocessed in the years to come. This facility features four main units: - Purification of plutonium nitrate; - Conversion into plutonium oxide; - PuO 2 conditioning into cans; Acid recovery. An extensive R and D program resulted in significant innovations. From a technological aspect the centrifugal extractor and the sub-critical tube bundle tank contributed to the reduction of the building dimensions. The extensive use of on-line analyses enables a more efficient follow-up of the process while minimizing the effluent production. On the other hand, the organization of the building which consists in grouping the rooms presenting the same risk of dispersal of nuclear materials also contributed to reduce the active zone volume. This facility, as any other facilities on the LA HAGUE site, will be remotely operated. (author)

Volume measurement system for plutonium nitrate solution and its uncertainty to be used for nuclear materials accountancy proved by demonstration over fifteen years

International Nuclear Information System (INIS)

Hosoma, Takashi

2010-10-01

An accurate volume measurement system for plutonium nitrate solution stored in an accountability tank with dip-tubes has been developed and demonstrated over fifteen years at the Plutonium Conversion Development Facility of the Japan Atomic Energy Agency. As a result of calibrations during the demonstration, it was proved that measurement uncertainty practically achieved and maintained was less than 0.1% (systematic character) and 0.15% (random) as one sigma which was half of the current target uncertainty admitted internationally. It was also proved that discrepancy between measured density and analytically determined density was less than 0.002 g·cm -3 as one sigma. These uncertainties include effects by long term use of the accountability tank where cumulative plutonium throughput is six tons. The system consists of high precision differential pressure transducers and a dead-weight tester, sequentially controlled valves for periodical zero adjustment, dampers to reduce pressure oscillation and a procedure to correct measurement biases. The sequence was also useful to carry out maintenances safely without contamination. Longevity of the transducer was longer than 15 years. Principles and essentials to determine solution volume and weight of plutonium, measurement biases and corrections, accurate pressure measurement system, maintenances and diagnostics, operational experiences, evaluation of measurement uncertainty are described. (author)

Water Solubility of Plutonium and Uranium Compounds and Residues at TA-55

International Nuclear Information System (INIS)

Reilly, Sean Douglas; Smith, Paul Herrick; Jarvinen, Gordon D.; Prochnow, David Adrian; Schulte, Louis D.; DeBurgomaster, Paul Christopher; Fife, Keith William; Rubin, Jim; Worl, Laura Ann

2016-01-01

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U 3 O 8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl 3 , and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate

Water Solubility of Plutonium and Uranium Compounds and Residues at TA-55

Energy Technology Data Exchange (ETDEWEB)

Reilly, Sean Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Jarvinen, Gordon D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Prochnow, David Adrian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Schulte, Louis D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; DeBurgomaster, Paul Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Fife, Keith William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Rubin, Jim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States; Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States

2016-06-13

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U₃O₈ (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl₃, and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a

PROCESS USING POTASSIUM LANTHANUM SULFATE FOR FORMING A CARRIER PRECIPITATE FOR PLUTONIUM VALUES

Science.gov (United States)

Angerman, A.A.

1958-10-21

A process is presented for recovering plutonium values in an oxidation state not greater than +4 from fluoride-soluble fission products. The process consists of adding to an aqueous acidic solution of such plutonium values a crystalline potassium lanthanum sulfate precipitate which carries the plutonium values from the solution.

United Kingdom experience in plutonium transportation

International Nuclear Information System (INIS)

1978-11-01

This paper describes the extensive experience of the UKAEA and British Nuclear Fuels Limited in transporting plutonium within the UK over the last 20 years and to destinations overseas since the early 1960s. The aspects covered include: the form of plutonium transported (nitrate, oxide, mixed oxide etc.); UK and international regulations (e.g. covering safety and safeguards matters); insurance; container design; mode of transport; physical protection; and a safety analysis. Costs are estimated to be largely independent of quantity

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

CONVERSION OF PLUTONIUM TRIFLUORIDE TO PLUTONIUM TETRAFLUORIDE

Science.gov (United States)

Fried, S.; Davidson, N.R.

1957-09-10

A large proportion of the trifluoride of plutonium can be converted, in the absence of hydrogen fluoride, to the tetrafiuoride of plutonium. This is done by heating plutonium trifluoride with oxygen at temperatures between 250 and 900 deg C. The trifiuoride of plutonium reacts with oxygen to form plutonium tetrafluoride and plutonium oxide, in a ratio of about 3 to 1. In the presence of moisture, plutonium tetrafluoride tends to hydrolyze at elevated temperatures and therefore it is desirable to have the process take place under anhydrous conditions.

Properties of concentrated plutonium nitrate solutions

International Nuclear Information System (INIS)

Gray, J.H.; Swanson, J.L.

1978-01-01

Selected properties were measured for solutions containing about 500 and 700 g/l plutonium (IV) in 4--5M nitric acid: density, viscosity, vapor pressure, boiling point, radiolytic gas (H 2 ) evolution rates, and corrosion rate on Ti and 304L stainless steel. Pu solubility was determined to be 550 to 800 g/l in 2.5 to 7M HNO 3 at ambient temperature and 820 to 860 g/l in 3M HNO 3 at 50 0 C

The gastrointestinal absorption of organically bound forms of plutonium in fed and fasted hamsters

International Nuclear Information System (INIS)

Harrison, J.D.; Cooper, J.R.; Bomford, J.A.; David, A.J.

1986-01-01

Values of about 0.005-0.01% were obtained for the absorption in fed hamsters of plutonium ingested as Pu 4+ citrate, isocitrate, phytate and malate complexes and Pu 3+ ascorbate compared with about 0.003-0.004% for Pu 4+ nitrate. Replacing drinking water with tea did not affect the result for Pu 4+ nitrate. Fasting hamsters for 8 h before the administration of plutonium citrate increased absorption to 0.1-0.2%. An extra period of fasting for 4 h after administration did not lead to a further increase in absorption. Similar values were also obtained when plutonium citrate was administered after a 24 h fast, followed either by immediate access to food or a further 4 h fast. In hamsters fasted for 24 h before administration of either Pu 3+ ascorbate or Pu 4+ nitrate, about 6-7 per cent of the ingested plutonium was retained in the gastrointestinal tract after one week. At three weeks after ingestion of Pu 3+ ascorbate, gastrointestinal retention had fallen 100-fold without an increase in absorption. (author)

In situ observation of plutonium transfer processes in the marine environment

International Nuclear Information System (INIS)

Guary, J.-C.; Fraizier, Andre

1975-09-01

A preliminary observation of plutonium transfer processes in the marine environment was carried out and showed that concentration of the radionuclide was lower when marine organisms stood at a higher trophic level. This observation supplemented by an investigation on contamination pathways showed that plutonium was not concentrated along the food chain and its uptake occured preferentially by direct contact of species with seawater, a process chiefly affecting producers and primary consumers. It appeared that the marine sediment was not a significant vector of plutonium transfer in burrowing species [fr

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)

Recovery and purification of uranium-234 from aged plutonium-238

International Nuclear Information System (INIS)

Keister, P.L.; Figgins, P.W.; Watrous, R.M.

1978-01-01

The current production methods used to recover and purify uranium-234 from aged plutonium-238 at Mound Laboratory are presented. The three chemical separation steps are described in detail. In the initial separation step, the bulk of the plutonium is precipitated as the oxalate. Successively lower levels of plutonium are achieved by anion exchange in nitrate media and by anion exchange in chloride media. The procedures used to characterize and analyze the final U 3 O 8 are given

Supported extractant membranes for americium and plutonium recovery

International Nuclear Information System (INIS)

Muscatello, A.C.; Navratil, J.D.; Killion, M.E.; Price, M.Y.

1987-01-01

Solid supported liquid membranes(SLM) are useful in transferring and concentrating americium and plutonium from nitrate solutions. Specifically, DHDECMP(dihexyl-N,N-diethylcarbamoylmethylphosphonate) supported on Accurel or Celgard polypropylene hollow fibers assembled in modular form transfers >95% of the americium and >70% of the plutonium from high nitrate (6.9 M), low acid (0.1 M) feeds into 0.25 M oxalic acid stripping solution. Membranes supporting TBP (tri-n-butylphosphate) also transfer these metal ions. Maximum permeabilities were observed to be 1 x 10 -3 cm sec -1 , similar to the values for other systems. The feed:strip volume ratio shows an inverse relationship to the fraction of metal ion transferred. Cation exchangers may be used to concentrate americium from the strip solution

Determination of free acid in plutonium (IV) solutions - thermometrically, potentiometrically

International Nuclear Information System (INIS)

Williams, T.L.; Tucker, G.M.; Huff, G.A.; Jordan, L.G.

1981-09-01

The thermometric titration technique was found to offer certain advantages over potentiometry in the determination of free acid in Pu(IV) solutions. The thermometric technique was applied to the determination of free acid in plutonium nitrate solutions using potassium fluoride to suppress the hydrolytic interference of plutonium(IV). The results indicate that 0.2 to 2.0 milliequivalents of free acid can be determined with acceptable bias and precision in solutions containing up to 30 milligrams of plutonium. In contrast, neither the thermometric nor the potentiometric technique was suitable for samples containing more than eight milligrams of plutonium complexed with potassium oxalate

EXPECTED IMPACT OF HANFORD PROCESSING ORGANICS OF PLUTONIUM DURING TANK WASTE SLUDGE RETRIEVAL

International Nuclear Information System (INIS)

TROYER, G.L.; WINTERS, W.I.

2004-01-01

This document evaluates the potential for extracting plutonium from Hanford waste tanks into residual organic solvents and how this process may have an impact on criticality specifications during the retrieval of wastes. The two controlling factors for concentrating plutonium are the solubility of the plutonium in the wastes and the extraction efficiency of the potential organic extractants that may be found in these wastes. Residual Hanford tank sludges contain plutonium in solid forms that are expected to be primarily insoluble Pu(IV) hydroxides. Evaluation of thermodynamic Pourbaix diagrams, documentation on solubility studies of various components in waste tank matrices, and actual analysis of plutonium in tank supernates all indicate that the solubility of Pu in the alkaline waste is on the order of 10 -6 M. Based on an upper limit plutonium solubility of 10 -5 M in high pH and a conservative distribution coefficient for organic extractants of a 0 for plutonium in 30% TBP at 0.07 M HNO 3 ), the estimated concentration for plutonium in the organic phase would be -7 M. This is well below the process control criteria. A significant increase in plutonium solubility or the E a o would have to occur to raise this concentration to the 0.01 M concern level for organics. Measured tank chemical component values, expected operating conditions, and the characteristics of the expected chemistry and extraction mechanisms indicate that concentration of plutonium from Hanford tank residual sludges to associated process organic extractants is significantly below levels of concern

Standard test method for plutonium by Iron (II)/Chromium (VI) amperometric titration

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method covers the determination of plutonium in unirradiated nuclear-grade plutonium dioxide, uranium-plutonium mixed oxides with uranium (U)/plutonium (Pu) ratios up to 21, plutonium metal, and plutonium nitrate solutions. Optimum quantities of plutonium to measure are 7 to 15 mg. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Dynamic process model of a plutonium oxalate precipitator

International Nuclear Information System (INIS)

Borgonovi, G.M.; Hammelman, J.E.; Miller, C.L.

1980-01-01

A dynamic model of a plutonium oxalate precipitator is developed to provide a means of predicting plutonium inventory on a continuous basis. The model is based on state-of-the-art crystallization equations, which describe nucleation and growth phenomena. The model parameters were obtained through the use of batch experimental data. The model has been used to study the approach to steady state, to investigate the response to input transients, and to simulate the control of the precipitation process. 12 refs

Evaluation of indigenous anion exchange resins for plutonium purification

International Nuclear Information System (INIS)

Kumaresan, R.; Sabharwal, K.N.; Srinivasan, T.G.; Vasudeva Rao, P.R.; Thite, B.S.; Ajithlal, R.T.; Sinalkar, Nitin; Dharampurikar, G.R.; Janardhanan, C.; Michael, K.M.; Vijayan, K.; Jambunathan, U.; Dey, P.K.

2004-01-01

Preliminary data with pure plutonium nitrate solution indicate that indigenous anion exchange resin can be used for the purification and concentration of plutonium. However, further studies are required to be conducted on larger scale with actual plant feed solutions before arriving to final conclusions. This includes repeated loading and elution cycles studies with the same bed and evaluation of the performance after each cycle

Development of an expert system for analysis of plutonium processing operations

International Nuclear Information System (INIS)

Boeringter, S.T.; Fasel, J.H.; Kornreich, D.E.

2001-01-01

At Los Alamos National Laboratory (LANL) an expert system has been developed for the analysis and assessment of plutonium processing operations. This system is based upon an object-oriented simulation environment specifically developed for the needs of nuclear material processing. The simulation environment, called the ''Process Modeling System'' (ProMoS), contains a library of over 250 plutonium-based unit process operations ranging from analytical chemistry, oxide operations, recycle and recovery, waste management, and component fabrication. (author)

Biological behaviour of plutonium inhaled by baboons as plutonium n-tributylphosphate complex. Comparison with ICRP models

International Nuclear Information System (INIS)

Metivier, H.; Duserre, C.; Rateau, G.; Legendre, N.; Masse, R.; Piechowski, J.; Menoux, B.

1989-01-01

In order to devise a model capable of calculating committed doses for workers contaminated by inhalation of plutonium tributylphosphate complex during reprocessing, we investigated the biokinetics of plutonium in baboons after inhalation of this chemical form. The animals were killed 0.6, 3, 15, 30, 90 and 365 days post inhalation. Urine and faeces were collected daily. After killing, the main organs were collected for chemical analysis. In order to improve our knowledge of the behaviour of systemic plutonium, three baboons were given an intravenous injection of Pu-TBP and were respectively killed 2, 30 and 365 days post injection. We observed that Pu-TBP could be classified as a W compound, with a half-time for lung clearance of 150 days. Urinary Pu excretion was 3 times higher than was expected from Durbin's model, suggesting that Pu introduced as Pu-TBP, is extremely mobile, and that the complex formed with blood proteins differs from the one formed after inhalation of plutonium nitrate. (author)

On-line estimator/detector design for a plutonium nitrate concentrator unit

International Nuclear Information System (INIS)

Candy, J.V.; Rozsa, R.B.

1979-04-01

In this report we consider the design of a nonlinear estimator to be used in conjunction with on-line detectors for a plutonium/concentrator. Using a complex state-of-the-art process model to simulate realistic data, we show that the estimator performance using a simplified process model is adequate over a wide range of operation. The estimator is used to simulate and characterize some on-line diversion detectors, i.e., detectors designed to indicate if some of the critical special nuclear material in process is stolen or diverted from the unit. Several different diversion scenarios are presented. Simulation results indicate that the estimators and detectors yielded reasonable performance for the scenarios investigated

Criticality of mixtures of plutonium and high enriched uranium

International Nuclear Information System (INIS)

Grolleau, E.; Lein, M.; Leka, G.; Maidou, B.; Klenov, P.

2003-01-01

This paper presents a criticality evaluation of moderated homogeneous plutonium-uranium mixtures. The fissile media studied are homogeneous mixtures of plutonium and high enriched uranium in two chemical forms: aqueous mixtures of metal and mixtures of nitrate solutions. The enrichment of uranium considered are 93.2wt.% 235 U and 100wt.% 235 U. The 240 Pu content in plutonium varies from 0wt.% 240 Pu to 12wt.% 240 Pu. The critical parameters (radii and masses of a 20 cm water reflected sphere) are calculated with the French criticality safety package CRISTAL V0. The comparison of the calculated critical parameters as a function of the moderator-to-fuel atomic ratio shows significant ranges in which high enriched uranium systems, as well as plutonium-uranium mixtures, are more reactive than plutonium systems. (author)

Treatment of plutonium contaminated ashes by electrogenerated Ag(II): a new, simple and efficient process

International Nuclear Information System (INIS)

Madic, C.; Saulze, J.L.; Bourges, J.; Lecomte, M.; Koehly, G.

1990-01-01

Incineration is a very attractive technique for managing plutonium contaminated solid wastes, allowing for large volume and mass reduction factors. After waste incineration, the plutonium is concentrated in the ashes and an efficient method must be designed for its recovery. To achieve this goal, a process based on the dissolution of plutonium in nitric solution under the agressive action of electrogenerated Ag(II) was developed. This process is very simple, requiring very few steps. Plutonium recovery yields up to 98% can be obtained and, in addition, the plutonium bearing solutions generated by the treatment can be processed by the PUREX technique for plutonium recovery. This process constitutes the basis for the development of industrial facilities: 1) a pilot facility is being built in MARCOULE (COGEMA, UP1 plant), to treat active ash in 1990; 2) an industrial facility will be built in the MELOX plant under construction at MARCOULE (COGEMA plant)

Chemical species of plutonium in Hanford radioactive tank waste

International Nuclear Information System (INIS)

Barney, G.S.

1997-01-01

Large quantities of radioactive wastes have been generated at the Hanford Site over its operating life. The wastes with the highest activities are stored underground in 177 large (mostly one million gallon volume) concrete tanks with steel liners. The wastes contain processing chemicals, cladding chemicals, fission products, and actinides that were neutralized to a basic pH before addition to the tanks to prevent corrosion of the steel liners. Because the mission of the Hanford Site was to provide plutonium for defense purposes, the amount of plutonium lost to the wastes was relatively small. The best estimate of the amount of plutonium lost to all the waste tanks is about 500 kg. Given uncertainties in the measurements, some estimates are as high as 1,000 kg (Roetman et al. 1994). The wastes generally consist of (1) a sludge layer generated by precipitation of dissolved metals from aqueous wastes solutions during neutralization with sodium hydroxide, (2) a salt cake layer formed by crystallization of salts after evaporation of the supernate solution, and (3) an aqueous supernate solution that exists as a separate layer or as liquid contained in cavities between sludge or salt cake particles. The identity of chemical species of plutonium in these wastes will allow a better understanding of the behavior of the plutonium during storage in tanks, retrieval of the wastes, and processing of the wastes. Plutonium chemistry in the wastes is important to criticality and environmental concerns, and in processing the wastes for final disposal. Plutonium has been found to exist mainly in the sludge layers of the tanks along with other precipitated metal hydrous oxides. This is expected due to its low solubility in basic aqueous solutions. Tank supernate solutions do not contain high concentrations of plutonium even though some tanks contain high concentrations of complexing agents. The solutions also contain significant concentrations of hydroxide which competes with other

The plutonium fuel cycles

International Nuclear Information System (INIS)

Pigford, T.H.; Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

Conversion of plutonium scrap and residue to boroilicate glass using the GMODS process

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.; Parker, G.W.; Rudolph, J.; Elam, K.R.; Ferrada, J.J.

1995-01-01

Plutonium scrap and residue represent major national and international concerns because (1) significant environmental, safety, and health (ES ampersand H) problems have been identified with their storage; (2) all plutonium recovered from the black market in Europe has been from this category; (3) storage costs are high; and (4) safeguards are difficult. It is proposed to address these problems by conversion of plutonium scrap and residue to a CRACHIP (CRiticality, Aerosol, and CHemically Inert Plutonium) glass using the Glass Material Oxidation and Dissolution System (GMODS). CRACHIP refers to a set of requirements for plutonium storage forms that minimize ES ampersand H concerns. The concept is several decades old. Conversion of plutonium from complex chemical mixtures and variable geometries into a certified, qualified, homogeneous CRACHIP glass creates a stable chemical form that minimizes ES ampersand H risks, simplifies safeguards and security, provides an easy-to-store form, decreases storage costs, and allows for future disposition options. GMODS is a new process to directly convert metals, ceramics, and amorphous solids to glass; oxidize organics with the residue converted to glass; and convert chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, and other materials to glass. GMODS is an enabling technology that creates new options. Conventional glassmaking processes require conversion of feeds to oxide-like forms before final conversion to glass. Such chemical conversion and separation processes are often complex and expensive

Factors Controlling Redox Speciation of Plutonium and Neptunium in Extraction Separation Processes

Energy Technology Data Exchange (ETDEWEB)

Paulenova, Alena [Principal Investigator; Vandegrift, III, George F. [Collaborator

2013-09-24

The objective of the project was to examine the factors controlling redox speciation of plutonium and neptunium in UREX+ extraction in terms of redox potentials, redox mechanism, kinetics and thermodynamics. Researchers employed redox-speciation extractions schemes in parallel to the spectroscopic experiments. The resulting distribution of redox species w studied uring spectroscopic, electrochemical, and spectro-electrochemical methods. This work reulted in collection of data on redox stability and distribution of redox couples in the nitric acid/nitrate electrolyte and the development of redox buffers to stabilize the desired oxidation state of separated radionuclides. The effects of temperature and concentrations on the redox behavior of neptunium were evaluated.

The Plutonium Temperature Effect Experimental Program

Energy Technology Data Exchange (ETDEWEB)

Haeck, Wim; Leclaire, Nicolas; Letang, Eric [IRSN, Fontenay-aux-Roses (France); Girault, Emmanuel; Fouillaud, Patrick [CEA, VALDUC (France)

2008-07-01

Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect but (up to now) no experimental program has confirmed this effect. The main goal of the French Plutonium Temperature Effect Experimental Program (or PU+ in short) is to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The experiments were conducted in the 'Apparatus B' facility at the CEA Valduc research centre in France and involved several sub-critical approach type of experiments using plutonium nitrate solutions with concentrations of 14.3, 15 and 20 g/l at temperatures ranging from 20 to 40 deg. C. A total number of 14 phase I experiments (consisting of independent subcritical approaches) have been performed (5 at 20 g/l, 4 at 15 g/l and 5 at 14.3 g/l) between 2006 and 2007. The impact of the uncertainties on the solution acidity and the plutonium concentration makes it difficult to clearly demonstrate the positive temperature effect, requiring an additional phase II experiment (in which the use of the same plutonium solution was ensured) from 22 to 28 deg. C performed in July 2007. This experiment has shown the existence of a positive temperature effect approx +2 pcm/deg. C (from 22 to 28 deg. C for a plutonium concentration of 14.3 g/l). (authors)

Applicability of hydroxylamine nitrate reductant in pulse-column contactors

International Nuclear Information System (INIS)

Reif, D.J.

1983-05-01

Uranium and plutonium separations were made from simulated breeder reactor spent fuel dissolver solution with laboratory-sized pulse column contactors. Hydroxylamine nitrate (HAN) was used for reduction of plutonium (1V). An integrated extraction-partition system, simulating a breeder fuel reprocessing flowsheet, carried out a partial partition of uranium and plutonium in the second contactor. Tests have shown that acceptable coprocessing can be ontained using HAN as a plutonium reductant. Pulse column performance was stable even though gaseous HAN oxidation products were present in the column. Gas evolution rates up to 0.27 cfm/ft 2 of column cross section were tested and found acceptable

Improving Efficiency with 3-D Imaging: Technology Essential in Removing Plutonium Processing Equipment from Plutonium Finishing Plant Gloveboxes

International Nuclear Information System (INIS)

Crow, Stephen H.; Kyle, Richard N.; Minette, Michael J.

2008-01-01

The Plutonium Finishing Plant at Hanford, Washington began operations in 1949 to process plutonium and plutonium products. Its primary mission was to produce plutonium metal, fabricate weapons parts, and stabilize reactive materials. These operations, and subsequent activities, were performed in remote production lines, consisting primarily of hundreds of gloveboxes. Over the years these gloveboxes and processes have been continuously modified. The plant is currently inactive and Fluor Hanford has been tasked to clean out contaminated equipment and gloveboxes from the facility so it can be demolished in the near future. Approximately 100 gloveboxes at PFP have been cleaned out in the past four years and about 90 gloveboxes remain to be cleaned out. Because specific commitment dates for this work have been established with the State of Washington and other entities, it is important to adopt work practices that increase the safety and speed of this effort. The most recent work practice to be adopted by Fluor Hanford D and D workers is the use of 3-D models to improve the efficiency of cleaning out radioactive gloveboxes at the plant. The use of 3-D models has significantly improved the work planning process by providing workers with a clear image of glovebox construction and composition, which is then used to determine cleanout methods and work sequences. The 3-D visual products enhance safety by enabling workers to more easily identify hazards and implement controls. In addition, the ability to identify and target the removal of radiological materials early in the D and D process provides substantial dose reduction for the workers

Plutonium and americium recovery from spent molten-salt-extraction salts with aluminum-magnesium alloys

International Nuclear Information System (INIS)

Cusick, M.J.; Sherwood, W.G.; Fitzpatrick, R.F.

1984-01-01

Development work was performed to determine the feasibility of removing plutonium and americium from spent molten-salt-extraction (MSE) salts using Al-Mg alloys. If the product buttons from this process are compatible with subsequent aqueous processing, the complex chloride-to-nitrate aqueous conversion step which is presently required for these salts may be eliminated. The optimum alloy composition used to treat spent 8 wt % MSE salts in the past yielded poor phase-disengagement characteristics when applied to 30 mol % salts. After a limited investigation of other alloy compositions in the Al-Mg-Pu-Am system, it was determined that the Al-Pu-Am system could yield a compatible alloy. In this system, experiments were performed to investigate the effects of plutonium loading in the alloy, excess magnesium, age of the spent salt on actinide recovery, phase disengagement, and button homogeneity. Experimental results indicate that 95 percent plutonium recoveries can be attained for fresh salts. Further development is required for backlog salts generated prior to 1981. A homogeneous product alloy, as required for aqueous processing, could not be produced

Ultra-small plutonium oxide nanocrystals: an innovative material in plutonium science.

Science.gov (United States)

Hudry, Damien; Apostolidis, Christos; Walter, Olaf; Janssen, Arne; Manara, Dario; Griveau, Jean-Christophe; Colineau, Eric; Vitova, Tonya; Prüssmann, Tim; Wang, Di; Kübel, Christian; Meyer, Daniel

2014-08-11

Apart from its technological importance, plutonium (Pu) is also one of the most intriguing elements because of its non-conventional physical properties and fascinating chemistry. Those fundamental aspects are particularly interesting when dealing with the challenging study of plutonium-based nanomaterials. Here we show that ultra-small (3.2±0.9 nm) and highly crystalline plutonium oxide (PuO2 ) nanocrystals (NCs) can be synthesized by the thermal decomposition of plutonyl nitrate ([PuO2 (NO3 )2 ]⋅3 H2 O) in a highly coordinating organic medium. This is the first example reporting on the preparation of significant quantities (several tens of milligrams) of PuO2 NCs, in a controllable and reproducible manner. The structure and magnetic properties of PuO2 NCs have been characterized by a wide variety of techniques (powder X-ray diffraction (PXRD), X-ray absorption fine structure (XAFS), X-ray absorption near edge structure (XANES), TEM, IR, Raman, UV/Vis spectroscopies, and superconducting quantum interference device (SQUID) magnetometry). The current PuO2 NCs constitute an innovative material for the study of challenging problems as diverse as the transport behavior of plutonium in the environment or size and shape effects on the physics of transuranium elements. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Some aspects of a technology of processing weapons grade plutonium to nuclear fuel

International Nuclear Information System (INIS)

Bibilashvili, Y.; Glagovsky, E.M.; Zakharkin, B.S.; Orlov, V.K.; Reshetnikov, F.G.; Rogozkin, B.G.; Soloni-N, M.I.

2000-01-01

The concept by Russia to use fissile weapons-grade materials, which are being recovered from nuclear pits in the process of disarmament, is based on an assessment of weapons-grade plutonium as an important energy source intended for use in nuclear power plants. However, in the path of involving plutonium excessive from the purposes of national safety into industrial power engineering there are a lot of problems, from which effectiveness and terms of its disposition are being dependent upon. Those problems have political, economical, financial and environmental character. This report outlines several technology problems of processing weapons-grade metallic plutonium into MOX-fuel for reactors based on thermal and fast neutrons, in particular, the issue of conversion of the metal into dioxide from the viewpoint of fabrication of pelletized MOX-fuel. The processing of metallic weapons-grade plutonium into nuclear fuel is a rather complicated and multi-stage process, every stage of which is its own production. Some of the stages are absent in production of MOX-fuel, for instance the stage of the conversion, i.e. transferring of metallic plutonium into dioxide of the ceramic quality. At this stage of plutonium utilization some tasks must be resolved as follows: I. As a result of the conversion, a material purified from ballast and radiogenic admixtures has to be obtained. This one will be applied to fabricate pelletized MOX-fuel going from morphological, physico-mechanical and technological properties. II. It is well known that metallic gallium, which is used as an alloying addition in weapons-grade plutonium, actively reacts with multiple metals. Therefore, an important issue is to study the effect of gallium on the technology of MOX-fuel production, quality of the pellets, as well as the interaction of gallium oxide with zirconium and steel shells of fuel elements depending upon the content of gallium in the fuel. The rate of the interaction of gallium oxide

Electrochemical processing of nitrate waste solutions

International Nuclear Information System (INIS)

Genders, D.; Weinberg, N.; Hartsough, D.

1992-01-01

The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F - ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions

Crystallization of sodium nitrate from radioactive waste

International Nuclear Information System (INIS)

Krapukhin, V.B.; Krasavina, E.P.; Pikaev, A.K.

1997-07-01

From the 1940s to the 1980s, the Institute of Physical Chemistry of the Russian Academy of Sciences (IPC/RAS) conducted research and development on processes to separate acetate and nitrate salts and acetic acid from radioactive wastes by crystallization. The research objective was to decrease waste volumes and produce the separated decontaminated materials for recycle. This report presents an account of the IPC/RAS experience in this field. Details on operating conditions, waste and product compositions, decontamination factors, and process equipment are described. The research and development was generally related to the management of intermediate-level radioactive wastes. The waste solutions resulted from recovery and processing of uranium, plutonium, and other products from irradiated nuclear fuel, neutralization of nuclear process solutions after extractant recovery, regeneration of process nitric acid, equipment decontamination, and other radiochemical processes. Waste components include nitric acid, metal nitrate and acetate salts, organic impurities, and surfactants. Waste management operations generally consist of two stages: volume reduction and processing of the concentrates for storage, solidification, and disposal. Filtration, coprecipitation, coagulation, evaporation, and sorption were used to reduce waste volume. 28 figs., 40 tabs

An MCNP model of glove boxes in a plutonium processing facility

International Nuclear Information System (INIS)

Dooley, D.E.; Kornreich, D.E.

1998-01-01

Nuclear material processing usually occurs simultaneously in several glove boxes whose primary purpose is to contain radioactive materials and prevent inhalation or ingestion of radioactive materials by workers. A room in the plutonium facility at Los Alamos National Laboratory has been slated for installation of a glove box for storing plutonium metal in various shapes during processing. This storage glove box will be located in a room containing other glove boxes used daily by workers processing plutonium parts. An MCNP model of the room and glove boxes has been constructed to estimate the neutron flux at various locations in the room for two different locations of the storage glove box and to determine the effect of placing polyethylene shielding around the storage glove box. A neutron dose survey of the room with sources dispersed as during normal production operations was used as a benchmark to compare the neutron dose equivalent rates calculated by the MCNP model

Solubility of Plutonium (IV) Oxalate During Americium/Curium Pretreatment

International Nuclear Information System (INIS)

Rudisill, T.S.

1999-01-01

ay of plutonium-241) in the dissolved precipitate, a value consistent with the recovery of europium, the americium surrogate.In a subsequent experiment, the plutonium solubility following an oxalate precipitation to simulate the preparation of a slurry feed for a batch melter was 21 mg/mL at 35 degrees C. The increase in solubility compared to the value measured during the pretreatment experiment was attributed to the increased nitrate concentration and ensuing increase in plutonium complexation. The solubility of the plutonium following a precipitant wash with 0.1M oxalic acid was unchanged. The recovery of plutonium from the precipitate slurry was greater than 97 percent allowing an estimation that approximately 92 percent of the plutonium in Tank 17.1 will report to the glass. The behavior of the lanthanides and soluble metal impurities was consistent with the behavior seen during the pretreatment experiment. A trace level material balance showed that 99.9 percent of the americium w as recovered from the precipitate slurry. The overall recovery of americium from the pretreatment and feed preparation processes was greater than 97 percent, which was consistent with the measured recovery of the europium surrogate

Adaptation of the IBM ECR [electric cantilever robot] robot to plutonium processing applications

International Nuclear Information System (INIS)

Armantrout, G.A.; Pedrotti, L.R.; Halter, E.A.; Crossfield, M.

1990-12-01

The changing regulatory climate in the US is adding increasing incentive to reduce operator dose and TRU waste for DOE plutonium processing operations. To help achieve that goal the authors have begun adapting a small commercial overhead gantry robot, the IBM electric cantilever robot (ECR), to plutonium processing applications. Steps are being taken to harden this robot to withstand the dry, often abrasive, environment within a plutonium glove box and to protect the electronic components against alpha radiation. A mock-up processing system for the reduction of the oxide to a metal was prepared and successfully demonstrated. Design of a working prototype is now underway using the results of this mock-up study. 7 figs., 4 tabs

Zirconolite glass-ceramics for plutonium immobilization: The effects of processing redox conditions on charge compensation and durability

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yingjie, E-mail: yzx@ansto.gov.au; Gregg, Daniel J.; Kong, Linggen; Jovanovich, Miodrag; Triani, Gerry

2017-07-15

Zirconolite glass-ceramic samples doped with plutonium have been prepared via hot isostatic pressing. The effects of processing redox and plutonium loadings on plutonium valences, the presence of cation vacancies, zirconolite phase compositions, microstructures and durability have been investigated. Either tetravalent or trivalent plutonium ions may be incorporated on the Ca-site of CaZrTi{sub 2}O{sub 7} zirconolite with the Ca-site cation vacancies and the incorporation of Al{sup 3+} ions on the Ti-site for charge compensation. Plutonium and gadolinium (as a neutron absorber) are predominantly partitioned in zirconolite phases leading to the formation of chemically durable glass-ceramics suitable for the immobilization of impure plutonium wastes arising from the nuclear fuel cycle. - Highlights: •Plutonium validations of zirconolite glass-ceramics. •Effects of processing redox and plutonium loading. •Zirconolite phase compositions and plutonium valences. •Cation vacancies and chemical durability.

Uranium/plutonium and uranium/neptunium separation by the Purex process using hydroxyurea

International Nuclear Information System (INIS)

Zhu Zhaowu; He Jianyu; Zhang Zefu; Zhang Yu; Zhu Jianmin; Zhen Weifang

2004-01-01

Hydroxyurea dissolved in nitric acid can strip plutonium and neptunium from tri-butyl phosphate efficiently and has little influence on the uranium distribution between the two phases. Simulating the 1B contactor of the Purex process by hydroxyurea with nitric acid solution as a stripping agent, the separation factors of uranium/plutonium and uranium/neptunium can reach values as high as 4.7 x 10 4 and 260, respectively. This indicates that hydroxyurea is a promising salt free agent for uranium/plutonium and uranium/neptunium separations. (author)

Separation of americium and plutonium from nuclear wastes by the TRUEX process

International Nuclear Information System (INIS)

Leonard, R.A.; Vandegrift, G.F.; Manry, C.W.

1986-01-01

Americium and plutonium can be removed from a transuranic (TRU) waste stream to <10 nCi/g by the TRUEX process. The resulting waste is nontransuranic, greatly reducing disposal costs. An overview is given of the TRUEX process and of centrifugal contactors used to implement this process. Then, a plan for the deployment of TRUEX at the Hanford Site is discussed. Finally, details are given on the proposed use of TRUEX to treat the liquid wastes from the Plutonium Finishing Plant at the Hanford Site

Analytical methods used in plutonium purification cycles by trilaurylamine; Methodes analytiques utilisees dans les cycles de purification du plutonium par la trilaurylamine

Energy Technology Data Exchange (ETDEWEB)

Perez, J J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1965-07-01

The utilisation of trilaurylamine as a solvent extractant for the purification of plutonium has entailed to perfect a set of analytical methods which involves, various techniques. The organic impurities of the solvent can be titrated by gas-liquid chromatography. The titration of the main degradation product, the di-laurylamine, can be accomplished also by spectro-colorimetry. Potentiometry is used for the analysis of the different salts of amine-nitrate-sulfate-bisulfate as also the extracted nitric acid. The determination of the nitrate in aqueous phase is carried out by constant current potentiometry. The range of application, the accuracy and the procedure of these analysis are related in the present report. (author) [French] L'utilisation de la trilaurylamine comme solvant d'extraction pour la purification du plutonium a necessite la mise au point d'un ensemble de methodes analytiques qui ressortent de techniques diverses. Les impuretes organiques du solvant peuvent etre dosees par chromatographie gaz-liquide. Le dosage du principal produit de degradation, la dilaurylamine, peut egalement etre effectue par spectrocolorimetrie. La potentiometrie est utilisee pour les analyses des differents sels d'amine: nitrate, sulfate, bisulfate, ainsi que de l'acide nitrique extrait. La determination des nitrates en phase aqueuse est executee par potentiometrie a courant impose. Le domaine d'application, la precision et le mode operatoire de ces analyses sont indiques dans le present rapport. (auteur)

A portable concentrator for processing plutonium containing solutions

International Nuclear Information System (INIS)

Chamberlain, D.B.; Conner, C.; Chen, L.

1995-01-01

This report describes a horizontal, compact agitated-film concentrator called a Rototherm, manufactured by Artisan Industries, Inc. which can be used to process aqueous solutions of radioactive wastes containing plutonium. The unit is designed to concentrate liquid streams to a high-solid content slurry

Influence of Acidity on Uranyl Nitrate Association in Aqueous Solutions: A Molecular Dynamics Simulation Study

International Nuclear Information System (INIS)

de Almeida, Valmor F.; Cui, Shengting; Khomami, Bamin; Ye, Xianggui; Smith, Rodney Bryan

2010-01-01

Uranyl ion complexation with water and nitrate is a key aspect of the uranium/plutonium extraction process. We have carried out a molecular dynamics simulation study to investigate this complexation process, including the molecular composition of the various complex species, the corresponding structure, and the equilibrium distribution of the complexes. The observed structures of the complexes suggest that in aqueous solution, uranyls are generally hydrated by 5 water molecules in the equatorial plane. When associating with nitrate ions, a water molecule is replaced by a nitrate ion, preserving the five-fold coordination and planar symmetry. Analysis of the pair correlation function between uranyl and nitrate suggests that nitrates bind to uranyl in aqueous solution mainly in a monodentate mode, although a small portion of bidentates occur. Dynamic association and dissociation between uranyls and nitrates take place in aqueous solution with a substantial amount of fluctuation in the number of various uranyl nitrate species. The average number of the uranyl mononitrate complexes shows a dependence on acid concentration consistent with equilibrium-constant analysis, namely, the concentration of [UO2NO3]+ increases with nitric acid concentration.

The nitrate to ammonia and ceramic (NAC) process

International Nuclear Information System (INIS)

Mattus, A.J.; Lee, D.D.

1993-01-01

A new low-temperature (50--60 degrees C) process for the reduction of nitrate or nitrite to ammonia gas in a stirred, ethylene glycol led reactor has been developed. The process has nearly completed 2 years of bench-top testing in preparation for a pilot-scale demonstration in the fall of 1994. The nitrate to ammonia and ceramic (NAC) process utilizes the active metal Al (in powder or shot form) in alkaline solution to convert nitrate to ammonia gas with the liberation of heat. Between 0.8 and 1.6 kg of Al per kilogram of sodium nitrate is required to convert solutions of between 3.1 and 6.2 M nitrate to near zero concentration. Prior to feeding Al to the reactor, 40 μm quartz is added based upon the total sodium content of the waste. Upon adding the Al, a by-product of gibbsite precipitates in the reactor as the ammonia leaves the solution. At the end of the reaction, the alumina-silica-based solids are dewatered, calcined, pressed, and sintered into a hard ceramic. Comparing the volume of the final ceramic product with the volume of the starting waste solution, we obtain an ∼70% volume reduction. This compares with an expected 50% volume increase if the waste were immobilized in cement-based grout. The process is being developed for use at Hanford, where as much as 125,000 tonnes of nitrate salts is stored in 4 million liter tanks. DOE may be able to shred radioactively contaminated scrap aluminum, and use this metal to feed the NAC reactor

Plutonium oxides analysis. Sulfur potentiometric analysis

International Nuclear Information System (INIS)

Anon.

Total sulfur determination (sulfur, sulfates, sulfides ...) in plutonium oxides, suitable for sulfate ion content between 0.003 percent to 0.2 percent, by dissolution in nitric hydrofluoric acid, nitrates elimination, addition of hydrochloric acid and reduction in hydrogen sulfide which is carried by an inert gas and neutralized by sodium hydroxide. Sodium sulfide is titrated with mercuric acetate by constant intensity potentiometry [fr

Plutonium gastrointestinal absorption by adults baboons

International Nuclear Information System (INIS)

Lataillade, G.; Madic, C.; Willemot, J.M.; Germain, P.; Colle, C.; Metivier, H.

1991-01-01

Gastrointestinal absorption of plutonium was investigated in baboons after ingestion of plutonium solution (oxidation states III; IV; V; VI), and plutonium incorporated in soya bean and winkles. We studied the effects of oxidation state and ingested mass for masses ranging from 0.35 μg to 51.6 x 10 +3 μg (4 x 10 -2 to 7776 μg of plutonium per kg of body weight). No clear increase in plutonium retention was shown for concentrations of plutonium smaller than 1 μg/kg. From 1 μg/kg to 1 mg/kg no effects of mass or oxidation state was observed and the mean fractional retention value was 10 -4 . For ingested masses higher than 1 mg/kg the fractional retention values respectively increased for Pu(V) and Pu(III) to (0.9 + 0.2) x 10 -2 and (7.4 + 4.1) x 10 -4 of the ingested mass. This increase might be due to the weak hydrolysis of these oxidation states which would increase gastrointestinal absorption by decrease of hydroxide formation. The fraction of plutonium retained after ingestion of soya bean was (3.0 + 0.5) x 10 -4 about 3 fold higher than the value for 238 Pu nitrate solution. No clear increase in plutonium retention was shown after ingestion of winkles containing 238 Pu. In conclusion, except for high masses of ingested Pu, the retention of which could reach 1% of the ingested dose, our results show that the gastrointestinal transfer factor of 10 -4 proposed by ICRP for gastrointestinal absorption of soluble form of Pu is acceptable, but 10 -3 would provide better safety margin [fr

Aqueous nitrate flowsheet optimization and enhancement using the ATLAS facility

International Nuclear Information System (INIS)

Schreiber, S.B.; Punjak, W.A.; Yarbro, S.L.

1993-08-01

The Advanced Testing Line for Actinide Separations (ATLAS) is a pilot plant of all aqueous nitrate plutonium recovery and purification operations within the Los Alamos Plutonium Facility. The main unit operations include dissolution, anion exchange, precipitations, evaporation, calcination, and waste stream polishing. In the current political environment, the emphasis has been redirected from the traditional goal of recovering a pure plutonium product to that of generating ''clean'' effluents while placing the plutonium into a form suitable for long term storage. The ATLAS facility is uniquely suited to fulfill this new role in the development and demonstration of new or revisited technologies. This report summarizes recent work in equipment improvements to the batch dissolver, an evaluation of homogeneous hydroxide precipitations, a demonstration of nitric acid recycle, and the preparation of neptunium and plutonium standards

Some plutonium IV polymers properties in Purex process

International Nuclear Information System (INIS)

Scoazec, H.; Pasquiou, J.Y.; Germain, M.

1990-01-01

The metabolism of plutonium polymers in fuel reprocessing using the Purex process with tributylphosphate as solvent, and its practical consequence in real operation conditions are examined. Precipitation with dibutylphosphoric acid, a solvent degradation product, occurs both in extraction and stripping units when polymers are present. (author)

Applications of molten salts in plutonium processing

International Nuclear Information System (INIS)

Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

1987-01-01

Plutonium is efficiently recovered from scrap at Los Alamos by a series of chemical reactions and separations conducted at temperatures ranging from 700 to 900 0 C. These processes usually employ a molten salt or salt eutectic as a heat sink and/or reaction medium. Salts for these operations were selected early in the development cycle. The selection criteria are being reevaluated. In this article we describe the processes now in use at Los Alamos and our studies of alternate salts and eutectics

The toxicological study of plutonium: tumor-inducing effect of plutonium in rats

International Nuclear Information System (INIS)

Chen Rusong; Zhao Yongchan; Wang Shoufang; Li Maohe

1992-09-01

The carcinogenic effects of Pu compounds through different routs of administration are introduced. The following results have been obtained: (1) Pu is really a powerful carcinogenic radionuclide. (2) Osteosarcoma can be induced in rats by plutonium nitrate at low level radiation. The incidence is from 38.9% to 42.9% in the 185 kBq/kg group. (3) Lung cancer can be induced by plutonium dioxide and the biological effect of TLN (thoracic lymph node) is significant. (4) Under the combined treatment of 239 Pu, 90 Sr and 144 Ce, the carcinogenic effects is much greater than their separative effect on rats. (5) By using the method of comparative toxicology, the risk of osteosarcoma induced by Pu in human is about 600/(10 6 man·cGy) which is extrapolated from animal to human. The above conclusions are important for evaluating the potential hazard of Pu to human beings

Properties of plutonium

International Nuclear Information System (INIS)

Ahn, Jin Su; Yoon, Hwan Ki; Min, Kyung Sik; Kim, Hyun Tae; Ahn, Jong Sung; Kwag, Eon Ho; Ryu, Keon Joong

1996-03-01

Plutonium has unique chemical and physical properties. Its uniqueness in use has led to rare publications, in Korea. This report covers physical aspects of phase change of metal plutonium, mechanical properties, thermal conductivity, etc, chemical aspects of corrosion, oxidation, how to produce plutonium from spent fuels by describing various chemical treatment methods, which are currently used and were used in the past. It also contains characteristics of the purex reprocessing process which is the most widely used nowadays. And show processes to purify and metalize from recovered plutonium solution. Detection and analysis methods are introduced with key pints for handling, critical safety, toxicity, and effects on peoples. This report gives not only a general idea on what plutonium is, rather than deep technical description, but also basic knowledge on plutonium production and safeguards diversion from the view point of nonproliferation. 18 refs. (Author) .new

Properties of plutonium

Energy Technology Data Exchange (ETDEWEB)

Ahn, Jin Su; Yoon, Hwan Ki; Min, Kyung Sik; Kim, Hyun Tae; Ahn, Jong Sung; Kwag, Eon Ho; Ryu, Keon Joong [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of)

1996-03-01

Plutonium has unique chemical and physical properties. Its uniqueness in use has led to rare publications, in Korea. This report covers physical aspects of phase change of metal plutonium, mechanical properties, thermal conductivity, etc, chemical aspects of corrosion, oxidation, how to produce plutonium from spent fuels by describing various chemical treatment methods, which are currently used and were used in the past. It also contains characteristics of the purex reprocessing process which is the most widely used nowadays. And show processes to purify and metalize from recovered plutonium solution. Detection and analysis methods are introduced with key pints for handling, critical safety, toxicity, and effects on peoples. This report gives not only a general idea on what plutonium is, rather than deep technical description, but also basic knowledge on plutonium production and safeguards diversion from the view point of nonproliferation. 18 refs. (Author) .new.

Baseline process description for simulating plutonium oxide production for precalc project

Energy Technology Data Exchange (ETDEWEB)

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

2017-10-26

Savannah River National Laboratory (SRNL) started a multi-year project, the PreCalc Project, to develop a computational simulation of a plutonium oxide (PuO₂) production facility with the objective to study the fundamental relationships between morphological and physicochemical properties. This report provides a detailed baseline process description to be used by SRNL personnel and collaborators to facilitate the initial design and construction of the simulation. The PreCalc Project team selected the HB-Line Plutonium Finishing Facility as the basis for a nominal baseline process since the facility is operational and significant model validation data can be obtained. The process boundary as well as process and facility design details necessary for multi-scale, multi-physics models are provided.

CONTAMINATED PROCESS EQUIPMENT REMOVAL FOR THE DECOMMISSIONG AND DECONTAMINATION OF THE 232-Z CONTAMINATED WASTE RECOVERY PROCESS FACILITY AT THE PLUTONIUM FINISHING PLANT

International Nuclear Information System (INIS)

HOPKINS, A.M.; MINETTE, M.J.; KLOS, D.B.

2007-01-01

This paper describes the unique challenges encountered and subsequent resolutions to accomplish the deactivation and decontamination of a plutonium ash contaminated building. The 232-Z Contaminated Waste Recovery Process Facility at the Plutonium Finishing Plant was used to recover plutonium from process wastes such as rags, gloves, containers and other items by incinerating the items and dissolving the resulting ash. The incineration process resulted in a light-weight plutonium ash residue that was highly mobile in air. This light-weight ash coated the incinerator's process equipment, which included gloveboxes, blowers, filters, furnaces, ducts, and filter boxes. Significant airborne contamination (over 1 million derived air concentration hours [DAC]) was found in the scrubber cell of the facility. Over 1300 grams of plutonium held up in the process equipment and attached to the walls had to be removed, packaged and disposed. This ash had to be removed before demolition of the building could take place

Distribution and identification of Plutonium(IV) species in tri-n-butyl phosphate/HNO3 extraction system containing acetohydroxamic acid

International Nuclear Information System (INIS)

Tkac, P.; Paulenova, A.; Vandegrift, G.F.; Krebs, J.F.

2009-01-01

There was a significant research progress achieved with the aim to modify conventional PUREX process by stripping of plutonium from the tri-n-butyl phosphate (TBP) extraction product in the form of non-extractable complexes upon addition of back-hold complexation agents. The present paper reports effects of such salt-free complexant, acetohydroxamic acid (HAHA), on distribution ratio of Pu(IV) under wide concentration of nitric acid and additional nitrate. General formula of plutonium species present in the organic phase can be described as Pu(OH) x (AHA) y (NO3) 4-x-y x 2TBP x wHNO 3 . (author)

Conversion of metal plutonium to plutonium dioxide by pyrochemical method

Energy Technology Data Exchange (ETDEWEB)

Panov, A.V.; Subbotin, V.G. [Russian Federal Nuclear Center, ALL-Russian Science and Research Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

2000-07-01

Report contains experimental results on metal plutonium of weapon origin samples conversion to plutonium dioxide by pyrochemical method. Circuits of processes are described. Their advantages and shortcomings are shown. Parameters of plutonium dioxide powders (phase and fraction compositions, poured density) manufactured by pyrochemical method in RFNC-VNIITF are shown as well. (authors)

Design of the Laboratory-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

Energy Technology Data Exchange (ETDEWEB)

Lumetta, Gregg J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meier, David E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Orton, Robert D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rapko, Brian M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smart, John E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-05-01

This report describes a design for a laboratory-scale capability to produce plutonium oxide (PuO₂) for use in identifying and validating nuclear forensics signatures associated with plutonium production, as well as for use as exercise and reference materials. This capability will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including PuO₂ dissolution, purification of the Pu by ion exchange, precipitation, and re-conversion to PuO₂ by calcination.

Biological nitrate removal processes from drinking water supply-a review.

Science.gov (United States)

Mohseni-Bandpi, Anoushiravan; Elliott, David Jack; Zazouli, Mohammad Ali

2013-12-19

This paper reviews both heterotrophic and autotrophic processes for the removal of nitrate from water supplies. The most commonly used carbon sources in heterotrophic denitrification are methanol, ethanol and acetic acid. Process performance for each feed stock is compared with particular reference nitrate and nitrite residual and to toxicity potential. Autotrophic nitrate removal has the advantages of not requiring an organic carbon source; however the slow growth rate of autotrophic bacteria and low nitrate removal rate have contributed to the fact that relatively few full scale plants are in operation at the present time.

Effects of various moderators on the critical mass of plutonium

International Nuclear Information System (INIS)

Doherty, A.L.

1986-01-01

The fissile material storage tanks in the Hanford Plutonium Critical Mass Laboratory (CML) in Richland, Washington, are presently being upgraded. During the design and planning phase of this modification, criticality analysis was necessary to compare potential moderator/absorber materials used as isolators between tanks. A parameter study was performed to assist in determining the appropriate moderator material to be used in the plutonium nitrate storage tank system in the mix room at the CML. Four moderator/absorber materials were identified as providing adequate isolation between the tanks

PRODUCTION OF PLUTONIUM METAL

Science.gov (United States)

Lyon, W.L.; Moore, R.H.

1961-01-17

A process is given for producing plutonium metal by the reduction of plutonium chloride, dissolved in alkali metal chloride plus or minus aluminum chloride, with magnesium or a magnesium-aluminum alloy at between 700 and 800 deg C and separating the plutonium or plutonium-aluminum alloy formed from the salt.

Development of technology for ammonium nitrate dissociation process

International Nuclear Information System (INIS)

Zakharkin, B.S.; Varykhanov, V.P.; Kucherenko, V.S.; Solov'yeva, L.N.; Revyakin, V.V.

2000-01-01

Ammonia and ammonium carbonate are frequently used as reagents in fuel production and processing of liquid radioactive wastes. In particular, liquid radioactive wastes that contain ammonium nitrate are generated during operations of metal precipitation. In closed vessels at elevated temperature, for example in evaporators or deposits in tubing, ammonium nitrate may explode due to generation of gaseous nitrogen oxides [2]. In this connection, steps have to be taken to rule out conditions that result in explosion. To do that, ammonium nitrate should be removed even prior to the initial stage of its formation. This report gives results of development of a method of dissociating ammonium nitrate

Plutonium uptake by a soil fungus and transport to its spores

International Nuclear Information System (INIS)

Beckert, W.F.; Au, F.H.F.

1976-01-01

Three concentrations of plutonium-238 nitrate, citrate and dioxide were each added to separate plates of malt agar buffered to pH 2.5 and 5.5 to determine the uptake of plutonium from these chemical forms and concentrations by a common soil fungus, Aspergillus niger. After inoculation and incubation, the aerial spores of Aspergillus niger were collected using a technique that excluded the possibility of cross-contamination of the spores by the culture media or by mycelial fragments. 238 Pu was taken up from all three chemical forms and transported to the aerial spores of Aspergillus niger at each concentration and at both pH levels. The specific activities of the spores grown at pH 5.5 were generally at least twice those of the spores grown at pH 2.5. The uptake of plutonium from the dioxide form was about one-third of that from the nitrate and citrate forms at both pH levels. The term 'transport factor' is used as a means to compare the transport of plutonium from the media to the fungal spores; the concentration-independent transport factor is defined as the specific activity of the spores divided by the specific activity of the dry culture medium. Though the transport factors were less than 1, which indicates discrimination against the transport of 238 Pu from the culture media to the spores, these findings suggest that this common soil fungus may be solubilizing soil-deposited plutonium and rendering it more biologically available for higher plants and animals. (author)

Processing plutonium-contaminated soil on Johnston Atoll

International Nuclear Information System (INIS)

Moroney, K.; Moroney, J. III; Turney, J.

1994-01-01

This article describes a cleanup project to process plutonium- and americium-contaminated soil on Johnston Atoll for volume reduction. Thermo Analytical's (TMA's) segmented gate system (SGS) for this remedial operation has been in successful on-site operation since 1992. Topics covered include the basis for development, a description of the Johnston Atoll; the significance of results; the benefits of the technology; applicability to other radiologically contaminated sites. 7 figs., 1 tab

Comparison of materials accounting in conversion and coconversion processes

International Nuclear Information System (INIS)

Dayem, H.A.

1981-01-01

Materials accounting systems performances are compared for plutonium nitrate-to-oxide conversion [Oxalate (III)] and uranium-plutonium coconversion (Coprecal and modified Coprecal). These processes have the same design basis plutonium throughput and achieve this throughput in parallel operating lines. However, the process line configurations differ. In comparing the materials loss detection sensitivities for the three processes, we find better materials loss detection sensitivity for the Oxalate (III) process than for either of the two Coprecal processes, better single-balance detection sensitivity for the original Coprecal than for the modified Coprecal, and better long-term detection sensitivity (> 1d) for the modified Coprecal than for the original Coprecal. Sensivity differences result from differences in in-process inventories, feeding arrangements, and scrap generation

Study of the physico-chemical agents influencing uranium and plutonium extraction by tributylphosphate in nitric media; Etude des facteurs physico-chimiques intervenant dans l'extraction de l'uranium et du plutonium par le phosphate de tributyle en milieu nitrique

Energy Technology Data Exchange (ETDEWEB)

Tarnero, M [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-03-01

The following different factors are reviewed: tributylphosphate concentration, nitric acid concentration, influence of non-extractable nitrates, simultaneous presence of uranium and plutonium, presence of some different ions, temperature, nature of the diluent, addition of a second active solvent (synergic or antagonistic effect), tributylphosphate and diluent degradation. (author) [French] On passe en revue les differents facteurs suivants: concentration en phosphate de tributyle, concentration en acide nitrique, influence des nitrates non-extractibles, presence simultanee d'uranium et de plutonium, presence d'ions divers, addition d'un second solvant actif (effet de synergie, ou effet antagoniste), degradation du phosphate de tributyle et des solvants inertes. (auteur)

Actinide metal processing

International Nuclear Information System (INIS)

Sauer, N.N.; Watkin, J.G.

1992-01-01

A process for converting an actinide metal such as thorium, uranium, or plutonium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is described together with a low temperature process for preparing an actinide oxide nitrate such as uranyl nitrate. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage

The controlled-potential coulometric determination of plutonium based upon cerium oxidation and the Pu022+/Pu4+ valency change

International Nuclear Information System (INIS)

Phillips, G.; Crossley, D.; Venkataramana, P.

1977-09-01

Conditions have been established enabling plutonium to be oxidised quantitatively to the hexavalent state in the working compartment of a controlled-potential coulometric cell using electrogenerated ceric ion or excess ceric nitrate. The excess ceric ion is reduced in situ electrochemically without reduction of the hexavalent plutonium. The plutonium is then determined controlled-potential coulometrically by reduction to Pu 3+ followed by oxidation to Pu 4+ . The first oxidation step is conducted in molar nitric acid solution containing sulphamic acid but the coulometric determination step is conducted in molar sulphuric acid solution. The results obtained in the coulometric determination step were less satisfactory following oxidation with electrogenerated ceric ion rather than with chemically added ceric nitrate. Using the recommended conditions, 6 mg quantities of plutonium can be determined with an accuracy of 100.06% and a precision of 0.12% (coefficient of variation). The behaviour of chromium, manganese and vanadium impurity is reported. (author)

Airborne Release of Particles in Overheating Incidents Involving Plutonium Metal and Compounds

Energy Technology Data Exchange (ETDEWEB)

Schwendiman, L. C.; Mishima, J.; Radasch, C. A. [Battelle Memorial Institute, Pacific Northwest Laboratory, Richland, WA (United States)

1968-12-15

Ever-increasing utilization of nuclear fuels will result in wide-scale plutonium recovery processing, reconstitution of fuels, transportation, and extensive handling of this material. A variety of circumstances resulting in overheating and fires involving plutonium may occur, releasing airborne particles. This work describes the observations from a study in which the airborne release of plutonium and its compounds was measured during an exposure of the material of interest containing plutonium to temperatures which may result from fires. Aerosol released from small cylinders of metallic plutonium ignited in air at temperatures from 410 to 650 Degree-Sign C ranged from 3 x 10{sup -6} to 5 x 10{sup -5} wt%. Particles smaller than 15{mu}m in diameter represented as much as 0.03% of the total released. Large plutonium pieces weighing from 456 to 1770 g were ignited and allowed to oxidize completely in air with a velocity of around 500 cm/sec. Release rates of from 0.0045 to 0.032 wt% per hour were found. The median mass diameter of airborne material was 4 {mu}m. Quenching the oxidation with magnesium oxide sand reduced the release to 2.9 X 10{sup -4} wt% per hour. Many experiments were carried out in which plutonium compounds as powders were heated at temperatures ranging from 700 to 1000 Degree-Sign C with several air flows. Release rates ranged from 5 x 10{sup -8} to 0.9 wt% per hour, depending upon the compound and the conditions imposed. The airborne release from boiling solutions of plutonium nitrate were roughly related to energy of boiling, and ranged from 4 x 10{sup -4} to 2 x 10{sup -1} % for the evaporation of 90% of the solution. The fraction airborne when combustibles contaminated with plutonium are burned is under study. The data reported can be used in assessing the consequences of off-standard situations involving plutonium and its compounds in fires. (author)

An environmentally benign plutonium processing future at Los Alamos

International Nuclear Information System (INIS)

Pillay, K.K.S.

1993-01-01

In recent years, the U.S. Department of Energy (DOE) has elevated environmental restoration and waste management to major mission areas, and it has established the reduction of wastes from DOE facilities as a major objective. The DOE facilities must now comply with all environmental regulations, including special regulations required of federal facilities. In recognition of this shift in philosophy, the plutonium processing facility at Los Alamos National Laboratory (LANL) has adopted the goal of becoming a facility that processes plutonium in a way that produces only environmentally benign waste. Becoming a facility with zero radionuclide and mixed-waste discharge is an extremely challenging goal and one that requires the technical contributions of a multidisciplinary team of experts. While all the technologies necessary to achieve this goal are not yet available, an extensive knowledge base does exist that can be applied to solving the remaining problems. Working toward this goal is a worthwhile endeavor, not only for LANL, but for the nuclear complex of the future

Fluorine and chlorine determination in mixed uranium-plutonium oxide fuel and plutonium dioxide

International Nuclear Information System (INIS)

Elinson, S.V.; Zemlyanukhina, N.A.; Pavlova, I.V.; Filatkina, V.P.; Tsvetkova, V.T.

1981-01-01

A technique of fluorine and chlorine determination in the mixed uranium-plutonium oxide fuel and plutonium dioxide, based on their simultaneous separation by means of pyrohydrolysis, is developed. Subsequently, fluorine is determined by photometry with alizarincomplexonate of lanthanum or according to the weakening of zirconium colouring with zylenol orange. Chlorine is determined using the photonephelometric method according to the reaction of chloride-ion interaction with silver nitrate or by spectrophotometric method according to the reaction with mercury rhodanide. The lower limit of fluorine determination is -6x10 -5 %, of chlorine- 1x10 -4 % in the sample of 1g. The relative mean quadratic deviation of the determination result (Ssub(r)), depends on the character of the material analyzed and at the content of nx10 -4 - nx10 -3 mass % is equal to from 0.05 to 0.32 for fluorine and from 0.11 to 0.35 for chlorine [ru

Improvement of INVS Measurement Uncertainty for Pu and U-Pu Nitrate Solution

Energy Technology Data Exchange (ETDEWEB)

Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Marlow, Johnna Boulds [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Makino, Risa [Japan Atomic Energy Agency (JAEA), Tokai (Japan); Nakamura, Hironbu [Japan Atomic Energy Agency (JAEA), Tokai (Japan)

2017-04-27

In the Tokai Reprocessing Plant (TRP) and the Plutonium Conversion Development Facility (PCDF), a large amount of plutonium nitrate solution which is recovered from light water reactor (LWR) and advanced thermal reactor (ATR), FUGEN are being stored. Since the solution is designated as a direct use material, the periodical inventory verification and flow verification are being conducted by Japan Safeguard Government Office (JSGO) and International Atomic Agency (IAEA).

Six-kilogram-scale electrorefining of plutonium metal

International Nuclear Information System (INIS)

Mullins, L.J.; Morgan, A.N.; Apgar, S.A. III; Christensen, D.C.

1982-09-01

The electrorefining of metallic plutonium scrap to produce high purity metal has been an established procedure at Los Alamos since 1964. This is a batch process and was limited to 4-kg plutonium because of criticality safety considerations. Improvements in critical mass measurements have permitted us to develop a process for 6-kg plutonium. The 6-kg process is now operational. The increased size of the process, together with other improvements which have been made, makes plutonium electrorefining the principal industrial tool for processing and purifying metallic plutonium scrap

Determination of uranium and plutonium by sequential potentiometric titration

International Nuclear Information System (INIS)

Kato, Yoshiharu; Takahashi, Masao

1976-01-01

The determination of uranium and plutonium in mixed oxide fuels has been developed by sequential potentiometric titration. A weighed sample of uranium and plutonium oxides is dissolved in a mixture of nitric and hydrofluoric acids and the solution is fumed with sulfuric acid. After the reduction of uranium and plutonium to uranium(IV) and plutonium(III) by chromium(II) sulfate, 5 ml of buffer solution (KCl-HCl, pH 1.0) is added to the solution. Then the solution is diluted to 30 ml with water and the pH of the solution is adjusted to 1.0 -- 1.5 with 1 M sodium hydroxide. The solution is stirred until the oxidation of the excess of chromium(II) by air is completed. After the removal of dissolved oxygen by bubbling nitrogen through the solution for 10 minutes, uranium (IV) is titrated with 0.1 N cerium(IV) sulfate. Then, plutonium is titrated by 0.02 N cerium(IV) sulfate. When a mixture of uranium and plutonium is titrated with 0.1 N potassium dichromate potentiometrically, the potential change at the end point of plutonium is very small and the end point of uranium is also unclear when 0.1 N potassium permanganate is used as a titrant. In the present method, nitrate, fluoride and copper(II) interfere with the determination of plutonium and uranium. Iron interferes quantitatively with the determination of plutonium but not of uranium. Results obtained in applying the proposed method to 50 mg of mixtures of plutonium and uranium ((7.5 -- 50))% Pu were accurate to within 0.15 mg of each element. (auth.)

Development of a low cost, low environmental impact process for disposal of nitrate wastes

International Nuclear Information System (INIS)

Napier, J.M.

1975-01-01

A uranium recycle process in the Y-12 Plant generates nitrate ions which must be discarded. Scrap enriched uranium is dissolved in nitric acid and solvent extracted to remove impurities from the uranium. Aluminum nitrate is also used in the process to remove the purified uranium from the solvent extraction process. Dilute nitric acid, aluminum nitrate, and metallic impurities must be discarded from this process. A program was started to develop a low cost, low environmental impact process for disposal of these nitrate wastes. Several disposal methods were considered. A process was selected which included: distillation and recycle of nitric acid; crystallization and recycle of aluminum nitrate; and biodegradation of the remaining nitrate waste solutions. For this presentation, only the biodegradation process will be discussed. A colony of Pseudomonas stutzeri, which is capable of using the nitrate ion as the oxygen supply, was used. An excess of organic material was used to insure that the maximum amount of nitrate was destroyed

Plutonium retention in dairy calves following ingestion of either in vivo labeled or in vitro labeled milk

International Nuclear Information System (INIS)

Sutton, W.W.; Patzer, R.G.; Hahn, P.B.; Potter, G.D.

1977-01-01

Holstein calves, 4 to 8 days of age, were used to compare the gastrointestinal uptake of plutonium-238 from either in vivo or in vitro plutonium-labeled milk. In vivo labeled milk was collected from intravenously dosed lactating adults and the in vitro labeled milk was prepared at the same nuclide concentration by adding citrate-buffered plutonium nitrate to uncontaminated cow's milk. Dosing was accomplished with individual plastic feeding buckets twice daily for 7 consecutive days. The calves were sacrificed 2 days after the final plutonium dose, at which time tissues were collected for nuclide analysis

Validation of KENO, ANISN and Hansen-Roach cross-section set on plutonium oxide and metal fuel system

International Nuclear Information System (INIS)

Matsumoto, Tadakuni; Yumoto, Ryozo; Nakano, Koh.

1980-01-01

In the previous report, the authors discussed the validity of KENO, ANISN and Hansen-Roach 16 group cross-section set on the critical plutonium nitrate solution systems with various geometries, absorbers and neutron interactions. The purpose of the present report is to examine the validity of the same calculation systems on the homogeneous plutonium oxide and plutonium-uranium mixed oxide fuels with various density values. Eleven experiments adopted for validation are summarized. First six experiments were performed at Pacific Northwest Laboratory of Battelle Memorial Institute, and the remaining five at Los Alamos Scientific Laboratory. The characteristics of core fuel are given, and the isotopic composition of plutonium, the relation between H/(Pu + U) atomic ratio and fuel density as compared with the atomic ratios of PuO 2 and mixed oxides in powder storage and pellet fabrication processes, and critical core dimensions and reflector conditions are shown. The effective multiplication factors were calculated with the KENO code. In case of the metal fuels with simple sphere geometry, additional calculations with the ANISN code were performed. The criticality calculation system composed of KENO, ANISN and Hansen-Roach cross-section set was found to be valid for calculating the criticality on plutonium oxide, plutonium-uranium mixed oxide, plutonium metal and uranium metal fuel systems as well as on plutonium solution systems with various geometries, absorbers and neutron interactions. There seems to remain some problems in the method for evaluating experimental correction. Some discussions foloow. (Wakatsuki, Y.)

An Improved Plutonium Trifluoride Precipitation Flowsheet

Energy Technology Data Exchange (ETDEWEB)

Harmon, H.D.

2001-06-26

This report discusses results of the plutonium trifluoride two-stage precipitation study. A series of precipitation experiments was used to identify the significant process variables affecting precipitation performance. A mathematical model of the precipitation process was developed which is based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter which can be used to control the performance of the plutonium trifluoride precipitation process. Recommendations have been made which will optimize the FB-Line plutonium trifluoride precipitation process.

An Improved Plutonium Trifluoride Precipitation Flowsheet

International Nuclear Information System (INIS)

Harmon, H.D.

2001-01-01

This report discusses results of the plutonium trifluoride two-stage precipitation study. A series of precipitation experiments was used to identify the significant process variables affecting precipitation performance. A mathematical model of the precipitation process was developed which is based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter which can be used to control the performance of the plutonium trifluoride precipitation process. Recommendations have been made which will optimize the FB-Line plutonium trifluoride precipitation process

Provision of NDA instrumentation for the control of operations on plutonium finishing and waste plants at the Sellafield nuclear fuel reprocessing facility

International Nuclear Information System (INIS)

Whitehouse, K.R.; Orr, C.H.

1995-01-01

On BNFL's Sellafield site a significant number of major plants are involved in the handling, processing and storage of plutonium in various forms including nitrate, oxide and mixed oxide (MOX). Other plants in operation or under construction treat and prepare for storage, plutonium bearing wastes in the form of plutonium contaminated materials -- PCM (transuranic waste -- TRU) or low level waste. Concurrently, a number of old plutonium handling plants are being decommissioned. The safety and cost effectiveness of these widely varying operations has been ensured by the development and installation of a wide range of special radiometric instrumentation. These systems based on a range of neutron counting and high resolution gamma spectrometric techniques -- singly or in combination -- enable BNFL to maintain a detailed and comprehensive picture of the disposition of plutonium within each plant and across the site. This paper describes an overview of the range of plant and paper prove waste measurement systems in this context, highlighting the specific roles of the Plutonium Inventory Measurement System (PIMS) for real time accountancy and the Decommissioning In-Situ Plutonium Inventory Monitor (DISPIM) for material control during decommissioning

Aqueous Nitrate Recovery Line at Los Alamos National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Finstad, Casey Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-06-15

This powerpoint is part of the ADPSM Plutonium Engineering Lecture Series, which is an opportunity for new hires at LANL to get an overview of work done at TA55. It goes into detail about the aqueous nitrate recovery line at Los Alamos National Laboratory.

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

Determination of uranium and plutonium in urine of people working with regenerated uranium

International Nuclear Information System (INIS)

Golutvina, M.M.; Ryzhova, E.A.

1987-01-01

Method of determining uranium and plutonium content in urine with their combined presence up to α-activity ratio Pu:U=1:100 is developed. The method is based on extraction chromatographic separation of nuclides using trimethyloctylammonium nitrate and their subsequent α-spectrometric determination. The coefficient of plutonium purification from uranium makes up 750. Chemical yield of Pu is 72±6%, U-76±8%. The method sensitivity is 0.2 decompositions per minute for a sample

Long time contamination from plutonium

Energy Technology Data Exchange (ETDEWEB)

Fueloep, M; Patzeltova, N; Ragan, P [Inst. of Preventive and Clinical Medicine, Bratislava (Slovakia); Matel, L [Comenius Univ., Bratislava (Slovakia). Department of Nuclear Chemistry

1996-12-31

Plutonium isotopes in the organism of the patient (who had participated in the liquidation works after the Chernobyl accident; for three month he had stayed in the epicenter, where he acted as a chauffeur driving a radioactive material to the place of destination) from urine were determined. For determination of the concentration of Pu-239, Pu-240 in urine a modified radiochemical method was used. After mineralization the sample was separated as an anion-nitrate complex with contact by the anion form of the resin in the column. The resin was washed by 8 M HNO{sub 3}, the 8 M HCl with 0.3 M HNO{sub 3} for removing the other radionuclides. The solution 0.36 M HCl with 0.01 M HF was used for the elution of plutonium. Using the lanthanum fluoride technique the sample was filtrated through a membrane filter. The plutonium was detected in the dry sample. The Pu-239 tracer was used for the evaluation of the plutonium separation efficiency. The alpha spectrometric measurements were carried out with a large area silicon detector. The samples were measured and evaluated in the energy region 4.98-5.18 MeV. The detection limit of alpha spectrometry measurements has been 0.01 Bq dm{sup -3}. The concentration of plutonium in the 24-hour urine was determined three times in the quarter year intervals. The results are: 54 mBq, 63.2 mBq, 53 mBq, with average 56,7 mBq. From the results of the analyses of plutonium depositions calculated according to ICRP 54 the intake of this radionuclide for the patient was 56.7 kBq. To estimate a committed effective dose (50 years) from the intake of plutonium was used a conversion factor 6.8.10{sup -5} Sv.Bq{sup -1} (class W). So the expressed committed effective dose received from the plutonium intake is 3.8 Sv. This number is relatively high and all the effective dose will be higher, because the patient was exposed to the other radionuclides too. (Abstract Truncated)

Plutonium Plant, Trombay

International Nuclear Information System (INIS)

Yadav, J.S.; Agarwal, K.

2017-01-01

The journey of Indian nuclear fuel reprocessing started with the commissioning of Plutonium Plant (PP) at Trombay on 22"n"d January, 1965 with an aim to reprocess the spent fuel from research reactor CIRUS. The basic process chosen for the plant was Plutonium Uranium Reduction EXtraction (PUREX) process. In seventies, the plant was subjected to major design modifications and replacement of hardware, which later met the additional demand from research reactor DHRUVA. The augmented plutonium plant has been operating since 1983. Experience gained from this plant was very much helpful to design future reprocessing plant in the country

Plutonium fires; Incendies de plutonium

Energy Technology Data Exchange (ETDEWEB)

Mestre, E.

1959-06-23

The author reports an information survey on accidents which occurred when handling plutonium. He first addresses accidents reported in documents. He indicates the circumstances and consequences of these accidents (explosion in glove boxes, fires of plutonium chips, plutonium fire followed by filter destruction, explosion during plutonium chip dissolution followed by chip fire). He describes hazards associated with plutonium fires: atmosphere and surface contamination, criticality. The author gives some advices to avoid plutonium fires. These advices concern electric installations, the use of flammable solvents, general cautions associated with plutonium handling, venting and filtration. He finally describes how to fight plutonium fires, and measures to be taken after the fire (staff contamination control, atmosphere control)

Continuous Material Balance Reconciliation for a Modern Plutonium Processing Facility

International Nuclear Information System (INIS)

CLARK, THOMASG.

2004-01-01

This paper describes a safeguards approach that can be deployed at any modern plutonium processing facility to increase the level of safeguards assurance and significantly reduce the impact of safeguards on process operations. One of the most perplexing problems facing the designers of plutonium processing facilities is the constraint placed upon the limit of error of the inventory difference (LEID). The current DOE manual constrains the LEID for Category I and II material balance areas to 2 per cent of active inventory up to a Category II quantity of the material being processed. For 239Pu a Category II quantity is two kilograms. Due to the large material throughput anticipated for some of the modern plutonium facilities, the required LEID cannot be achieved reliably during a nominal two month inventory period, even by using state-of-the-science non-destructive assay (NDA) methods. The most cost-effective and least disruptive solution appears to be increasing the frequency of material balance closure and thus reducing the throughput being measured during each inventory period. Current inventory accounting practices and systems can already provide the book inventory values at any point in time. However, closing the material balance with measured values has typically required the process to be cleaned out, and in-process materials packaged and measured. This process requires one to two weeks of facility down time every two months for each inventory, thus significantly reducing productivity. To provide a solution to this problem, a non-traditional approach is proposed that will include using in-line instruments to provide measurement of the process materials on a near real-time basis. A new software component will be developed that will operate with the standard LANMAS application to provide the running material balance reconciliation, including the calculation of the inventory difference and variance propagation. The combined measurement system and software

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

Long time contamination from plutonium

International Nuclear Information System (INIS)

Fueloep, M.; Patzeltova, N.; Ragan, P.; Matel, L.

1995-01-01

Plutonium isotopes in the organism of the patient (who had participated in the liquidation works after the Chernobyl accident; for three month he had stayed in the epicenter, where he acted as a chauffeur driving a radioactive material to the place of destination) from urine were determined. For determination of the concentration of Pu-239, Pu-240 in urine a modified radiochemical method was used. After mineralization the sample was separated as an anion-nitrate complex with contact by the anion form of the resin in the column. The resin was washed by 8 M HNO 3 , the 8 M HCl with 0.3 M HNO 3 for removing the other radionuclides. The solution 0.36 M HCl with 0.01 M HF was used for the elution of plutonium. Using the lanthanum fluoride technique the sample was filtrated through a membrane filter. The plutonium was detected in the dry sample. The Pu-239 tracer was used for the evaluation of the plutonium separation efficiency. The alpha spectrometric measurements were carried out with a large area silicon detector. The samples were measured and evaluated in the energy region 4.98-5.18 MeV. The detection limit of alpha spectrometry measurements has been 0.01 Bq dm -3 . The concentration of plutonium in the 24-hour urine was determined three times in the quarter year intervals. The results are: 54 mBq, 63.2 mBq, 53 mBq, with average 56,7 mBq. From the results of the analyses of plutonium depositions calculated according to ICRP 54 the intake of this radionuclide for the patient was 56.7 kBq. To estimate a committed effective dose (50 years) from the intake of plutonium was used a conversion factor 6.8.10 -5 Sv.Bq -1 (class W). So the expressed committed effective dose received from the plutonium intake is 3.8 Sv. This number is relatively high and all the effective dose will be higher, because the patient was exposed to the other radionuclides too. For example the determination of the rate radionuclides Am-241/Pu-239,Pu-240 was 32-36 % in the fallout after the Chernobyl

Plutonium storage phenomenology

International Nuclear Information System (INIS)

Szempruch, R.

1995-12-01

Plutonium has been produced, handled, and stored at Department of Energy (DOE) facilities since the 1940s. Many changes have occurred during the last 40 years in the sources, production demands, and end uses of plutonium. These have resulted in corresponding changes in the isotopic composition as well as the chemical and physical forms of the processed and stored plutonium. Thousands of ordinary food pack tin cans have been used successfully for many years to handle and store plutonium. Other containers have been used with equal success. This paper addressees the exceptions to this satisfactory experience. To aid in understanding the challenges of handling plutonium for storage or immobilization the lessons learned from past storage experience and the necessary countermeasures to improve storage performance are discussed

Treatment of plutonium process residues by molten salt oxidation

Energy Technology Data Exchange (ETDEWEB)

Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J. [Los Alamos National Lab., NM (United States); Heslop, M. [Naval Surface Warfare Center (United States). Indian Head Div.; Wernly, K. [Molten Salt Oxidation Corp. (United States)

1999-04-01

Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible {sup 238}Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na{sub 2}SO{sub 4}, Na{sub 3}PO{sub 4} and NaAsO{sub 2} or Na{sub 3}AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the {sup 238}Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox.

Treatment of plutonium process residues by molten salt oxidation

International Nuclear Information System (INIS)

Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J.; Heslop, M.

1999-01-01

Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible 238 Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na 2 SO 4 , Na 3 PO 4 and NaAsO 2 or Na 3 AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the 238 Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox

Stop plutonium; Stop plutonium

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-02-01

This press document aims to inform the public on the hazards bound to the plutonium exploitation in France and especially the plutonium transport. The first part is a technical presentation of the plutonium and the MOX (Mixed Oxide Fuel). The second part presents the installation of the plutonium industry in France. The third part is devoted to the plutonium convoys safety. The highlight is done on the problem of the leak of ''secret'' of such transports. (A.L.B.)

Coordinated safeguards for materials management in a nitrate-to-oxide conversion facility

International Nuclear Information System (INIS)

Dayem, H.A.; Cobb, D.D.; Dietz, R.J.; Hakkila, E.A.; Kern, E.A.; Shipley, J.P.; Smith, D.B.; Bowersox, D.F.

1977-09-01

The conceptual design of a materials management system for safeguarding special nuclear materials in a plutonium nitrate-to-oxide conversion facility is developed and evaluated. Dynamic material balances are drawn from information provided by nondestructive-analysis techniques, process-control instrumentation, and conventional chemical analyses augmented by process-monitoring devices. Powerful statistical methods, cast in the framework of decision analysis and applied to unit-process accounting areas, ensure adequate spatial and temporal quantification of possible diversion with minimal process disruption. Modeling and simulation techniques assist in evaluating the sensitivity of the system to various diversion schemes and in comparing safeguards strategies. Features that would improve the safeguardability of the conversion process are discussed

Process control and safeguards system plutonium inventory conrol for MOX fuel facility

International Nuclear Information System (INIS)

Mishima, T.; Aoki, M.; Muto, T.; Amanuma, T.

1979-01-01

The plutonium inventory control (PINC) system is a real-time material accountability control system that is expected to be applied to a new large-scale plutonium fuel production facility for both fast breeder reactor and heavy water reactor at the Power Reactor and Nuclear Development Corporation. The PINC is basically a system for material control but is expected to develop into a whole facility control system, including criticality control, process control, quality control, facility protection, and so forth. Under PINC, every process and storage area is divided into a unit area, which is the smallest unit for both accountability and process control. Item and material weight automatically are accounted for at every unit area, and data are simultaneously treated by a computer network system. Sensors necessary for the system are being developed. 9 figures

The effect of oxidation state on the absorption of ingested or inhaled plutonium

International Nuclear Information System (INIS)

Sullivan, M.F.; Gorham, L.S.; Ryan, J.L.

1980-01-01

Rats received plutonium nitrate (as 239 Pu(IV) or 239 Pu(VI) by gavage or nose-only exposure. Some of the animals were deprived of food before and/or after exposure. Results obtained indicate that there was increased retention of plutonium after exposure to 239 Pu(VI) in comparison with 239 Pu(IV). The absence of food either before and/or after inhalation exposure had no effect on the amount of 239 Pu retained by the liver and carcass. It was concluded that plutonium in its hexavalent state may under certain conditions be more readily absorbed from the GI tract, but that these conditions are unlikely to occur in human exposure. (H.K.)

Thermodynamics of dehydration process of uranyl nitrate pentahydrate of thorium

International Nuclear Information System (INIS)

Khamidov, F.A.; Mirsaidov, I.U.; Nazarov, K.M.; Nasriddinov, S.K.; Badalov, A.B.

2010-01-01

Present article is devoted to thermodynamics of dehydration process of uranyl nitrate pentahydrate of thorium. The results of researches of dehydration process of uranyl nitrate pentahydrate of thorium Th(NO_3)_4·5H_2O were considered. The researches were carried out by means of tensimeter method with membrane zero-manometer under equilibrium conditions and at 300-450 K temperature ranges. The thermodynamic characteristics of dehydration process of initial crystalline hydrate was defined.

Test of a sample container for shipment of small size plutonium samples with PAT-2

International Nuclear Information System (INIS)

Kuhn, E.; Aigner, H.; Deron, S.

1981-11-01

A light-weight container for the air transport of plutonium, to be designated PAT-2, has been developed in the USA and is presently undergoing licensing. The very limited effective space for bearing plutonium required the design of small size sample canisters to meet the needs of international safeguards for the shipment of plutonium samples. The applicability of a small canister for the sampling of small size powder and solution samples has been tested in an intralaboratory experiment. The results of the experiment, based on the concept of pre-weighed samples, show that the tested canister can successfully be used for the sampling of small size PuO 2 -powder samples of homogeneous source material, as well as for dried aliquands of plutonium nitrate solutions. (author)

Study of an automatic dosing of neptunium in the industrial process of separation neptunium 237-plutonium 238

International Nuclear Information System (INIS)

Ros, Pierre

1973-01-01

The objective is to study and to adapt a method of automatic dosing of neptunium to the industrial process of separation and purification of plutonium 238, while taking the information quality and economic aspects into account. After a recall of some generalities on the production of plutonium 238, and the process of separation plutonium-neptunium, the author addresses the dosing of neptunium. The adopted measurement technique is spectrophotometry (of neptunium, of neptunium peroxide) which is the most flexible and economic to adapt to automatic control. The author proposes a project of chemical automatic machine, and discusses the complex (stoichiometry, form) and some aspects of neptunium dosing (redox reactions, process control) [fr

Biological transfer of plutonium via in vivo labeled goat's milk. Final report

International Nuclear Information System (INIS)

Sutton, W.W.; Mullen, A.A.; Lloyd, S.R.; Mosley, R.E.

1976-03-01

The long physical and biological half-life and high relative toxicity have dictated that considerable effort be devoted to quantifying plutonium transport through the various trophic levels. Despite the fact that biological transport of plutonium has been studied for many years, quantitative values for its transfer to milk, and its subsequent uptake by suckling animals have not been established. Three lactating goats were given intravenous injections of citrate-buffered plutonium nitrate at a rate of 75 microcuries per animal per day for three consecutive days. In all three goats approximately one percent of the total plutonium dose was transferred to the milk by the fifth post-treatment day. Plutonium retained by the tissues was deposited primarily in the liver and bone. In vitro plutonium-labeled milk was also fed to groups of rats and juvenile goats. Tissue concentrations of plutonium from juvenile goats which had received either in vivo or in vitro labeled milk were somewhat variable. Due possibly to this, within group variability and the small number of animals per group (two) there were no clearly discernible differences between treatments. The only comparison point to show a consistent trend was the observation that, as expected, juvenile rats retained more of the ingested dose than the adult animals

Chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions

International Nuclear Information System (INIS)

Anon.

1981-01-01

These analytical procedures are designed to show whether a given material meets the purchaser's specifications as to plutonium content, effective fissile content, and impurity content. The following procedures are described in detail: plutonium by controlled-potential coulometry; plutonium by amperometric titration with iron(II); free acid by titration in an oxalate solution; free acid by iodate precipitation-potentiometric titration method; uranium by Arsenazo I spectrophotometric method; thorium by thorin spectrophotometric method; iron by 1,10-phenanthroline spectrophotometric method; chloride by thiocyanate spectrophotometric method; fluoride by distillation-spectrophotometric method; sulfate by barium sulfate turbidimetric method; isotopic composition by mass spectrometry; americium-241 by extraction and gamma counting; americium-241 by gamma counting; gamma-emitting fission products, uranium, and thorium by gamma-ray spectroscopy; rare earths by copper spark spectrochemical method; tungsten, niobium (columbium), and tantalum by spectrochemical method; simple preparation by spectrographic analysis for general impurities

The use of plutonium rapides surregenerateurs aspects techniques et economiques

International Nuclear Information System (INIS)

Guillet, H.; Delayre, R.; Mougniot, J.C.; Ferrari, A.

1977-01-01

Nuclear energy production utilizing U 235 and U 238 inevitably results in the formation of plutonium. Some of this is directly used by the reactor in power production. Some reactors, e.g. the Candu type, burn most of their plutonium ''in situ''. However the surplus quantity of plutonium produced is increasing, and by 1990 the world stock of plutonium is predicted to be about 1000 tons (300 tons in Europe and 400 in the USA). This represents approximately 0.1 Q of potential power, where Q=10 21 joules. Proposals for dealing with this plutonium include its storage, either as irradiated fuel or as a refined substance (plutonium nitrate or oxide); its use in thermal reactors as a substitute for U 235 ; its use in fast breeder reactors which can act as consumers of plutonium as well as producers and which can therefore regulate the world's plutonium stocks. Leaving aside the question of storage, certain technical conditions must be fulfilled. It is essential that reprocessing plants should be operational and available in adequate numbers. An industry able to make use of plutonium should be developed with large capacity units. There seem to be no problems with plutonium use in reactors: conclusive experiments have proved the accuracy of calculations regarding plutonium recycling in thermal reactors and a number of fast breeders have proved the possibility of using plutonium in them. Experience acquired during the past ten years permits us to deal with the safety and safeguards problem of using plutonium. Many economic questions remain, however, including questions affecting the utilization of plutonium, such as the cost of regenerating irradiated fuel and of making fuel, and questions affecting the strategy of using plutonium in different ways, such as the price at which it should be sold. Also the cost of producing electricity using plutonium is not yet clear. It is unavoidable that plutonium will be used in nuclear power development. The technical and economic problems

Advanced local dose rate calculations with the Monte Carlo code MCNP for plutonium nitrate storage containers

International Nuclear Information System (INIS)

Quade, U.

1994-01-01

Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de

Decontaminaion of metals containing plutonium and americium

International Nuclear Information System (INIS)

Seitz, M.G.; Gerding, T.J.; Steindler, M.J.

1979-06-01

Melt-slagging (melt-refining) techniques were evaluated as a decontamination and consolidation step for metals contaminated with oxides of plutonium and americium. Experiments were performed in which mild steel, stainless steel, and nickel contaminated with oxides of plutonium and americium were melted in the presence of silicate slags of various compositions. The metal products were low in contamination, with the plutonium and americium strongly fractionated to the slags. Partition coefficients (plutonium in slag/plutonium in steel) of 7 x 10 6 were measured with boro-silicate slag and of 3 x 10 6 with calcium, magnesium silicate slag. Decontamination of metals containing as much as 14,000 ppM plutonium appears to be as efficient as for metals with plutonium levels of 400 ppM. Staged extraction, that is, a remelting of processed metal with clean slag, results in further decontamination of the metal. The second extraction is effective with either resistance-furnace melting or electric-arc melting. Slag adhering to the metal ingots and in defects within the ingots is in the important contributors to plutonium retained in processed metals. If these sources of plutonium are controlled, the melt-refining process can be used on a large scale to convert highly contaminated metals to homogeneous and compact forms with very low concentrations of plutonium and americium. A conceptual design of a melt-refining process to decontaminate plutonium- and americium-contaminated metals is described. The process includes single-stage refining of contaminated metals to produce a metal product which would have less than 10 nCi/g of TRU-element contamination. Two plant sizes were considered. The smaller conceptual plant processes 77 kg of metal per 8-h period and may be portable.The larger one processes 140 kg of metal per 8-h period, is stationary, and may be near te maximum size that is practical for a metal decontamination process

Selective permeation of plutonium(IV) through a supported liquid membrane containing tri-iso-amyl phosphate as an ionophore

International Nuclear Information System (INIS)

Shukla, J.P.; Kedari, C.S.; Dharmapurikar, G.R.

1998-01-01

Selective ionophoric mobility of plutonium with ease of concentration upgradation from aqueous nitrate solutions was investigated. A thin flat-sheet supported liquid membrane (SLM) impregnated with tri-iso-amyl phosphate (TAP) was used. Accurel polypropylene hydrophobic microporous membrane 'Enka' was tested as the solid polymeric support. The source phase generally contained extremely dilute (ca. 10 -6 mol/dm -3 ) to moderately concentrated plutonium(IV) nitrate solutions (ca. 10 -3 mol/dm -3 ) in about 4 mol/dm -3 HNO 3 . Membrane permeability and selectivity dependency on variables like nitric acid concentration in the source phase, carrier concentration, receiving phase composition, etc. were systematically evaluated. More than 90% pertraction of plutonium could be easily accomplished in single run employing a feed solution consisting of about 1 mg/dm -3 Pu and 4 mol/dm -3 HNO 3 , carrier concentration of 0.8 mol/dm -3 TAP/dodecane; the receiving phase was 0.5 mol/dm -3 sodium carbonate or 0.5 mol/dm -3 ascorbic acid. The selective diffusivity of plutonium(IV) was observed from various effluents originating from fuel reprocessing operations. Reusability of membrane supports was also found to be satisfactory. (author)

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

Hold-up monitoring system for plutonium process tanks

International Nuclear Information System (INIS)

Zhu Rongbao; Jin Huimin; Tan Yajun

1994-01-01

The development of hold-up monitoring system for plutonium process tanks and a calculation method for α activities deposited in containers and inner walls of pipe are described. The hardware of monitoring system consists of a portable HPGe detector, a φ50 mm x 60 mm NaI(Tl) detector, γ-ray tungsten collimators, ORTEC92X Spectrum Master and an AST-286 computer. The software of system includes Maestro Tm for Window3 and a PHOUP1 hold-up application software for user. The Monte-Carlo simulation calculation supported by MCNP software is performed for the probability calculation of all the unscattering γ-rays reaching to the detection positions from the source terms deposited in the complicated tanks. A measurement mean value for different positions is used to minimize the effect of heterogeneous distribution of source term. The sensitivity is better than 3.7 x 10 6 Bq/kg (steel) for a plutonium simulation source on a 3-8 mm thick steel plate surrounded by 0.8 x 10 -10 C/kg·s γ field from long-life fission products

Production of Plutonium Metal from Aqueous Solutions

Energy Technology Data Exchange (ETDEWEB)

Orth, D.A.

2003-01-16

The primary separation of plutonium from irradiated uranium by the Purex solvent extraction process at the Savannah River Plant produces a dilute plutonium solution containing residual fission products and uranium. A cation exchange process is used for concentration and further decontamination of the plutonium, as the first step in the final preparation of metal. This paper discusses the production of plutonium metal from the aqueous solutions.

Electrolytic treatment of liquid waste containing ammonium nitrate

International Nuclear Information System (INIS)

Komori, R.; Ogawa, N.; Ohtsuka, K.; Ohuchi, J.

1981-01-01

A study was made on the safe decomposition of ammonium nitrate, which is the main component of α-liquid waste from plutonium fuel facilities, by means of electrolytic reduction and thermal decomposition. In the first stage, ammonium nitrate is reduced to ammonium nitrite by electrolytic reduction using an electrolyser with a cation exchange membrane as a diaphragm. In the second stage, ammonium nitrite is decomposed to N 2 and H 2 O. The alkaline region and a low temperature are preferable for electrolytic reduction and the acidic region and high temperature for thermal decomposition. A basis was established for an ammonium nitrate treatment system in aqueous solution through the operation of a bench-scale unit, and the operating data obtained was applied to the basic design of a 10-m 3 /a facility. (author)

Preliminary process simulation and analysis of GMODS: Processing of plutonium surplus materials

International Nuclear Information System (INIS)

Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.; Forsberg, C.W.; Maliyekkel, A.T.

1996-01-01

To address growing concerns in the areas of arms control, control of fissile materials, waste management, and environment and health, the US Department of Energy is studying and evaluating various options for the control and disposal of surplus fissile materials (SFMs). One of the options under consideration is the Glass Material Oxidation and Dissolution System (GMODS) which directly converts plutonium-bearing materials such as metals, ceramics, and organics into a durable-high-quality glass for long-term storage or a waste form for disposal. This study undertook the development of a computer simulation of the GMODS process using FLOW. That computer simulation was used to perform an assessment of how GMODS would handle the treatment of plutonium, rich scrap (RS) and lead scrap (LS), and identify critical process parameters. Among the key process parameters affecting the glass formation were processing temperatures, additives, and the effects of varying them on the final product. This assessment looked at the quantity of glass produced, the quality of the final glass form, and the effect of blending different groups of the feed streams on the glass produced. The model also provided a way to study the current process assumptions and determine in which areas more experimental studies are required. The simulation showed that the glass chemistry postulated in the models is workable. It is expected that the glass chemistry assumed during the modeling process can be verified by the results of the laboratory experiments that are currently being conducted relating to the GMODS process.Further waste characterization, especially of the SFM waste streams not studied in this report, will provide more nearly accurate results and give a more detailed evaluation of the GMODS process

Preliminary process simulation and analysis of GMODS: Processing of plutonium surplus materials

Energy Technology Data Exchange (ETDEWEB)

Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.; Forsberg, C.W. [Oak Ridge National Lab., TN (United States); Maliyekkel, A.T. [Oak Ridge Associated Universities, TN (United States)

1996-01-02

To address growing concerns in the areas of arms control, control of fissile materials, waste management, and environment and health, the US Department of Energy is studying and evaluating various options for the control and disposal of surplus fissile materials (SFMs). One of the options under consideration is the Glass Material Oxidation and Dissolution System (GMODS) which directly converts plutonium-bearing materials such as metals, ceramics, and organics into a durable-high-quality glass for long-term storage or a waste form for disposal. This study undertook the development of a computer simulation of the GMODS process using FLOW. That computer simulation was used to perform an assessment of how GMODS would handle the treatment of plutonium, rich scrap (RS) and lead scrap (LS), and identify critical process parameters. Among the key process parameters affecting the glass formation were processing temperatures, additives, and the effects of varying them on the final product. This assessment looked at the quantity of glass produced, the quality of the final glass form, and the effect of blending different groups of the feed streams on the glass produced. The model also provided a way to study the current process assumptions and determine in which areas more experimental studies are required. The simulation showed that the glass chemistry postulated in the models is workable. It is expected that the glass chemistry assumed during the modeling process can be verified by the results of the laboratory experiments that are currently being conducted relating to the GMODS process.Further waste characterization, especially of the SFM waste streams not studied in this report, will provide more nearly accurate results and give a more detailed evaluation of the GMODS process.

Design-Only Conceptual Design Report: Plutonium Immobilization Plant

International Nuclear Information System (INIS)

DiSabatino, A.; Loftus, D.

1999-01-01

This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

Weapons-grade plutonium dispositioning. Volume 4

International Nuclear Information System (INIS)

Sterbentz, J.W.; Olsen, C.S.; Sinha, U.P.

1993-06-01

This study is in response to a request by the Reactor Panel Subcommittee of the National Academy of Sciences (NAS) Committee on International Security and Arms Control (CISAC) to evaluate the feasibility of using plutonium fuels (without uranium) for disposal in existing conventional or advanced light water reactor (LWR) designs and in low temperature/pressure LWR designs that might be developed for plutonium disposal. Three plutonium-based fuel forms (oxides, aluminum metallics, and carbides) are evaluated for neutronic performance, fabrication technology, and material and compatibility issues. For the carbides, only the fabrication technologies are addressed. Viable plutonium oxide fuels for conventional or advanced LWRs include plutonium-zirconium-calcium oxide (PuO 2 -ZrO 2 -CaO) with the addition of thorium oxide (ThO 2 ) or a burnable poison such as erbium oxide (Er 2 O 3 ) or europium oxide (Eu 2 O 3 ) to achieve acceptable neutronic performance. Thorium will breed fissile uranium that may be unacceptable from a proliferation standpoint. Fabrication of uranium and mixed uranium-plutonium oxide fuels is well established; however, fabrication of plutonium-based oxide fuels will require further development. Viable aluminum-plutonium metallic fuels for a low temperature/pressure LWR include plutonium aluminide in an aluminum matrix (PuAl 4 -Al) with the addition of a burnable poison such as erbium (Er) or europium (Eu). Fabrication of low-enriched plutonium in aluminum-plutonium metallic fuel rods was initially established 30 years ago and will require development to recapture and adapt the technology to meet current environmental and safety regulations. Fabrication of high-enriched uranium plate fuel by the picture-frame process is a well established process, but the use of plutonium would require the process to be upgraded in the United States to conform with current regulations and minimize the waste streams

Solubility of plutonium and waste evaporation

International Nuclear Information System (INIS)

Karraker, D.G.

1993-01-01

Chemical processing of irradiated reactor elements at the Savannah River Site separates uranium, plutonium and fission products; fission products and process-added chemicals are mixed with an excess of NaOH and discharged as a basic slurry into large underground tanks for temporary storage. The slurry is composed of base-insoluble solids that settle to the bottom of the tank; the liquid supemate contains a mixture of base-soluble chemicals--nitrates, nitrites aluminate, sulfate, etc. To conserve space in the waste tanks, the supemate is concentrated by evaporation. As the evaporation proceeds, the solubilities of some components are exceeded, and these species crystallize from solution. Normally, these components are soluble in the hot solution discharged from the waste tank evaporator and do not crystallize until the solution cools. However, concern was aroused at West Valley over the possibility that plutonium would precipitate and accumulate in the evaporator, conceivably to the point that a nuclear accident was possible. There is also a concern at SRS from evaporation of sludge washes, which arise from washing the base-insoluble solids (open-quote sludge close-quote) with ca. 1M NaOH to reduce the Al and S0 4 -2 content. The sludge washes of necessity extract a low level of Pu from the sludge and are evaporated to reduce their volume, presenting the possibility of precipitating Pu. Measurements of the solubility of Pu in synthetic solutions of similar composition to waste supernate and sludge washes are described in this report

Determination of uranium and plutonium in metal conversion products from electrolytic reduction process

International Nuclear Information System (INIS)

Lee, Chang Heon; Suh, Moo Yul; Joe, Kih Soo; Sohn, Se Chul; Jee, Kwang Young; Kim, Won Ho

2005-01-01

Chemical characterization of process materials is required for the optimization of an electrolytic reduction process in which uranium dioxide, a matrix of spent PWR fuels, is electrolytically reduced to uranium metal in a medium of LiCl-Li 2 O molten at 650 .deg. C. A study on the determination of fissile materials in the uranium metal products containing corrosion products, fission products and residual process materials has been performed by controlled-potential coulometric titration which is well known in the field of nuclear science and technology. Interference of Fe, Ni, Cr and Mg (corrosion products), Nd (fission product) and LiCl molten salt (residual process material) on the determination of uranium and plutonium, and the necessity of plutonium separation prior to the titration are discussed in detail. Under the analytical condition established already, their recovery yields are evaluated along with analytical reliability

Volume reduction and plutonium recovery in alpha wastes by cryogenic crushing and lixiviation

International Nuclear Information System (INIS)

Arnal, T.; Pajot, J.

1986-06-01

The industry of plutonium generates solid alpha wastes of medium activity called ''technological wastes''. They are mainly produced during the fabrication and reprocessing of nuclear reactor fuels and they are of a wide variety i.e: vinyl bags, gloves, glass, steel materials used in glove box operation, etc... These wastes contain relevant residual quantities of uranium and plutonium in the form of oxides or nitrates, reaching up to several dozen grams per cubic meter. Up to the beginning of the eighties, they were conditionned without any treatment and stored as such on the production site. However, for an economic and safe storage, recovering of the plutonium contained in these waste streams and reduction of their volume is of obvious importance. At the plutonium ''Complexe de Fabrication des Combustibles de Cadarache'' was developed a new technical solution of this problem that combines cryogenic crushing of the solid waste and plutonium recovery from the crushed material by chemical lixiviation. The first results obtained in applying this system on the industrial scale are reported briefly

Plutonium separation by reduction stripping. Application to processing of mixed oxide (U,Pu)O2 fuel fabrication wastes

International Nuclear Information System (INIS)

Arnal, Thierry; Cousinou, Gerard; Ganivet, Michel.

1978-11-01

A procedure is described for separating plutonium from a uranium VI and plutonium IV mixture contained in an organic phase (tributyl phosphate diluted in dodecane). This separation is obtained by extracting the plutonium III using two organic reducers: hydrazine and paraminophenol. Paraminophenol has excellent reducing qualities, similar to those of ferrous sulphamate, but has the added advantage of not contaminating extracted plutonium. This procedure is currently used in processing production wastes from mixed oxide (U,Pu)O 2 fuels; the installation using this procedure is described in detail in this paper. Operating results show the remarkable efficiency of this procedure: the separated plutonium and uranium mass flows have been increased to 185 and 350 g.h -1 respectively; the uranium contains less than 0.1 ppm of plutonium on completion of the purification cycle [fr

METHOD OF PRODUCING PLUTONIUM TETRAFLUORIDE

Science.gov (United States)

Tolley, W.B.; Smith, R.C.

1959-12-15

A process is presented for preparing plutonium tetrafluoride from plutonium(IV) oxalate. The oxalate is dried and decomposed at about 300 deg C to the dioxide, mixed with ammonium bifluoride, and the mixture is heated to between 50 and 150 deg C whereby ammonium plutonium fluoride is formed. The ammonium plutonium fluoride is then heated to about 300 deg C for volatilization of ammonium fluoride. Both heating steps are preferably carried out in an inert atmosphere.

Influence of chemical form, feeding regimen, and animal species on the gastrointestinal absorption of plutonium

International Nuclear Information System (INIS)

Bhattacharyya, M.H.; Larsen, R.P.; Cohen, N.; Ralston, L.G.; Oldham, R.D.; Moretti, E.S.; Ayres, L.

1985-01-01

We evaluated the effect of chemical form and feeding regimen on the gastrointestinal (GI) absorption of plutonium in adult mice at plutonium concentrations relevant to the establishment of drinking water standards. Mean fractional GI absorption values in fasted adult mice were: Pu(VI) bicarbonate, 15 x 10 -4 ; Pu(IV) bicarbonate, 20 x 10 -4 ; Pu(IV) nitrate (pH2), 17 x 10 -4 ; Pu(IV) citrate, 24 x 10 -4 ; and Pu(IV) polymer, 3 x 10 -4 . Values in fed adult mice were: Pu(VI) bicarbonate, 1.4 x 10 -4 ; Pu(IV) polymer, 0.3 x 10 -4 . Pu(VI) is the oxidation state in chlorinated drinking waters and Pu(IV) is the oxidation state in many untreated natural waters. To assess the validity of extrapolating data from mice to humans, we also determined the GI absorption of Pu(VI) bicarbonate in adult baboons with a dual-isotope method that does not require animal sacrifice. Fractional GI absorption values obtained by this method were 23 +- 10 x 10 -4 for fasted baboons (n=5) and 1.4 +- 0.9 x 10 -4 for fed baboons (n=3). We have so far validated this method in one baboon and are currently completing validation in two additional animals. At low plutonium concentrations, plutonium oxidation state [Pu(VI) vs Pu(IV)] and administration medium (bicarbonate vs nitrate vs citrate) had little effect on the GI absorption of plutonium in mice. Formation of Pu(IV) polymers and animal feeding decreased the GI absorption of plutonium 5- to 10-fold. The GI absorption of Pu(VI) bicarbonate in both fed and fasted adult baboons appeared to be the same as in fed and fasted adult mice, respectively. 17 refs., 2 tabs

Plutonium roundtable discussion

International Nuclear Information System (INIS)

Penneman, R.A.

1982-01-01

The roundtable discussion began with remarks by the chairman who pointed out the complicated nature of plutonium chemistry. Judging from the papers presented at this symposium, he noticed a pattern which indicated to him the result of diminished funding for investigation of basic plutonium chemistry and funding focused on certain problem areas. Dr. G.L. silver pointed to plutonium chemists' erroneous use of a simplified summary equation involving the disproportionation of Pu(EV) and their each of appreciation of alpha coefficients. To his appreciation of alpha coefficients. To his charges, Dr. J.T. Bell spoke in defense of the chemists. This discussion was followed by W.W. Schulz's comments on the need for experimental work to determine solubility data for plutonium in its various oxidation states under geologic repository conditions. Discussion then turned to plutonium pyrachemical process with Dana C. Christensen as the main speaker. This paper presents edited versions of participants' written version

Analytical methods used in plutonium purification cycles by trilaurylamine

International Nuclear Information System (INIS)

Perez, J.J.

1965-01-01

The utilisation of trilaurylamine as a solvent extractant for the purification of plutonium has entailed to perfect a set of analytical methods which involves, various techniques. The organic impurities of the solvent can be titrated by gas-liquid chromatography. The titration of the main degradation product, the di-laurylamine, can be accomplished also by spectro-colorimetry. Potentiometry is used for the analysis of the different salts of amine-nitrate-sulfate-bisulfate as also the extracted nitric acid. The determination of the nitrate in aqueous phase is carried out by constant current potentiometry. The range of application, the accuracy and the procedure of these analysis are related in the present report. (author) [fr

Experiences of shipper-receiver differences in plutonium oxide transactions

International Nuclear Information System (INIS)

Swinburn, K.A.; McGowan, I.R.

1976-01-01

Plutonium oxide has the property of absorbing moisture, carbon dioxide etc., to the extent that it is necessary to take special steps to determine the plutonium content in such a way as to avoid shipper-receiver differences. This paper describes a method used internationally for several years for correcting for any absorption. The consignee witnesses the weighing of the bulk material (D kg) at the time of blending and sampling at the consignor's Works, and also the weighing of the sample (C g). Immediately before analysis the sample is reweighed (B g) and assayed for plutonium (A%). The weight of plutonium is then (A/100)x(B/C)xD kg. The accepted weight is the average between consignor's and consignee's figures after both have used the formula indicated. It goes on to give actual assay and fissile content figures obtained by BNFL and its consignees for the international transactions already referred to, and concludes that the method of approach results cumulatively in a negligible shipper-receiver difference. Individual transactions are shown to be commercially acceptable and, it is considered, also acceptable for safeguards purposes. It is suggested that for safeguarded material the safeguards inspectors might also be able to be present at, and themselves witness, the blending, sampling and weighing operations, and at the same time take their own samples for analysis. It is further suggested that the principle could be extended to accoutancy for plutonium in plutonium nitrate solutions provided leakage does not occur from the sample vials. (author)

The first milligrams of plutonium

International Nuclear Information System (INIS)

Goldschmidt, B.

1996-01-01

This paper relates the discovery of the different plutonium chemical extraction processes in their historical context. The first experiments started during the second world war in 1942 with the American ''Metallurgical Laboratory'' project which brought together Arthur Compton, Enrico Fermi and Glenn Seaborg. During the same period, a competitive English-Canadian project, the ''Montreal Project'', was carried out to test different plutonium solvent extraction techniques. The author participated in both projects and joined the CEA in 1946, where he was in charge of the uranium and plutonium chemistry. By the end of 1949, his team could isolate the first milligrams of French plutonium from uranium oxide pellets of the ZOE reactor. In the beginning of 1952 he developed with his team the PUREX process. (J.S.)

Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

International Nuclear Information System (INIS)

LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

1996-01-01

The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-2 1). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

International Nuclear Information System (INIS)

LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

1996-01-01

The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-21). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

Optimum Condition for Plutonium Electrodeposition Process in Radiochemistry and Environment Laboratory, Nuclear Malaysia

International Nuclear Information System (INIS)

Yii, Mei-Wo; Abdullah Siddiqi Ismail

2014-01-01

Determination of alpha emitting plutonium radionuclides such as Pu-238, Pu-239 and Pu-240 concentrations inside a sample require lots of radiochemistry purification process to separate them from other interfering alpha emitters. These pure isotopes are then been electrodeposited onto a stainless steel disc and quantified using alpha spectrometry counter. In Radiochemistry and Environment Laboratory (RAS), Nuclear Malaysia, the quantification is done by comparing these isotopes with the recovery of known amount plutonium tracer, Pu-242, that been added into the sample prior analysis. This study been conducted to find the optimum conditions for the electrolysis process used at RAS. Four variable parameters that may interfere the percentage recovery of tracer hence the current, cathode to anode distance, pH and electrolysis duration had been identify and studied. Study was carry out using Pu-242 standard solution and the deposition disc was counted using Zinc Sulphite (silver) counter. Studies outcome suggested that the optimum conditions to reduce plutonium ion happens at 1-1.1 ampere of current, 3-5 mm of electrodes distance, pH 2.2-2.5 and a minimal electrolysis duration of 2 hours. (author)

The establishment of in-process plutonium mass equation in Rokkasho Reprocessing Plant

International Nuclear Information System (INIS)

Yamaya, Kosuke; Ebata, Takashi; Yamazaki, Yoshihiro; Kawai, Akio; Iwamoto, Tomonori

2008-01-01

At Rokkasho Reprocessing Plant (RRP), Active Test (AT) using actual spent fuels for the final confirmation of the equipment and the system has been performed toward the commercial operation. From the safeguards viewpoint, performance of material accountancy equipment is confirmed and data for evaluating parameters of the inspection equipment is obtained by making use of the AT period. RRP is applied to Near Real Time material Accountancy (NRTA). Under the NRTA scheme, the inventory at a cut-off time during process operation needs to be accounted for. There are some un-measurable inventories of plutonium in the process, which will be calculated from inventory estimation equations. The amount of these plutonium inventories calculated from the equations is so large that it is essential to improve the inventory estimation equations to be quite accurate. Therefore, correctness of the inventory estimation equations is evaluated by using process operation data obtained during AT. This paper describes the results of evaluating the inventory estimation equations by using the process operation data and the NRTA procedure under continuous operating condition as well. (author)

Materials measurement and accounting in an operating plutonium conversion and purification process. Phase I. Process modeling and simulation

International Nuclear Information System (INIS)

Thomas, C.C. Jr.; Ostenak, C.A.; Gutmacher, R.G.; Dayem, H.A.; Kern, E.A.

1981-04-01

A model of an operating conversion and purification process for the production of reactor-grade plutonium dioxide was developed as the first component in the design and evaluation of a nuclear materials measurement and accountability system. The model accurately simulates process operation and can be used to identify process problems and to predict the effect of process modifications

Adsorption and diffusion of plutonium in soil

International Nuclear Information System (INIS)

Brown, D.A.

The behavior of plutonium (Pu) was studied in three soils that varied in texture, CEC, pH, organic matter content and mineralogy (Fuquay, Muscatine, Burbank). Two isotopes, 238 Pu and 239 Pu, were used in order to detect Pu over a range of several orders of magnitude. Unless added in a chelated form, Pu was added to the soil as a nitrate in .01 N HNO 3 to simulate the release of acidic waste on the soil and to prevent rapid Pu hydrolysis or polymerization

The nitrate to ammonia and ceramic (NAC) process: A newly developed low-temperature technology

International Nuclear Information System (INIS)

Mattus, A.J.; Lee, D.D.

1993-01-01

Bench-top feasibility studies with Hanford single-shell tank (SST) simulants, using a new low-temperature (50-60 degrees C) process for converting nitrate to ammonia and ceramic, have conclusively shown that between 90 and 99% of the nitrate at Hanford can be readily converted to ammonia. In this process, aluminum powders or shot can be used to convert alkaline, nitrate-based supernate to ammonia and an alumina-silica-based ceramic solid. The process may actually be able to utilize already contaminated aluminum scrap metal from various US DOE sites to effect the conversion. The final nitrate-free ceramic product can be calcined, pressed, and sintered like any other ceramic. Based upon the starting volumes of 6.2 and 3.1 M sodium nitrate solution (probable supernate concentrations resulting from salt-cake/sludge removal from the Hanford SSTs), volume reductions as high as 70% are currently obtained, compared with an expected 40 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. These data were used to cost a batch facility with a production rate of 1200 kilograms of nitrate per hour for processing all the Hanford SST waste over 20 years. Our process cost analysis indicates that between $2.01 and 2.66 will be required to convert each kilogram of nitrate. Based upon 1957 literature, these costs are one-third to one-half of the processing costs quoted for electrolytic and thermal processes

Plutonium Disposition Now exclamation point

International Nuclear Information System (INIS)

Buckner, M.R.

1995-01-01

A means for use of existing processing facilities and reactors for plutonium disposition is described which requires a minimum capital investment and allows rapid implementation. The scenario includes interim storage and processing under IAEA control, and fabrication into MOX fuel in existing or planned facilities in Europe for use in operating reactors in the two home countries. Conceptual studies indicate that existing Westinghouse four-loop designs can safety dispose of 0.94 MT of plutonium per calendar year. Thus, it would be possible to consume the expected US excess stockpile of about 50 MT in two to three units of this type, and it is highly likely that a comparable amount of the FSU excess plutonium could be deposed of in a few VVER-1000's. The only major capital project for this mode of plutonium disposition would be the weapons-grade plutonium processing which could be done in a dedicated international facility or using existing facilities in the US and FSU under IAEA control. This option offers the potential for quick implementation at a very low cost to the governments of the two countries

Plutonium Immobilization Can Loading Conceptual Design

Energy Technology Data Exchange (ETDEWEB)

Kriikku, E.

1999-05-13

'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

Plutonium Immobilization Can Loading Conceptual Design

International Nuclear Information System (INIS)

Kriikku, E.

1999-01-01

'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

Electrochemical processing of nitrate waste solutions

Energy Technology Data Exchange (ETDEWEB)

Genders, D.; Weinberg, N.; Hartsough, D. (Electrosynthesis Co., Inc., Cheektowaga, NY (United States))

1992-10-07

The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F[sup [minus

Progress on plutonium stabilization

Energy Technology Data Exchange (ETDEWEB)

Hurt, D. [Defense Nuclear Facilities Safety Board, Washington, DC (United States)

1996-05-01

The Defense Nuclear Facilities Safety Board has safety oversight responsibility for most of the facilities where unstable forms of plutonium are being processed and packaged for interim storage. The Board has issued recommendations on plutonium stabilization and has has a considerable influence on DOE`s stabilization schedules and priorities. The Board has not made any recommendations on long-term plutonium disposition, although it may get more involved in the future if DOE develops plans to use defense nuclear facilities for disposition activities.

Progress on plutonium stabilization

International Nuclear Information System (INIS)

Hurt, D.

1996-01-01

The Defense Nuclear Facilities Safety Board has safety oversight responsibility for most of the facilities where unstable forms of plutonium are being processed and packaged for interim storage. The Board has issued recommendations on plutonium stabilization and has has a considerable influence on DOE's stabilization schedules and priorities. The Board has not made any recommendations on long-term plutonium disposition, although it may get more involved in the future if DOE develops plans to use defense nuclear facilities for disposition activities

Electrochemical processing of nitrate waste solutions. Phase 2, Final report

Energy Technology Data Exchange (ETDEWEB)

Genders, D.; Weinberg, N.; Hartsough, D. [Electrosynthesis Co., Inc., Cheektowaga, NY (US)

1992-10-07

The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F{sup {minus}} ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions.

Monitoring nitrite and nitrate residues in frankfurters during processing and storage.

Science.gov (United States)

Pérez-Rodríguez, M L; Bosch-Bosch, N; Garciá-Mata, M

1996-09-01

Frankfurter-type sausages were prepared in a pilot plant with different concentrations of NaNO(2) (75, 125 or 250 ppm) combined or not with 200 ppm KNO(3). A meat system, free of curing agents, was also used as control. Nitrite and nitrate levels were tested in various processing steps and over 120 days storage at 3 °C of the vacuum-packaged frankfurters. Little influence of the originally added nitrite level on the amount of nitrate formed was observed. Important losses of nitrite and nitrate were due to cooking. Thereafter about 50% of the nitrite added initially remained in this form in all samples (39, 59 and 146 ppm, respectively) and between 10 and 15% as nitrate. When only nitrate was initially added, formation of nitrite after cooking was observed (maximum level 43 ppm NaNO(2)). Formulations prepared with both nitrate and nitrite showed no significant differences (p nitrite or nitrate counterparts. A good correlation among nitrite and nitrate levels and storage time was showed by multiple linear regression analysis. It is concluded that the use of nitrate in combination with nitrite in cooked meat products seems to have little technological significance and adds to the total body burden of nitrite.

Decorporation approach following rat lung contamination with a moderately soluble compound of plutonium using local and systemic Ca-DTPA combined chelation

International Nuclear Information System (INIS)

Gremy, Olivier; Bruel, Sylvie; Renault, Daniel; Van der Meeren, Anne; Tsapis, Nicolas

2012-01-01

Decorporation efficacy of prompt pulmonary delivery of DTPA dry powder was assessed following lung contamination with plutonium nitrate and compared to an intravenous injection of DTPA solution and a combined administration of both DTPA compounds. In addition, efficacy of a delayed treatment was assessed. In case of either early or late administration, insufflated DTPA was more efficient than intravenously injected DTPA in reducing the plutonium lung burden due to its high local concentration. Prompt treatment with DTPA powder was also more effective in limiting extrapulmonary deposits by removing the early transportable fraction of plutonium from lungs prior its absorption into blood. Translocation of DTPA from lungs to blood may also contribute to the decrease in extrapulmonary retention, as shown by reduced liver deposit after delayed pulmonary administration of DTPA. Efficacy of DTPA dry powder was further increased by the combined intravenous administration of DTPA solution for reducing extrapulmonary deposits of plutonium and promoting its urinary excretion. According to our results, the most effective treatment protocol for plutonium decorporation was the early pulmonary delivery of DTPA powder supplemented by an intravenous injection of DTPA solution. Following inhalation of plutonium as nitrate chemical form, this combined chelation therapy should provide a more effective method of treatment than conventional intravenous injection alone. At later stages following lung contamination, pulmonary administration of DTPA should also be considered as the treatment of choice for decreasing the lung burden. (authors)

Modeling of Diffusion of Plutonium in Other Metals and of Gaseous Species in Plutonium-Based Systems

International Nuclear Information System (INIS)

Cooper, Bernard R.; Gayanath W. Fernando; Beiden, S.; Setty, A.; Sevilla, E.H.

2004-01-01

Establish standards for temperature conditions under which plutonium, uranium, or neptunium from nuclear wastes permeates steel, with which it is in contact, by diffusion processes. The primary focus is on plutonium because of the greater difficulties created by the peculiarities of face-centered-cubic-stabilized (delta) plutonium (the form used in the technology generating the waste)

Eutrophication decrease: Phosphate adsorption processes in presence of nitrates.

Science.gov (United States)

Boeykens, Susana P; Piol, M Natalia; Samudio Legal, Lisa; Saralegui, Andrea B; Vázquez, Cristina

2017-12-01

Eutrophication causes aquatic environment degradation as well as serious problems for different purposes of water uses. Phosphorus and nitrogen, mainly as phosphate and nitrate respectively, are considered responsible for eutrophication degradation. The focus of this work was the study of adsorption processes for decreasing phosphate and nitrate concentrations in bi-component aqueous systems. Dolomite and hydroxyapatite were selected as low-cost adsorbents. Obtained results showed that both adsorbents have high capacity for phosphate adsorption which the presence of nitrate does not modify. Hydroxyapatite proved to be the most efficient adsorbent, however, it showed a low percentage of desorption and few possibilities of reuse. Dolomite, on the other hand, allows a desorption of the adsorbed material that favours its reuse. Copyright © 2017 Elsevier Ltd. All rights reserved.

Soil nitrate reducing processes - drivers, mechanisms for spatial variation, and significance for nitrous oxide production.

Science.gov (United States)

Giles, Madeline; Morley, Nicholas; Baggs, Elizabeth M; Daniell, Tim J

2012-01-01

The microbial processes of denitrification and dissimilatory nitrate reduction to ammonium (DNRA) are two important nitrate reducing mechanisms in soil, which are responsible for the loss of nitrate ([Formula: see text]) and production of the potent greenhouse gas, nitrous oxide (N(2)O). A number of factors are known to control these processes, including O(2) concentrations and moisture content, N, C, pH, and the size and community structure of nitrate reducing organisms responsible for the processes. There is an increasing understanding associated with many of these controls on flux through the nitrogen cycle in soil systems. However, there remains uncertainty about how the nitrate reducing communities are linked to environmental variables and the flux of products from these processes. The high spatial variability of environmental controls and microbial communities across small sub centimeter areas of soil may prove to be critical in determining why an understanding of the links between biotic and abiotic controls has proved elusive. This spatial effect is often overlooked as a driver of nitrate reducing processes. An increased knowledge of the effects of spatial heterogeneity in soil on nitrate reduction processes will be fundamental in understanding the drivers, location, and potential for N(2)O production from soils.

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

Review of major plutonium pyrochemical technology

International Nuclear Information System (INIS)

Moser, W.S.; Navratil, J.D.

1983-01-01

The past twenty years have seen significant growth in the development and application of pyrochemical technology for processing of plutonium. For particular feedstocks and specific applications, non-aqueous high-temperature processes offer key advantages over conventional hydrometallurgical systems. Major processes in use today include: (1) direct oxide reduction for conversion of PuO 2 to metal, (2) molten salt extraction for americium removal from plutonium, (3) molten salt electrorefining for Pu purification, and (4) hydriding to remove plutonium from host substrates. This paper reviews current major pyrochemical processes from the classical calcination-hydrofluorination-bomb reduction sequence through new techniques under development. Each process is presented and brief descriptions of production equipment are given. 47 references, 5 figures

Plutonium decontamination studies using Reverse Osmosis

International Nuclear Information System (INIS)

Plock, C.E.; Travis, T.N.

1980-01-01

Water in batches of 45 gallons each, from a creek crossing the Rocky Flats Plant, was transferred to the Reverse Osmosis (RO) laboratory for experimental testing. The testing involved using RO for plutonium decontamination. For each test, the water was spiked with plutonium, had its pH adjusted, and was then processed by RO. At a water recovery level of 87%, the plutonium decontamination factors ranged from near 100 to 1200, depending on the pH of the processed water

Uranium-plutonium fuel for fast reactors

International Nuclear Information System (INIS)

Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

1996-01-01

Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

Plutonium recovery from carbonate wash solutions

International Nuclear Information System (INIS)

Gray, J.H.; Reif, D.J.; Chostner, D.F.; Holcomb, H.P.

1991-01-01

540Periodically higher than expected levels of plutonium are found in carbonate solutions used to wash second plutonium cycle solvent. The recent accumulation of plutonium in carbonate wash solutions has led to studies to determine the cause of that plutonium accumulation, to evaluate the quality of all canyon solvents, and to develop additional criteria needed to establish when solvent quality is acceptable. Solvent from three canyon solvent extraction cycles was used to evaluate technology required to measure tributyl phosphate (TBP) degradation products and was used to evaluate solvent quality criteria during the development of plutonium recovery processes. 1 fig

Characterization of the deviation of the ideality of concentrated electrolytic solutions: plutonium 4 and uranium 4 nitrate salts study; Contribution a la caracterisation de l'ecart a l'idealite des solutions concentrees d'electrolytes: application aux cas de nitrates de plutonium (4) et d'uranium (4)

Energy Technology Data Exchange (ETDEWEB)

Charrin, N

2000-07-01

The purpose of this work was to establish a new binary data base by compiling the activity coefficients of plutonium and uranium at oxidation state +IV to better account for media effects in the liquid-liquid extraction operations implemented to reprocess spent nuclear fuel. Chapter 1: first reviews the basic thermodynamic concepts before describing the issues involved in acquiring binary data for the tetravalent actinides. The difficulties arise from two characteristics of this type of electrolyte: its radioactive properties (high specific activity requiring nuclearization of the experimental instrumentation) and its physicochemical properties (strong hydrolysis). After defining the notion of fictive binary data, an approach based on the thermodynamic concept of simple solutions is described in which the activity coefficient of an aqueous phase constituent is dependent on two parameters: the water activity of the system and the total concentration of dissolved constituents. The method of acquiring fictive binary electrolyte data is based on water activity measurements for ternary or quaternary actinide mixtures in nitric acid media, and binary data for nitric acid. The experimental value is then correlated with the characteristics of the fictive binary solution of the relevant electrolyte. Chapter 2: proposes more reliable binary data for nitric acid than the published equivalents, the disparities of which are discussed. The validation of the method described in Chapter 1 for acquiring fictive binary data is then addressed. The test electrolyte, for which binary data are available in the literature, is thorium(IV) nitrate. The method is validated by comparing the published binary data obtained experimentally for binary solutions with the data determined for the ternary Th(NO{sub 3}){sub 4}/HNO{sub 3}/H{sub 2}O system investigated in this study. The very encouraging results of this comparison led us to undertake further research in this area. Chapter 3 discusses

Plutonium spectrophotometric analysis

International Nuclear Information System (INIS)

Esteban, A.; Cassaniti, P.; Orosco, E.H.

1990-01-01

Plutonium ions in solution have absorption spectra so different that it is possible to use them for analytical purposes. Detailed studies have been performed in nitric solutions. Some very convenient methods for the determination of plutonium and its oxidation states, especially the ratios Pu(III):Pu(IV) and Pu(IV):Pu(VI) in a mixture of both, have been developed. These methods are described in this paper, including: a) Absorption spectra for plutonium (III), (IV), (VI) and mixtures. b) Relative extinction coefficients for the above mentioned species. c) Dependences of the relative extinction coefficients on the nitric acid concentration and the plutonium VI deviation from the Beer-Lambert law. The developed methods are simple and rapid and then, suitable in process control. Accuracy is improved when relative absorbance measurements are performed or controlled the variables which have effect on the spectra and extinction coefficients. (Author) [es

Plutonium estimation in the process solutions and oxide dissolved audit samplers by potentiometry using memo titrator

International Nuclear Information System (INIS)

Kumaraguru, K.; Shukla, Y.D.; Vijayan, K.; Ramamoorthy, N.; Jambunathan, U.; Kapoor, S.C.

1990-01-01

Potentiometric method is employed by using memotitrator coupled with combined electrode for the estimation of plutonium. The estimations are carried out on the process samples and the acid dissolved samples for auditing, in the concentration range of 5 g/l to 20 g/l. The chemical procedure is: i)oxidising plutonium to higher oxidation state by silver oxide, ii)reducing the same by adding excess ferrous, and iii)titrating potassium dichromate against the unreacted ferrous. The plutonium content is computed from ferrous consumed in the reaction. The average percentage error of the method is +/-0.27. The values obtained are in close agreement with those obtained by coulometry. (author)

Investigation of plutonium (4) hydroxoformates

International Nuclear Information System (INIS)

Andryushin, V.G.; Belov, V.A.; Galaktionov, S.V.; Kozhevnikov, P.B.; Matyukha, V.A.; Shmidt, V.S.

1982-01-01

Deposition processes of plutonium (4) hydroxoformates in the system Pu(NO 3 ) 4 -HNO 3 -HCoOH-N6 4 OH-H 2 O have been studied in pH range 0.2-10.7 at total plutonium concentration in the system 100 g/l. It is shown that under the conditions plutonium (4) hydrolysis takes place with the formation of hydroxoformates. A local maximum of plutonium (4) hydroxoformate solubility in the range pH=3.8-4.8, which is evidently conditioned by the formation of soluble formate complex of plutonium in the region, is pointed out. The basic plutonium (4) formates of the composition PuOsub(x)(OH)sub(y)(COOH)sub(4-2x-y)xnHsub(2)O, where 1,3 >=x >= 0.7, 1.7 >= y >= 1.0 and n=1.5-7.0, are singled out, their thermal stability being studied. Density of the crystals and plutonium dioxide, formed during their thermal decomposition, is measured. It is established that for plutonium (4) hydroxoformates common regularities of the influence of salt composition (OH - -, CHOO - - and H 2 O-group numbers in the mulecule) on position of temperature decomposition effects and on the density of compounds, which have been previously found during the study of thorium and plutonium hydroxosalts are observed. It is shown that the density of plutonium dioxide decreases with the increase of hydration and hydrolysis degree of the initial plutonium hydroxoformate

Hydroxylamine Nitrate Decomposition under Non-radiological Conditions

Energy Technology Data Exchange (ETDEWEB)

McFarlane, Joanna [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delmau, Laetitia Helene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DePaoli, David W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mattus, Catherine H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Phelps, Clarice E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Roach, Benjamin D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-07-01

Hydroxylamine nitrate (HAN) is used to reduce Pu(IV) to Pu(III) in the separation of plutonium from uranium. HAN becomes unstable under certain conditions and has been known to explode, causing injury to humans including death. Hence, it is necessary to deactivate HAN once the reduction of plutonium is finished. This report reviews what is known about the chemistry of HAN and various methods to achieve a safe decomposition. However, there are areas where more information is needed to make a decision about the handling of HAN in reprocessing of nuclear fuel. Experiments have demonstrated a number of non-radiolytic ways to safely decompose HAN, including heating in HNO3, photolytic oxidation in the presence of H2O2, and the addition of a metal such as Fe(III) that will oxidize the HAN.

The development and testing of the new flowsheets for the plutonium purification of the Purex process

Energy Technology Data Exchange (ETDEWEB)

Bugrov, K.V.; Korotaev, V.G.; Korchenkin, K.K.; Logunov, M.V.; Ludin, S.A.; Mashkin, A.N.; Melentev, A.B.; Samarina, N.S. [FSUE ' PAMayak' , Lenin st., 35, Ozersk 456780 (Russian Federation)

2016-07-01

In order to improve the extraction flowsheet of RT-1 Plant two versions of plutonium purification unit flowsheet were developed: a flowsheet with stabilization of Pu(IV)-Np(IV) valence pair and Pu, Np co-recovery, and a flowsheet with stabilization of Pu(IV)-Np(V) valence pair and Pu recovery. The task related to stabilization of the valence pair of the target components in the required state was solved with the use of reactants already applied at RT-1 Plant, namely, hydrogen peroxide, hydrazine nitrate and catalyst (Fe). Both flowsheets were adapted for the plant purification facility with minimum modifications of the equipment, and passed the full scale industrial testing. As a result of this work, reduction in volume and salt content of the raffinate was achieved. (authors)

Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

International Nuclear Information System (INIS)

Brownson, D.A.; Hanson, D.J.; Blackman, H.S.

1993-06-01

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition

Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

Energy Technology Data Exchange (ETDEWEB)

Brownson, D.A.; Hanson, D.J.; Blackman, H.S. [and others

1993-06-01

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition.

Modelling of uranium/plutonium splitting in purex process

International Nuclear Information System (INIS)

Boullis, B.; Baron, P.

1987-06-01

A mathematical model simulating the highly complex uranium/plutonium splitting operation in PUREX process has been achieved by the french ''Commissariat a l'Energie Atomique''. The development of such a model, which includes transfer and redox reactions kinetics for all the species involved, required an important experimental work in the field of basis chemical data acquisition. The model has been successfully validated by comparison of its results with those of specific trials achieved (at laboratory scale), and with the available results of the french reprocessing units operation. It has then been used for the design of french new plants splitting operations

Waste minimization and the goal of an environmentally benign plutonium processing facility: A strategic plan

International Nuclear Information System (INIS)

Pillay, K.K.S.

1994-02-01

To maintain capabilities in nuclear weapons technologies, the Department of Energy (DOE) has to maintain a plutonium processing facility that meets all the current and emerging standards of environmental regulations. A strategic goal to transform the Plutonium Processing Facility at Los Alamos into an environmentally benign operation is identified. A variety of technologies and systems necessary to meet this goal are identified. Two initiatives now in early stages of implementation are described in some detail. A highly motivated and trained work force and a systems approach to waste minimization and pollution prevention are necessary to maintain technical capabilities, to comply with regulations, and to meet the strategic goal

Plutonium in nature; Le plutonium dans la nature

Energy Technology Data Exchange (ETDEWEB)

Madic, C.

1994-12-31

Plutonium in nature comes from natural sources and anthropogenic ones. Plutonium at the earth surface comes principally from anthropogenic sources. It is easily detectable in environment. The plutonium behaviour in environment is complex. It seems necessary for the future to reduce releases in environment, to improve predictive models of plutonium behaviour in geosphere, to precise biological impact of anthropogenic plutonium releases.

Plutonium(IV) peroxide formation in nitric medium and kinetics Pu(VI) reduction by hydrogen peroxide

International Nuclear Information System (INIS)

Maillard, C.; Adnet, J.M.

2001-01-01

Reduction of plutonium (VI) to Pu(IV) with hydrogen peroxide is a step in industrial processes used to purify plutonium nitrate solutions. This operation must be carefully controlled, in order to avoid any formation of the Pu(IV) peroxide green precipitate and to obtain exclusively Pu(IV). This led us to study the acidity and Pu and H 2 O 2 concentrations influences on the precipitate appearance and to perform a Pu(VI) reduction kinetic study on a wide range of acidities ([HNO 3 ]: 0.5 to 8 M), plutonium concentrations ([Pu(VI)]: 0.1 to 0.8 M) and [H 2 O 2 ]/[Pu(VI)] ratio (from 1 to 8). Thus, the domain of Pu(IV) peroxide formation and the reactional paths were established. With the exception of 0.5 M nitric acid medium, the kinetic curves show two distinct regims: the first one corresponds to an induction period where the Pu(VI) concentration doesn't change, the second corresponds to a linear decrease of Pu(VI). An increase of the temperature greatly accelerates the Pu(VI) reduction rate while [H 2 O 2 ]/[Pu(VI)] has almost no influence. The Pu(VI) total reduction time decreases when initial concentration of plutonium increases. By increasing nitric acid concentration from 0.5 M to 6 M, the total Pu(VI) reduction time decreases. This time increases when [HNO 3 ] varies from 6 M to 8 M. (orig.)

Strategies for the plutonium utilization

International Nuclear Information System (INIS)

Zouain, D.M.; Lima, J.O.V.; Sakamoto, L.H.

1981-11-01

A review of the activities involving plutonium (its recycle, utilization and technological status and perspectives) is done. These informations are useful for an economic viability study for the plutonium utilization in thermal reactors (recycling) and in fast breeders reactor (FBR), trying to collect the major number of informations about these subjects. The initial phase describes the present status and projections of plutonium accumulation and requirements. Then, the technological process are described and some strategies are analyzed. (E.G.) [pt

Soil nitrate reducing processes drivers, mechanisms for spatial variation, and significance for nitrous oxide production

OpenAIRE

Giles, M.; Morley, N.; Baggs, E.M.; Daniell, T.J.

2012-01-01

The microbial processes of denitrification and dissimilatory nitrate reduction to ammonium\\ud (DNRA) are two important nitrate reducing mechanisms in soil, which are responsible for\\ud the loss of nitrate (NO−\\ud 3 ) and production of the potent greenhouse gas, nitrous oxide (N2O).\\ud A number of factors are known to control these processes, including O2 concentrations and\\ud moisture content, N, C, pH, and the size and community structure of nitrate reducing organisms\\ud responsible for the ...

The Mayak Worker Dosimetry System (MWDS 2013): soluble plutonium retention in the lungs of an occupationally exposed USTUR case

International Nuclear Information System (INIS)

Tolmachev, S. Y.; Avtandilashvili, M.; Martinez, F.; Thomas, E. M.; Miller, F. L.; Nielsen, C. E.; Puncher, M.; Morgan, W. F.; Birchall, A.

2017-01-01

For the first time, plutonium retention in human upper airways was investigated based on the dosimetric structure of the human respiratory tract proposed by the International Commission on Radiological Protection (ICRP). This paper describes analytical work methodology, case selection criteria, and summarizes findings on soluble (ICRP 68 Type M material) plutonium distribution in the lungs of a former nuclear worker occupationally exposed to plutonium nitrate [ 239 Pu(NO 3 ) 4 ]. Thirty-eight years post-intake, plutonium was found to be uniformly distributed between bronchial (BB), bronchiolar (bb) and alveolar-interstitial (AI) dosimetric compartments as well as between the left and right lungs. 239+240 Pu and 238 Pu total body activity was estimated to be 2333 ± 23 and 42.1 ± 0.7 Bq, respectively. The results of this work provide key information on the extent of plutonium binding in the upper airways of the human respiratory tract. (authors)

Pyrochemical recovery of plutonium fluoride reduction slag

International Nuclear Information System (INIS)

Christensen, D.C.; Rayburn, J.A.

1983-07-01

A process was developed for the pyrochemical recovery of plutonium from residues resulting from the PuF 4 reduction process. The process involves crushing the CaF 2 slag and dissolving it at 800 0 C in a CaCl 2 solvent. The plutonium, which exists either as finely divided metal or as incompletely reduced fluoride salt, is reduced to metal and/or allowed to coalesce as a massive button in the bottom of the reaction crucible. The recovery of plutonium in a 1-day cycle averaged 96%; all of the resulting residues were discardable

Direct spectrophotometric analysis of low level Pu (III) in Pu(IV) nitrate solution

International Nuclear Information System (INIS)

Mageswaran, P.; Suresh Kumar, K.; Kumar, T.; Gayen, J.K.; Shreekumar, B.; Dey, P.K.

2010-01-01

Among the various methods demonstrated for the conversion of plutonium nitrate to its oxide, the oxalate precipitation process either as Pu (III) or Pu (IV) oxalate gained wide acceptance. Since uranous nitrate is the most successful partitioning agent used in the PUREX process for the separation of Pu from the bulk amount of U, the Pu (III) oxalate precipitation of the purified nitrate solution will not give required decontamination from U. Hence Pu IV oxalate precipitation process is a better option to achieve the end user's specified PuO 2 product. Prior to the precipitation process, ensuring of the Pu (IV) oxidation state is essential. Hence monitoring of the level of Pu oxidation state either Pu (III) or Pu (IV) in the feed solution plays a significant role to establish complete conversion of Pu (III). The method in vogue to estimate Pu(lV) content is extractive radiometry using Theonyl Trifluoro Acetone (TTA). As the the method warrants a sample preparation with respect to acidity, a precise measurement of Pu (IV) without affecting the Pu(III) level in the feed sample is difficult. Present study is focused on the exploration of direct spectrophotometry using an optic fiber probe of path length of 40mm to monitor the low level of Pu(III) after removing the bulk Pu(lV) which interfere in the Pu(III) absorption spectrum, using TTA-TBP synergistic mixture without changing the sample acidity

Electrochemical studies of plutonium(IV) complexes in aqueous nitrate solutions

International Nuclear Information System (INIS)

Kim, Seong-Yun; Asakura, Toshihide; Morita, Yasuji

2005-01-01

Electrochemistry has been used to investigate the behavior of plutonium (IV) in 1-7 M HNO 3 solutions. These Pu(IV) complexes were found to be reduced quasi-reversibly to Pu(III) species. The formal redox potentials (E 0 ) for Pu(IV)/Pu(III) couples were determined to be +0.721, +0.712, +0.706, +0.705, +0.704, 0.694, and +0.696 V (vs. Ag/AgCl(SSE)) for Pu(IV) complexes in 1, 2, 3, 4, 5, 6, 7 M HNO 3 solutions, respectively. These results indicate that the reduction product of Pu(IV) is Pu(III), which is considerably stable in HNO 3 solution. (author)

Soil nitrate reducing processes – drivers, mechanisms for spatial variation, and significance for nitrous oxide production

Science.gov (United States)

Giles, Madeline; Morley, Nicholas; Baggs, Elizabeth M.; Daniell, Tim J.

2012-01-01

The microbial processes of denitrification and dissimilatory nitrate reduction to ammonium (DNRA) are two important nitrate reducing mechanisms in soil, which are responsible for the loss of nitrate (NO3−) and production of the potent greenhouse gas, nitrous oxide (N2O). A number of factors are known to control these processes, including O2 concentrations and moisture content, N, C, pH, and the size and community structure of nitrate reducing organisms responsible for the processes. There is an increasing understanding associated with many of these controls on flux through the nitrogen cycle in soil systems. However, there remains uncertainty about how the nitrate reducing communities are linked to environmental variables and the flux of products from these processes. The high spatial variability of environmental controls and microbial communities across small sub centimeter areas of soil may prove to be critical in determining why an understanding of the links between biotic and abiotic controls has proved elusive. This spatial effect is often overlooked as a driver of nitrate reducing processes. An increased knowledge of the effects of spatial heterogeneity in soil on nitrate reduction processes will be fundamental in understanding the drivers, location, and potential for N2O production from soils. PMID:23264770

Soil nitrate reducing processes – drivers, mechanisms for spatial variation and significance for nitrous oxide production

Directory of Open Access Journals (Sweden)

Madeline Eleanore Giles

2012-12-01

Full Text Available The microbial processes of denitrification and dissimilatory nitrate reduction to ammonium (DNRA are two important nitrate reducing mechanisms in soil, which are responsible for the loss of nitrate (NO3-¬ and production of the potent greenhouse gas, nitrous oxide (N2O. A number of factors are known to control these processes, including O2 concentrations and moisture content, N, C, pH and the size and community structure of nitrate reducing organisms responsible for the processes. There is an increasing understanding associated with many of these controls on flux through the nitrogen cycle in soil systems. However, there remains uncertainty about how the nitrate reducing communities are linked to environmental variables and the flux of products from these processes. The high spatial variability of environmental controls and microbial communities across small sub cm areas of soil may prove to be critical in determining why an understanding of the links between biotic and abiotic controls has proved elusive. This spatial effect is often overlooked as a driver of nitrate reducing processes. An increased knowledge of the effects of spatial heterogeneity in soil on nitrate reduction processes will be fundamental in understanding the drivers, location and potential for N2O production from soils.

Treatability study of arsenic, fluoride and nitrate from drinking water by adsorption process

International Nuclear Information System (INIS)

Abbas, N.; Irfan, M.; Butt, M.T.

2014-01-01

Natural contamination of nitrate, fluoride, arsenic and dissolved salts in ground water sources is the main health menace at present in different parts of Pakistan. The metalloids especially arsenic, fluoride and nitrate pose severe health hazards to human being. The present research work investigated the removal techniques for arsenic, fluoride and nitrate from drinking water by adsorption process. Ion exchange resins, activated carbon and activated alumina were used for removal of selected contaminants. These adsorbents were evaluated by comparing their removal efficiency as well as requisite operator skills. The result of activated alumina was found good as compared to activated carbon, mix bed resins and ion exchange resins (IRA-400) for maximum removal of arsenic, nitrate and fluoride. The removal efficiency of arsenic, fluoride and nitrate were found 96%, 99%, 98% respectively in case of activated alumina. The advantage of adsorption process is easy to use and relatively cheaper as compared to other treatment methodologies. (author)

Base case short-term Pu-nitrate store. 1. Pt

International Nuclear Information System (INIS)

1978-11-01

This paper describes the interim storage of plutonium nitrate between the reprocessing plant and the conversion area of a mixed oxide fuel fabrication plant, which are assumed to be co-located. The capacity of the store is assumed to be 6 tons Pu. A general description of the major handling, criticality, maintenance and physical protection philosophy is given, together with comments on its safety and impact on the environment

Cation exchange process for recovery of plutonium from laboratory solutions containing chloride

International Nuclear Information System (INIS)

Gray, L.W.

1978-10-01

A cation exchange technique was developed for the separation of plutonium from laboratory solutions containing either Pu(III) or Pu(III)--Pu(IV) mixtures in acidic solutions containing chloride ions. The procedure consists of adjusting the acid concentration to less than one molar and adjusting the valence of the plutonium ion to the (III) state, if necessary. The adjusted solution is fed to a cation exchange column and washed with distilled water to remove residual chlorides from the column. Plutonium is then eluted from the column with 5M nitric acid containing 0.34M sulfamic acid. This procedure was used to separate plutonium from 1.2M chloride solution on a production-scale column. Typical plutonium recovery was 99.97%, while greater than 96% of the original chloride was rejected

Solvent anode for plutonium purification

International Nuclear Information System (INIS)

Bowersox, D.F.; Fife, K.W.; Christensen, D.C.

1986-01-01

The purpose of this study is to develop a technique to allow complete oxidation of plutonium from the anode during plutonium electrorefining. This will eliminate the generation of a ''spent'' anode heel which requires further treatment for recovery. Our approach is to employ a solvent metal in the anode to provide a liquid anode pool throughout electrorefining. We use molten salts and metals in ceramic crucibles at 700 0 C. Our goal is to produce plutonium metal at 99.9% purity with oxidation and transfer of more than 98% of the impure plutonium feed metal from the anode into the salt and product phases. We have met these criteria in experiments on the 100 to 1000 g scale. We plan to scale our operations to 4 kg of feed plutonium and to optimize the process parameters

Removal of zirconium and niobium activities from plutonium nitrate during plutonium reconversion process

International Nuclear Information System (INIS)

Ajithlal, R.T.; Rakshe, P.R.; Kumaraguru, K.

2010-01-01

Present investigation deals with quality improvement of Pu solutions after ion exchange cycle of Purex process. In order to improve the decontamination factor of Pu with respect to fission products zirconium ( 95 Zr) and niobium ( 95 Nb), Pu-Product solution was precipitated as oxalate at different compositions of nitric acid with stoichiometric and hyper-stoichiometric amount of oxalic acid. The Pu-oxalate so precipitated was washed with respective feed solutions of oxalic and nitric acid mixture, similar to feed conditions. Fission product activities in the feed, supernatant and the washes were analysed for gross gamma activity and individual fission products by Multichannel analyzer using HPGe-detector. A solution comprising of 4M HNO 3 + 0.2M excess oxalic acid precipitation with excess amount of washing yielded effective decontamination of the Pu product. (author)

Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

International Nuclear Information System (INIS)

Tingey, Joel M.; Jones, Susan A.

2005-01-01

Eighteen plutonium oxide samples originating from the Plutonium Finishing Plant (PFP) on the Hanford Site were analyzed to provide additional data on the suitability of PFP thermally stabilized plutonium oxides and Rocky Flats oxides as alternate feedstock to the Mixed Oxide Fuel Fabrication Facility (MFFF). Radiochemical and chemical analyses were performed on fusions, acid leaches, and water leaches of these 18 samples. The results from these destructive analyses were compared with nondestructive analyses (NDA) performed at PFP and the acceptance criteria for the alternate feedstock. The plutonium oxide materials considered as alternate feedstock at Hanford originated from several different sources including Rocky Flats oxide, scrap from the Remote Mechanical C-Line (RMC) and the Plutonium Reclamation Facility (PRF), and materials from other plutonium conversion processes at Hanford. These materials were received at PFP as metals, oxides, and solutions. All of the material considered as alternate feedstock was converted to PuO2 and thermally stabilized by heating the PuO2 powder at 950 C in an oxidizing environment. The two samples from solutions were converted to PuO2 by precipitation with Mg(OH)2. The 18 plutonium oxide samples were grouped into four categories based on their origin. The Rocky Flats oxide was divided into two categories, low- and high-chloride Rocky Flats oxides. The other two categories were PRF/RMC scrap oxides, which included scrap from both process lines and oxides produced from solutions. The two solution samples came from samples that were being tested at Pacific Northwest National Laboratory because all of the plutonium oxide from solutions at PFP had already been processed and placed in 3013 containers. These samples originated at the PFP and are from plutonium nitrate product and double-pass filtrate solutions after they had been thermally stabilized. The other 16 samples originated from thermal stabilization batches before canning at

Uncertainty propagation for the coulometric measurement of the plutonium concentration in CRM126 solution provided by JAEA

Energy Technology Data Exchange (ETDEWEB)

Morales-Arteaga, Maria [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-11-07

This GUM WorkbenchTM propagation of uncertainty is for the coulometric measurement of the plutonium concentration in a Pu standard material (C126) supplied as individual aliquots that were prepared by mass. The C126 solution had been prepared and as aliquoted as standard material. Samples are aliquoted into glass vials and heated to dryness for distribution as dried nitrate. The individual plutonium aliquots were not separated chemically or otherwise purified prior to measurement by coulometry in the F/H Laboratory. Hydrogen peroxide was used for valence adjustment.

Plutonium helps probe protein, superconductor

International Nuclear Information System (INIS)

Anon.

1990-01-01

Scientists are finding that plutonium can be a useful research tool that may help them answer important questions in fields as diverse as biochemistry and solid-state physics. This paper reports that U.S. research involving plutonium is confined to the Department of Energy's national laboratories and centers around nuclear weapons technology, waste cleanup and disposal, and health effects. But at Los Alamos National Laboratory, scientists also are using plutonium to probe the biochemical behavior of calmodulin, a key calcium-binding protein that mediates calcium-regulated processes in biological systems. At Argonne National Laboratory, another team is trying to learn how a superconductor's properties are affected by the 5f electrons of an actinide like plutonium

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

International Nuclear Information System (INIS)

Govindan, P.; Palamalai, A.; Vijayan, K.S.; Subba Rao, R.V.; Venkataraman, M.; Natarajan, R.

2013-01-01

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0-8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0-8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5-7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing. (author)

Preliminary studies on the spatial-temporal microdistribution of inhaled soluble plutonium in the lungs of dogs

International Nuclear Information System (INIS)

Cho, M.W.; Dagle, G.E.

1987-01-01

The pulmonary microdistribution of inhaled soluble plutonium in four beagle dogs was studied in autoradiographs of histologic sections and transmission electron micrographs of lungs. Dogs were exposed to a single nose-only aerosol of 239 Pu nitrate with a post-exposure time ranging from 1 month to 42 months. At one month after the exposure, the plutonium was dispersed throughout the lung section, with a higher percentage of the activity found on alveolar macrophages and alveolar septa. However, a nonrandom localization of the plutonium was observed as time passed. The focal concentrations were primarily in nodular or diffuse interstitial fibrotic tissues typically contiguous with subpleural, peribronchial, or perivascular areas. More than 50% of the total activity was in the form of single-tracks at one month exposure, and this percentage increased with time. In summary, this preliminary study suggests an initial random dispersion of soluble plutonium with increased concentration of activity to nonrandom focal locations with time

Plutonium waste incineration using pyrohydrolysis

International Nuclear Information System (INIS)

Meyer, M.L.

1991-01-01

Waste generated by Savannah River Site (SRS) plutonium operations includes a contaminated organic waste stream. A conventional method for disposing of the organic waste stream and recovering the nuclear material is by incineration. When the organic material is burned, the plutonium remains in the incinerator ash. Plutonium recovery from incinerator ash is highly dependent on the maximum temperature to which the oxide is exposed. Recovery via acid leaching is reduced for a high fired ash (>800 degree C), while plutonium oxides fired at lower decomposition temperatures (400--800 degrees C) are more soluble at any given acid concentration. To determine the feasibility of using a lower temperature process, tests were conducted using an electrically heated, controlled-air incinerator. Nine nonradioactive, solid, waste materials were batch-fed and processed in a top-heated cylindrical furnace. Waste material processing was completed using a 19-liter batch over a nominal 8-hour cycle. A processing cycle consisted of 1 hour for heating, 4 hours for reacting, and 3 hours for chamber cooling. The water gas shift reaction was used to hydrolyze waste materials in an atmosphere of 336% steam and 4.4% oxygen. Throughput ranged from 0.14 to 0.27 kg/hr depending on the variability in the waste material composition and density

Plutonium

International Nuclear Information System (INIS)

Watson, G.M.

1976-01-01

Discovery of the neutron made it easy to create elements which do not exist in nature. One of these is plutonium, and its isotope with mass number 239 has nuclear properties which make it both a good fuel for nuclear power reactors and a good explosive for nuclear weapons. Since it was discovered during a war the latter characteristic was put to use, but it is now evident that use of plutonium in a particular kind of nuclear reactor, the fast breeder reactor, will allow the world's resources of uranium to last for millennia as a major source of energy. Plutonium is very radiotoxic, resembling radium in this respect. Therefore the widespread introduction of fast breeder reactors to meet energy demands can be contemplated only after assurances on two points; that adequate control of the radiological hazard resulting from the handling of very large amounts of plutonium can be guaranteed, and that diversion of plutonium to illicit use can be prevented. The problems exist to a lesser degree already, since all types of nuclear reactor produce some plutonium. Some plutonium has already been dispersed in the environment, the bulk of it from atmospheric tests of nuclear weapons. (author)

Plutonium-239 and americium-241 uptake by plants from soil. Final report

International Nuclear Information System (INIS)

Brown, K.W.

1979-03-01

Alfalfa was grown in soil contaminated with plutonium-239 dioxide (239PuO2) at a concentration of 29.7 nanocuries per gram (nCi/g). In addition to alfalfa, radishes, wheat, rye, and tomatoes were grown in soils contaminated with americium-241 nitrate (241Am(NO3)3) at a concentration of 189 nCi/g. The length of exposure varied from 52 days for the radishes to 237 days for the alfalfa. The magnitude of plutonium incorporation by the alfalfa as indicated by the concentration ratio, 0.0000025, was similar to previously reported data using other chemical forms of plutonium. The results did indicate, however, that differences in the biological availability of plutonium isotopes do exist. All of the species exposed to americium-241 assimilated and translocated this radioisotope to the stem, leaf, and fruiting structures. The magnitude of incorporation as signified by the concentration ratios varied from 0.00001 for the wheat grass to 0.0152 for the radishes. An increase in the uptake of americium also occurred as a function of time for four of the five plant species. Evidence indicates that the predominant factor in plutonium and americium uptake by plants may involve the chelation of these elements in soils by the action of compounds such as citric acid and/or other similar chelating agents released from plant roots

Risk of transporting plutonium dioxide and liquid plutonium nitrate by truck and rail

International Nuclear Information System (INIS)

Williams, L.D.; Hall, R.J.

1976-07-01

Results are summarized of the risk assessments of shipping PuO 2 powder and liquid Pu nitrate by truck and rail in the U.S. In the analysis method used, the system is described, potential release sequences are identified and evaluated (fault tree used), and the system risk is assessed. It is concluded that: there is little difference in risk in shipping PuO 2 by rail and by truck; there also is little change in risk in shipping PuO 2 by rail and by truck; there also is little change in risk for liquid shipment; the vermiculite loss is somewhat less important in rail shipment; and the response of the L-10 container to crush is more important in rail transport

Plutonium Finishing Plant safety evaluation report

International Nuclear Information System (INIS)

1995-01-01

The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE's independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91

Automation of plutonium spectrophotometry

International Nuclear Information System (INIS)

Perez, J.J.; Boisde, G.; Goujon de Beauvivier, M.; Chevalier, G.; Isaac, M.

1980-01-01

Instrumentation was designed and constructed for automatic control of plutonium by molecular absorption spectrophotometry, on behalf of the reprocessing facilities, to meet two objectives: on-line measurement, of the valency state of plutonium, on by-pass, with the measured concentration covering the process concentration range up to a few mg.l -1 ; laboratory measurement of plutonium adjusted to valency VI, with operation carried out using a preparative system meeting the required containment specifications. For this two objectives, the photometer, optical cell connections are made by optical fibers resistant to β, γ radiation. Except this characteristic the devices are different according to the quality required for result [fr

Variations of uranium and plutonium coprocessing as proliferation-resistant alternatives to the classical purex process

International Nuclear Information System (INIS)

Buckham, J.A.; Sumner, W.B.

1979-08-01

Evaluation of these alternatives for processing LWR fuel has led to the following conclusions: (1) None of the alternaives provide a pure, technical solution which completely eliminates the potential for proliferation of nuclear weapons by utilizing plutonium from the light water reactors. (2) The heat spike alternative appears feasible and provides the most effective method of rendering the LWR plutonim unattractive for weapons use. (3) The low-DF process alternate would require demonstration to: (a) determine the reliability of the in-cell recycle streams which are used to prevent reversion of the process for purification of plutonium, and (b) verify the fission product decontamination factors. (4) The alternates evaluated have no significant impacts on the design of waste treatment facilities, although the required capacities of high-level solid waste processing and high-level liquid waste storage can be significantly altered. (5) The impact of these alternate processes on fuel fabrication and other aspects of the fuel cycle requires additional evaluation

BenzoDODA grafted polymeric resin—Plutonium selective solid sorbent

Energy Technology Data Exchange (ETDEWEB)

Ruhela, R., E-mail: riteshr@barc.gov.in [Materials Processing Division, Materials Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Panja, S., E-mail: surajit@barc.gov.in [Fuel Reprocessing Division, Nuclear Fuels Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Singh, A.K. [Materials Processing Division, Materials Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Dhami, P.S.; Gandhi, P.M. [Fuel Reprocessing Division, Nuclear Fuels Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

2016-11-15

Highlights: • BenzoDODA grafted polymeric resin was synthesized and evaluated for sorption of Pu(IV). • Fast sorption kinetics for ‘Pu(IV)’. • Ease of back extraction of ‘Pu’ form loaded resin. • Ease of recyclability and fair stability in HNO{sub 3} medium. - Abstract: A new ligand grafted polymeric resin (BenzoDODA SDVB) was synthesized by covalently attaching plutonium selective ligand (BenzoDODA) on to styrene divinyl benzene (SDVB) polymer matrix. BenzoDODA SDVB resin was evaluated for separation and recovery of plutonium(IV) from nitric acid medium. Sorption of Pu(IV) was found to decrease with the increase in nitric acid concentration, with very small sorption above 7.0 M HNO{sub 3}. Sorption kinetics was fast enough to achieve the equilibrium within 60 min of contact where the kinetic data fitted well to pseudo-second-order model. Sorption isotherm data fitted well to Langmuir model suggesting chemical interaction between the BenzoDODA moiety and plutonium(IV) ions. Sorption studies with some of representative radionuclides of high level waste showed that BenzoDODA SDVB is selective and therefore could be a promising solid sorbent for separation and recovery of plutonium. Further, the theoretical calculations done on BenzoDODA SDVB resin suggested Pu(NO{sub 3}){sub 4}·BenzoDODA (1:1) sorbed complex conformed to generally observed square antiprism geometry of the plutonium complexes, with contributions from oxygen atoms of four nitrate ions as well as from four oxygen atoms present in BenzoDODA (two phenolic ether oxygen atoms and two carbonyl oxygen atoms of amidic moiety).

Chemical Disposition of Plutonium in Hanford Site Tank Wastes

Energy Technology Data Exchange (ETDEWEB)

Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Susan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-05-07

This report examines the chemical disposition of plutonium (Pu) in Hanford Site tank wastes, by itself and in its observed and potential interactions with the neutron absorbers aluminum (Al), cadmium (Cd), chromium (Cr), iron (Fe), manganese (Mn), nickel (Ni), and sodium (Na). Consideration also is given to the interactions of plutonium with uranium (U). No consideration of the disposition of uranium itself as an element with fissile isotopes is considered except tangentially with respect to its interaction as an absorber for plutonium. The report begins with a brief review of Hanford Site plutonium processes, examining the various means used to recover plutonium from irradiated fuel and from scrap, and also examines the intermediate processing of plutonium to prepare useful chemical forms. The paper provides an overview of Hanford tank defined-waste–type compositions and some calculations of the ratios of plutonium to absorber elements in these waste types and in individual waste analyses. These assessments are based on Hanford tank waste inventory data derived from separately published, expert assessments of tank disposal records, process flowsheets, and chemical/radiochemical analyses. This work also investigates the distribution and expected speciation of plutonium in tank waste solution and solid phases. For the solid phases, both pure plutonium compounds and plutonium interactions with absorber elements are considered. These assessments of plutonium chemistry are based largely on analyses of idealized or simulated tank waste or strongly alkaline systems. The very limited information available on plutonium behavior, disposition, and speciation in genuine tank waste also is discussed. The assessments show that plutonium coprecipitates strongly with chromium, iron, manganese and uranium absorbers. Plutonium’s chemical interactions with aluminum, nickel, and sodium are minimal to non-existent. Credit for neutronic interaction of plutonium with these absorbers

Uncertainty propagation for the coulometric measurement of the plutonium concentration in MOX-PU4.

Energy Technology Data Exchange (ETDEWEB)

None, None

2017-11-07

This GUM WorkbenchTM propagation of uncertainty is for the coulometric measurement of the plutonium concentration in a Pu standard material (C126) supplied as individual aliquots that were prepared by mass. The C126 solution had been prepared and as aliquoted as standard material. Samples are aliquoted into glass vials and heated to dryness for distribution as dried nitrate. The individual plutonium aliquots were not separated chemically or otherwise purified prior to measurement by coulometry in the F/H Laboratory. Hydrogen peroxide was used for valence adjustment. The Pu assay measurement results were corrected for the interference from trace iron in the solution measured for assay. Aliquot mass measurements were corrected for air buoyancy. The relative atomic mass (atomic weight) of the plutonium from X126 certoficate was used. The isotopic composition was determined by thermal ionization mass spectrometry (TIMS) for comparison but not used in calculations.

Two-stage precipitation of plutonium trifluoride

International Nuclear Information System (INIS)

Luerkens, D.W.

1984-04-01

Plutonium trifluoride was precipitated using a two-stage precipitation system. A series of precipitation experiments identified the significant process variables affecting precipitate characteristics. A mathematical precipitation model was developed which was based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter that can be used to control particle characteristics

Optimizing Electrocoagulation Process for the Removal of Nitrate From Aqueous Solution

Directory of Open Access Journals (Sweden)

Dehghani

2016-01-01

Full Text Available Background High levels of nitrate anion are frequently detected in many groundwater resources in Fars province. Objectives The present study aimed to determine the removal efficiency of nitrate from aqueous solutions by electrocoagulation process using aluminum and iron electrodes. Materials and Methods A laboratory-scale batch reactor was conducted to determine nitrate removal efficiency using the electrocoagulation method. The removal of nitrate was determined at pH levels of 3, 7, and 11, different voltages (15, 20, and 30 V, and operation times of 30, 60, and 75 min, respectively. Data were analyzed using the SPSS software version 16 (Chicago, Illinois, USA and Pearson’s correlation coefficient was used to analyze the relationship between the parameters. Results Results of the present study showed that the removal efficiency was increased from 27% to 86% as pH increased from 3 to 11 at the optimal condition of 30 V and 75 min operation time. Moreover, by increasing the reaction time from 30 V to 75 min the removal efficiency was increased from 63% to 86%, respectively (30 V and pH = 11. Pearson’s correlation analysis showed that there was a significant relationship between removal efficiency and voltage and reaction time as well (P < 0.01. Conclusions In conclusion, the electrocoagulation process can be used for removing nitrate from water resources because of high efficiency, simplicity, and relatively low cost.

Stop plutonium

International Nuclear Information System (INIS)

2003-02-01

This press document aims to inform the public on the hazards bound to the plutonium exploitation in France and especially the plutonium transport. The first part is a technical presentation of the plutonium and the MOX (Mixed Oxide Fuel). The second part presents the installation of the plutonium industry in France. The third part is devoted to the plutonium convoys safety. The highlight is done on the problem of the leak of ''secret'' of such transports. (A.L.B.)

Age difference in deposition of plutonium in organs of rats and the estimation of distribution in humans

Energy Technology Data Exchange (ETDEWEB)

Fukuda, Satoshi; Iida, Haruzo [National Inst. of Radiological Sciences, Chiba (Japan)

2000-05-01

Differences in plutonium distribution in various organs, particularly the bones, of rats injected at different ages were examined in order to aid in estimating plutonium distribution in humans. Comparisons were made between rats and humans based on the bone histomorphometric and mineral density data. Male and female rats of three ages (3, 12, and 24 months old), respectively, received an injection of plutonium nitrate by two dose modalities; a fixed amount of plutonium without regard to age, sex, or body weight; per g of body weight. The rats were killed 2 weeks after the injection of plutonium. The amounts of plutonium deposited in the organs varied without regard to the body or organ weight; those in the skeleton increased from 3 to 12 months, reaching a peak at 12 months, but then decreased, along with the age-related changes in the bone surface, volume, and mineral density. Those in the liver, spleen and kidney decreased despite the body weight gain with age in both sexes. Age-related differences in the deposition of plutonium in humans were estimated based on the bone data characteristics obtained from the histomorphometry and bone mineral density for corresponding of ages between rats and humans. The results indicate that age is the most important factor in estimating the distribution of plutonium deposition in the early period after plutonium exposure, and that body or organ weight is not always a useful indicator, particularly in the aged. (author)

Plutonium dissolution from Rocky Flats Plant incinerator ash

International Nuclear Information System (INIS)

Delegard, C.H.

1985-06-01

Rockwell Hanford Operations (Rockwell) soon will commence recovery of plutonium from Rocky Flats Plant incinerator ash. In preparation for this processing, Rockwell undertook literature and laboratory studies to identify, select and optimize plutonium dissolution methods for treating the ash. Ash reburning, followed by dissolution in nitric acid containing calcium fluoride, was selected as the processing method for the ash. Recommended values of process parameters were identified. Using the selected process, 99.5% plutonium recovery was achieved, leaving about 12.7 wt % heel residue for an equal weight composite of the three ashes tested. 15 refs., 26 figs

Search for plutonium salt deposits in the plutonium extraction batteries of the Marcoule plant (1963)

International Nuclear Information System (INIS)

Bouzigues, H.; Reneaud, J.M.

1963-01-01

This report describes a method and a special apparatus making it possible to detach the insoluble plutonium salt deposits in the extraction chain of an irradiated fuel treatment plant. The process chosen allows the detection, in the extraction batteries or in the highly active chemical engineering equipment, of plutonium quantities of a few grains. After four years operation it has been impossible to detect measurable quantities of plutonium in any part of the extraction chain. The results have been confirmed by visual examinations carried out with a specially constructed endoscope. (authors) [fr

Impact of food processing and storage conditions on nitrate content in canned vegetable-based infant foods.

Science.gov (United States)

Tamme, T; Reinik, M; Roasto, M; Meremäe, K; Kiis, A

2009-08-01

The nitrate and nitrite contents were determined in canned vegetable-based infant foods of five varieties. Furthermore, changes in nitrate content during industrial processing were studied. Samples were taken from raw materials, homogenized mixtures, and final products after sterilization, and then analyzed for nitrate and nitrite content by high-pressure liquid chromatography. Processing steps preceding heat treatment, such as vegetable peeling and washing, decreased the nitrate concentrations in the range of 17 to 52%. During processing, the nitrate content in canned infant foods decreased 39 to 50%, compared with nitrate concentration in the raw-vegetable mixture. The final nitrate concentration in infant foods depends mainly on the initial nitrate content of the raw-vegetable mixture. The effect of storage time (24 and 48 h) and temperature (4 to 6 degrees C and 20 to 22 degrees C) on nitrate and nitrite content in opened canned infant-food samples was studied. After 24 h of storage at refrigerated and room temperatures, the mean nitrate content increased on average by 7 and 13%, and after 48 h of storage by 15 and 29%, respectively. The nitrite content in all analyzed samples was below the quantification limit. Storage requirements of industrial manufacturers must be followed strictly. Opened can foods, stored under refrigerated conditions, have to be consumed within 2 days, as recommended by manufacturers. The infant-food producers must pay more attention to the quality of raw materials. Nitrate content analyses should be added as compulsory tests to the quality assurance programs.

An experimental study of factors in the recovery of plutonium from combustible wastes treated by incineration, pyrolysis and other processes

International Nuclear Information System (INIS)

Bamber, D.C.; McDonald, L.A.; Roberts, W.G.; Sutcliffe, P.W.; Wilkins, J.D.

1984-01-01

The work described in this report is concerned with the incineration and pyrolysis of plutonium-contaminated combustible wastes, the leaching of the ashes and chars and the subsequent treatment of the leach solutions. A range of ashes and chars have been prepared from a range of plutonium-contaminated materials covering a variety of combustible materials (e.g. PVC, neoprene, Hypalon) and plutonium contaminants [e.g. PuO 2 , Pu(NO 3 ) 4 , (U, Pu)O 2 ]. Treatment temperatures in the range of 550-900 0 C have been investigated, the best results being obtained at or below 700 0 C with pyrolysis followed by char oxidation being the favoured process. A number of methods for treatment of the leach solutions have been considered and some have been investigated experimentally. Extraction of plutonium and americium with tributylphosphate (TBP) from a leach solution conditioned to 0.1 M H/+5 M NO 3 - has been studied. The key stage has been found to be the conditioning step where precautions must be taken to ensure that plutonium-containing precipitates and non-extractable plutonium are not formed. Consideration has also been given to treatment of the americium containing raffinates from a high acid TBP extraction and some methods have been investigated. A range of simple washing experiments have been carried out in order to compare the process with incineration/pyrolysis

Plutonium metal preparation and purification at Los Alamos, 1984

International Nuclear Information System (INIS)

Christensen, D.C.; Williams, J.D.; McNeese, J.A.; Fife, K.W.

1984-01-01

Plutonium metal preparation and purification are well established at Los Alamos. Metal is prepared by calcothermic reduction of both PuF 4 and PuO 2 . Metal is purified by halide slagging, casting, and electrorefining. The product from the production sequence is ultrapure plutonium metal. All of the processes involve high temperature operation and all but casting involve molten salt media. Development efforts are fourfold: (1) recover plutonium values from residues; (2) reduce residue generation through process improvements and changes; (3) recycle of reagents, and (4) optimize and integrate all processes into a close-loop system. Plutonium residues are comprised of oxides, chlorides, colloidal metal suspensions, and impure metal heels. Pyrochemical recovery techniques are under development to address each residue. In addition, we are looking back at each residue generation step and are making process changes to reduce plutonium content in each residue. Reagent salt is the principle media used in pyrochemical processing. The regeneration and recycle of these reagents will both reduce our waste handling and operating expense. The fourth area, process optimization, involves both existing processes and new process developments. A status of efforts in all four of these areas will be summarized

Plutonium immobilization program - Cold pour Phase 1 test results

International Nuclear Information System (INIS)

Hamilton, L.

2000-01-01

The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

Plutonium Immobilization Program - Cold pour Phase 1 test results

International Nuclear Information System (INIS)

Hamilton, L.

2000-01-01

The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

THE DEACTIVATION, DECONTAMINATION AND DECOMMISSIONING OF THE PLUTONIUM FINISHING PLANT, A FORMER PLUTONIUM PROCESSING FACILITY AT DOE'S HANFORD SITE

International Nuclear Information System (INIS)

CHARBONEAU, S.L.

2006-01-01

The Plutonium Finishing Plant (PFP) was constructed as part of the Manhattan Project during World War II. The Manhattan Project was developed to usher in the use of nuclear weapons to end the war. The primary mission of the PFP was to provide plutonium used as special nuclear material (SNM) for fabrication of nuclear devices for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race and later the processing of fuel grade mixed plutonium-uranium oxide to support DOE's breeder reactor program. In October 1990, at the close of the production mission for PFP, a shutdown order was prepared by the Department of Energy (DOE) in Washington,; DC--and issued to the Richland DOE field office. Subsequent to the shutdown order, a team from the Defense Nuclear Facilities Safety Board (DNFSB) analyzed the hazards at PFP associated with the continued storage of certain forms of plutonium solutions and solids. The assessment identified many discrete actions that were required to stabilize the different plutonium forms into stable form and repackage the material in high integrity containers. These actions were technically complicated and completed as part of the PFP nuclear material stabilization project between 1995 and early 2005. The completion of the stabilization project was a necessary first step in deactivating PFP. During stabilization, DOE entered into negotiations with the U.S. Environmental Protection Agency (EPA) and the State of Washington and established milestones for the Deactivation and Decommissioning (DandD) of the PFP. The DOE and its contractor, Fluor Hanford (Fluor), have made great progress in deactivating, decontaminating and decommissioning the PFP at the Hanford Site as detailed in this paper. Background information covering the PFP DandD effort includes descriptions of negotiations with the State of Washington concerning consent

Nondestructive assay of plutonium residue in horizontal storage tanks

International Nuclear Information System (INIS)

Marsh, S.F.

1985-01-01

Aqueous plutonium recovery and purification processes often involve the temporary storage of plutonium solutions in holding tanks. Because plutonium is known to precipitate from aqueous solutions under certain conditions, there is a continuing need to assay emptied tanks for plutonium residue. A portable gamma spectrometer system, specifically designed for this purpose, provides rapid assay of such plutonium residues in horizontal storage tanks. A means is thus available for the nondestructive analysis of these tanks on a regular schedule to ensure that significant deposits of plutonium are not allowed to accumulate. 5 figs

INTEGRATED BIOREACTOR SYSTEM FOR THE TREATMENT OF CYANIDE, METALS AND NITRATES IN MINE PROCESS WATER

Science.gov (United States)

An innovative biological process is described for the tratment of cyanide-, metals- and nitrate-contaminated mine process water. The technology was tested for its ability to detoxify cyanide and nitrate and to immobilize metals in wastewater from agitation cyanide leaching. A pil...

Analysis of civilian processing programs in reduction of excess separated plutonium and high-enriched uranium

International Nuclear Information System (INIS)

Persiani, P.J.

1995-01-01

The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. The analysis addresses several options in reducing the excess separated plutonium and HEU, and the consequences on nonproliferation and safeguards policy assessments resulting from the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials

Status summary of chemical processing development in plutonium-238 supply program

Energy Technology Data Exchange (ETDEWEB)

Collins, Emory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Benker, Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wham, Robert M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DePaoli, David W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delmau, Laetitia Helene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sherman, Steven R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-10-01

This document summarizes the status of development of chemical processing in the Plutonium-238 Supply Program (PSP) near the end of Demonstration 1. The objective of the PSP is “to develop, demonstrate, and document a production process that meets program objectives and to prepare for its operation” (Frazier et al. 2016). Success in the effort includes establishing capability using the current infrastructure to produce Np targets for irradiation in Department of Energy research reactors, chemically processing the irradiated targets to separate and purify the produced Pu and transferring the PuO₂ product to Los Alamos National Laboratory (LANL) at an average rate of 1.5 kg/y.

PUBG; purex solvent extraction process model. [IBM3033; CDC CYBER175; FORTRAN IV

Energy Technology Data Exchange (ETDEWEB)

Geldard, J.F.; Beyerlein, A.L.

PUBG is a chemical model of the Purex solvent extraction system, by which plutonium and uranium are recovered from spent nuclear fuel rods. The system comprises a number of mixer-settler banks. This discrete stage structure is the basis of the algorithms used in PUBG. The stages are connected to provide for countercurrent flow of the aqueous and organic phases. PUBG uses the common convention that has the aqueous phase enter at the lowest numbered stage and exit at the highest one; the organic phase flows oppositely. The volumes of the mixers are smaller than those of the settlers. The mixers generate a fine dispersion of one phase in the other. The high interfacial area is intended to provide for rapid mass transfer of the plutonium and uranium from one phase to the other. The separation of this dispersion back into the two phases occurs in the settlers. The species considered by PUBG are Hydrogen (1+), Plutonium (4+), Uranyl Oxide (2+), Plutonium (3+), Nitrate Anion, and reductant in the aqueous phase and Hydrogen (1+), Uranyl Oxide (2+), Plutonium (4+), and TBP (tri-n-butylphosphate) in the organic phase. The reductant used in the Purex process is either Uranium (4+) or HAN (hydroxylamine nitrate).IBM3033;CDC CYBER175; FORTRAN IV; OS/MVS or OS/MVT (IBM3033), NOS 1.3 (CDC CYBER175); The IBM3033 version requires 150K bytes of memory for execution; 62,000 (octal) words are required by the CDC CYBER175 version..

Carcinogenesis and Inflammatory Effects of Plutonium-Nitrate Retention in an Exposed Nuclear Worker and Beagle Dogs

International Nuclear Information System (INIS)

Nielsen, Christopher E.; Wang, Xihai; Robinson, Robert J.; Brooks, Antone L.; Lovaglio, Jamie A.; Patton, Kristin M.; McComish, Stacey; Tolmachev, Sergei Y.; Morgan, William F.

2014-01-01

The genetic and inflammatory response pathways elicited following plutonium exposure in archival lung tissue of an occupationally exposed human and experimentally exposed beagle dogs were investigated. These pathways include: tissue injury, apoptosis and gene expression modifications related to carcinogenesis and inflammation. In order to determine which pathways are involved, multiple lung samples from a plutonium exposed worker (Case 0269), a human control (Case 0385), and plutonium exposed beagle dogs were examined using histological staining and immunohistochemistry. Examinations were performed to identify target tissues at risk of radiation-induced fibrosis, inflammation, and carcinogenesis. Case 0269 showed interstitial fibrosis in peripheral and subpleural regions of the lung, but no pulmonary tumors. In contrast, the dogs with similar and higher doses showed pulmonary tumors primarily in brochiolo-alveolar, peripheral and subpleural alveolar regions. The TUNEL assay showed slight elevation of apoptosis in tracheal mucosa, tumor cells, and nuclear debris was present in the inflammatory regions of alveoli and lymph nodes of both the human and the dogs. The expression of apoptosis and a number of chemokine/cytokine genes was slightly but not significantly elevated in protein or gene levels compared to that of the control samples. In the beagles, mucous production was increased in the airway epithelial goblet cells and glands of trachea, and a number of chemokine/cytokine genes showed positive immunoreactivity. This analysis of archival tissue from an accidentally exposed worker and in a large animal model provides valuable information on the effects of long-term retention of plutonium in the respiratory tract and the histological evaluation study may impact mechanistic studies of radiation carcinogenesis

Safeguarding the Plutonium Fuel Cycle

International Nuclear Information System (INIS)

Johnson, S.J.; Lockwood, D.

2013-01-01

In developing a Safeguards Approach for a plutonium process facility, two general diversion and misuse scenarios must be addressed: 1) Unreported batches of undeclared nuclear material being processed through the plant and bypassing the accountancy measurement points, and 2) The operator removing plutonium at a rate that cannot be detected with confidence due to measurement uncertainties. This paper will look at the implementation of international safeguards at plutonium fuel cycle facilities in light of past lessons learned and current safeguards approaches. It will then discuss technical areas which are currently being addressed as future tools to improve on the efficiency of safeguards implementation, while maintaining its effectiveness. The discussion of new improvements will include: safeguards by design (SBD), process monitoring (PM), measurement and monitoring equipment, and data management. The paper is illustrated with the implementation of international safeguards at the Rokkasho Reprocessing Plant in Japan and its accountancy structure is detailed. The paper is followed by the slides of the presentation

Remote handling in the Plutonium Immobilization Project: Plutonium conversion and first stage immobilization

International Nuclear Information System (INIS)

Brault, J.R.

2000-01-01

Since the break up of the Soviet Union at the end of the Cold War, the United States and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

Plutonium working group report on environmental, safety and health vulnerabilities associated with the Department's plutonium storage. Volume I: Summary

International Nuclear Information System (INIS)

1994-11-01

At the conclusion of the Cold War, the Department of Energy (DOE) stopped plutonium processing for nuclear weapons production. Facilities used for that purpose now hold significant quantities of plutonium in various forms. Unless properly stored and handled, plutonium can present environment, safety and health (ES ampersand H) hazards. Improperly stored plutonium poses a variety of hazards. When containers or packaging fail to fully protect plutonium metal from exposure to air, oxidation can occur and cause packaging failures and personnel contamination. Contamination can also result when plutonium solutions leak from bottles, tanks or piping. Plutonium in the form of scrap or residues generated by weapons production are often very corrosive, chemically reactive and difficult to contain. Buildings and equipment that are aging, poorly maintained or of obsolete design contribute to the overall problem. Inadvertent accumulations of plutonium of any form in sufficient quantities within facilities can result in nuclear criticality events that could emit large amounts of radiation locally. Contamination events and precursors of criticality events are causing safety and health concerns for workers at the Department's plutonium facilities. Contamination events also potentially threaten the public and the surrounding environment

Search for plutonium salt deposits in the plutonium extraction batteries of the Marcoule plant (1963); Recherche de depots de sels de plutonium dans les batteries d'extraction du plutonium de l'usine de Marcoule (1963)

Energy Technology Data Exchange (ETDEWEB)

Bouzigues, H; Reneaud, J M [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1963-07-01

This report describes a method and a special apparatus making it possible to detach the insoluble plutonium salt deposits in the extraction chain of an irradiated fuel treatment plant. The process chosen allows the detection, in the extraction batteries or in the highly active chemical engineering equipment, of plutonium quantities of a few grains. After four years operation it has been impossible to detect measurable quantities of plutonium in any part of the extraction chain. The results have been confirmed by visual examinations carried out with a specially constructed endoscope. (authors) [French] Ce rapport decrit une methode et un montage special permettant de detacher les accumulations de sels de plutonium insolubles dans les chaines d'extraction d'une usine de traitement de combustible irradie. Le procede retenu permet de reperer, dans des batteries d'extraction ou dans l'appareillage de genie chimique fortement actif, des masses de plutonium de quelques grammes. Apres quatre annees de fonctionnement, il n'a pas ete possible de deceler des quantites ponderables de plutonium en aucun endroit de la chaine d'extraction. Ces resultats ont ete confirmes par les examens visuels effectues a l'aide d'un endoscope concu specialement pour cet usage. (auteurs)

Evaluating sources and processing of nonpoint source nitrate in a small suburban watershed in China

Science.gov (United States)

Han, Li; Huang, Minsheng; Ma, Minghai; Wei, Jinbao; Hu, Wei; Chouhan, Seema

2018-04-01

Identifying nonpoint sources of nitrate has been a long-term challenge in mixed land-use watershed. In the present study, we combine dual nitrate isotope, runoff and stream water monitoring to elucidate the nonpoint nitrate sources across land use, and determine the relative importance of biogeochemical processes for nitrate export in a small suburban watershed, Longhongjian watershed, China. Our study suggested that NH4+ fertilizer, soil NH4+, litter fall and groundwater were the main nitrate sources in Longhongjian Stream. There were large changes in nitrate sources in response to season and land use. Runoff analysis illustrated that the tea plantation and forest areas contributed to a dominated proportion of the TN export. Spatial analysis illustrated that NO3- concentration was high in the tea plantation and forest areas, and δ15N-NO3 and δ18O-NO3 were enriched in the step ponds. Temporal analysis showed high NO3- level in spring, and nitrate isotopes were enriched in summer. Study as well showed that the step ponds played an important role in mitigating nitrate pollution. Nitrification and plant uptake were the significant biogeochemical processes contributing to the nitrogen transformation, and denitrification hardly occurred in the stream.

Process for the preparation of uranium dioxide

International Nuclear Information System (INIS)

Watt, G.W.; Baugh, D.W. Jr.

1981-01-01

A method for the preparation of actinide dioxides using actinide nitrate hexahydrates as starting materials is described. The actinide nitrate hexahydrate is reacted with sodium dithionite, and the product is heated in the absence of oxygen to obtain the dioxide. Preferably, the actinide is uranium, plutonium or neptunium. (LL)

Solvent extraction process development for high plutonium fuel cycles

Energy Technology Data Exchange (ETDEWEB)

Anil Kumar, R; Selvaraj, P G; Natarajan, R; Raman, V R [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

1994-06-01

The purification of high plutonium bearing irradiated fuels using 30% TBP in dodecane diluent requires precise determination of concentration profiles during steady state, transient and process upset conditions. Mathematical models have been developed and a computer code is in use for determining Pu-U concentration profiles in a solvent extraction equipment in a typical reprocessing plant. The process parameters have been optimised for recovery of U and Pu and decontamination from the fission products. This computer code is used to analyse the extraction flow sheets of fuels of two typical Pu-U compositions encountered in Indian fast breeder programme. The analysis include the effect of uncertainty in equilibrium condition prediction by the model and the variation of flows of streams during plant operation. The studies highlight the margin available to avoid second organic phase formation and adjustments required in the process flowsheet. (author). 7 refs., 7 figs., 2 tabs.

Destruction of nitrates, organics, and ferrocyanides by hydrothermal processing

International Nuclear Information System (INIS)

Robinson, J.M.; Foy, B.R.; Dell'Orco, P.C.; Anderson, G.; Archuleta, F.; Atencio, J.; Breshears, D.; Brewer, R.; Eaton, H.; McFarland, R.; McInroy, R.; Reynolds, T.; Sedillo, M.; Wilmanns, E.; Buelow, S.J.

1993-01-01

This work targets the remediation of the aqueous mixed wastes stored in the underground tanks at the Department of Energy site in Hanford, Washington via hydrothermal processing. The feasibility of destroying the nitrate, organic, and ferrocyanide components of the wastes under supercritical and near critical conditions (623 degree K to 873 degree K, 22.1 MPa to 103.4 MPa) is addressed. A novel method was developed for determining the solubility of nitrate salts in supercritical water solutions at pressures ranging from 24.8 MPa to 30.3 MPa (3600 psi to 4400 psi) and temperatures from 723 degree K to 798 degree K. Sodium nitrate solubilities ranged from 293 mg/kg at 24.8 MPa and 798 degree K to 1963 mg/kg at 30.3 MPa and 723 degree K. Solubility was found to vary directly with pressure, and inversely with temperature. An empirical relationship was developed for the estimation of sodium nitrate solubility at water densities between 0.08 and 0.16 kg/L and temperatures between 723 degree K and 798 degree K. A small volume batch reactor equipped with optical diagnostics was used to monitor the phase behavior of a diluted variant of a tank 101-SY simulant. Preliminary results suggest that a single phase is formed at 83 MPa at 773 degree K

Release and dispersion of overheated liquids from plutonium-nitrate transfer containers

International Nuclear Information System (INIS)

Seehars, H.D.; Hochrainer, D.; Spiekermann, M.

1985-01-01

Potential traffic accidents of 18B-transfer containers with liquid Pu-nitrate during road transportation may induce an exposure of the Ti vessel itself to a fire due to the ignition of the leaking fuel up to a critical level, causing the burst of the vessel and the more or less complete release of the contents in form of liquid aerosol particles. A report is given on experiments with the original Ti vessels and a quadrivalent Ce-nitrate solution used as a substitute with similar physico-chemical properties as the Pu-nitrate solution. Total release of mass as well as of the respirable particle mass fraction is strongly dependent on the orientation of the vessel. Maximum release, connected with a high bursting pressure and the total destruction of the vessel, is observed in case of the vertical orientation of the vessel. According to the weak temporal variability of the low wind speed (between 3 and 4.5 m/s) and direction parallel to the centre of the measuring area strong horizontal variations of the inhalation hazard occured in the range less than 50 m from the origin, while spatially homogeneous inhalation hazards were observed in the range of more than 50 m up to 200 m, almost independent on the orientation of the vessel. The extremely high total particle mass fractions between 1.6 and 8.6 mg at distances up to 50 m are noticeable. Scanning electron microscope analysis and electron probe microanalysis of Ce particles deposited up to distances of 100 m from the origin, indicated their deposition in the liquid state. (orig./HP)

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)

Experimental Studies on Plutonium Kinetics in Marine Biota

International Nuclear Information System (INIS)

Fowler, S.; Heyraud, M.; Beasley, T.M.

1976-01-01

Laboratory experiments were undertaken to measure plutonium flux through marine organisms and to clarify the pathways by which this important element is cycled in the marine environment. The use of a specially prepared isotope, plutonium-237, allowed measurements to be made with standard Nal(Tl) scintillation techniques. Mussels, shrimp and worms were allowed to accumulate plutonium-237 from sea water for up to 25 days. Accumulation by shrimp was relatively slow and the degree of uptake was strongly influenced by moulting. Cast moults contained large fractions of the shrimps ' plutonium content, indicating the high affinity of plutonium for surface areas. Only small amounts of the isotope in the moult are lost to water; hence, moulting is considered to be an important biological parameter in the biogeochemical cycling of plutonium. Mussels attained higher concentration factors than shrimp with most of the accumulated isotope (> 80%) located in the shell. Byssus threads often contained large fractions of the mussels' plutonium-237 content and reached concentration factors as high as 4100. Worms readily accumulated plutonium-237 in either the +4 or +6 state, reaching concentration factors of approximately 200, Retention studies indicated a relatively slow loss of plutonium-237 from all animals studied. In the case of mussels, a computed half-time for a large fraction of the animals plutonium content was of the order of 2 years. The more rapid loss from shrimp (Tb 1/2 = 1.5 months) was due principally to the large fraction of plutonium lost at moult. Food chain studies with shrimp indicated that tissue build-up via plutonium ingestion would be a slow process. Total excretion was not entirely a result of passing contaminated food through the gut; approximately 15% of the ingested plutonium was removed from the contaminated food and subsequently excreted by processes other than defaecation of labelled food. Ratios of four different plutonium isotopes used in

Experimental studies on plutonium kinetics in marine biota

International Nuclear Information System (INIS)

Fowler, S.; Heyraud, M.; Beasley, T.M.

1975-01-01

Laboratory experiments were undertaken to measure plutonium flux through marine organisms and to clarify the pathways by which this important element is cycled in the marine environment. The use of a specially prepared isotope, plutonium-237, allowed measurements to be made with standard NaI(Tl) scintillation techniques. Mussels, shrimp and worms were allowed to accumulate plutonium-237 from seawater for up to 25 days. Accumulation by shrimp was relatively slow and the degree of uptake was strongly influenced by moulting. Cast moults contained large fractions of the shrimps' plutonium content, indicating the high affinity of plutonium for surface areas. Only small amounts of the isotope in the moult are lost to water; hence, moulting is considered to be an important biological parameter in the biogeochemical cycling of plutonium. Mussels attained higher concentration factors than shrimp with most of the accumulated isotope (>80%) located in the shell. Byssus threads often contained large fractions of the mussels' plutonium-237 content and reached concentration factors as high as 4100. Worms readily accumulated plutonium-237 in either the +4 or +6 state, reaching concentration factors of approximately 200. Retention studies indicated a relatively slow loss of plutonium-237 from all animals studied. In the case of mussels, a computed half-time for a large fraction of the animals' plutonium content was of the order of 2 years. The more rapid loss from shrimp (Tbsub(1/2)=1.5 months) was due principally to the large fraction of plutonium lost at moult. Food chain studies with shrimp indicated that tissue build-up via plutonium ingestion would be a slow process. Total excretion was not entirely a result of passing contaminated food through the gut; approximately 15% of the ingested plutonium was removed from the contaminated food and subsequently excreted by processes other than defaecation of labelled food. Ratios of four different plutonium isotopes used in the

Studying of the dehydration process of uranyl nitrate hexahydrate (Unh)

International Nuclear Information System (INIS)

Badalov, A.; Kamalov, J.J.; Homidov, B.J.; Mirsaidov, I.U.; Eshbekov, N.R.

2005-01-01

By the tensimeteric method is studying the dehydration process of uranyl nitrate hexahydrate (Unh). It is shown, that the temperature interval 300-400 K in equilibrium conditions the dehydration process of Unh runs in three stages. According to the equations of dependence of saturated steam pressure from temperature, the thermodynamic characteristics of each stage of the dehydration process of Unh are calculated

Plutonium-236 traces determination in plutonium-238 by α spectrometry

International Nuclear Information System (INIS)

Acena, M.L.; Pottier, R.; Berger, R.

1969-01-01

Two methods are described in this report for the determination of plutonium-236 traces in plutonium-238 by a spectrometry using semi-conductor detectors. The first method involves a direct comparison of the areas under the peaks of the α spectra of plutonium-236 and plutonium-238. The electrolytic preparation of the sources is carried out after preliminary purification of the plutonium. The second method makes it possible to determine the 236 Pu/ 238 Pu ratio by comparing the areas of the α peaks of uranium-232 and uranium-234, which are the decay products of the two plutonium isotopes respectively. The uranium in the source, also deposited by electrolysis, is separated from a 1 mg amount of plutonium either by a T.L.A. extraction, or by the use of ion-exchange resins. The report ends with a discussion of the results obtained with plutonium of two different origins. (authors) [fr

Direct reduction of plutonium from dicesium hexachloroplutonate

International Nuclear Information System (INIS)

Averill, W.A.; Boyd, T.E.

1991-01-01

The Rocky Flats Plant produces dicesium hexachloroplutonate (DCHP) primarily as a reagent in the molten salt extraction of americium from plutonium metal. DCHP is precipitated from aqueous chloride solutions derived from the leaching of process residues with a high degree of selectivity. DCHP is a chloride salt of plutonium, while the traditional aqueous precipitate is a hydrated oxide. Plutonium metal preparation from the oxide involves either the conversion of oxide to a halide followed by metallothermic reduction or direct reduction of the oxide using a flux. Either method generates at least three times as much radioactively contaminated waste as metal produced. Plutonium concentration by DCHP precipitation, however, produces a chloride salt that can be reduced using calcium metal at a temperature of approximately 1000K. In this paper the advantages and limitations of this process are discussed

Nitrate control strategies in an activated sludge wastewater treatment process

Energy Technology Data Exchange (ETDEWEB)

Shen, Wenhao; Tao, Erpan; Chen, Xiaoquan; Liu, Dawei [South China University of Technology, Guangzhou (China); Liu, Hongbin [Kyung Hee University, Yongin (Korea, Republic of)

2014-03-15

We studied nitrate control strategies in an activated sludge wastewater treatment process (WWTP) based on the activated sludge model. Two control strategies, back propagation for proportional-integral-derivative (BP-PID) and adaptive-network based fuzzy inference systems (ANFIS), are applied in the WWTP. The simulation results show that the simple local constant setpoint control has poor control effects on the nitrate concentration control. However, the ANFIS (4*1) controller, which considers not only the local constant setpoint control of the nitrate concentration, but also three important indices in the effluent--ammonia concentration, total suspended sludge concentration and total nitrogen concentration--demonstrates good control performance. The results also prove that ANFIS (4*1) controller has better control performance than that of the controllers PI, BP-PID and ANFIS (2*1), and that the ANFIS (4*1) controller is effective in improving the effluent quality and maintaining the stability of the effluent quality.

Geomorphology of plutonium in the Northern Rio Grande

Energy Technology Data Exchange (ETDEWEB)

Graf, W.L. [Arizona Univ., Tempe, AZ (United States). Dept., of Geography

1993-03-01

Nearly all of the plutonium in the natural environment of the Northern Rio Grande is associated with soils and sediment, and river processes account for most of the mobility of these materials. A composite regional budget for plutonium based on multi-decadal averages for sediment and plutonium movement shows that 90 percent of the plutonium moving into the system is from atmospheric fallout. The remaining 10 percent is from releases at Los Alamos. Annual variation in plutonium flux and storage exceeds 100 percent. The contribution to the plutonium budget from Los Alamos is associated with relatively coarse sediment which often behaves as bedload in the Rio Grande. Infusion of these materials into the main stream were largest in 1951, 1952, 1957, and 1968. Because of the schedule of delivery of plutonium to Los Alamos for experimentation and weapons manufacturing, the latter two years are probably the most important. Although the Los Alamos contribution to the entire plutonium budget was relatively small, in these four critical years it constituted 71--86 percent of the plutonium in bedload immediately downstream from Otowi.

Geomorphology of plutonium in the Northern Rio Grande

International Nuclear Information System (INIS)

Graf, W.L.

1993-03-01

Nearly all of the plutonium in the natural environment of the Northern Rio Grande is associated with soils and sediment, and river processes account for most of the mobility of these materials. A composite regional budget for plutonium based on multi-decadal averages for sediment and plutonium movement shows that 90 percent of the plutonium moving into the system is from atmospheric fallout. The remaining 10 percent is from releases at Los Alamos. Annual variation in plutonium flux and storage exceeds 100 percent. The contribution to the plutonium budget from Los Alamos is associated with relatively coarse sediment which often behaves as bedload in the Rio Grande. Infusion of these materials into the main stream were largest in 1951, 1952, 1957, and 1968. Because of the schedule of delivery of plutonium to Los Alamos for experimentation and weapons manufacturing, the latter two years are probably the most important. Although the Los Alamos contribution to the entire plutonium budget was relatively small, in these four critical years it constituted 71--86 percent of the plutonium in bedload immediately downstream from Otowi

Plutonium controversy

International Nuclear Information System (INIS)

Gofman, J.W.

1976-01-01

If the world chooses to seek a solution to the energy dilemma through nuclear energy, the element plutonium will become an article of commerce to be handled in quantities of thousands of tonnes annually. Plutonium is a uniquely potent inhalation carcinogen, the potential induction of lung cancer dwarfing other possible toxic effects. For reasons to be presented here, it is the author's opinion that plutonium's carcinogenicity has been very seriously underestimated. If one couples the corrected carcinogenicity with the probable degree of industrial containment of the plutonium, it appears that the commercialization of a plutonium-based energy economy is not an acceptable option for society. Sagan's statement that ''the experience of 30 years supports the contention that plutonium can be used safely'' is manifestly indefensible. No meaningful epidemiological study of plutonium-exposed workers for that 30-year period has ever been done. Since thousands of those possibly exposed have left the industry and are not even available to follow-up, it is doubtful that any meaningful study of ''the experience of 30 years'' will ever be accomplished

Plutonium working group report on environmental, safety and health vulnerabilities associated with the Department's plutonium storage. Volume 2, Appendix A: Process and protocol

International Nuclear Information System (INIS)

1994-09-01

This appendix contains documentation prepared by the Plutonium ES and H Vulnerability Working Group for conducting the Plutonium ES and H Vulnerability Assessment and training the assessment teams. It has the following five parts. (1) The Project Plan describes the genesis of the project, sets forth the goals, objectives and scope, provides definitions, the projected schedule, and elements of protocol. (2) The Assessment Plan provides a detailed methodology necessary to guide the many professionals who have been recruited to conduct the DOE-wide assessment. It provides guidance on which types and forms of plutonium are to be considered within the scope of the assessment, and lays out the assessment methodology to be used. (3) The memorandum from the Project to Operations Office Managers provides the protocol and direction for participation in the assessment by external stakeholders and members of the public; and the guidance for the physical inspection of plutonium materials in storage. (4) The memorandum from the Project to the assessment teams provides guidance for vulnerability screening criteria, vulnerability evaluation and prioritization process, and vulnerability quantification for prioritization. (5) The Team Training manual was used at the training session held in Colorado Springs on April 19--21, 1994 for all members of the Working Group Assessment Teams and for the leaders of the Site Assessment Teams. The goal was to provide the same training to all of the individuals who would be conducting the assessments, and thereby provide consistency in the conduct of the assessments and uniformity in reporting of the results. The training manual in Section A.5 includes supplemental material provided to the attendees after the meeting

Preparation of hexavalent plutonium and its determination in the presence of tetravalent plutonium; Preparation de plutonium hexavalent et dosage en presence de plutonium tetravalent

Energy Technology Data Exchange (ETDEWEB)

Corpel, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Corpel, J [Institut du Radium, 75 - Paris (France)

1958-07-01

In order to study the eventual reduction of plutonium from the VI-valent state to the IV-valent state, in sulphuric medium, under the influence of its own {alpha} radiation or of the {gamma}-rays from a cobalt-60 source, we have developed a method for preparing pure hexavalent plutonium and two methods for determining solutions containing tetravalent and hexavalent plutonium simultaneously. Hexavalent plutonium was prepared by anodic oxidation at a platinum electrode. Study of the oxidation yield as a function of various factors has made it possible to define experimental conditions giving complete oxidation. For concentrations in total plutonium greater than 1.5 x 10{sup -3} M, determination of the two valencies IV and VI was carried out by spectrophotometry at two wavelengths. For lower concentrations, the determination was done by counting, after separation of the tetravalent plutonium in the form of fluoride in the presence of a carrier. (author) [French] Afin d'etudier l'eventuelle reduction du plutonium de l'etat de valence VI a l'etat de valence IV, en milieu sulfurique sous l'influence de son propre rayonnement {alpha} ou des rayons {gamma} d'une source de cobalt-60, nous avons mis au point une methode de preparation de plutonium hexavalent pur et deux methodes de dosage des solutions contenant simultanement du plutonium tetravalent et du plutonium hexavalent. Nous avons prepare le plutonium hexavalent par oxydation anodique au contact d'une electrode de platine. L'etude de rendement de l'oxydation en fonction des divers facteurs nous a permis de definir des conditions experimentales donnant une oxydation complete. Pour des concentrations en plutonium total superieures a 1,5.10{sup -3} M, le dosage des deux valences IV et VI a ete realise par spectrophotometrie a deux longueurs d'onde. Pour des concentrations inferieures, le dosage a ete effectue par comptage apres separation du plutonium tetravalent sous la forme du fluorure en presence d'un entraineur

Plutonium in nature

International Nuclear Information System (INIS)

Madic, C.

1994-01-01

Plutonium in nature comes from natural sources and anthropogenic ones. Plutonium at the earth surface comes principally from anthropogenic sources. It is easily detectable in environment. The plutonium behaviour in environment is complex. It seems necessary for the future to reduce releases in environment, to improve predictive models of plutonium behaviour in geosphere, to precise biological impact of anthropogenic plutonium releases

A review of plutonium oxalate decomposition reactions and effects of decomposition temperature on the surface area of the plutonium dioxide product

International Nuclear Information System (INIS)

Orr, R.M.; Sims, H.E.; Taylor, R.J.

2015-01-01

Plutonium (IV) and (III) ions in nitric acid solution readily form insoluble precipitates with oxalic acid. The plutonium oxalates are then easily thermally decomposed to form plutonium dioxide powder. This simple process forms the basis of current industrial conversion or ‘finishing’ processes that are used in commercial scale reprocessing plants. It is also widely used in analytical or laboratory scale operations and for waste residues treatment. However, the mechanisms of the thermal decompositions in both air and inert atmospheres have been the subject of various studies over several decades. The nature of intermediate phases is of fundamental interest whilst understanding the evolution of gases at different temperatures is relevant to process control. The thermal decomposition is also used to control a number of powder properties of the PuO_2 product that are important to either long term storage or mixed oxide fuel manufacturing. These properties are the surface area, residual carbon impurities and adsorbed volatile species whereas the morphology and particle size distribution are functions of the precipitation process. Available data and experience regarding the thermal and radiation-induced decompositions of plutonium oxalate to oxide are reviewed. The mechanisms of the thermal decompositions are considered with a particular focus on the likely redox chemistry involved. Also, whilst it is well known that the surface area is dependent on calcination temperature, there is a wide variation in the published data and so new correlations have been derived. Better understanding of plutonium (III) and (IV) oxalate decompositions will assist the development of more proliferation resistant actinide co-conversion processes that are needed for advanced reprocessing in future closed nuclear fuel cycles. - Highlights: • Critical review of plutonium oxalate decomposition reactions. • New analysis of relationship between SSA and calcination temperature. • New SEM

Standard test method for quantitative determination of americium 241 in plutonium by Gamma-Ray spectrometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

1994-01-01

1.1 This test method covers the quantitative determination of americium 241 by gamma-ray spectrometry in plutonium nitrate solution samples that do not contain significant amounts of radioactive fission products or other high specific activity gamma-ray emitters. 1.2 This test method can be used to determine the americium 241 in samples of plutonium metal, oxide and other solid forms, when the solid is appropriately sampled and dissolved. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Simultaneous nitrate reduction and acetaminophen oxidation using the continuous-flow chemical-less VUV process as an integrated advanced oxidation and reduction process

Energy Technology Data Exchange (ETDEWEB)

Moussavi, Gholamreza, E-mail: moussavi@modares.ac.ir; Shekoohiyan, Sakine

2016-11-15

Highlights: • Simultaneous advanced oxidation and reduction processes were explored in VUV system. • Complete reduction of nitrate to N{sub 2} was achieved at the presence of acetaminophen. • Complete degradation of acetaminophen was achieved at the presence of nitrate. • Over 95% of acetaminophen was mineralized in the VUV photoreactor. • VUV is a chemical-less advanced process for treating water emerging contaminants. - Abstract: This work was aimed at investigating the performance of the continuous-flow VUV photoreactor as a novel chemical-less advanced process for simultaneously oxidizing acetaminophen (ACT) as a model of pharmaceuticals and reducing nitrate in a single reactor. Solution pH was an important parameter affecting the performance of VUV; the highest ACT oxidation and nitrate reduction attained at solution pH between 6 and 8. The ACT was oxidized mainly by HO· while the aqueous electrons were the main working agents in the reduction of nitrate. The performance of VUV photoreactor improved with the increase of hydraulic retention time (HRT); the complete degradation of ACT and ∼99% reduction of nitrate with 100% N{sub 2} selectivity achieved at HRT of 80 min. The VUV effluent concentrations of nitrite and ammonium at HRT of 80 min were below the drinking water standards. The real water sample contaminated with the ACT and nitrate was efficiently treated in the VUV photoreactor. Therefore, the VUV photoreactor is a chemical-less advanced process in which both advanced oxidation and advanced reduction reactions are accomplished. This unique feature possesses VUV photoreactor as a promising method of treating water contaminated with both pharmaceutical and nitrate.

Simultaneous nitrate reduction and acetaminophen oxidation using the continuous-flow chemical-less VUV process as an integrated advanced oxidation and reduction process

International Nuclear Information System (INIS)

Moussavi, Gholamreza; Shekoohiyan, Sakine

2016-01-01

Highlights: • Simultaneous advanced oxidation and reduction processes were explored in VUV system. • Complete reduction of nitrate to N_2 was achieved at the presence of acetaminophen. • Complete degradation of acetaminophen was achieved at the presence of nitrate. • Over 95% of acetaminophen was mineralized in the VUV photoreactor. • VUV is a chemical-less advanced process for treating water emerging contaminants. - Abstract: This work was aimed at investigating the performance of the continuous-flow VUV photoreactor as a novel chemical-less advanced process for simultaneously oxidizing acetaminophen (ACT) as a model of pharmaceuticals and reducing nitrate in a single reactor. Solution pH was an important parameter affecting the performance of VUV; the highest ACT oxidation and nitrate reduction attained at solution pH between 6 and 8. The ACT was oxidized mainly by HO· while the aqueous electrons were the main working agents in the reduction of nitrate. The performance of VUV photoreactor improved with the increase of hydraulic retention time (HRT); the complete degradation of ACT and ∼99% reduction of nitrate with 100% N_2 selectivity achieved at HRT of 80 min. The VUV effluent concentrations of nitrite and ammonium at HRT of 80 min were below the drinking water standards. The real water sample contaminated with the ACT and nitrate was efficiently treated in the VUV photoreactor. Therefore, the VUV photoreactor is a chemical-less advanced process in which both advanced oxidation and advanced reduction reactions are accomplished. This unique feature possesses VUV photoreactor as a promising method of treating water contaminated with both pharmaceutical and nitrate.

Critical and subcritical parameters of the system simulating plutonium metal dissolution

International Nuclear Information System (INIS)

Vasilev, Yury Yu.; Ryazanov, Boris G.; Sviridov, Victor I.; Mozhayeva, Lubov I.

2003-01-01

Dissolution of plutonium metal was simulated using the Monte Carlo computer code to calculate criticality safety limits for the process. Calculations were made for the constant masses of plutonium charged to the dissolving vessel considering distribution of plutonium in metal and solution phases. Critical parameters and limits were calculated as a function of dissolving vessel volume and plutonium metal mass. 240 Pu content was assumed to be from 0% to 10% (mass). Critical parameters were evaluated for the system with a water reflector. Results of this paper may be used in the designing process equipment for plutonium metal dissolution. (author)

Tracing discharges of plutonium and technetium from nuclear processing plants by ultra-sensitive accelerator mass spectrometry

International Nuclear Information System (INIS)

Fifield, L.K.; Hausladen, P.A.; Cresswell, R.G.; Di Tada, M.L.; Day, J.P.; Carling, R.S.; Oughton, D.H.

1999-01-01

Historical discharges of plutonium from the Russian nuclear processing plant at Mayak in the Urals have been traced in sediments, soils and river water using ultra-sensitive detection of plutonium isotopes by accelerator mass spectrometry (AMS). Significant advantages of AMS over other techniques are its very high sensitivity. which is presently ∼10 6 atoms (1 μBq), and its ability to determine the 240 Pu/ 239 Pu ratio. The latter is a sensitive indicator of the source of the plutonium, being very low (1-2%) for weapons grade plutonium, and higher (∼ 20%) for plutonium from civil reactors or fallout from nuclear weapons testing. Since this ratio has changed significantly over the years of discharges from Mayak, a measurement can provide important information about the source of plutonium at a particular location. Similar measurements have been performed on samples from the Kara Sea which contains a graveyard of nuclear submarines from the former Soviet Union. AMS techniques have also been developed for detection of 99 Tc down to levels of a few femtograms. This isotope is one of the most prolific fission products and has a very long half-life of 220 ka. Hundreds of kg have been discharged from the nuclear reprocessing plant at Sellafield in the UK. While there may be public health issues associated with these discharges which can be addressed with AMS, these discharges may also constitute a valuable oceanographic tracer experiment in this climatically-important region of the world's oceans. Applications to date have included a human uptake study to assess long-term retention of 99 Tc in the body, and a survey of seaweeds from northern Europe to establish a baseline for a future oceanographic study

Measurements of plutonium in environmental samples

International Nuclear Information System (INIS)

D'Alberti, F.; Risposi, L.

1996-01-01

Within the activities connected with the start up of the PETRA Laboratory (Processo per l'Estrazione di Terre Rare ed Attinidi, i.e. process for extraction of rare earths and actinides), the Radiation Protection Unit of the J.R.C.-Ispra has carried out a well planned set of experimental measurements aimed at evaluating the zero point of the isotopes of plutonium in environmental samples by alfa spectrometry. After the International Moratorium in 1963, no release of plutonium has occurred in the environment apart from the burn up of SNAP 9A satellite in April 1964. Since then the plutonium concentration in air and in fallout samples has been continuously decreasing requiring, therefore, optimization of both instrumentation and experimental measurement procedures in order to obtain better sensibilities. In this work, the experimental methodology followed at the J.R.C.-Ispra for measurements of plutonium concentration in air, deposition and soil is described and the plutonium behaviour in these samples is reported and discussed starting from 1961

Measurements of plutonium in environmental samples

Energy Technology Data Exchange (ETDEWEB)

D' Alberti, F; Risposi, L [Instituto di Fisica Applicata, University of Milan, Milan (Italy)

1996-01-01

Within the activities connected with the start up of the PETRA Laboratory (Processo per l'Estrazione di Terre Rare ed Attinidi, i.e. process for extraction of rare earths and actinides), the Radiation Protection Unit of the J.R.C.-Ispra has carried out a well planned set of experimental measurements aimed at evaluating the zero point of the isotopes of plutonium in environmental samples by alfa spectrometry. After the International Moratorium in 1963, no release of plutonium has occurred in the environment apart from the burn up of SNAP 9A satellite in April 1964. Since then the plutonium concentration in air and in fallout samples has been continuously decreasing requiring, therefore, optimization of both instrumentation and experimental measurement procedures in order to obtain better sensibilities. In this work, the experimental methodology followed at the J.R.C.-Ispra for measurements of plutonium concentration in air, deposition and soil is described and the plutonium behaviour in these samples is reported and discussed starting from 1961.

In-line measurement of plutonium and americium in mixed solutions

International Nuclear Information System (INIS)

Li, T.K.

1981-01-01

A solution assay instrument (SAI) has been developed at the Los Alamos National Laboratory and installed in the plutonium purification and americium recovery process area in the Los Alamos Plutonium Processing Facility. The instrument is designed for accurate, timely, and simultaneous nondestructive analysis of plutonium and americium in process solutions that have a wide range of concentrations and Am/Pu ratios. For a 25-mL sample, the assay precision is 5 g/L within a 2000-s count time

Flexible process options for the immobilisation of residues and wastes containing plutonium

International Nuclear Information System (INIS)

Stewart, M.W.A.; Moricca, S.A.; Day, R. A.; Begg, B. D.; Scales, C. R.; Maddrell, E. R.; Eilbeck, A. B.

2007-01-01

Residues and waste streams containing plutonium present unique technical, safety, regulatory, security, and socio-political challenges. In the UK these streams range from lightly plutonium contaminated materials (PCM) through to residue s resulting directly from Pu processing operations. In addition there are potentially stocks of Pu oxide powders whose future designation may be either a waste or an asset, due to their levels of contamination making their reuse uneconomic, or to changes in nuclear policy. While waste management routes exist for PCM, an immobilisation process is required for streams containing higher levels of Pu. Such a process is being developed by Nexia Solutions and ANSTO to treat and immobilise Pu waste and residues currently stored on the Sellafield site. The characteristics of these Pu waste streams are highly variable. The physical form of the Pu waste ranges from liquids, sludges, powders/granules, to solid components (e.g., test fuels), with the Pu present as an ion in solution, as a salt, metal, oxide or other compound. The chemistry of the Pu waste streams also varies considerably with a variety of impurities present in many waste streams. Furthermore, with fissile isotopes present, criticality is an issue during operations and in the store or repository. Safeguards and security concerns must be assessed and controlled. The process under development, by using a combination of tailored waste form chemistry combined with flexible process technology aims to develop a process line to handle a broad range of Pu waste streams. It aims to be capable of dealing with not only current arisings but those anticipated to arise as a result of future operations or policy changes. (authors)

Managing plutonium in Britain. Current options

International Nuclear Information System (INIS)

1998-01-01

This is the report of a two day meeting to discuss issues arising from the reprocessing of plutonium and production of mixed oxide nuclear fuels in Britain. It was held at Charney Manor, near Oxford, on June 25 and 26, 1998, and was attended by 35 participants, including government officials, scientists, policy analysts, representatives of interested NGO's, journalists, a Member of Parliament, and visiting representatives from the US and Irish governments. The topic of managing plutonium has been a consistent thread within ORG's work, and was the subject of one of our previous reports, CDR 12. This particular seminar arose out of discussions earlier in the year between Dr. Frank Barnaby and the Rt. Hon. Michael Meacher MP, Minister for the Environment. With important decisions about the management of plutonium in Britain pending, ORG undertook to hold a seminar at which all aspects of the subject could be aired. A number of on-going events formed the background to this initiative. The first was British Nuclear Fuels' [BNFL] application to the Environment Agency to commission a mixed oxide fuel [MOX] plant at Sellafield. The second was BNFL's application to vary radioactive discharge limits at Sellafield. Thirdly, a House of Lords Select Committee was in process of taking evidence, on the disposal of radioactive waste. Fourthly, the Royal Society, in a recent report entitled Management of Separated Plutonium, recommended that 'the Government should commission a comprehensive review... of the options for the management of plutonium'. Four formal presentations were made to the meeting, on the subjects of Britain's plutonium policy, commercial prospects for plutonium use, problems of plutonium accountancy, and the danger of nuclear terrorism, by experts from outside the nuclear industry. It was hoped that the industry's viewpoint would also be heard, and BNFL were invited to present a paper, but declined on the grounds that they were 'currently involved in a formal

Preconceptual design for separation of plutonium and gallium by ion exchange

International Nuclear Information System (INIS)

DeMuth, S.F.

1997-01-01

The disposition of plutonium from decommissioned nuclear weapons, by incorporation into commercial UO 2 -based nuclear reactor fuel, is a viable means to reduce the potential for theft of excess plutonium. This fuel, which would be a combination of plutonium oxide and uranium oxide, is referred to as a mixed oxide (MOX). Following power generation in commercial reactors with this fuel, the remaining plutonium would become mixed with highly radioactive fission products in a spent fuel assembly. The radioactivity, complex chemical composition, and large size of this spent fuel assembly, would make theft difficult with elaborate chemical processing required for plutonium recovery. In fabricating the MOX fuel, it is important to maintain current commercial fuel purity specifications. While impurities from the weapons plutonium may or may not have a detrimental affect on the fuel fabrication or fuel/cladding performance, certifying the effect as insignificant could be more costly than purification. Two primary concerns have been raised with regard to the gallium impurity: (1) gallium vaporization during fuel sintering may adversely affect the MOX fuel fabrication process, and (2) gallium vaporization during reactor operation may adversely affect the fuel cladding performance. Consequently, processes for the separation of plutonium from gallium are currently being developed and/or designed. In particular, two separation processes are being considered: (1) a developmental, potentially lower cost and lower waste, thermal vaporization process following PuO 2 powder preparation, and (2) an off-the-shelf, potentially higher cost and higher waste, aqueous-based ion exchange (IX) process. While it is planned to use the thermal vaporization process should its development prove successful, IX has been recommended as a backup process. This report presents a preconceptual design with material balances for separation of plutonium from gallium by IX

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

Plutonium oxide dissolution

International Nuclear Information System (INIS)

Gray, J.H.

1992-01-01

Several processing options for dissolving plutonium oxide (PuO 2 ) from high-fired materials have been studied. The scoping studies performed on these options were focused on PuO 2 typically generated by burning plutonium metal and PuO 2 produced during incineration of alpha contaminated waste. At least two processing options remain applicable for dissolving high-fired PuO 2 in canyon dissolvers. The options involve solid solution formation of PuO 2 With uranium oxide (UO 2 ) and alloying incinerator ash with aluminum. An oxidative dissolution process involving nitric acid solutions containing a strong oxidizing agent, such as cerium (IV), was neither proven nor rejected. This uncertainty was due to difficulty in regenerating cerium (IV) ions during dissolution. However, recent work on silver-catalyzed dissolution of PuO 2 with persulfate has demonstrated that persulfate ions regenerate silver (II). Use of persulfate to regenerate cerium (IV) or bismuth (V) ions during dissolution of PuO 2 materials may warrant further study

Erosional losses of fallout plutonium

International Nuclear Information System (INIS)

Foster, G.R.; Hakonson, T.E.

1987-01-01

Plutonium from fallout after atmospheric explosion of nuclear weapons in the 1950's and 1960s is being redistributed over the landscape by soil erosion and carried on sediment by streams to oceans. Erosion rates computed with the Universal Soil Loss Equation for more than 200,000 sample points on nonfederal land across the US were used to estimate plutonium removal rates by soil erosion. On the average, only about 4% of the eroded sediment reaches the outlet of a major river. The remaining sediment is deposited en route, and because deposition is a selective process, the sediment is enriched in fine particles having the highest concentration of plutonium because of the element's strong association with clay and silt-sized sediment. Estimated enrichment ratios, sediment delivery ratios, and erosion rates were used to estimate annual delivery of fallout plutonium. These estimates ranged from 0.002% of the initial fallout plutonium inventory for the Savannah River basin to 0.01% for the Columbia River basin, to 0.02% for the Hudson and Rio Grande River basins, to 0.08% for the Mississippi River basin. If the deposition of plutonium had been uniformly 1 mCi/km 2 , the estimated plutonium activity on suspended sediment would range from about 7 fCi/g of sediment of the Savannah River basin, to 9 fCi/g for the Mississippi River basin, to 12 fCi/g for the Hudson River basin, to 14 fCi/g for the Columbia and Rio Grande River basins. 45 references, 2 figures, 17 tables

Method of processing nitrate-containing radioactive liquid wastes

International Nuclear Information System (INIS)

Ogawa, Norito; Nagase, Kiyoharu; Otsuka, Katsuyuki; Ouchi, Jin.

1983-01-01

Purpose: To efficiently concentrate nitrate-containing low level radioactive liquid wastes by electrolytically dialyzing radioactive liquid wastes to decompose the nitrate salt by using an electrolytic cell comprising three chambers having ion exchange membranes and anodes made of special materials. Method: Nitrate-containing low level radioactive liquid wastes are supplied to and electrolytically dialyzed in a central chamber of an electrolytic cell comprising three chambers having cationic exchange membranes and anionic exchange membranes made of flouro-polymer as partition membranes, whereby the nitrate is decomposed to form nitric acid in the anode chamber and alkali hydroxide compound or ammonium hydroxide in the cathode chamber, as well as concentrate the radioactive substance in the central chamber. Coated metals of at least one type of platinum metal is used as the anode for the electrolytic cell. This enables efficient industrial concentration of nitrate-containing low level radioactive liquid wastes. (Yoshihara, H.)

Field test of plutonium and thorium contaminated clay soils from the Mound Site using the ACT*DE*CON Process

International Nuclear Information System (INIS)

Johnson, J.O.; Swift, N.A.; Church, R.H.; Neff, R.A.

1998-01-01

A treatability test was run during the summer and fall of 1997 to demonstrate the effectiveness of ACT*DE*CON for removing plutonium and thorium from the clay soils around Mound. ACT*DE*CON is a proprietary solution patented by Selentec. The process utilized a highly selective dissolution of the contaminants by the use of a chemical wash. The pilot scale process involved pretreatment of the soil in an attrition scrubber with ACT*DE*CON solution. This blended solution was then passed through a counter-current extraction chamber where additional contact with ACT*DE*CON solution occurred, followed by a rinse cycle. During this process sand was added to aid contact of the solution with the soil particles. The sand is removed during the rinse step and reused. The chelating agent is separated from the contaminant and recycled back into the process, along with the reverse osmosis permeate. The resulting solution can be further treated to concentrate the contaminant. Three different types of environmental soils were tested -- plutonium and thorium contaminated soils with the natural clay content, and plutonium contaminated soils with a high percentage of fine clay particles. The goal of these tests was to reduce the plutonium levels from several hundreds of pCi/g to between 25 and 75 pCi/g and the thorium from a couple hundred pCi/g to less than 5 pCi/g. The results of these four tests are presented along with a discussion of the operating parameters and the lessons learned relating to full scale implementation at Mound as well as other potential applications of this process

Los Alamos DP West Plutonium Facility decontamination project

International Nuclear Information System (INIS)

Garde, R.; Cox, E.J.; Valentine, A.M.

1982-01-01

The DP West Plutonium Facility operated by the Los Alamos National Laboratory, Los Alamos, New Mexico, was decontaminated between April 1978 and April 1981. The facility was constructed in 1944 to 1945 to produce plutonium metal and fabricate parts for nuclear weapons. It was continually used as a plutonium processing and research facility until mid-1978. Decontamination operations included dismantling and removing gloveboxes and conveyor tunnels; removing process systems, utilities, and exhaust ducts; and decontaminating all remaining surfaces. This report describes glovebox and conveyor tunnel separations, decontamination techniques, health and safety considerations, waste management procedures, and costs of the operation

Nondestructive analysis of plutonium contaminated soil

International Nuclear Information System (INIS)

Smith, H.E.; Taylor, L.H.

1977-01-01

Plutonium contaminated soil is currently being removed from a covered liquid waste disposal trench near the Pu Processing facility on the Hanford Project. This soil with the plutonium is being mined using remote techniques and equipment. The mined soil is being packaged for placement into retrievable storage, pending possible recovery. To meet the requirements of criticality safety and materials accountability, a nondestructive analysis program has been developed to determine the quantity of plutonium in each packing-storage container. This paper describes the total measurement program: equipment systems, calibration techniques, matrix assumption, instrument control program and a review of laboratory operating experience

Plutonium

International Nuclear Information System (INIS)

Anon.

1995-01-01

Plutonium, which was obtained and identified for the first time in 1941 by chemist Glenn Seaborg - through neutron irradiation of uranium 238 - is closely related to the history of nuclear energy. From the very beginning, because of the high radiotoxicity of plutonium, a tremendous amount of research work has been devoted to the study of the biological effects and the consequences on the environment. It can be said that plutonium is presently one of the elements, whose nuclear and physico-chemical characteristics are the best known. The first part of this issue is a survey of the knowledge acquired on the subject, which emphasizes the sanitary effects and transfer into the environment. Then the properties of plutonium related to energy generation are dealt with. Fissionable, like uranium 235, plutonium has proved a high-performance nuclear fuel. Originally used in breeder reactors, it is now being more and more widely recycled in light water reactors, in MOX fuel. Reprocessing, recycling and manufacturing of these new types of fuel, bound of become more and more widespread, are now part of a self-consistent series of operations, whose technical, economical, industrial and strategical aspects are reviewed. (author)

A phenomenological model for improving understanding of the ammonium nitrate agglomeration process

Directory of Open Access Journals (Sweden)

Videla Leiva Alvaro

2016-01-01

Full Text Available Ammonium nitrate is intensively used as explosive in the mining industry as the main component of ANFO. The ammonium nitrate is known to be a strong hygroscopic crystal matter which generates problems due to the creation of water bridges between crystals leading later to nucleation and crystallization forming an agglomerated solid cake. The agglomeration process damages the ammonium nitrate performance and is undesirable. Usually either organic or inorganic coatings are used to control agglomeration. In the present work a characterization method of humidity adsorption of the ammonium nitrate crystal was performed under laboratory conditions. Several samples were exposed into a defined humidity in a controlled chamber during 5 hours after which the samples were tested to measure agglomeration as the resistance force to compression. A clear relation was found between coating protection level, humidity and agglomeration. Agglomeration can be then predicted by a phenomenological model based of combination of the mono-layer BET adsorption and CNT nucleation models.

Learning more about plutonium

International Nuclear Information System (INIS)

2005-01-01

This document offers chemical, metallurgical and economical information on the plutonium, a hard white radioelement. It deals also on the plutonium formation in the earth, the plutonium use in the nuclear industry, the plutonium in the environment and the plutonium toxicity. (A.L.B.)

Engineering Options Assessment Report. Nitrate Salt Waste Stream Processing

Energy Technology Data Exchange (ETDEWEB)

Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-13

This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 above-ground UNS, and 79 candidate below-ground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing

Energy Technology Data Exchange (ETDEWEB)

Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-18

This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 aboveground UNS, and 79 candidate belowground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

Using solvent extraction to process nitrate anion exchange column effluents

International Nuclear Information System (INIS)

Yarbro, S.L.

1987-10-01

Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), a new organophosphorous extractant, and a new centrifugal mixer-settler both recently developed at Argonne were evaluated for their potential use in the recovery of actinides from nitrate anion exchange column effluents. The performance of the extractant was evaluated by measuring the extraction coefficient values as a function of acid and salt concentration. Additional performance parameters include extraction coefficient behavior as a function of the total metal concentration in the organic phase, and comparison of different stripping and organic scrubbing techniques. A simulated effluent stream was used to evaluate the performance of the centrifugal mixer-settlers by comparing experimental and calculated interstage concentration profiles. Both the CMPO extractant and the centrifugal mixer-settlers have potential for processing nitrate column effluents, particularly if the stripping behavior can be improved. Details of the proposed process are presented in the flowsheet and contactor design analyses

Using solvent extraction to process nitrate anion exchange column effluents

Energy Technology Data Exchange (ETDEWEB)

Yarbro, S.L.

1987-10-01

Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), a new organophosphorous extractant, and a new centrifugal mixer-settler both recently developed at Argonne were evaluated for their potential use in the recovery of actinides from nitrate anion exchange column effluents. The performance of the extractant was evaluated by measuring the extraction coefficient values as a function of acid and salt concentration. Additional performance parameters include extraction coefficient behavior as a function of the total metal concentration in the organic phase, and comparison of different stripping and organic scrubbing techniques. A simulated effluent stream was used to evaluate the performance of the centrifugal mixer-settlers by comparing experimental and calculated interstage concentration profiles. Both the CMPO extractant and the centrifugal mixer-settlers have potential for processing nitrate column effluents, particularly if the stripping behavior can be improved. Details of the proposed process are presented in the flowsheet and contactor design analyses.

Plutonium-241 processing: from impure oxide to high purity metal target disks

International Nuclear Information System (INIS)

Conner, W.V.; Baaso, D.L.

1975-01-01

The preparation of three plutonium-241 metal target disks, using a precision casting technique, is described. The disks were 0.625 inch in diameter and 0.125, 0.025, and 0.010 inch thick. All three disks were prepared simultaneously in a single casting. The variation in thickness of each disk was within +-1 percent of the disk's average thickness. The plutonium-241 was highly pure, and the finished disks contained a total of only 297 parts per million of detectable impurities. Purification of the plutonium oxide ( 241 PuO 2 ) and the conversion of the purified 241 PuO 2 to metal are also described. (U.S.)

A review of plutonium oxalate decomposition reactions and effects of decomposition temperature on the surface area of the plutonium dioxide product

Energy Technology Data Exchange (ETDEWEB)

Orr, R.M.; Sims, H.E.; Taylor, R.J., E-mail: robin.j.taylor@nnl.co.uk

2015-10-15

Plutonium (IV) and (III) ions in nitric acid solution readily form insoluble precipitates with oxalic acid. The plutonium oxalates are then easily thermally decomposed to form plutonium dioxide powder. This simple process forms the basis of current industrial conversion or ‘finishing’ processes that are used in commercial scale reprocessing plants. It is also widely used in analytical or laboratory scale operations and for waste residues treatment. However, the mechanisms of the thermal decompositions in both air and inert atmospheres have been the subject of various studies over several decades. The nature of intermediate phases is of fundamental interest whilst understanding the evolution of gases at different temperatures is relevant to process control. The thermal decomposition is also used to control a number of powder properties of the PuO{sub 2} product that are important to either long term storage or mixed oxide fuel manufacturing. These properties are the surface area, residual carbon impurities and adsorbed volatile species whereas the morphology and particle size distribution are functions of the precipitation process. Available data and experience regarding the thermal and radiation-induced decompositions of plutonium oxalate to oxide are reviewed. The mechanisms of the thermal decompositions are considered with a particular focus on the likely redox chemistry involved. Also, whilst it is well known that the surface area is dependent on calcination temperature, there is a wide variation in the published data and so new correlations have been derived. Better understanding of plutonium (III) and (IV) oxalate decompositions will assist the development of more proliferation resistant actinide co-conversion processes that are needed for advanced reprocessing in future closed nuclear fuel cycles. - Highlights: • Critical review of plutonium oxalate decomposition reactions. • New analysis of relationship between SSA and calcination temperature. �

Processing plutonium-contaminated soild for volume reduction using the segmented gate system

International Nuclear Information System (INIS)

Moroney, K.S.; Moroney, J.D.; Turney, J.M.; Doane, R.W.

1994-01-01

TMA/Eberline has developed and demonstrated an effective method for removing mixed plutonium and americium contamination from a coral soil matrix at the Defense Nuclear Agency's Johnston Atoll site. TMA's onsite soil processing for volume reduction is ongoing at a rate of over 2000 metric tons per week. The system uses arrays of sensitive radiation detectors coupled with sophisticated computer software developed by Eberline Instrument Corporation. The proprietary software controls four soil sorting units operating in parallel that utilize TMA's unique Segmented Gate System technology to remove radiologically contaminated soil from a moving supply on conveyor belts. Clean soil is released for use elsewhere on the island. Contaminated soil is diverted to either a metal drum for collecting higher activity open-quotes hotclose quotes particles (>5000 Becquerels), or to a supplementary soil washing process designed to remove finely divided particles of dispersed low level contamination. Site contamination limits specify maximum dispersed radioactivity of no more than 500 Becquerels per kilogram of soil averaged over no more than 0.1 cubic meter. Results of soil processing at this site have been excellent. After processing over 50,000 metric tons, the volume of contaminated material that would have required expensive special handling, packaging, and disposal as radioactive waste has been successfully reduced by over 98 percent. By mid-January 1994, nearly three million kiloBecquerels of plutonium/americium contamination had been physically separated from the contaminated feed by TMA's Segmented Gate System, and quality control sampling showed no radioactivity above release criteria in the open-quotes cleanclose quotes soil pile

SAFETY STUDIES TO MEASURE EXOTHERMIC REACTIONS OF SPENT PLUTONIUM DECONTAMINATION CHEMICALS USING WET AND DRY DECONTAMINATION METHODS

International Nuclear Information System (INIS)

HOPKINS, A.M.; JACKSON, G.W.; MINETTE, M.; EWALT, J.; COOPER, T.; SCOTT, P.; JONES, S.; SCHEELEY, R.

2005-01-01

The Plutonium Finishing Plant (PFP) at the Hanford site in Eastern Washington is currently being decommissioned by Fluor Hanford. Chemicals being considered for dccontamination of gloveboxes in PFP include cerium (IV) nitrate in a nitric acid solution, and proprietary commercial solutions that include acids and sequestering agents. Aggressive chemicals are commonly used to remove transuranic contaminants from process equipment to allow disposal of the equipment as low level waste. Fluor's decontamination procedure involves application of chemical solutions as a spray on the contaminated surfaces, followed by a wipe-down with rags. Alternatively, a process of applying oxidizing Ce IV ions contained in a gel matrix and vacuuming a dry gel material is being evaluated. These processes effectively transfer the transuranic materials to rags or a gel matrix which is then packaged as TRU waste and disposed

Appraisal of BWR plutonium burners for energy centers

International Nuclear Information System (INIS)

Williamson, H.E.

1976-01-01

The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

Civil plutonium management

International Nuclear Information System (INIS)

Sicard, B.; Zaetta, A.

2004-01-01

During 1960 and 1970 the researches on the plutonium recycling in fast neutrons reactors were stimulated by the fear of uranium reserves diminishing. At the beginning of 1980, the plutonium mono-recycling for water cooled reactors is implementing. After 1990 the public opinion concerning the radioactive wastes management and the consequences of the disarmament agreements between Russia and United States, modified the context. This paper presents the today situation and technology associated to the different options and strategical solutions of the plutonium management: the plutonium use in the world, the neutronic characteristics, the plutonium effect on the reactors characteristics, the MOX behavior in the reactors, the MOX fabrication and treatment, the possible improvements to the plutonium use, the concepts performance in a nuclear park. (A.L.B.)

Acid-digestion treatment of plutonium-containing waste

International Nuclear Information System (INIS)

Wieczorek, H.; Kemmler, G.; Krause, H.

1981-01-01

The Radioactive Acid-Digestion Test Unit (RADTU) has been constructed at Hanford to demonstrate the application of the acid-digestion process for treating combustible transuranic wastes and scrap materials. The RADTU, with its original tray digestion vessel, has recently completed a six-month campaign processing potentially contaminated non-glovebox wastes from a Hanford plutonium facility. During this campaign, it processed 2100 kg largely cellulosic wastes at an average sustained processing rate of 3 kg/h as limited by the acid-waste contact and the water boil-off rate from the acid feeds. The on-line operating efficiency was nearly 50% on a twelve-hour day, five-day week basis. Following this campaign, a new annular high-rate digester has been installed for testing. In preliminary tests with simulated wastes, the new digester demonstrated a sustained capacity of 10 kg/h with greatly improved intimacy of contact between the digestion acid and the waste. The new design also doubles the heat-transfer surface, which is expected to provide at least twice the water boil-off rate of the previous tray digester design. Following shakedown testing with simulated and low-level wastes, the new unit will be used to process combustible plutonium scrap and waste from Hanford plutonium facilities for the purposes of volume reduction, plutonium recovery, and stabilization of the final waste form. (author)

Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

International Nuclear Information System (INIS)

Hanus, D.; Kleykamp, H.

1982-01-01

Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

Energy Technology Data Exchange (ETDEWEB)

Kyser, E.; King, W.

2012-04-25

Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after {approx}4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

Energy Technology Data Exchange (ETDEWEB)

Kyser, E. A.; King, W. D.

2012-07-31

Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after ~4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

Nitrates of rare earths

International Nuclear Information System (INIS)

Komissarova, L.N.; Pushkina, L.Ya.

1984-01-01

The systematization of experimental data with account of the last achievements in the field of studying the RE nitrate properties is realized. The methods of production, solubility in aqueous solutions structure, thermodynamic characteristics and thermal stability of nitrate hydrates, RE anhydrous and basic nitrates are considered. The data on RE nirtrate complexing in aqueous solutions are given. Binary nitrates, nitrate solvates and RE nitrate adducts with organic compounds are described. The use of RE nitrates in the course of RE production, in the processes of separation and fine cleaning of RE preparations is considered

Design of an integrated non-destructive plutonium assay facility

International Nuclear Information System (INIS)

Moore, C.B.

1984-01-01

The Department of Energy requires improved technology for nuclear materials accounting as an essential part of new plutonium processing facilities. New facilities are being constructed at the Savannah River Plant by the Du Pont Company, Operating Contractor, to recover plutonium from scrap and waste material generated at SRP and other DOE contract processing facilities. This paper covers design concepts and planning required to incorporate state-of-the-art plutonium assay instruments developed at several national laboratories into an integrated, at-line nuclear material accounting facility operating in the production area. 3 figures

Photochemical technique for reduction of uranium and subsequently plutonium in the Purex process

International Nuclear Information System (INIS)

Goldstein, M.; Barker, J.J.; Gangwer, T.

1976-09-01

A photochemical modification of the Purex process is described in which a purified side stream of UO 2 ++ ion is reduced to U +4 outside the radioactive area of the reprocessing plant. The U +4 is then cycled back to step 2 of the Purex process to reduce the plutonium and effect separation within the partitioning column. This process is shown to be very energy efficient and compatible with existing conventional lamp technology. Preliminary cost estimates of the energy requirements for photon production are essentially negligible. Conceptual systems and photochemical reactor designs are presented. Potential benefits of this system are discussed

Natural hazards that may trigger a radiological release from a plutonium processing facility

Energy Technology Data Exchange (ETDEWEB)

Selvidge, J. E.

1977-04-28

Calculations show the probability of a tornado striking a plutonium area at Rocky Flats is 2.2 x 10/sup -4/ per year. The source term (expected value of plutonium release) should such an event occur is calculated at 3.3 x 10/sup -7/ grams. The source term for high-velocity, downslope winds is higher--2.2 x 10/sup -3/ grams. The probability of a meteorite that weighs one or more pounds (453 grams) striking a plutonium area is estimated at 8.88 x 10/sup -7/ per year. Because of this small probability and the remote chance that a plutonium release would occur even if a meteorite hit occurred, the hazard from meteorite impact is considered negligible. Conservative assumptions result in all calculated frequencies being almost certainly too high. Empirical observations have indicated lower frequencies than those calculated.

Natural hazards that may trigger a radiological release from a plutonium processing facility

International Nuclear Information System (INIS)

Selvidge, J.E.

1977-01-01

Calculations show the probability of a tornado striking a plutonium area at Rocky Flats is 2.2 x 10 -4 per year. The source term (expected value of plutonium release) should such an event occur is calculated at 3.3 x 10 -7 grams. The source term for high-velocity, downslope winds is higher--2.2 x 10 -3 grams. The probability of a meteorite that weighs one or more pounds (453 grams) striking a plutonium area is estimated at 8.88 x 10 -7 per year. Because of this small probability and the remote chance that a plutonium release would occur even if a meteorite hit occurred, the hazard from meteorite impact is considered negligible. Conservative assumptions result in all calculated frequencies being almost certainly too high. Empirical observations have indicated lower frequencies than those calculated

Characteristics of airborne plutonium resuspended from near-background aged surface-sources

International Nuclear Information System (INIS)

Sehmel, G.A.

1982-11-01

Plutonium content in samples of airborne solids collected at five Hanford sites was determined in several experiments directed toward investigating resuspension processes for aged surface sources. Though airborne plutonium concentrations are extremely low, radiochemical technique sensitivities allow plutonium characterization to be considered as a function of host-particle diameter in samples of airborne solids. Plutonium concentrations and activity densities are a function of aerodynamic particle diameter, sampling height, wind speed, wind direction and plutonium isotopic ratios

The testing of a method for dosing plutonium by {alpha}-counting in the presence of strong concentrations of salts or of uranium; Essai d'une methode de dosage du plutonium par comptage {alpha} en presence de fortes concentrations en sels ou en uranium

Energy Technology Data Exchange (ETDEWEB)

Fontaine, A M; Baude-Malafosse, L M; Cunq, M J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

This report describes a method for dosing small quantities of plutonium in a solution having a high concentration of salts. It shows the possibility of dosing up to 5.10{sup -3} {mu}g of Pu in the presence of 10 mg of NaNO{sub 3} with out decreasing the counting-rate. The only error possible is that in the counting. It is also possible to dose 10{sup -3} {mu}g of Pu in the presence of 1,7 mg of uranyl nitrate. (author) [French] Ce rapport decrit une methode de dosage de faibles quantites de plutonium dans une solution de forte concentration en sels. Il montre la possibilite de doser jusqu'a 5.10{sup -3} {mu}g de Pu en presence de 10 mg de NO{sub 3}Na sans diminution du taux de comptage. La seule erreur que l'on puisse faire est l'erreur de comptage. On peut aussi doser 10{sup -3} {mu}g de Pu en presence de 1,7 mg de nitrate d'uranyle. (auteur)

Installation places of criticality accident detectors in the plutonium conversion development facility

International Nuclear Information System (INIS)

Sanada, Yukihisa; Tsujimura, Norio; Shimizu, Yoshio; Izaki, Kenji; Furuta, Sadaaki

2008-01-01

At the Plutonium Conversion Development Facility (PCDF) in the Nuclear Fuel Cycle Engineering Laboratories, the co-conversion technologies to purify the mixed plutonium and uranium nitrate solution discharged from a reprocessing plant have been developed. The probability of a criticality accident in PCDF is extremely low. However, the criticality accident alarm system (CAAS) has been in place since 1982 to reduce the radiation dose to workers in case of such a rare criticality accident. The CAAS contains criticality accident detector units (CADs), one unit consisting of three plastic scintillation detectors, and using the 2 out of 3 voting system for the purpose of high reliability. Currently, eight CADs are installed in PCDF evaluating the dose using a simple equation allowing for a safety margin. The purpose of this study is to show the determination procedures for the adequate relocation of the CADs which adequately ensures safety in PCDF. (author)

A portable concentrator for processing plutonium

International Nuclear Information System (INIS)

Chamberlain, D.B.; Conner, C.; Chen, L.

1995-01-01

A horizontal, agitated film concentrator designed to concentrate liquid streams to a high solid content slurry is briefly described. The Rototherm unit is being studied for use at US Department of Energy facilities to handle large quantities of aqueous plutonium solutions. Capabilities for evaporating more than 98% of the water present in a single pass have been demonstrated. Decontamination factors of 10 6 to 10 7 are expected. The unit may also be useful for recycling aqueous waste treatment reagents from the decontamination of gaseous diffusion plants