Characterization and chemistry of fission products released from LWR fuel under accident conditions

International Nuclear Information System (INIS)

Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

1984-01-01

Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

Experience of iodine, caesium and noble gas release from AGR failures

International Nuclear Information System (INIS)

Chapman, C.J.; Harris, A.M.; Phillips, M.E.

1985-01-01

In the event of a fuel failure in an Advanced Gas Cooled Reactor (AGR), the quantity of fission products available for release to the environment is determined by the transport of fission products in the UO 2 fuel, by the possible retention of fission products in the fuel can interspace and by the deposition of fission products on gas circuit surfaces ('plate-out'). The fission products of principal radiological concern are radioactive caesium (Cs-137 and Cs-134) and iodine (principally I-131). Results are summarised of a number of experiments which were designed to study the release of these fission products from individual fuel failures in the prototype AGR at Windscale. Results are also presented of fission product release from failures in commercial AGRs. Comparisons of measured releases of caesium and iodine relative to the release of the noble gas fission products show that, for some fuel failures, there is a significant retention of caesium and iodine within the fuel can interspace. Under normal conditions circuit deposition reduces caesium and iodine gas concentrations by several orders of magnitude. Differing release behaviour of caesium and iodine from the failures is examined together with subsequent deposition within the sampling equipment. These observations are important factors which must be considered in developing an understanding of the mechanisms involved in circuit deposition. (author)

Release of fission products from irradiated aluminide fuel at high temperature

International Nuclear Information System (INIS)

Shibata, Toshikazu; Kanda, Keiji; Mishima, Kaichiro; Tamai, Tadaharu; Hayashi, Masatoshi; Snelgrove, James L.; Stahl, David; Matos, James E.; Travelli, Armando; Case, F. Neil; Posey, John C.

1983-01-01

Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Al x . The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

Release of fission products from irradiated aluminide fuel at high temperature

International Nuclear Information System (INIS)

Shibata, T.; Kanda, K.; Mishima, K.

1982-01-01

Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500 0 C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1 0 C on the plates. The next release at 585 0 C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 0 C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed

Development of Silver-exchanged Adsorbents for the Removal of Fission Iodine from Alkaline Dissolution

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Woon; Lee, Seung Kon; Lee, Su Seung; Lee, Jun Sig [KAERI, Daejeon (Korea, Republic of); Kim, Sang Wook [Dongguk University, Gyeongju (Korea, Republic of)

2016-05-15

{sup 99} Mo is extracted from the filtrate solution through column-based multistep separation and purification process. In the process, removal of radio-impurities from the solution is essential to acquire high-quality fission {sup 99} Mo. Iodine is the main impurity having about 15% of total radioactivity among the whole fission products. Most of the iodine exists in the caustic dissolution as iodide form. In this study, silver-exchanged adsorbent is used to adsorb iodide from the solution. Adsorbed iodide can be recovered and recycled for radiopharmaceuticals. Compound is dried again. After heating ascorbic acid solution, solution is added to dried compound. Heat the mixture. After removing supernatant, the mixture is washed with hot distilled water and then cool distilled water in the order named. Finally, the mixture is heated and then recovering by using the sieve. In this study, silver-exchanged adsorbent is used to adsorb iodide from the solution. Adsorbed iodide can be recovered and recycled for radiopharmaceuticals. Silver-doped DAW-70 alumina by using silver mirror reaction is less impurities and simpler than method using ascorbic acid.

Immobilization of iodine in concrete

Science.gov (United States)

Clark, Walter E.; Thompson, Clarence T.

1977-04-12

A method for immobilizing fission product radioactive iodine recovered from irradiated nuclear fuel comprises combining material comprising water, Portland cement and about 3-20 wt. % iodine as Ba(IO.sub.3).sub.2 to provide a fluid mixture and allowing the fluid mixture to harden, said Ba(IO.sub.3).sub.2 comprising said radioactive iodine. An article for solid waste disposal comprises concrete prepared by this method. BACKGROUND OF THE INVENTION This invention was made in the course of, or under a contract with the Energy Research and Development Administration. It relates in general to reactor waste solidification and more specifically to the immobilization of fission product radioactive iodine recovered from irradiated nuclear fuel for underground storage.

Iodine volatility

International Nuclear Information System (INIS)

Beahm, E.C.; Shockley, W.E.

1984-01-01

The ultimate aim of this program is to couple experimental aqueous iodine volatilities to a fission product release model. Iodine partition coefficients, for inorganic iodine, have been measured during hydrolysis and radiolysis. The hydrolysis experiments have illustrated the importance of reaction time on iodine volatility. However, radiolysis effects can override hydrolysis in determining iodine volatility. In addition, silver metal in radiolysis samples can react to form silver iodide accompanied by a decrease in iodine volatility. Experimental data are now being coupled to an iodine transport and release model that was developed in the Federal Republic of Germany

Release of fission products in transients

International Nuclear Information System (INIS)

Christensen, H.; Lundqwist, R.

1979-07-01

A station for automatic sampling of coolant has been put in operation at the Oskarshamn-1 reactor. The release of 131 J and other fission products in spikes in connection with reactor trips and scheduled shutdowns has been measured. A model developed at General Electric has been used to predict the spike release in Oskarshamn-1 and the predicted values have been compared with experimental values. Literature data of iodine spikes in BWR and PWR have been reviewed. (author)

Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO2 burnup

International Nuclear Information System (INIS)

Appelhans, A.D.; Turnbull, J.A.; White, R.J.

1983-01-01

The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500 0 C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO 2 of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines

TMI-2 [Three Mile Island] fission product inventory program: FY-85 status report

International Nuclear Information System (INIS)

Langer, S.; Croney, S.T.; Akers, D.W.; Russell, M.L.

1986-11-01

This report presents the status of the TMI-2 fission product inventory program through May 1985. The fission product inventory program is an assessment of the location of fission products distributed in the plant as a result of the TMI-2 accident. Included in this report are principal results of samples from the reactor building where most of the mobile fission products (i.e., radiocesium and iodine) are expected to be found. The data are now complete enough for most reactor components; therefore, it is possible to direct the balance of the examination and sampling program to areas and components where it is likely to be most productive. Those areas are the reactor core and the reactor building basement, with emphasis on the currently unsampled portions of the core

Analytical measurements of fission products during a severe nuclear accident

Science.gov (United States)

Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

2018-01-01

The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

Analytical measurements of fission products during a severe nuclear accident

Directory of Open Access Journals (Sweden)

Doizi D.

2018-01-01

Full Text Available The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

Continuous fission-product monitor system at Oyster Creek. Final report

International Nuclear Information System (INIS)

Collins, L.L.; Chulick, E.T.

1980-10-01

A continuous on-line fission product monitor has been installed at the Oyster Creek Nuclear Generating Station, Forked River, New Jersey. The on-line monitor is a minicomputer-controlled high-resolution gamma-ray spectrometer system. An intrinsic Ge detector scans a collimated sample line of coolant from one of the plant's recirculation loops. The minicomputer is a Nuclear Data 6620 system. Data were accumulated for the period from April 1979 through January 1980, the end of cycle 8 for the Oyster Creek plant. Accumulated spectra, an average of three a day, were stored on magnetic disk and subsequently analyzed for fisson products, Because of difficulties in measuring absolute detector efficiency, quantitative fission product concentrations in the coolant could not be determined. Data for iodine fission products are reported as a function of time. The data indicate the existence of fuel defects in the Oyster Creek core during cycle 8

Development of assessment technology for hydrogen burn and fission product behavior in containment

International Nuclear Information System (INIS)

Kim, S. B.; Kim, J. T.; Ha, K. S.; Hong, S. W.; Song, Y. M.; Park, J. H.; Cho, Y. R.; Kang, H. S.

2012-04-01

Analysis tools for hydrogen burn was established to resolve the hydrogen issues in containment. To validate CFX commercial CFD(computational fluid dynamics) code, the hydrogen combustion experiments such as FLAME and ENACEFF for reactor containment were analyzed. And OpenFOAM hydrogen combustion code was developed and validated. Experiments for the flame propagation characteristics in IRWST and the run-up-distance for DDT(Deflagration to detonation transition) were performed and analytical model was evaluated to evaluation of the performance of hydrogen mitigation system, that is, PAR(Passive auto-catalistic re-combiner) To improvement of the fission product modelling in containment, separate analysis module for Iodine behavior and its application tool of K-IODIP (Korea IODIne Package) were developed. PHEBUS FPT-3 analysis was performed to validate MELCOR code. And also the characteristics of fission product behaviors in Future Reactors(GEN-IV) were compared

Fission product retention during faults involving steam generator tube rupture

International Nuclear Information System (INIS)

Rodliffe, R.S.

1983-08-01

In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

Studies of short-lived products of spallation fission reactions at TRIUMF

CERN Document Server

Bischoff, G; D'Auria, J M; Dautet, H; Lee, J K P; Pate, B D; Wiesehahn, W

1976-01-01

The gas-jet recoil transport technique has been used to transport products from spallation and fission reactions from a target chamber to a shielded location for nuclear spectroscopic studies. These involve X- beta - gamma coincidence measurements and (shortly) time- of-flight mass spectroscopy. It has been deduced that the proton beam at present intensities has no appreciable effect on the ability of ethylene and other cluster-producing gases to transport radioactivity. Preliminary results will be presented for shortlived fission products from uranium, and for spallation products of iodine and argon. The latter were obtained from the bombardment of gas and aerosol targets mixed with the transporting gas in the target chamber, which appears to be a generally useful technique.

The release of fission products from uranium metal: a review

International Nuclear Information System (INIS)

Minshall, P.C.

1989-03-01

The literature on the release of fission products as gaseous species from irradiated uranium metal in oxidising atmospheres has been reviewed. Release of actinides and of fission products as spalled particulate were not considered. Data is given on the release in air, carbon dioxide, steam and mixtures of steam and air. The majority of data discussed lie between 800 and 1200 0 C though some results for xenon, krypton and iodine releases below 800 0 C are given. Two measures of fission product release are discussed: the release fraction, F(tot), which is the ratio of the total release to the initial inventory, and the fractional release, F(ox), which is the fraction released from the oxidised metal. The effect of burn-up, atmosphere and temperature on F(tot) and F(ox) is examined and the conditions under which the release fraction, F(tot) is proportional to the extent of oxidation discussed. (author)

Cumulative fission yield of Ce-148 produced by thermal-neutron fission of U-235

International Nuclear Information System (INIS)

Hasan, A.A.

1984-12-01

Cumulative fission yield of 148 cesium isotopes and some other fission products produced by thermal-neutron fission of 235 uranium is determined by Germanium/Lithium spectroscopic methods. The measuremets were done at Tsing-Hua open pool reactor using 3 to 4 mg of 93.15% enriched 235 uranium samples. Gamma rays are assigned to the responsible fission products by matching gamma rays energies and half lives. Fission rate is calculated by fission track method. Cumulative fission yields of 148 cesium, 90 krypton, 130 iodine, 144 lanthanum, 89 krypton, 136 xenon, 137 xenon and 140 cesium are calculated. This values are compared with previously predicted values and showed good agreement. 21 Ref

Sensibility analysis of the effect of various key parameters on fission product concentration (Mass Number 133 to 138)

International Nuclear Information System (INIS)

Sola, A.

1978-01-01

An analytical sensitivity analysis has been made of the effect of various parameters on the evaluation of fission product concentration. Such parameters include cross-sections, decay constants, branching ratios, fission yields, flux and time. The formulae are applied to isotopes of the iodine, xenon, caesium and barium series. The agreement between analytically obtained data and that derived from a computer-evaluated model is good, suggesting that the analytical representation includes all the important parameters useful to the evaluation of the fission product concentrations

Fission product behaviour in the Peach Bottom and Fort St. Vrain HTGRs

International Nuclear Information System (INIS)

Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

1981-01-01

Actual operating data from the Peach Bottom (PB) and Fort St. Vrain (FSV) High-Temperature Gas-Cooled Reactors (HTGRs) have been compared with code predictions to assess the validity of the methods used to predict the behaviour of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design. The PB primary circuit after seven years of operation was exceptionally clean. A fuel element purge system virtually eliminated the release of fission gases into the primary coolant circuit. Extensive examinations at end-of-life revealed that only Cs and trace amounts of Sr had plated out in the circuit. Their plateout distributions were in excellent agreement with PAD code predictions. Most of the deposited activity was associated with carbonaceous surface films which resulted from occasional small inleakages of lubricating oil. Primary circuit activities in FSV during the first cycle were also very low. Noble gas activity was about 1% of the design limit; and the circulating iodines were at least one order of magnitude below the limit, although the measurement uncertainties are significant. The plateout per pass of the iodine isotopes increased with decreasing half-life (the value for I-131 is about 1% per pass) as predicted with the PADLOC code. Gamma scanning of two helium circulators indicated very low plateout activities. Iodine-131 was the principal fission product observed, along with small amounts of Cs-134, Cs-137, and Ba/La-140. (author)

Iodine filters in nuclear installations

International Nuclear Information System (INIS)

Wilhelm, J.G.

1982-01-01

The present report discusses the significance for environmental exposure of the iodine released with the gaseous effluents of nuclear power stations and reprocessing plants in relation to releases of other airborne radionuclides. Iodine filtration processes are described. The release pathways and the composition of airborne fission product iodine mixtures and their bearing on environmental exposure are discussed on the basis of measured fission product iodine emissions. The sorbents which can be used for iodine filtration, their removal efficiencies and range of applications are dealt with in detail. The particular conditions governing iodine removal, which are determined by the various gaseous iodine species, are illustrated on the basis of experimentally determined retention profiles. Particular attention is given to the limitations imposed by temperature, humidity, radiation and filter poisoning. The types of filter normally used are described, their advantages and drawbacks discussed, the principles underlying their design are outlined and the sources of error indicated. The methods normally applied to test the efficiency of various iodine sorbents are described and assessed. Operating experience with iodine filters, gathered from surveillance periods of many years, is supplemented by a large number of test results and the findings of extensive experiments. Possible ways of prolonging the permissible service lives of iodine filters are discussed and information is given on protective measures. The various iodine removal processes applied in reprocessing plants are described and compared with reference to efficiency and cost. The latest developments in filter technology in reprocessing plants are briefly outlined

Fission products collecting devices

International Nuclear Information System (INIS)

Matsumoto, Hiroshi

1979-01-01

Purpose: To enable fission products trap with no contamination to coolants and cover gas by the provision of a fission products trap above the upper part of a nuclear power plant. Constitution: Upon fuel failures in a reactor core, nuclear fission products leak into coolants and move along the flow of the coolants to the coolants above the reactor core. The fission products are collected in a trap container and guided along a pipeline into fission products detector. The fission products detector monitors the concentration of the fission products and opens the downstream valve of the detector when a predetermined concentration of the fission products is detected to introduce the fission products into a waste gas processing device and release them through the exhaust pipe. (Seki, T.)

Iodine oxides in large-scale THAI tests

International Nuclear Information System (INIS)

Funke, F.; Langrock, G.; Kanzleiter, T.; Poss, G.; Fischer, K.; Kühnel, A.; Weber, G.; Allelein, H.-J.

2012-01-01

Highlights: ► Iodine oxide particles were produced from gaseous iodine and ozone. ► Ozone replaced the effect of ionizing radiation in the large-scale THAI facility. ► The mean diameter of the iodine oxide particles was about 0.35 μm. ► Particle formation was faster than the chemical reaction between iodine and ozone. ► Deposition of iodine oxide particles was slow in the absence of other aerosols. - Abstract: The conversion of gaseous molecular iodine into iodine oxide aerosols has significant relevance in the understanding of the fission product iodine volatility in a LWR containment during severe accidents. In containment, the high radiation field caused by fission products released from the reactor core induces radiolytic oxidation into iodine oxides. To study the characteristics and the behaviour of iodine oxides in large scale, two THAI tests Iod-13 and Iod-14 were performed, simulating radiolytic oxidation of molecular iodine by reaction of iodine with ozone, with ozone injected from an ozone generator. The observed iodine oxides form submicron particles with mean volume-related diameters of about 0.35 μm and show low deposition rates in the THAI tests performed in the absence of other nuclear aerosols. Formation of iodine aerosols from gaseous precursors iodine and ozone is fast as compared to their chemical interaction. The current approach in empirical iodine containment behaviour models in severe accidents, including the radiolytic production of I 2 -oxidizing agents followed by the I 2 oxidation itself, is confirmed by these THAI tests.

Immobilization of iodine in concrete

International Nuclear Information System (INIS)

Clark, W.E.; Thompson, C.T.

1977-01-01

A method for immobilizing fission product radioactive iodine recovered from irradiated nuclear fuel comprises combining material comprising water, Portland cement and about 3 to 20 wt percent iodine as Ba(IO 3 ) 2 to provide a fluid mixture and allowing the fluid mixture to harden, said Ba(IO 3 ) 2 comprising said radioactive iodine. An article for solid waste disposal comprises concrete prepared by this method. 10 claims, 2 figures

The effect of stimulated fission products on the structure and the mechanical properties of zircaloy

International Nuclear Information System (INIS)

Holub, F.

1982-01-01

The objective of investigation was to study the long-term effects of individual simulated fission products on the mechanical properties and the structure of Zircaloy. Tensile Test specimens of Zircaloy were annealed with important simulated fission products at 350 0 C up to 10,000 hours and at higher temperatures (500, 700 0 C) up to 2,000 hours. The principal methods of investigation on annealed Zircaloy specimens were tension tests at room temperature and at 400 0 C, scanning electron microscopy and microprobe technique, X-ray diffraction, X-ray fluorescence, optical metallography. The action of fission products at normal temperatures of reactor operation will give rise to a small enhancement of strength and a small drop of ductility of the fuel cladding material only. At high fuel pin temperatures which may be realized under abnormal operation conditions, some of the fission products potentially will produce detrimental consequences on the integrity of fuel pins. The most effective fission products will be: lanthanum oxide, followed by the earth alkaline oxides and the other rare earth oxides, molybdenum, iodine and cadmium

The Phebus Fission Product and Source Term International Programmes

International Nuclear Information System (INIS)

Clement, B.; Zeyen, R.

2005-01-01

The international Phebus FP programme, initiated in 1988 is one of the major research programmes on light water reactors severe accidents. After a short description of the facility and of the test matrix, the main outcomes and results of the first four integral tests are provided and analysed. Several results were unexpected and some are of importance for safety analyses, particularly concerning fuel degradation, cladding oxidation, chemical form of some fission products, especially iodine, effect of control rod materials on degradation and chemistry, iodine behaviour in the containment. Prediction capabilities of calculation tools have largely been improved as a result of this research effort. However, significant uncertainties remain for a number of phenomena, requiring detailed physical analysis and implementation of improved models in codes, sustained by a number of separate-effect experiments. This is the subject of the new Source Term programme for a better understanding of the phenomenology on important safety issues, in accordance with priorities defined in the EURSAFE project of the 5 th European framework programme aiming at reducing the uncertainties on Source Term analyses. It covers iodine chemistry, impact of boron carbide control rods degradation and oxidation, air ingress situations and fission product release from fuel. Regarding the interpretation of Phebus, an international co-operation has been established since over ten years, particularly helpful for the improvement and common understanding of severe accident phenomena. Few months ago, the Phebus community was happy to welcome representatives of a large number of organisations from the following new European countries: the Czech republic, Hungary, Lithuania, Slovakia, Slovenia and also from Bulgaria and Romania. (author)

Iodine-129 separation and determination by neutron activation analysis

International Nuclear Information System (INIS)

Bate, L.C.; Stokely, J.R.

1982-01-01

This paper describes a method for analysis of iodine-129 in fission product mixtures originating from fuel reprocessing studies and low-level wastes. The method utilizes conventional iodine valence adjustment and solvent extraction techniques to chemically separate iodine-129 from most fission products. The iodine-129 is determined by neutron irradiation and measurement of the 12.4 hour iodine-130 produced by the neutron capture reaction. Special techniques were devised for neutron irradiation of iodine-129 samples in the pneumatic tube irradiation facilities at the High Flux Isotope (HFIR) and Oak Ridge Research (ORR) Reactors. Chemically separated iodine-129 is adsorbed on an anion exchange resin column made from an irradiation container. The loaded resin is then irradiated in either of the pneumatic facilities to produce iodine-130. Sensitivity of the analysis with the HFIR facility (flux: 5x10 14 n/cm 2 /s) and a 100 second irradiation time is approximately 0.03 nanograms. Samples up to 250 ml in volume can be easily processed. (author)

Polarographic determination of Iodide and Iodate, in Solutions Coming from Aerosols in Fission Products Containment Studies in Nuclear Power Stations

International Nuclear Information System (INIS)

Sanchez, M.; Ballesteros, O.; Fernandez, M.; Clavero, M.A.; Gonzalez, A.M.

2000-01-01

A polarographic method is described for the iodine species determination, iodide and iodate in water solutions. the iodate can be determined by differential pulse polarography. Calibration curves and the detection and determination limits have been obtained. Iodides is oxidized to iodate with sodium hypochlorite and the excess of oxidizing agent is destroyed with sodium sulphide. The concentration of iodide is calculated as the difference between the concentration of iodate in the sample before and after the oxidation. As an application, species of iodine in samples coming from the experimental plants GIRS (Gaseous Iodine Removal by Sprays) of Nuclear Fission Department of the CIEMAT, dedicated to fission products containment studies in nuclear power station, were determined. (Author) 10 refs

Development of Silver-exchanged Adsorbents for the Removal of Fission Iodine from Alkaline Dissolution

International Nuclear Information System (INIS)

Kim, Taewoon; Lee, Seung-Kon; Lee, Suseung; Lee, Jun Sig

2015-01-01

Most of the iodine exists in the caustic dissolution as iodide form. KAERI is developing LEU-based fission 99 Mo production process which is connected to the new research reactor, which is being constructed in Kijang, Busan, Korea. In KAERI process, silver-exchanged adsorbent is used to adsorb iodide from the solution. Adsorbed iodide can be recovered and recycled for radiopharmaceuticals. In KAERI process, silver-exchanged adsorbent is used to adsorb iodide from the solution. Adsorbed iodide can be recovered and recycled for radiopharmaceuticals. Synthesis of silver-doped alumina is conducted in two ways. One is using the ascorbic acid as a reducing agent. However, this method is impossible to control

Disposal of Iodine-129

International Nuclear Information System (INIS)

Morgan, M.T.; Moore, J.G.; Devaney, H.E.; Rogers, G.C.; Williams, C.; Newman, E.

1978-01-01

One of the problems to be solved in the nuclear waste management field is the disposal of radioactive iodine-129, which is one of the more volatile and long-lived fission products. Studies have shown that fission products can be fixed in concrete for permanent disposal. Current studies have demonstrated that practical cementitious grouts may contain up to 18% iodine as barium iodate. The waste disposal criterion is based on the fact that harmful effects to present or future generations can be avoided by isolation and/or dilution. Long-term isolation is effective in deep, dry repositories; however, since penetration by water is possible, although unlikely, release was calculated based on leach rates into water. Further considerations have indicated that sea disposal on or in the ocean floor may be a more acceptable alternative

Interactions of fission product vapours with aerosols

Energy Technology Data Exchange (ETDEWEB)

Benson, C G; Newland, M S [AEA Technology, Winfrith (United Kingdom)

1996-12-01

Reactions between structural and reactor materials aerosols and fission product vapours released during a severe accident in a light water reactor (LWR) will influence the magnitude of the radiological source term ultimately released to the environment. The interaction of cadmium aerosol with iodine vapour at different temperatures has been examined in a programme of experiments designed to characterise the kinetics of the system. Laser induced fluorescence (LIF) is a technique that is particularly amenable to the study of systems involving elemental iodine because of the high intensity of the fluorescence lines. Therefore this technique was used in the experiments to measure the decrease in the concentration of iodine vapour as the reaction with cadmium proceeded. Experiments were conducted over the range of temperatures (20-350{sup o}C), using calibrated iodine vapour and cadmium aerosol generators that gave well-quantified sources. The LIF results provided information on the kinetics of the process, whilst examination of filter samples gave data on the composition and morphology of the aerosol particles that were formed. The results showed that the reaction of cadmium with iodine was relatively fast, giving reaction half-lives of approximately 0.3 s. This suggests that the assumption used by primary circuit codes such as VICTORIA that reaction rates are mass-transfer limited, is justified for the cadmium-iodine reaction. The reaction was first order with respect to both cadmium and iodine, and was assigned as pseudo second order overall. However, there appeared to be a dependence of aerosol surface area on the overall rate constant, making the precise order of the reaction difficult to assign. The relatively high volatility of the cadmium iodide formed in the reaction played an important role in determining the composition of the particles. (author) 23 figs., 7 tabs., 22 refs.

Interactions of fission product vapours with aerosols

International Nuclear Information System (INIS)

Benson, C.G.; Newland, M.S.

1996-01-01

Reactions between structural and reactor materials aerosols and fission product vapours released during a severe accident in a light water reactor (LWR) will influence the magnitude of the radiological source term ultimately released to the environment. The interaction of cadmium aerosol with iodine vapour at different temperatures has been examined in a programme of experiments designed to characterise the kinetics of the system. Laser induced fluorescence (LIF) is a technique that is particularly amenable to the study of systems involving elemental iodine because of the high intensity of the fluorescence lines. Therefore this technique was used in the experiments to measure the decrease in the concentration of iodine vapour as the reaction with cadmium proceeded. Experiments were conducted over the range of temperatures (20-350 o C), using calibrated iodine vapour and cadmium aerosol generators that gave well-quantified sources. The LIF results provided information on the kinetics of the process, whilst examination of filter samples gave data on the composition and morphology of the aerosol particles that were formed. The results showed that the reaction of cadmium with iodine was relatively fast, giving reaction half-lives of approximately 0.3 s. This suggests that the assumption used by primary circuit codes such as VICTORIA that reaction rates are mass-transfer limited, is justified for the cadmium-iodine reaction. The reaction was first order with respect to both cadmium and iodine, and was assigned as pseudo second order overall. However, there appeared to be a dependence of aerosol surface area on the overall rate constant, making the precise order of the reaction difficult to assign. The relatively high volatility of the cadmium iodide formed in the reaction played an important role in determining the composition of the particles. (author) 23 figs., 7 tabs., 22 refs

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

IMPACT OF THE CHEMICAL FORM OF IN-CONTAINMENT SOURCE ON FISSION PRODUCT RELEASE FROM WWER-1000/V-320 TYPE NPP CONTAINMENT DURING LOCA

Directory of Open Access Journals (Sweden)

Adam Kecek

2016-12-01

Full Text Available Nuclear power plant accidents may be followed by a release of fission products into the environment. This release is dependent on several phenomena, such as chemistry, pressure, type of the accident etc. The aim of this paper is to assess the impact of the chemical form of iodine on the fission product release into the environment.

Fission product release modelling for application of fuel-failure monitoring and detection - An overview

Energy Technology Data Exchange (ETDEWEB)

Lewis, B.J., E-mail: lewibre@gmail.com [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario, K7K 7B4 (Canada); Chan, P.K.; El-Jaby, A. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario, K7K 7B4 (Canada); Iglesias, F.C.; Fitchett, A. [Candesco Division of Kinectrics Inc., 26 Wellington Street East, 3rd Floor, Toronto, Ontario M5E 1S2 (Canada)

2017-06-15

A review of fission product release theory is presented in support of fuel-failure monitoring analysis for the characterization and location of defective fuel. This work is used to describe: (i) the development of the steady-state Visual-DETECT code for coolant activity analysis to characterize failures in the core and the amount of tramp uranium; (ii) a generalization of this model in the STAR code for prediction of the time-dependent release of iodine and noble gas fission products to the coolant during reactor start-up, steady-state, shutdown, and bundle-shifting manoeuvres; (iii) an extension of the model to account for the release of fission products that are delayed-neutron precursors for assessment of fuel-failure location; and (iv) a simplification of the steady-state model to assess the methodology proposed by WANO for a fuel reliability indicator for water-cooled reactors.

Dry method for recycling iodine-loaded silver zeolite

International Nuclear Information System (INIS)

Thomas, T.R.; Staples, B.A.; Murphy, L.P.

1978-01-01

Fission product iodine is removed from a waste gas stream and stored by passing the gas stream through a bed of silver-exchanged zeolite until the zeolite is loaded with iodine, passing dry hydrogen gas through the bed to remove the iodine and regenerate the bed, and passing the hydrogen stream containing the hydrogen iodide thus formed through a lead-exchanged zeolite which absorbs the radioactive iodine from the gas stream and permanently storing the lead-exchanged zeolite loaded with radioactive iodine

Sensibility analysis of the effect of various key parameters on fission product concentration mass number 127 to 132 and Xe - 133 m)

International Nuclear Information System (INIS)

Sola, A.

1978-01-01

An analytical sensitivity analysis has been made of the effect' of various parameters on the evaluation of fission product concentration. Such parameters include cross sections, decay constants, branching ratios, fission yields, flux and time. The formulae are applied to isotopes of the Antimony, Tellurium, Iodine and Xenon series. The agreement between analytically obtained data and that derived from a computer evaluated model is good, suggesting that the analytical representation includes all the important parameters useful to the evaluation of the fission product concentrations

Fission product chemistry in severe nuclear reactor accidents, specialists' meeting at JRC-Ispra, 15-17 January 1990

International Nuclear Information System (INIS)

Nichols, A.L.

1990-05-01

A specialists' meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions). (author)

[Fission product yields of 60 fissioning reactions]. Final report

International Nuclear Information System (INIS)

Rider, B.F.

1995-01-01

In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

Prediction of iodine activity peak during refuelling

International Nuclear Information System (INIS)

Hozer, Z.; Vajda, N.

2001-01-01

The increase of fission product activities in the primary circuit of a nuclear power plant indicates the existence of defects in some fuel rods. The power change leads to the cooling down of the fuel and results in the fragmentation of the UO 2 pellets, which facilitates the release of fission products from the intergranular regions. Furthermore the injection of boric acid after shutdown will increase the primary activity, due to the solution of deposited fission products from the surface of the core components. The calculation of these phenomena usually is based on the evaluation of activity measurements and power plant data. The estimation of iodine spiking peak during reactor transients is based on correlation with operating parameters, such as reactor power and primary pressure. The approach used in the present method was applied for CANDU reactors. The VVER-440 specific correlations were determined using the activity measurements of the Paks NPP and the data provided by the Russian fuel supplier. The present method is used for the evaluation of the iodine isotopes, as well as the noble gases. A numerical model has been developed for iodine spiking simulation and has been validated against several shutdown transients, measured at Paks NPP. (R.P.)

Fission product plateout and liftoff in the MHTGR primary system: A review

International Nuclear Information System (INIS)

Wichner, R.P.

1991-04-01

A review is presented of the technical basis for predicting radioactivity release resulting from depressurization of an MHTGR primary system. Consideration is restricted to so called dry events with no involvement of the steam system. The various types of deposition mechanisms effective for iodine, cesium, strontium, and silver are discussed in terms of their chemical characteristics and the nature of the materials in the primary system. Emphasis is given to iodine behavior, including means for estimating the quantity available for release, the types of plateout locations in the primary system, and the effect of dust on distribution and release. The behavior of fission products cesium, strontium, and silver in such accidents is presented qualitatively. A major part of the review deals with expected dust levels, types, and transport. Available information on the level and nature of dust in the HTGR primary system is reviewed. A summary is presented of dust deposition and liftoff mechanisms. It was concluded that recent approaches to dust liftoff modeling, based on turbulent burst concepts for removal from surfaces, probably offer advantages over the current shear ratio approach. This study concludes that iodine releases from dry depressurization events are likely to be extremely low, on the order of millicuries, due to a predictably low degree of chemical desorption, a low degree of dust liftoff, and a low involvement of iodine with dust. It was also concluded that deposition mechanisms controlling the distribution of fission product material in the primary system, and hence also controlling the degree of liftoff, depend strongly on the chemical nature of the individual elements. Therefore contrary to the current practice, both plateout and liftoff models should reflect those unique chemical and physical properties. 56 refs., 16 figs., 23 tabs

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1984-09-01

This is the tenth issue of a report series on Fission Product Data, which informs us about all the activities in this field, which are planned, ongoing, or have recently been completed. The types of activities included are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission), neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products, lumped fission product data (decay heat, absorption, etc.). There is also a section with recent references relative to fission product nuclear data

Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

International Nuclear Information System (INIS)

Bourgeois, P.; Stora, J.P.

1979-01-01

The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

Chemical thermodynamics of iodine species in the HTGR fuel particle

International Nuclear Information System (INIS)

Lindemer, T.B.

1982-09-01

The iodine-containing species in an intact fuel particle in the high-temperature gas-cooled reactor (HTGR) have been calculated. Assumptions include: (1) attainment of chemical thermodynamic equilibrium among all species in the open porosity of the particle, primarily in the buffer layer; and (2) fission-product concentrations in proportion to their yields. The primary gaseous species is calculated to be cesium iodide; in carbide-containing fuels, gaseous barium iodide may exhibit equivalent pressures. The condensed iodine-containing phase is usually cesium iodide, but in carbide-containing fuels, barium iodide may be stable instead. Absorption of elemental iodine on the carbon in the particle appears to be less than or equal to 10 -4 μg I/g C. The fission-product-spectra excess of cesium over iodine would generally be adsorbed on the carbon, but may form Cs 2 MoO 4 under some circumstances

Deep Bed Iodine Sorbent Testing FY 2011 Report

International Nuclear Information System (INIS)

Soelberg, Nick; Watson, Tony

2011-01-01

Nuclear fission results in the production of fission products (FPs) and activation products that increasingly interfere with the fission process as their concentrations increase. Some of these fission and activation products tend to evolve in gaseous species during used nuclear fuel reprocessing. Analyses have shown that I129, due to its radioactivity, high potential mobility in the environment, and high longevity (half life of 15.7 million years), can require control efficiencies of up to 1,000x or higher to meet regulatory emission limits. Deep-bed iodine sorption testing has been done to evaluate the performance of solid sorbents for capturing iodine in off-gas streams from nuclear fuel reprocessing plants. The objectives of the FY 2011 deep bed iodine sorbent testing are: (1) Evaluate sorbents for iodine capture under various conditions of gas compositions and operating temperature (determine sorption efficiencies, capacities, and mass transfer zone depths); and (2) Generate data for dynamic iodine sorption modeling. Three tests performed this fiscal year on silver zeolite light phase (AgZ-LP) sorbent are reported here. Additional tests are still in progress and can be reported in a revision of this report or a future report. Testing was somewhat delayed and limited this year due to initial activities to address some questions of prior testing, and due to a period of maintenance for the on-line GC. Each test consisted of (a) flowing a synthetic blend of gases designed to be similar to an aqueous dissolver off-gas stream over the sorbent contained in three separate bed segments in series, (b) measuring each bed inlet and outlet gas concentrations of iodine and methyl iodide (the two surrogates of iodine gas species considered most representative of iodine species expected in dissolver off-gas), (c) operating for a long enough time to achieve breakthrough of the iodine species from at least one (preferably the first two) bed segments, and (d) post-test purging

HEU and LEU MTR fuel elements as target materials for the production of fission molybdenum

International Nuclear Information System (INIS)

Sameh, A.A.; Bertram-Berg, A.

1993-01-01

The processing of irradiated MTR-fuels for the production of fission nuclides for nuclear medicine presents a significantly increasing task in the field of chemical separation technology of high activity levels. By far the most required product is MO-99, the mother nuclide of Tc-99m which is used in over 90% of the organ function tests in nuclear medicine. Because of the short half life of Mo-99 (66 h) the separation has to be carried out from shortly cooled neutron irradiated U-targets. The needed product purity, the extremely high radiation level, the presence of fission gases like xenon-133 and of volatile toxic isotopes such as iodine-131 and its compounds in kCi-scale require a sophisticated process technology

Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

International Nuclear Information System (INIS)

Verfondern, K.; Mueller, D.

1991-01-01

Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

Analysis of Dust and Fission Products in PBMR Turbine

International Nuclear Information System (INIS)

Stempniewicz, M.M.; Wessels, D.

2014-01-01

A 400 MWth direct cycle Pebble Bed Modular reactor was under development in South Africa. The work performed included design and safety analyses. In HTR/PBMR, graphite dust is generated during normal reactor operation due to pebble-to-pebble scratching. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released, although in very small quantities, during normal operation. This paper presents a model and analyses of the PBMR turbine with the SPECTRA code. The purpose of the present work was to estimate the amount and distribution of deposited dust and the fission products, namely cesium, iodine, and silver, during plant life-time, which was assumed to be 40 full-power years. The performed work showed that after 40 years of plant life-time deposited layers are very small. The largest deposition is of course observed on the dust filters. Apart from the dust filters, the largest dust deposition is observed on the: • Outer Casing (inner walls) • Turbine Rotor Cooling Cavity (inner walls) • HPC Cold Cooling Gas Header (inner walls) This is caused by relatively low gas velocities in these volumes. The low velocities allow a continuous build-up of the dust layer. About 90% of cesium, 40% of iodine, and 99.9% of silver is adsorbed on the metallic structures of the turbine. The sorption rate increases along the turbine due to decreasing temperatures. In case of cesium and iodine the highest concentrations are observed in the last stage (stage 12) of the turbine. In the case of silver the sorption is so large that the silver vapor is significantly depleted in the last stages of the turbine. This is a reason for having a maximum in silver concentration in the stage 10. In the following stages the concentration decreases due to very small silver vapor fraction in the gas. (author)

Production of adsorbent from palm shell for radioactive iodine scrubbing process

International Nuclear Information System (INIS)

Mohamad Azman Che Mat Isa; Ku Halim Ku Hamid; Muhd Noor Muhd Yunus; Mohamad Puad Abu; Abdul Halim Badaruddin; Mohammad Nizammudin Abd Aziz; Muhd Ridwan Abdul Rahim

2010-01-01

The biggest biomass source in Malaysia comes from oil palm industry. According to the statistic of year 2004, Malaysia produced 40 million tones per year of biomass which 30 million tones of biomass originated from the oil palm industries. Therefore, the biomass waste such as palm kernel shell can be used to produce granular adsorbent for radioactive materials. For that reason, a newly system, called Rocking Kiln - Fluidized Bed (RK - FB) was developed to utilize large amount of the biomass to produce high value added product. Charcoal or chemically produced activated carbon could be produced by using the kiln. Washing process was introduced to remove particles, minerals and volatile matters from charcoal produced and then would create more surface area in the adsorbent by creating more active sites. In this research, the adsorbent produced was used to scrub iodine 131. In nuclear power reactor, iodine isotope 131 is produced during nuclear fission, and this elementary radioactive iodine may pollute exhaust air streams that could cause thyroid cancer. For removal of radioactive iodine, normally a potassium iodide - impregnated activated carbon (KI - AC) is used. Thus, a process will be developed to produce KI - AC and this product will be used to calculate the efficiency to remove the radioactive iodine 131.The results obtain show that adsorbent produced has a high potential to be used in radioactive adsorbing and likely more economics. This paper will elaborate further the experimental set-up of in Kiln - Fluidized Bed (RK - FB), adsorbent quality and radioactive scrubbing process. (author)

Chemisorption of organic iodine compounds forming from fission isotopes of radioactive iodine

International Nuclear Information System (INIS)

Tot, G.; Galina, F.; Zel'd, E.

1977-01-01

Studied is ethyl iodine adsorption, labelled by iodine 131, on palladium black and on aluminium oxide activized by palladium. The desorption of adsorbed iodine in the temperature range of 20-600 deg C by the mass spectroscopy and thermal gravimetric methods was investigated. At the ethyl iodine and palladium interaction the bond between carbon and iodine in the ethyl iodine molecule breaks down and extracting iodine reacts with palladium, forming a stable compound at high temperatures. Desorption of adsorbed iodine is insignificant up to the temperatures of 250-300 deg C. Thus, sorbents, containing palladium, may be successfully applied for iodine absorption from the organic iodine compounds. These compounds spontaneously appear from the iodine fragment ratio isotopes during their interaction with some environmental organic impurities

Fission product transport at Three Mile Island

International Nuclear Information System (INIS)

Owen, D.E.; Cox, T.E.; Broughton, J.M.

1985-01-01

Three Mile Island Unit 2 radionuclide analyses are reviewed and summarized in order to determine how fission products moved to various parts of the reactor system at the time of the accident. Despite high fuel temperatures and major core damage, the core retained a very large fraction of most radionuclides. Reactor coolant, either remaining in the primary system or released to various sumps and tanks, retained significant quantities of cesium and iodine. Noble gases were effectively retained within the containment building with the exception of small releases to the environment. Long-term deposition and retention on vessel, piping, and bulding surfaces were insignificant for all isotopes examined. The measured partitioning of radionuclides within these systems is tabulated and recommendations for additional analyses are presented

Measurement of Fission Product Yields from Fast-Neutron Fission

Science.gov (United States)

Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

2014-09-01

One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

Technical bases for estimating fission product behavior during LWR accidents. Technical report

International Nuclear Information System (INIS)

1981-06-01

The objective of this report is to provide the Nuclear Regulatory Commission and the public with a description of the best technical information currently available for estimating the release of radioactive material during postulated reactor accidents, and to identify where gaps exist in our knowledge. This report focuses on those low probability-high consequence accidents involving severe damage to the reactor core and core meltdown that dominate the risk to the public. Furthermore, in this report particular emphasis is placed on the accident behavior of radioactive iodine, as (1) radioiodine is predicted to be a major contributor to public exposure, (2) current regulatory accident analysis procedures focus on iodine, and (3) several technical issues have been raised recently about the magnitude of iodine release. The generation, transport, and attenuation of aerosols were also investigated in some detail to assess their effect on fission product release estimates and to determine the performance of engineered safety features under accident conditions exceeding their design bases

Fission product behavior during the first two PBF severe fuel damage tests

International Nuclear Information System (INIS)

Osetek, D.J.; Cronenberg, A.W.; Hobbins, R.R.; Vinjamuri, K.

1984-01-01

The results of the first two severe fuel damage tests performed in the Power Burst Facility are assessed in terms of fission product release and chemical behavior. On-line gamma spectroscopy and grab sample data indicate limited release during solid-phase fuel heatup. Analysis indicates that the fuel morphology conditions for the trace-irradiated fuel employed in these two tests limit initial release. Only upon high temperature fuel restructuring and liquefaction is significant release indicated. Chemical equilibrium predictions, based on steam oxidation or reduction conditions, indicate I to be the primary iodine species during trnsport in the steam environment of the first test and CsI to be the primary species during transport in the hydrogen environment of the second test. However, the higher steam flow rate conditions of the first test transported the released iodine through the sample system; whereas, low-hydrogen flow rate of the second test apparently allowed the vast majority of iodine-bearing compounds to plateout during transport

LWR severe accident simulation: Iodine behaviour in FPT2 experiment and advances on containment iodine chemistry

Energy Technology Data Exchange (ETDEWEB)

Girault, N., E-mail: nathalie.girault@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Bosland, L. [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Dickinson, S. [National Nuclear Laboratory, Harwell, Oxon OX11 0QT (United Kingdom); Funke, F. [AREVA NP Gmbh, PO Box 1109, 91001 Erlangen (Germany); Guentay, S. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Herranz, L.E. [Centro des Investigaciones Energeticas, MedioAmbiantales y Tecnologicas, av. Complutense 2, 28040 Madrid (Spain); Powers, D. [Sandia National Laboratories, New Mexico, PO Box 5800, Albuquerque, NM 87185 (United States)

2012-02-15

Highlights: Black-Right-Pointing-Pointer Short term gaseous iodine fraction can be produced either in primary circuit or on containment condensing surfaces. Black-Right-Pointing-Pointer Gaseous radiolytic reactions convert volatile iodine into non-volatile iodine oxide particulates. Black-Right-Pointing-Pointer Alkaline and evaporating sump decrease the iodine volatility in containment. Black-Right-Pointing-Pointer Release of volatile iodine from containment surfaces explained the long term stationary residual gaseous iodine concentration. - Abstract: The Phebus Fission Product (FP) Program studies key phenomena of severe accidents in water-cooled nuclear reactors. In the framework of the Phebus program, five in-pile experiments have been performed that cover fuel rod degradation and behaviour of fission products released via the coolant circuit into the containment vessel. The focus of this paper is on iodine behaviour during the Phebus FPT2 test. FPT2 used a 33 GWd/t uranium dioxide fuel enriched to 4.5%, re-irradiated in situ for 7 days to a burn-up of 130 MWd/t. This test was performed to study the impact of steam-poor conditions and boric acid on the fission product chemistry. For the containment vessel, more specifically, the objective was to study iodine chemistry in an alkaline sump under evaporating conditions. The iodine results of the Phebus FPT2 test confirmed many of the essential features of iodine behaviour in the containment vessel provided by the first two Phebus tests, FPT0 and FPT1. These are the existence of an early gaseous iodine fraction, the persistence of low gaseous iodine concentrations and the importance of the sump in suppressing the iodine partitioning from sump to atmosphere. The main new insights provided by the Phebus FPT2 test were the iodine desorption from stainless steel walls deposits and the role of the evaporating sump in further iodine depletion in the containment atmosphere. The current paper presents an interpretation of

Current status of the FASTGRASS/PARAGRASS models for fission product release from LWR fuel during normal and accident conditions

International Nuclear Information System (INIS)

Rest, J.; Zawadski, S.A.; Piasecka, M.

1983-10-01

The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wide range of fuel operating conditions, and can be regarded as a state-of-the-art model based on our current level of understanding of fission product behavior. PARAGRASS is an extremely efficient, mechanistic computer code with the capability of modeling steady-state and transient fission-product behavior. The models in PARAGRASS are based on the more detailed ones in FASTGRASS. PARAGRASS updates for the FRAPCON (PNL), FRAP-T (INEL), and SCDAP (INEL) codes have recently been completed and implemented. Results from an extensive FASTGRASS verification are presented and discussed for steady-state and transient conditions. In addition, FASTGRASS predictions for fission product release rate constants are compared with those in NUREG-0772. 21 references, 13 figures

Comparison of the ENIGMA code with experimental data on thermal performance, stable fission gas and iodine release at high burnup

Energy Technology Data Exchange (ETDEWEB)

Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

1997-08-01

The predictions of the ENIGMA code have been compared with data from high burn-up fuel experiments from the Halden and RISO reactors. The experiments modelled were IFA-504 and IFA-558 from Halden and the test II-5 from the RISO power burnup test series. The code has well modelled the fuel thermal performance and has provided a good measure of iodine release from pre-interlinked fuel. After interlinkage the iodine predictions remain a good fit for one experiment, but there is significant overprediction for a second experiment (IFA-558). Stable fission gas release is also well modelled and the predictions are within the expected uncertainly band throughout the burn-up range. This report presents code predictions for stable fission gas release to 40GWd/tU, iodine release measurements to 50GWd/tU and thermal performance (fuel centre temperature) to 55GWd/tU. Fuel ratings of up to 38kW/m were modelled at the high burn-up levels. The code is shown to accurately or conservatively predict all these parameters. (author). 1 ref., 6 figs.

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, G.

1976-05-01

The purpose of this series is to inform scientists working on Fission Product Nuclear Data, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. This report consists of reproductions of essentially unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat

Computer program FPIP-REV calculates fission product inventory for U-235 fission

Science.gov (United States)

Brown, W. S.; Call, D. W.

1967-01-01

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1981-06-01

This is the seventh issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The sixth issue of this series has been published in June 1980 as INDC(NDS)-113/G+P. The present issue includes contributions which were received by NDS between 1 August 1980 and 25 May 1981

Study on collection efficiency of fission products by spray: Experimental device and modelling

Energy Technology Data Exchange (ETDEWEB)

Ducret, D.; Roblot, D.; Vendel, J. [Institut de Protection et de Surete Nucleaire, Gif-Sur-Yvette (France); Billarand, Y. [ECCO Pharmacie et Chimie, Neuilly (France)

1997-08-01

Consequences of an hypothetical overheating reactor accident in nuclear power plants can be limited by spraying cold water drops into containment building. The spray reduces the pressure and the temperature levels by condensation of steam and leads to the washout of fission products (aerosols and gaseous iodine). The present study includes a large program devoted to the evaluation of realistic washout rates. An experimental device (named CARAIDAS) was designed and built in order to determine the collection efficiency of aerosols and iodine absorption by drops with representative conditions of post-accident atmosphere. This experimental device is presented in the paper and more particularly: (1) the experimental enclosure in which representative thermodynamic conditions can be achieved, (2) the monosized drops generator, the drops diameter measurement and the drops collector, (3) the cesium iodide aerosols generator and the aerosols measurements. Modelling of steam condensation on drops aerosols collection and iodine absorption are described. First experimental and code results on drops and aerosols behaviour are compared. 8 refs., 18 figs.

Fission product yields

International Nuclear Information System (INIS)

Valenta, V.; Hep, J.

1978-01-01

Data are summed up necessary for determining the yields of individual fission products from different fissionable nuclides. Fractional independent yields, cumulative and isobaric yields are presented here for the thermal fission of 235 U, 239 Pu, 241 Pu and for fast fission (approximately 1 MeV) of 235 U, 238 U, 239 Pu, 241 Pu; these values are included into the 5th version of the YIELDS library, supplementing the BIBFP library. A comparison is made of experimental data and possible improvements of calculational methods are suggested. (author)

Fission-product yields for thermal-neutron fission of curium-243

International Nuclear Information System (INIS)

Breederland, D.G.

1982-01-01

Cumulative fission yields for 25 gamma rays emitted during the decay of 23 fission products produced by thermal-neutron fission of 243 Cm have been determined. Using Ge(Li) spectroscopy, 33 successive pulse-height spectra of gamma rays emitted from a 77-ng sample of 243 Cm over a period of approximately two and one-half months were analyzed. Reduction of these spectra resulted in the identification and matching of gamma-ray energies and half-lives to specific radionuclides. Using these results, 23 cumulative fission-product yields were calculated. Only those radionuclides having half-lives between 6 hours and 65 days were observed. Prior to this experiment, no fission-product yields had been recorded for 243 Cm

Fission-product source terms

International Nuclear Information System (INIS)

Lorenz, R.A.

1981-01-01

This presentation consists of a review of fission-product source terms for light water reactor (LWR) fuel. A source term is the quantity of fission products released under specified conditions that can be used to calculate the consequences of the release. The source term usually defines release from breached fuel-rod cladding but could also describe release from the primary coolant system, the reactor containment shell, or the site boundary. The source term would be different for each locality, and the chemical and physical forms of the fission products could also differ

Fission product behaviour in severe accidents

International Nuclear Information System (INIS)

Jokiniemi, J.; Auvinen, A.; Maekynen, J.; Valmari, T.

1998-01-01

The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

Grain boundary sweeping and dissolution effects on fission product behavior under severe fuel damage accident conditions

International Nuclear Information System (INIS)

Rest, J.

1985-10-01

The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behavior considers the migration and coalescence of fission gas bubbles in either molten uranium, or a zircaloy-uranium eutectic melt. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally irradiated fuel are highlighted

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1983-08-01

This is the ninth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The eighth issue of this series has been published in July 1982 as INDC(NDS)-130. The present issue includes contributions which were received by NDS between 1 August 1982 and 25 June 1983

Fission product release profiles from spherical HTR fuel elements at accident temperatures

International Nuclear Information System (INIS)

Schenk, W.; Pitzer, D.; Nabielek, H.

1986-10-01

A total of 22 fuel elements with modern TRISO particles has been tested in the temperature range 1500-2500 0 C. Additionally, release profiles of iodine and other isotopes have been obtained with seven UO 2 samples at 1400-1800 0 C. For heating times up to 100 hours at the maximum temperature, the following results are pertinent to HTR accident conditions: Ag 110 m is the only fission products to be released at 1200-1600 0 C by diffusion through intact SiC, but it is of low significance in accident assessments; cesium, iodine, strontium, and noble gas releases up to 1600 0 C are solely due to various forms of contamination; at 1700-1800 0 C, corrosion induced SiC defects cause the release of Cs, Sr, I/Xe/Kr; above 2000 0 C, thermal decomposition of the silicon carbide layer sets in while pyrocarbons still remain intact. Around 1600 0 C, the accident specific contribution of cesium, strontium, iodine, and noble gases is negligible. (orig./HP) [de

Behaviour of short-lived fission products within operating UO2 fuel elements

International Nuclear Information System (INIS)

Hastings, I.J.; Hunt, C.E.L.; Lipsett, J.J.

1983-01-01

We have carried out experiments using a ''sweep gas'' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 500 mm long and contained fuel of density 10.65-10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. In tests at linear powers of 45 and 60 kW/m to maximum burnups of 70 MW.h/kg U, the species measured directly at the spectrometer were generally the short-lived xenons and kryptons. We did not observe iodine or bromine during normal operation. However, we have deduced the behaviour of I-133 and I-135 from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against lambda (decay constant) or effective lambda for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. Our inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5x10 -3 . The ANS 5.4 release correlation gives calculated results in good agreement with our measurements. (author)

Diffusion of cesium and iodine in compressed IG-110 graphite compacts

International Nuclear Information System (INIS)

Carter, L.M.; Brockman, J.D.; Robertson, J.D.; Loyalka, S.K.

2016-01-01

Nuclear graphite grade IG-110 is currently used in the High Temperature Engineering Test Reactor (HTTR) in Japan for certain permanent and replaceable core components, and is a material of interest in general. Therefore, transport parameters for fission products in this material are needed. Measurement of diffusion through pressed compacts of IG-110 graphite is experimentally attractive because they are easy to prepare with homogeneous distributions of fission product surrogates. In this work, we measured diffusion coefficients for Cs and I in pressed compacts made from IG-110 powder in the 1079–1290 K temperature range, and compared them to those obtained in as-received IG-110. - Highlights: • A method for analysis of fission product diffusion in graphite by ICP-MS was applied to pressed IG-110 graphite compacts containing cesium and iodine. • Diffusion coefficients for cesium and iodine were obtained. • The measurement design simulates HTGR conditions of high temperature and flowing helium.

Storage of spent fuels: implementation of a research program on the risk of waste container rupture due to stress corrosion induced by fission products

International Nuclear Information System (INIS)

Parise, M.; Walle, E.; Foct, J.

2001-01-01

The following topics were dealt with: research programm on stress corrosion of spent fuel casks materials due to fission products, such as iodine, chemical interactions with zirconium, chemical aspects of stress corrosion, rupture risk assessment

Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

Energy Technology Data Exchange (ETDEWEB)

Lemmens, K., E-mail: klemmens@sckcen.be [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); González-Robles, E.; Kienzler, B. [Karlsruhe Institute of Technology Institute for Nuclear Waste Disposal (KIT-INE), PO Box 3640, D-76021 Karlsruhe (Germany); Curti, E. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Serrano-Purroy, D. [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, PO Box 2340, D-76125 Karlsruhe (Germany); Sureda, R.; Martínez-Torrents, A. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Roth, O. [Studsvik, Nuclear AB, 611 82 Nyköping (Sweden); Slonszki, E. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary); Mennecart, T. [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Günther-Leopold, I. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Hózer, Z. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary)

2017-02-15

The instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45–63 GWd/t{sub HM} and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride – bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H{sub 2} atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways. - Highlights: • Leach tests were performed to study the instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel. • In these tests, the fission gas release given by the operator was a pessimistic estimator of the iodine and cesium release. • Iodine and cesium release is proportional to linear power rating beyond 200 W cm{sup −1}. • Closure of the fuel-cladding gap at high burn-up slows down the release. • The release rate decreases following an exponential equation.

Grain boundary sweeping and dissolution effects on fission product behaviour under severe fuel damage accident conditions

International Nuclear Information System (INIS)

Rest, J.

1986-01-01

The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behaviour considers the migration and coalescence of fission gas bubbles in either molten uranium, or a Zircaloy-Uranium eutectic melt. Results of the analyses demonstrate that intragranular fission product behavior during the tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. Whereas fuel liquefaction leads to an enhanced release of fission products in trace-irradiated fuel, the occurrence of fuel liquefaction in normally-irradiated fuel can degrade fission product release. This phenomenon is due in part to reduced gas-bubble mobilities in a viscous medium as compared to vapor transport, and in part to a degradation of grain growth rates and the subsequent decrease in grain-boundary sweeping of intragranular fission products into the liquified lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and normally-irradiated fuel. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, G.

1975-01-01

This is the first issue of a report series on Fission Product Nuclear Data (FPND), published every six months by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). Its purpose is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat. The present issue includes contributions which were received by NDS before 1 November 1975

Radioactive iodine releases from nuclear power plant, (2)

International Nuclear Information System (INIS)

Naritomi, Mitsuo

1974-01-01

Internal radiation dose through the respiratory intake of fission products is predominantly due to radioactive iodine not only at the time of reactor accidents but also in normal operation of nuclear facilities. Technological studies in this field have thus been quite active to this day. With the rapid advance of nuclear power generation in recent years, the efforts to reduce environmental release of radioactive iodine and to enhance environmental safety are all the more emphasized. Experiences in the Japan Atomic Energy Research Institute during past about six years are described concerning the radioactive iodine release to the atmosphere in 131 I production and the measures taken to reduce the release. Then, problems are expounded regarding the radioactive iodine release at the time of reactor accidents and in spent fuel reprocessing. (Mori, K.)

Progress in fission product nuclear data

International Nuclear Information System (INIS)

Lammer, M.

1982-07-01

This is the eighth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. Therefore, the IAEA cannot be held responsible for the information contained nor for any consequences resulting from the use of this information. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The seventh issue of this series has been published in July 1981 as INDC(NDS)-116. The present issue includes contributions which were received by NDS between 1 August 1981 and 15 June 1982

Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

Energy Technology Data Exchange (ETDEWEB)

B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

2008-07-01

A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

Fission product detection

International Nuclear Information System (INIS)

Liatard, E.; Akrouf, S.; Bruandet, J.F

1987-01-01

The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells [fr

Extraction chromatography of fission products

International Nuclear Information System (INIS)

Bonnevie-Svendsen, M.; Goon, K.

1978-01-01

Various cases of using extraction chromatography during analysis of fission products are reviewed. The use of the extraction chromatography method is considered while analysing reprocessed products of nuclear fuel for quantitative radiochemical analysis and control of fission product and actinoide separation during extraction and their chemical state in production solutions. The method is used to obtain pure fractions of typical burnup monitors (neodymium, molybdenum, cerium, cesium, europium, lanthanides) during determination of nuclear fuel burnup degree. While studying the nature of nuclear reactions the method is used to separate quickly short-life isotopes, to purify β-radiator fractions before measuring their half-life periods, to enrich isotopes forming with low output during fission. Examples of using extraction chromatography are given to separate long half-life or stable fission products from spent solutions, to control environment object contamination

Experiments on The Retention of The Fission Product Iodine in Nuclear Reactor Accidents

International Nuclear Information System (INIS)

Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

2004-01-01

The paper reports PSI participation in the ICHEMM Project (Iodine Chemistry and Mitigation Mechanisms) of the EU 5th Framework Programme, which has the objective of identifying new possible mitigation processes, or accident management measures, favouring the conversion of volatile iodine species, especially organic iodine released during a severe accident at a Nuclear Power Plant (NPP), into nonvolatile forms. Significant effort has been expended, firstly to develop analytical methods for determining concentrations of the common iodine species (iodide, iodate, molecular iodine and methyl iodide), and secondly to provide missing and important information on decomposition phenomena and involved reactions. Experiments have been conducted at PSI to obtain data on thermal (hydrolysis) and radiolytic decomposition of CH 3 I in aqueous solution under conditions as close as possible to those in an anticipated accident. Radioactive-labelled methyl iodide was used to monitor and quantify the decomposition and to determine the overall mass balance. Results have shown that CH 3 I hydrolysis at 90 o C will dominate over radiolysis at the dose rates (max. 0.4 Gy.s -1 ) employed in the experiments. Much increased decomposition rates of CH 3 I have been achieved with chemical reagents (additives), such as sodium thiosulphate or ammonium sulphide. After completion of the ICHEMM Project, further advances have been made from experiments carried out on additives (especially in combination with Aliquat 336) in terms of a fast and complete retention of iodine. The data from these experiments are of direct relevance to severe accident management for both PWR and BWR systems. The impact of the results on possible strategies for management of radioactive waste has also been investigated. (author)

Experiments on The Retention of The Fission Product Iodine in Nuclear Reactor Accidents

Energy Technology Data Exchange (ETDEWEB)

Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B

2004-03-01

The paper reports PSI participation in the ICHEMM Project (Iodine Chemistry and Mitigation Mechanisms) of the EU 5th Framework Programme, which has the objective of identifying new possible mitigation processes, or accident management measures, favouring the conversion of volatile iodine species, especially organic iodine released during a severe accident at a Nuclear Power Plant (NPP), into nonvolatile forms. Significant effort has been expended, firstly to develop analytical methods for determining concentrations of the common iodine species (iodide, iodate, molecular iodine and methyl iodide), and secondly to provide missing and important information on decomposition phenomena and involved reactions. Experiments have been conducted at PSI to obtain data on thermal (hydrolysis) and radiolytic decomposition of CH{sub 3}I in aqueous solution under conditions as close as possible to those in an anticipated accident. Radioactive-labelled methyl iodide was used to monitor and quantify the decomposition and to determine the overall mass balance. Results have shown that CH{sub 3}I hydrolysis at 90{sup o}C will dominate over radiolysis at the dose rates (max. 0.4 Gy.s{sup -1}) employed in the experiments. Much increased decomposition rates of CH{sub 3}I have been achieved with chemical reagents (additives), such as sodium thiosulphate or ammonium sulphide. After completion of the ICHEMM Project, further advances have been made from experiments carried out on additives (especially in combination with Aliquat 336) in terms of a fast and complete retention of iodine. The data from these experiments are of direct relevance to severe accident management for both PWR and BWR systems. The impact of the results on possible strategies for management of radioactive waste has also been investigated. (author)

In-Plant Fission Product Behavior in SGTR Accident with Long-Term SBO

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Woon; Han, Seok Jung; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

2015-05-15

The off-site AC power was recovered in 9 days after the accident in the NPS. Therefore safety injection by fire pump truck with fresh water or seawater is only available in the Fukushima accident. However, safety injection by fire pump truck is not always effective due to the high pressure of RPV inside or leakages of alternative water injection flow paths. In the SBO situations in pressurized water reactor plant (PWR), turbine driven auxiliary feedwater (TD-AFW) pump can inject water to the secondary side of steam generator. However, turbine inlet steam flow control valve cannot work properly when loss of vital DC power occurs. Vital DC power is designed to be maintained during 4 or 8 hours in the SBO conditions. In this paper motor-driven and turbine driven AFW pumps are all assumed to be not working at time 0 sec as a worst case assumption. Iodine pool-scrubbing can occur in the secondary side of the faulted steam generator. However, iodine pool-scrubbing in the secondary side of the faulted steam generator is assumed not to be working, due to the assumption of the loss of DC battery for turbine inlet flow control valve. Iodine pool-scrubbing is one of the long-term research issues in safety assessment of nuclear power plant severe accident. PHEBUS FPT series and THAI experiment projects are typical projects on the resolving source term issues in severe accident of nuclear power plants. However, iodine retention by pool scrubbing is still a debating issue. In such containment bypass sequences, fission products can be released out to environment directly from RCS without retention or deposition in containment structures. SGTR is one of the hazardous accident scenarios in the typical PSA, because SGTR induces a large release amount of source term to environment directly. A key operation strategy was the isolation of the broken reactor coolant system loop from the intact loop. Typical core degradation in SGTR scenarios occurs with multiple failures of the isolation

Investigation of short-living fission products from the spontaneous fission of Cf-252

International Nuclear Information System (INIS)

Klonk, H.

1976-01-01

In this paper, a method of separating and measuring fission products of Cf-252 is presented. The measurement was achieved by means of γ-spectrometry and thus provides a quantitative analysis with a good separation of the fission products with respect to both atomic number Z and mass number A. The separation of the fission products from the fission source was achieved by means of solid traps. An automatic changing apparatus made it possible to keep irradiation and measuring times short, so even very short-lived fission products could be registered. The quantitative evaluation of primary fission products was made possible by correction according to Bateman equations. With that, the yields of single nuclides and the dispersion of charge can be determined. (orig./WL) [de

Radiolytical oxidation of gaseous iodine by beta radiation

International Nuclear Information System (INIS)

Kaerkelae, Teemu; Auvinen, Ari; Kekki, Tommi; Kotiluoto, Petri; Lyyraenen, Jussi; Jokiniemi, Jorma

2015-01-01

Iodine is one of the most radiotoxic fission product released from fuel during a severe nuclear power plant accident. Within the containment building, iodine compounds can react e.g. on the painted surfaces and form gaseous organic iodides. In this study, it was found out that gaseous methyl iodide (CH 3 I) is oxidised when exposed to beta radiation in an oxygen containing atmosphere. As a result, nucleation of aerosol particles takes place and the formation of iodine oxide particles is suggested. These particles are highly hygroscopic. They take up water from the air humidity and iodine oxides dissolve within the droplets. In order to mitigate the possible source term, it is of interest to understand the effect of beta radiation on the speciation of iodine.

Radiolytical oxidation of gaseous iodine by beta radiation

Energy Technology Data Exchange (ETDEWEB)

Kaerkelae, Teemu; Auvinen, Ari; Kekki, Tommi; Kotiluoto, Petri; Lyyraenen, Jussi [VTT Technical Research Centre of Finland, Espoo (Finland); Jokiniemi, Jorma [VTT Technical Research Centre of Finland, Espoo (Finland); Eastern Finland Univ., Kuopio (Finland)

2015-07-01

Iodine is one of the most radiotoxic fission product released from fuel during a severe nuclear power plant accident. Within the containment building, iodine compounds can react e.g. on the painted surfaces and form gaseous organic iodides. In this study, it was found out that gaseous methyl iodide (CH{sub 3}I) is oxidised when exposed to beta radiation in an oxygen containing atmosphere. As a result, nucleation of aerosol particles takes place and the formation of iodine oxide particles is suggested. These particles are highly hygroscopic. They take up water from the air humidity and iodine oxides dissolve within the droplets. In order to mitigate the possible source term, it is of interest to understand the effect of beta radiation on the speciation of iodine.

Study of the production of neutron-rich isotope beams issuing from fissions induced by fast neutrons; Etude de la production de faisceaux riches en neutrons par fission induite par neutrons rapides

Energy Technology Data Exchange (ETDEWEB)

Lau, Ch

2000-09-15

This work is a contribution to the PARRNe project (production of radioactive neutron-rich isotopes). This project is based on the fission fragments coming from the fission of 238-uranium induced by fast neutrons. The fast neutron flux is produced by the collisions of deutons in a converter. Thick targets of uranium carbide and liquid uranium targets have been designed in order to allow a quick release of fission fragments. A device, able to trap on a cryogenic thimble rare gas released by the target, has allowed the production of radioactive nuclei whose half-life is about 1 second. This installation has been settled to different deuton accelerators in the framework of the European collaboration SPIRAL-2. A calibration experiment has proved the feasibility of fixing an ISOL-type isotope separator to a 15 MV tandem accelerator, this installation can provide 500 nA deutons beams whose energy is 26 MeV and be a valuable tool for studying fast-neutron induced fission. Zinc, krypton, rubidium, cadmium, iodine, xenon and cesium beams have been produced in this installation. The most intense beams reach 10000 nuclei by micro-coulomb for 26 MeV deutons. An extra gain of 2 magnitude orders can be obtained by using a more specific ion source and by increasing the thickness of the target. Another extra gain of 2 magnitude orders involves 100 MeV deutons.

On-line measurement of gaseous iodine species during a PWR severe accident

Energy Technology Data Exchange (ETDEWEB)

Haykal, I.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, 91191 Gif sur Yvette Cedex, (France); Perrin, A. [CNRS-University of Paris Est and Paris 7, Laboratoire Inter-Universitaire des Systemes Atmospheriques, 94010 Creteil, (France); Vincent, B. [University of Burgundy, Laboratoire de physique, CNRS UMR 5027, 9, Avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Manceron, L. [Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); Mejean, G. [University of Joseph Fourier in Grenoble, Laboratoire de Spectrometrie Physique-CNRS UMR 5588, 38402 Saint Martin d' Heres, (France); Ducros, G. [CEA Cadarache, CEA, DEN, Departement d' Etudes des Combustibles, 13108 Saint-Paul-lez-Durance cedex, (France)

2015-07-01

A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still not clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)

Transient fission-product release during reactor shutdown and startup

International Nuclear Information System (INIS)

Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

1997-12-01

Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

Fission Product Library and Resource

Energy Technology Data Exchange (ETDEWEB)

Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-09-29

Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

Attachment of gaseous fission products to aerosols

International Nuclear Information System (INIS)

Skyrme, G.

1985-01-01

Accidents may occur in which the integrity of fuel cladding is breached and volatile fission products are released to the containment atmosphere. In order to assess the magnitude of the subsequent radiological hazard it is necessary to know the transport behaviour of such fission products. It is frequently assumed that the fission products remain in the gaseous phase. There is a possibility, however, that they may attach themselves to particles and hence substantially modify their transport properties. This paper provides a theoretical assessment of the conditions under which gaseous fission products may be attached to aerosol particles. Specific topics discussed are: the mass transfer of a gaseous fission product to an isolated aerosol particle in an infinite medium; the rate at which the concentration of fission products in the gas phase diminishes within a container as a result of deposition on a population of particles; and the distribution of deposited fission product between different particle sizes in a log-normal distribution. It is shown that, for a given mass, small particles are more efficient for fission product attachment, and that only small concentrations of such particles may be necessary to achieve rapid attachment. Conditions under which gaseous fission products are not attached to particles are also considered, viz, the competing processes of deposition onto the containment walls and onto aerosol particles, and the possibility of the removal of aerosols from the containment by various deposition processes, or agglomeration, before attachment takes place. (author)

The coupled kinetics of grain growth and fission product behavior in nuclear fuel under degraded-core accident conditions

International Nuclear Information System (INIS)

Rest, J.

1985-01-01

The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, and cesium release from (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests (performed at Oak Ridge National Laboratory) and (2) trace-irratiated LWR fuel during severe-fuel-damage (SFD) tests (performed in the PBF reactor in Idaho). A theory of grain boundary sweeping of gas bubbles has been included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of determining whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away. In addition, as FASTGRASS-VFP provides for a mechanistic calculation of intra- and intergranular fission product behavior, the coupled calculation between fission gas behavior and grain growth is kinetically comprehensive. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. The effect of fuel oxidation by steam on fission product and grain growth behavior is also considered. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted. (orig.)

Progress in fission product nuclear data. No. 14

International Nuclear Information System (INIS)

Lammer, M.

1994-06-01

This is the 14th issue of a report series on Fission Product Nuclear Data published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of fission product yields, neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data from neutron induced and spontaneous fission, lumped fission product data. The first part of the report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. The third part contains requirements for further measurements

Progress in fission product nuclear data. No. 13

International Nuclear Information System (INIS)

Lammer, M.

1990-11-01

This is the 13th issue of a report series published by the Nuclear Data Section of the IAEA. The types of activities included are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission), neutron reaction cross-sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products and bumped fission product data (decay heat, absorption, etc.). The first part of the report consists of unaltered original data which the authors have sent to IAEA/NDS. The second part contains some recent references relative to fission product nuclear data, which were not covered by the contributions submitted, and selected papers from conferences. Part 3 contains requirements for further measurements

Fission-product release during accidents

International Nuclear Information System (INIS)

Hunt, C.E.L.; Cox, D.S.

1991-09-01

One of the aims when managing a reactor accident is to minimize the release of radioactive fission products. Release is dependent not only on the temperature, but also on the partial pressure of oxygen. Strongly oxidizing atmospheres, such as those that occurred during the Chernobyl accident, released semi-volatile elements like ruthenium, which has volatile oxides. At low temperatures, UO 2 oxidization to U 3 O 8 can result in extensive breakup of the fuel, resulting in the release of non-volatile fission products as aerosols. Under less oxidizing conditions, when hydrogen accumulates from the zirconium-water reaction, the resulting low oxygen partial pressure can significantly reduce these reactions. At TMI-2, only the noble gases and volatile fission products were released in significant quantities. A knowledge of the effect of atmosphere as well as temperature on the release of fission products from damaged reactor cores is therefore a useful, if not necessary, component of information required for accident management

Grain boundary sweeping and liquefaction-induced fission product behavior in nuclear fuel under severe-core damage accident conditions

International Nuclear Information System (INIS)

Rest, J.

1984-05-01

The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from: (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests performed at Oak Ridge National Laboratory; and (2) trace-irradiated and high-burnup LWR fuel during severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. Whereas fuel liquefaction leads to an enhanced release of fission products in trace-irradiated fuel, the occurrence of fuel liquefaction in high-burnup fuel can degrade fission product release. This phenomenon is due in part to reduced gas-bubble mobilities in a viscous medium as compared to vapor transport, and in part to a degradation of grain growth rates and the subsequent decrease in grain-boundary sweeping of intragranular fission products into the liquefied lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and high-burnup fuel. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted

Modelling iodine behaviour using LIRIC 3.0

Energy Technology Data Exchange (ETDEWEB)

Wren, J C; Glowa, G A; Ball, J M [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-01

The overall objective of the iodine chemistry research program at the Whiteshell Laboratories of AECL is to develop and validate the LIRIC (Library of Iodine Reactions In Containment) model. The model, once validated, is intended as either a stand-alone analytical tool or for incorporation into a code for licensing analyses of fission-product behaviour in containment. LIRIC is currently being used to assess the role and importance of individual phenomena on iodine volatility under reactor accident conditions and, thus, help to establish priorities within the iodine research program. The LIRIC model has undergone significant alterations since it was last reported (LIRIC 2.0), mainly as a result of considerable development in understanding of iodine behaviour over the last few years. The new version, LIRIC 3.0, has been used to simulate various results from the Radioiodine Test Facility (RTF) with reasonable success, although under somewhat limited conditions.

Chemical Production using Fission Fragments

International Nuclear Information System (INIS)

Dawson, J. K.; Moseley, F.

1960-01-01

Some reactor design considerations of the use of fission recoil fragment energy for the production of chemicals of industrial importance have been discussed previously in a paper given at the Second United Nations International Conference on the Peaceful Uses of Atomic Energy [A/Conf. 15/P.76]. The present paper summarizes more recent progress made on this topic at AERE, Harwell. The range-energy relationship for fission fragments is discussed in the context of the choice of fuel system for a chemical production reactor, and the experimental observation of a variation of chemical effect along the length of a fission fragment track is described for the irradiation of nitrogen-oxygen mixtures. Recent results are given on the effect of fission fragments on carbon monoxide-hydrogen gas mixtures and on water vapour. No system investigated to date shows any outstanding promise for large-scale chemical production. (author) [fr

Behavior of solid fission products in irradiated fuel

International Nuclear Information System (INIS)

Song, Ung Sup; Jung, Yang Hong; Kim, Hee Moon; Yoo, Byun Gok; Kim, Do Sik; Choo, Yong Sun; Hong, Kwon Pyo

2004-01-01

Many fission products are generated by fission events in UO 2 fuel under irradiation in nuclear reactor. Concentration of each fission product is changed by conditions of neutron energy spectrum, fissile material, critical thermal power, irradiation period and cooling time. Volatile materials such as Cs and I, the fission products, degrade nuclear fuel rod by the decrease of thermal conductivity in pellet and the stress corrosion cracking in cladding. Metal fission products (white inclusion) make pellet be swelled and decrease volume of pellet by densification. It seems that metal fission products are filled in the pore in pellet and placed between UO 2 lattices as interstitial. In addition, metal oxide state may change structural lattice volume. Considering behavior of fission products mentioned above, concentration of them is important. Fission products could be classified as bellows; solid solution in matrix : Sr, Zr, Nb, Y, La, Ce, Pr, Nd, Pm, Sm - metal precipitates : Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sb, Te - oxide precipitates : Ba, Zr, Nb, Mo, (Rb, Cs, Te) - volatile and gases : Kr, Xe, Br, I, (Rb, Cs, Te)

Recovery and use of fission product noble metals

International Nuclear Information System (INIS)

Jensen, G.A.; Rohmann, C.A.; Perrigo, L.D.

1980-06-01

Noble metals in fission products are of strategic value. Market prices for noble metals are rising more rapidly than recovery costs. A promising concept has been developed for recovery of noble metals from fission product waste. Although the assessment was made only for the three noble metal fission products (Rh, Pd, Ru), there are other fission products and actinides which have potential value

A study of fission product transport from failed fuel during N reactor postulated accidents

International Nuclear Information System (INIS)

Hagrman, D.L.

1989-09-01

This report presents a study of fission product transport behavior in N Reactor during a severe accident. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The current report is an extension to a previous interum study that has added an aerosol formation model, replaced an older aerosol deposition model with an improved correlation, and incorporated results of a revised analysis of the process tubes. The LACE LA1 and LA3 tests are used to assess the revised model applied to determine aerosol deposition. The study concludes that a cesium iodide aerosol is likely to form near the downstream end of the process tubes. Transport of most of the released cesium and iodine as well as less volatile material depends on the behavior of this aerosol and the behavior is sensitive to several parameters that are not well known. If the environment is very clean and effluent flow is sufficient to support oxidation of the zircaloy and uranium of the process tubes, almost none of the aerosol deposits in the riser. Reduction of the effluent flow or the presence of high concentrations of aerosols of very low volatile material like zirconium, uranium, or their oxides causes deposition of the fission products in the riser piping. 24 refs., 18 figs., 11 tabs

RSAC, Gamma Doses, Inhalation and Ingestion Doses, Fission Products Inventory after Fission Products Release

International Nuclear Information System (INIS)

Richardson, L.C.

1967-01-01

1 - Description of problem or function: RSAC generates a fission product inventory from a given set of reactor operating conditions and then computes the external gamma dose, the deposition gamma dose, and the inhalation-ingestion dose to critical body organs as a result of exposure to these fission products. Program output includes reactor operating history, fission product inventory, dosages, and ingestion parameters. 2 - Method of solution: The fission product inventory generated by the reactor operating conditions and the inventory remaining at various times after release are computed using the equations of W. Rubinson in Journal of Chemical Physics, Vol. 17, pages 542-547, June 1949. The external gamma dose and the deposition gamma dose are calculated by determining disintegration rates as a function of space and time, then integrating using Hermite's numerical techniques for the spatial dependence. The inhalation-ingestion dose is determined by the type and quantity of activity inhaled and the biological rate of decay following inhalation. These quantities are integrated with respect to time to obtain the dosage. The ingestion dose is related to the inhalation dose by an input constant

The future implications of some long-lived fission product nuclides discharged to the environment in fuel reprocessing wastes

International Nuclear Information System (INIS)

Bryant, P.M.; Jones, J.A.

1972-12-01

Current reprocessing practice leads to the discharge to the environment of virtually all the krypton-85 and tritium, and a large fraction of the iodine-129, formed as fission products in reactor fuel. As nuclear power programmes expand the global inventory of these long-lived nuclides is increasing. The radiological significance of these discharges is assessed in terms of radiation exposure of various population groups during the next few decades. The results of this assessment show that krypton-85 will give higher dose rates than tritium or iodine-129, but that on conventional radiological protection criteria these do not justify taking action to remove krypton-85 from reprocessing plant effluents before the 21st century. (author)

Study of the behaviour of tetracycline as fission products extracting agent

International Nuclear Information System (INIS)

Cunha, I.I.L.

1983-01-01

Both spectrophotometric and potentiometric titration techniques were used to show the formation of complexes between tetracycline and the elements: zirconium, uranium, molybdenum, strontium, barium and ruthenium. It has been verified that tetracycline does not form complexes with cesium, tellurium and iodine. Those techniques have also been used to determine the sites on the tetracycline molecule at which ions may be bound. The behaviour of tetracycline as an extracting agent for those elements, as well as for niobium and technetium has been studied and the influence of the acidity of the aqueous phase upon extraction of the elements mentioned has been considered. Extraction experiments were carried out in the presence of chloride, perchlorate, nitrate and sulfate ions. Studies have been made to determine whether or not the complex extracted into organic phase is really the complex formed between tetracycline and the elements considered as well as to determine the time of shaking necessary so that the equilibrium between the phases is attained. Based on all information obtained from extraction experiments made for uranium and the fission products Zr-95, Nb-95, Ce-141, La-140, Ru-103, Ba-140 and Cs-137, the possibility of using tetracycline for separating those fission products from each other and from uranium has been studies and a scheme for simultaneous separation of those elements has been proposed. The same study has been made for I-131, Tc-99m, Mo-99, Te-132, Np-239 and uranium. The method described is applicable to the separation of some fission products existing in solutions at tracer levels, and not to be used in nuclear fuel reprocessing or any other industrial application. (Author) [pt

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Fission products in glasses. Pt. 2

International Nuclear Information System (INIS)

De, A.K.; Luckscheiter, B.; Malow, G.; Schiewer, E.

1977-09-01

Glass ceramics of different composition with high leach and impact resistance can be produced for fission product solidification. In contrast to commercial glass products, they consist of a number of crystalline phases and a residual glass phase. The major crystalline phase allows a classification into celsian, diopside, encryptite, and perovskite ceramics. They all are of special importance as host phases for long-lived fission products. The paper reports on relations between product composition and melting properties, viscosity, crystallization properties, and fixation capability for fission products. Further investigations deal with dimensional stability, impact resistance, thermal expansion, and thermal conductivity. The properties of the ceramics are compared with those of the basic products. The problems still to be solved with regard to further improvement and application of these products are discussed. (RB) [de

Fission-product retention in HTGR fuels

International Nuclear Information System (INIS)

Homan, F.J.; Kania, M.J.; Tiegs, T.N.

1982-01-01

Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

Release of fission products by a rod with a manufacturing defect (BREAK). The BOUFFON 08 J 4 experiment

International Nuclear Information System (INIS)

Chenebault, P.; Kurka, G.

The BOUFFON 08 experiment consisted of the circuit irradiation in the Siloe reactor of the thermosiphon and the 15 x 15 type rod with a seal defect simulating a faulty solder (weld). The conditions of power, temperature, and pressure were typical of PWR conditions. The fission products released by the faulty rod and the presence of a water cooling circuit have been analyzed using samples. The γ-activity was essentially due to the volatile fission products; it progressively increased over the course of 25 days of irradiation by the rod and reaches 30 to 35 curies per kW of nuclear waste. Over the course of the test, the activities radiated by the tube circuits were one to two orders higher than those that uniquely correspond to the nuclides that were present in the water: the iodine and noble gases were deposited on the walls and this deposit was continuously resupplied by the circulating water. At the end of irradiation, secondary ruptures were apparent at the level of the fuel; these were revealed by a gross increase in the concentrations of solid fission products and by the abnormal distribution of radioactive nuclides in the fuel

Dosimetric measurement of the disintegration rate of fission products

International Nuclear Information System (INIS)

Solymosi, J.; Nagy, L.G.; Zagyvai, P.

1992-01-01

Investigations on the disintegration rate of fission products of 238 U and 239 Pu are presented. The intensity of the β-and γ-radiation of fission products were measured continously in an interval of 1-1300 hours following the fission, offering the possibility for determining the general and specific characteristics of the individual fission products. A universal measuring procedure was elaborated for the rapid in situ determination of the dosimetric features of fission products, which is suitable for the accurate evaluation and prediction of external absorbed dose even in case of fission products of various origin and unknown composition. (author) 6 refs.; 7 figs.; 1 tab

Progress in fission product nuclear data. Information about activities in the field of measurements and compilations/evaluations of fission product nuclear data (FPND)

International Nuclear Information System (INIS)

Lammer, G.

1978-07-01

This is the fourth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.)

Study of the production of neutron-rich isotope beams issuing from fissions induced by fast neutrons

International Nuclear Information System (INIS)

Lau, Ch.

2000-01-01

This work is a contribution to the PARRNe project (production of radioactive neutron-rich isotopes). This project is based on the fission fragments coming from the fission of 238-uranium induced by fast neutrons. The fast neutron flux is produced by the collisions of deutons in a converter. Thick targets of uranium carbide and liquid uranium targets have been designed in order to allow a quick release of fission fragments. A device, able to trap on a cryogenic thimble rare gas released by the target, has allowed the production of radioactive nuclei whose half-life is about 1 second. This installation has been settled to different deuton accelerators in the framework of the European collaboration SPIRAL-2. A calibration experiment has proved the feasibility of fixing an ISOL-type isotope separator to a 15 MV tandem accelerator, this installation can provide 500 nA deutons beams whose energy is 26 MeV and be a valuable tool for studying fast-neutron induced fission. Zinc, krypton, rubidium, cadmium, iodine, xenon and cesium beams have been produced in this installation. The most intense beams reach 10000 nuclei by micro-coulomb for 26 MeV deutons. An extra gain of 2 magnitude orders can be obtained by using a more specific ion source and by increasing the thickness of the target. Another extra gain of 2 magnitude orders involves 100 MeV deutons

Fission product induced swelling of U–Mo alloy fuel

International Nuclear Information System (INIS)

Kim, Yeon Soo; Hofman, G.L.

2011-01-01

Highlights: ► We measured fuel swelling of U–Mo alloy by fission products at temperatures below 250 °C. ► We quantified the swelling portion of U–Mo by fission gas bubbles. ► We developed an empirical model as a function of fission density. - Abstract: Fuel swelling of U–Mo alloy was modeled using the measured data from samples irradiated up to a fission density of ∼7 × 10 27 fissions/m 3 at temperatures below ∼250 °C. The overall fuel swelling was measured from U–Mo foils with as-fabricated thickness of 250 μm. Volume fractions occupied by fission gas bubbles were measured and fuel swelling caused by the fission gas bubbles was quantified. The portion of fuel swelling by solid fission products including solid and liquid fission products as well as fission gas atoms not enclosed in the fission gas bubbles is estimated by subtracting the portion of fuel swelling by gas bubbles from the overall fuel swelling. Empirical correlations for overall fuel swelling, swelling by gas bubbles, and swelling by solid fission products were obtained in terms of fission density.

Contribution to the study of the behaviour, in the urban environment, during the runoff of rainwater, of the fission products emitted during a nuclear accident

International Nuclear Information System (INIS)

Pioch, M.

1993-01-01

In the context of research into the environmental consequences of a serious accident occurring on a pressurized water reactor, this paper concerns the experimental study of behaviour of five fission products (caesium, strontium, iodine, ruthenium and tellurium) in the urban environment under the action of rainwater. Stable or radioactive multiple-element aerosols were produced. Their physicochemical characteristics and their solubility in rainwater were studied. Caesium and rubidium forms solutions totally and quickly, while strontium is partially soluble (approximately 50 %) and iodine is only slightly soluble. The behaviour of fission products on five urban surfaces was then studied. Batch experiments showed that the retention of dissolved forms of radioelements varied according to the material. The reactions involved are ion exchange reactions. The presence of certain ions in water (in particular NH 4 + ) increase the desorption of radioelements. Using a laboratory rainfall simulator, the re-entrainment of fission products by rainwater was examined. Two modes of deposition and two intensities of rainfall were simulated. The desorption of radioelements is greater after wet deposition and remobilization is reduced by an increase in intensity of rainfall. An addition of NH 4 + in water is especially effective in the case of wet depositions. Suggestions are made in order to improve experimental protocols and continue the research. (author). 75 refs., 51 figs., 69 tabs., 14 appends

Insights into the control of the release of iodine, cesium, strontium and other fission products in the containment by severe accident management

International Nuclear Information System (INIS)

2000-03-01

This document is intended to provide a management-level overview of the technical bases for accident management activities to attenuate releases of radioactive materials in the very unlikely event of a severe nuclear power reactor accident - activities known commonly as management of severe accident source terms. Such activities are natural complements to accident management activities directed at arresting or slowing accident progression. Abbreviated, qualitative discussions are presented in the document on the more important severe nuclear reactor accidents, the nature of radioactive material releases during accidents, natural processes that act to attenuate the amount of radioactive material that can escape a power plant, and the physical and chemical principles used in engineered systems to further attenuate radioactive releases during accidents. At the end of each section of the report, an annotated bibliography is provided. These bibliographies are intended to serve as introductions to the vast literature pertinent to all aspects of accident management including the management of radioactive source terms. Finally, it must be noted that much of the presentation has been made from the perspective of conventional pressurized water reactors and boiling water reactors. Many important details will be different for other types of reactors or for reactors with special features. Readers are asked to do the mental manipulations necessary to apply the ideas discussed here to the particular circumstances and features of their own reactors. The report is based on the following outline: - a brief discussion of fission product sources; fission product characteristics; chemical compounds; - transport and deposition of fission products; brief description of different deposition and agglomeration processes; - retention of fission products; re-evaporation, resuspension, etc.; - discussion of various possibilities to enhance the removal of fission products from the containment

Simulation of Fission Product Liftoff Behavior During Depressurization Transients

International Nuclear Information System (INIS)

Tak, Nam-il; Yoon, Churl; Lee, Sung Nam

2016-01-01

As one of crucial technologies for the NHDD project, the development of the GAMMA-FP code is on-going. The GAMMA-FP code is targeted for fission product transport analysis under accident conditions. A well-known experiment named COMEDIE considered two important phenomena, i.e., fission product plateout and liftoff, for fission product transport within the primary circuit of a prismatic high temperature gas cooled reactor. The accumulated fission products on the structural material via the plateout can be liftoff during a blowdown phase after a pipe break accident. Since the fission product liftoff can increase a radioactivity risk, it is important to predict the amount of fission product liftoff during depressurization accidents. In this work, a model for fission product liftoff is implemented into the GAMMA-FP code and the GAMMA-FP code with the implemented model is validated using the COMEDIE blowdown test data. The results of GAMMA-FP show that the GAMMA-FP code can reliably simulate a pressure transient during blowdown phase after a pipe break accident. In addition, a reasonable amount of fission product liftoff was predicted by the GAMMA-FP code. The maximum difference between the measured and predicted liftoff fraction was less than a factor of 10. More in-depth study is required to increase the accuracy of prediction for a fission product liftoff

Fission 99Mo production technology

International Nuclear Information System (INIS)

Miao Zengxing; Luo Zhifu; Ma Huimin; Liang Yufu; Yu Ningwen

2003-01-01

This paper describes a production technology of fission 99 Mo in the Department Isotope, CIAE. The irradiation target is tubular U-Al alloy containing highly enriched uranium. The target is irradiated in the swimming pool reactor core. The neutron flux is about 4x10 13 /cm 2 .sec. The production scale is 3.7-7.4 TBq (100-200Ci) of fission 99 Mo per batch. Total recovery of 99 Mo is more than 70%. The production practice proves that the process and equipment are safe and reliable. (author)

Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

Science.gov (United States)

Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rundberg, R. S.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

2014-09-01

A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

Fission-product SiC reaction in HTGR fuel

International Nuclear Information System (INIS)

Montgomery, F.

1981-01-01

The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

An analysis of the additional fission product release phenomena

International Nuclear Information System (INIS)

Takeda, Tsuneo; Nagai, Hitoshi

1978-09-01

The additional fission product release behavior through a defect hole on the cladding of fuel rods has been studied qualitatively with a computer program CODAC-ARFP. The additional fission product release phenomena are described as qualitative evaluation. The additional fission product release behavior in coolant temperature and pressure fluctuations and in reactor start-up and shut-down depends on coolant water flow behavior into and from the free space of fuel rods through a defect hole. Based on the results of evaluations, the experimental results with an inpile water loop OWL-1 are described in detail. The estimation methods of fission product quantity in the free space and fission product release ratio (quantity released into the coolant/quantity in the free space before beginning of release) are necessary for analysis of the fission product release behavior; the estimation method of water flow through a defect hole is also necessary. In development of the above estimation methods, outpile and capsule experiments supporting the additional fission product release experiments are required. (author)

Development of fission Mo-99 production technology

International Nuclear Information System (INIS)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2000-05-01

Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production

Development of fission Mo-99 production technology

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2000-05-01

Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production.

ENDF/B-5. Fission Product Yields File

International Nuclear Information System (INIS)

Schwerer, O.

1985-10-01

The ENDF/B-5 Fission Product Yields File contains a complete set of independent and cumulative fission product yields, representing the final data from ENDF/B-5 as received at the IAEA Nuclear Data Section in June 1985. Yields for 11 fissioning nuclides at one or more neutron incident energies are included. The data are available costfree on magnetic tape from the IAEA Nuclear Data Section. (author). 4 refs

Iodine behaviour in the SLOWPOKE nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Bekeris, P A; Evans, G J [Toronto Univ., ON (Canada). Dept. of Chemical Engineering and Applied Chemistry

1994-12-31

The purpose of this project is to measure and attempt to explain the presence and volatility of iodine isotopes present as fission products in the SLOWPOKE-2 reactor. Liquid sampling and extraction procedures developed indicated that approximately 40% of the reactor iodine is in the form of iodate (IO{sub 3}{sup -}), and 60% is in the form of iodide (I{sup -}). No appreciable amount in non-polar forms such as molecular iodine (I{sub 2}) or organic iodides (RI) were detected. This goes contrary to past expectations that all of the iodine in the liquid phase would be in the form of I{sup -}. In addition partition coefficients for I-131 were determined as 2-6x10{sup 6} at a neutral pH. Kr-88 is suspected as a possible interfering isotope in the measurement of I-131 in the liquid and gas phases. (author). 9 refs., 2 tabs., 2 figs.

Model for fission-product calculations

International Nuclear Information System (INIS)

Smith, A.B.

1984-01-01

Many fission-product cross sections remain unmeasurable thus considerable reliance must be placed upon calculational interpolation and extrapolation from the few available measured cross sections. The vehicle, particularly for the lighter fission products, is the conventional optical-statistical model. The applied goals generally are: capture cross sections to 7 to 10% accuracies and inelastic-scattering cross sections to 25 to 50%. Comparisons of recent evaluations and experimental results indicate that these goals too often are far from being met, particularly in the area of inelastic scattering, and some of the evaluated fission-product cross sections are simply physically unreasonable. It is difficult to avoid the conclusion that the models employed in many of the evaluations are inappropriate and/or inappropriately used. In order to alleviate the above unfortunate situations, a regional optical-statistical (OM) model was sought with the goal of quantitative prediction of the cross sections of the lighter-mass (Z = 30-51) fission products. The first step toward that goal was the establishment of a reliable experimental data base consisting of energy-averaged neutron total and differential-scattering cross sections. The second step was the deduction of a regional model from the experimental data. It was assumed that a spherical OM is appropriate: a reasonable and practical assumption. The resulting OM then was verified against the measured data base. Finally, the physical character of the regional model is examined

Iodine chemistry at severe accidents. A review and evaluation of the state-of-the-art in the field. APRI 5 report. Part I: Iodine chemistry at hypothetical severe accidents. A review of the state-of-the-art 2003. Part II: A comparison of our knowledge on iodine chemistry and fission products with the current models used in MAAP 4.0.5

International Nuclear Information System (INIS)

Liljenzin, Jan-Olov

2005-01-01

The current report tries to summarize and analyze the state-of-the-art on Iodine chemistry relevant to the conditions expected during severe accidents in nuclear power plants. This has made it necessary to compare a considerable amount of data, new as well as old, in order to try to find the reasons behind some changes in the expected chemical behaviour of Iodine. In a few cases this has been far from simple. Many numerical values are given in this report. However, me numbers given should not be used in a non-critical way because they are often deduced from measurements whose interpretation depends on various kinds of systematic differences and assumptions with regard to technique, 'known' constants, and models applied. The most important observation today is that one can no longer uncritically assume that iodine is only released and transported as cesium iodide. The considerable effect that control rod material (including other construction materials) can have on the way in which an accident develops and on its iodine chemistry is clearly seen from the results of the experiments performed within the PHEBUS FP project. The second part of the report evaluates new knowledge on Iodine chemistry and Iodine behaviour of importance in severe nuclear reactor accidents. Also some new information regarding the behaviour and chemistry of other fission products has been collected. In the light of this information, the current modelling of Iodine behaviour in the MAAP code version 4.0.5 has been investigated. No modelling errors have been found. However, some of the equations used to calculate the vapour pressure of the components in the AlC-alloy used in PWR control rods give questionable results. An error in the MAAP manual was found which should be corrected. Finally, some suggestions are given for future improvements in the modelling of severe accidents used in MAAP for both BWRs and PWRs

Systematics of Fission-Product Yields

International Nuclear Information System (INIS)

Wahl, A.C.

2002-01-01

Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z F = 90 thru 98, mass number A F = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru ∼200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from ∼ 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron (∼ fission spectrum) induced fission reactions

Systematics of Fission-Product Yields

Energy Technology Data Exchange (ETDEWEB)

A.C. Wahl

2002-05-01

Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from {approx} 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron ({approx} fission spectrum) induced fission reactions.

The role of fission products in whole core accidents

Energy Technology Data Exchange (ETDEWEB)

Baker, A R [FRSD, UKAEA, RNPDE, Risley, Warrington (United Kingdom); Teague, H J [SRD, UKAEA, Culcheth, Warrington (United Kingdom)

1977-07-01

The review of the role of fission products in whole-core accidents falls into two parts. Firstly, there is a discussion of the hypothetical accidents usually considered in the UK and how they are dealt with. Secondly, there is a discussion of individual topics where fission products are known to be important or might be so. There is a brief discussion of the UK work on the establishment of an equation of state for unirradiated fuel and how this might be extended to incorporate fission product effects. The main issue is the contribution of fission products to the effective vapour pressure and the experimental programme on the pulsed reactor VIPER investigates this. Fission products may influence the probability of occurrence and the severity of MFCIs. Finally, the fission product effects in the pre-disassembly, disassembly and recriticality stages of an accident are discussed. (author)

Impact of fuel chemistry on fission product behaviour

International Nuclear Information System (INIS)

Poortmans, C.; Van Uffelen, P.; Van den Berghe, S.

1999-01-01

The report contains a series of papers presented at SCK-CEN's workshop on the impact of fuel chemistry on fission product behaviour. Contributing authors discuss different processes affecting the behaviour of fission products in different types of spent nuclear fuel. In addition, a number of papers discusses the behaviour of actinides and fission products released from spent fuel and vitrified high-level waste in geological disposal conditions

Fission product and aerosol behaviour within the containment

International Nuclear Information System (INIS)

Beard, A.M.; Benson, C.G.; Bowsher, B.R.; Dickinson, S.; Nichols, A.L.

1990-04-01

Experimental studies have been undertaken to characterise the behaviour of fission products in the containment of a pressurised water reactor during a severe accident. The following aspects of fission product transport have been studied: (a) aerosol nucleation, (b) vapour transport processes, (c) chemical forms of high-temperature vapours, (d) interaction of fission product vapours with aerosols generated from within the reactor core, (e) resuspension processes, (f) chemistry in the containment. Chemical effects have been shown to be important in defining and quantifying fission product source terms in a wide range of accident sequences. Both the chemical forms of the fission product vapours and their interactions with reactor materials aerosols could have a major effect on the magnitude and physicochemical forms of the radioactive emission from a severe reactor accident. Only the main conclusions are presented in this summary document; detailed technical aspects of the work are described in separate reports listed in the annex

Calculated apparent yields of rare gas fission products

International Nuclear Information System (INIS)

Delucchi, A.A.

1975-01-01

The apparent fission yield of the rare gas fission products from four mass chains is calculated as a function of separation time for six different fissioning systems. A plot of the calculated fission yield along with a one standard deviation error band is given for each rare gas fission product and for each fissioning system. Those parameters in the calculation that were major contributors to the calculated standard deviation at each separation time were identified and the results presented on a separate plot. To extend the usefulness of these calculations as new and better values for the input parameters become available, a third plot was generated for each system which shows how sensitive the derived fission yield is to a change in any given parameter used in the calculation. (U.S.)

Photofission observations in reactor environments using selected fission-product yields

International Nuclear Information System (INIS)

Gold, R.; Ruddy, F.H.; Roberts, J.H.

1982-01-01

A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized

Separation of short-lived fission products

International Nuclear Information System (INIS)

Tamai, Tadaharu; Ohyoshi, Emiko; Ohyoshi, Akira; Kiso, Yoshiyuki; Shinagawa, Mutsuaki.

1976-01-01

A rbief review is presented on the various methods of separation available for both gaseous and liquid states, for the separation of short-lived fission products formed by binary fission of neutron irradiated uranium. The means available for gaseous state are the hot atom reaction, the hydride method and on-line mass separation. For liquid state, use can be made of precipitation, ionic or atomic exchange, solvent extraction and paper electrophoresis. Particular reference is made to electrophoretic separation of ions produced by fission in aqueous solution of uranium. The principle of electrophoretic separation and the procedures for separating the element of interest from the other fission products are outlined, with reference made to the results obtained with the method by the present authors. The elements in question are alkalines, alkaline earths, rare earths, halogens, selenium and

Thermodynamic analysis of volatile organometallic fission products

International Nuclear Information System (INIS)

Auxier II, J.D.; Hall, H.L.; Cressy, Derek

2016-01-01

The ability to perform rapid separations in a post nuclear weapon detonation scenario is an important aspect of national security. In the past, separations of fission products have been performed using solvent extraction, precipitation, etc. The focus of this work is to explore the feasibility of using thermochromatography, a technique largely employed in superheavy element chemistry, to expedite the separation of fission products from fuel components. A series of fission product complexes were synthesized and the thermodynamic parameters were measured using TGA/DSC methods. Once measured, these parameters were used to predict their retention times using thermochromatography. (author)

Yields of fission products produced by thermal-neutron fission of 229Th

International Nuclear Information System (INIS)

Dickens, J.K.; McConnell, J.W.

1983-01-01

Absolute yields have been determined for 47 gamma rays emitted in the decay of 37 fission products representing 25 mass chains created during thermal-neutron fission of 229 Th. Using a Ge(Li) detector, spectra were obtained of gamma rays emitted between 15 min and 0.4 yr after very short irradiations by thermal neutrons of a 15-μg sample of 229 Th. On the basis of measured gamma-ray yields and known nuclear data, yields for cumulative production of 37 fission products were deduced. The absolute overall normalization uncertainty is 235 U, we postulate a simple functional dependence sigma = sigma(Z/sub p/), and using this dependence obtain values of Z/sub p/(A) for 15 mass chains created during fission of 229 Th. Values of Z/sub p/(A) were estimated for other mass chains based upon results of a recent study of Z/sub p/(A). Charge distributions determined using the deduced mass distribution and the deduced sets of Z/sub p/(A) and sigma(Z/sub p/) are in very good agreement with recent measurements, exhibiting a pronounced even-odd effect in elemental yields. These results may be used to predict unmeasured yields for 229 Th fission

NEACRP thermal fission product benchmark

International Nuclear Information System (INIS)

Halsall, M.J.; Taubman, C.J.

1989-09-01

The objective of the thermal fission product benchmark was to compare the range of fission product data in use at the present time. A simple homogeneous problem was set with 200 atoms H/1 atom U235, to be burnt up to 1000 days and then decay for 1000 days. The problem was repeated with 200 atoms H/1 atom Pu239, 20 atoms H/1 atom U235 and 20 atoms H/1 atom Pu239. There were ten participants and the submissions received are detailed in this report. (author)

Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

Science.gov (United States)

Le Gall, C.; Geiger, E.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

2017-11-01

Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

Transport properties of fission product vapors

International Nuclear Information System (INIS)

Im, K.H.; Ahluwalia, R.K.

1983-07-01

Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors

Polarographic determination of Iodide and Iodate, in Solutions Coming from Aerosols in Fission Products Containment Studies in Nuclear Power Stations; Determinacion Polarografica de Especies de Iodo (Ioduro y Iodato) en Soluciones Procedentes de Aerosoles, para Estudios de Contencion de Productos de Fision en Centrales Nucleares

Energy Technology Data Exchange (ETDEWEB)

Sanchez, M.; Ballesteros, O.; Fernandez, M.; Clavero, M.; Gonzalez, A. M. [Ciemat, Madrid (Spain)

2000-07-01

A polarographic method is described for the iodine species determination, iodide and iodate in water solutions. the iodate can be determined by differential pulse polarography. Calibration curves and the detection and determination limits have been obtained. Iodides is oxidized to iodate with sodium hypochlorite and the excess of oxidizing agent is destroyed with sodium sulphide. The concentration of iodide is calculated as the difference between the concentration of iodate in the sample before and after the oxidation. As an application, species of iodine in samples coming from the experimental plants GIRS (Gaseous Iodine Removal by Sprays) of Nuclear Fission Department of the CIEMAT, dedicated to fission products containment studies in nuclear power station, were determined. (Author) 10 refs.

Iodine behaviour in severe accidents

Energy Technology Data Exchange (ETDEWEB)

Dutton, L M.C.; Grindon, E; Handy, B J; Sutherland, L [NNC Ltd., Knutsford (United Kingdom); Bruns, W G; Sims, H E [AEA Technology, Harwell (United Kingdom); Dickinson, S [AEA Technology, Winfrith (United Kingdom); Hueber, C; Jacquemain, D [IPSN/CEA, Cadarache, Saint Paul-Lez-Durance (France)

1996-12-01

A description is given of analyses which identify which aspects of the modelling and data are most important in evaluating the release of radioactive iodine to the environment following a potential severe accident at a PWR and which identify the major uncertainties which affect that release. Three iodine codes are used namely INSPECT, IODE and IMPAIR, and their predictions are compared with those of the PSA code MAAP. INSPECT is a mechanistic code which models iodine behaviour in the aqueous aerosol, spray water and sump water, and the partitioning of volatile species between the aqueous phases and containment gas space. Organic iodine is not modelled. IODE and IMPAIR are semi-empirical codes which do not model iodine behaviour in the aqueous aerosol, but model organic iodine. The fault sequences addressed are based on analyses for the Sizewell `B` design. Two types of sequence have been analysed.: (a) those in which a major release of fission products from the primary circuit to the containment occur, e.g. a large LOCAS, (b) those where the release by-passes the containment, e.g. a leak into the auxiliary building. In the analysis of the LOCA sequences where the pH of the sump is controlled to be a value of 8 or greater, all three codes predict that the oxidation of iodine to produce gas phase species does not make a significant contribution to the source term due to leakage from the reactor building and that the latter is dominated by iodide in the aerosol. In the case where the pH of the sump is not controlled, it is found that the proportion of gas phase iodine increases significantly, although the cumulative leakage predicted by all three codes is not significantly different from that predicted by MAAP. The radiolytic production of nitric acid could be a major factor in determining the pH, and if the pH were reduced, the codes predict an increase in gas phase iodine species leaked from the containment. (author) 4 figs., 7 tabs., 13 refs.

Correlation of recent fission product release data

International Nuclear Information System (INIS)

Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

1989-01-01

For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

JNDC nuclear data library of fission products

International Nuclear Information System (INIS)

Tasaka, Kanji; Ihara, Hitoshi; Akiyama, Masatsugu; Yoshida, Tadashi; Matumoto, Zyun-itiro; Nakasima, Ryuzo

1983-10-01

The JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library for 1172 fission products is described in this report. The gross theory of beta decay has been used extensively for estimating unknown decay data and also some of known decay data with poor accuracy. The calculated decay powers of fission products using the present library show excellent agreement with the latest measurements at ORNL (Oak Ridge National Laboratory), LANL (Los Alamos National Laboratory) and UTT (University of Tokyo, Tokai) for cooling times shorter than 10 3 s after irradiation. The calculated decay powers by the existing libraries showed systematic deviations at short cooling times; the calculated beta and gamma decay powers after burst fission were smaller than the experimental results for cooling times shorter than 10 s, and in the cooling time range 10 to 10 3 s the beta-decay power was larger than the measured values and the gamma decay power smaller than the measured results. The present JNDC FP nuclear data library resolved these discrepancies in the short cooling time ranges. The decay power of fission products has been calculated for ten fission types and the results have been fitted by an analytical function with 31 exponentials. This permits the easy application of the present results of decay power calculations to a LOCA (Loss-of-Coolant Accident) analysis of a light water reactor and so on. (author)

Volatilization and reaction of fission products in flowing steam

International Nuclear Information System (INIS)

Johnson, I.; Steidl, D.V.; Johnson, C.E.

1985-01-01

The principal risk to the public from nuclear power plants derives from the highly radioactive atoms (fission products) generated as energy is produced in the nuclear fuel. The revolatilization of fission products from reactor system surfaces due to self-heating by radioactive decay has become a complicating factor in the source-term redefinition effort. It has had a major impact on calculations of fission product distributions in accident safety analyses. The focus of this research effort was to investigate the volatilization and transport of fission products and control rod materials in a flowing gaseous steam-hydrogen mixture. Fission product and control rod materials in various combinations were studied including CsI, CsOH, TeO 2 , SrO, Ag, In, Cd and Mn. The vaporization behavior of the deposits were characterized with respect to vaporization rates, chemical species and downstream transport behavior

The behavior of fission products during nuclear rocket reactor tests

International Nuclear Information System (INIS)

Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

1991-01-01

Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

Guidelines on the use of stable iodine as a prophylactic measure during nuclear emergencies

International Nuclear Information System (INIS)

1995-09-01

Among the fission products that may be released in a power reactor accident the radio iodines are unique in that the dose received by persons exposed by inhalation of radio iodines in the gaseous plume may be substantially reduced. GMA-9 provides guidance on the medical aspects of the use of stable iodine compounds as a prophylactic measure in the event of a nuclear accident. A review of the physiologic basis for the use of stable iodine as a prophylactic measure and the effects of radiation on the thyroid gland are provided. Logistic factors that must be considered to provide the optimum level of radiological protection and medical safety are also addressed. 71 refs., 6 tabs., 2 figs

Guidelines on the use of stable iodine as a prophylactic measure during nuclear emergencies

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

Among the fission products that may be released in a power reactor accident the radio iodines are unique in that the dose received by persons exposed by inhalation of radio iodines in the gaseous plume may be substantially reduced. GMA-9 provides guidance on the medical aspects of the use of stable iodine compounds as a prophylactic measure in the event of a nuclear accident. A review of the physiologic basis for the use of stable iodine as a prophylactic measure and the effects of radiation on the thyroid gland are provided. Logistic factors that must be considered to provide the optimum level of radiological protection and medical safety are also addressed. 71 refs., 6 tabs., 2 figs.

Fuel morphology effects on fission product release

International Nuclear Information System (INIS)

Osetek, D.J.; Hartwell, J.K.; Cronenberg, A.W.

1986-01-01

Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations for the observed differences are offered that relate fuel morphology changes to the releases

ELSA: A simplified code for fission product release calculations

International Nuclear Information System (INIS)

Manenc, H.; Notley, M.J.

1996-01-01

During a light water reactor severe accident, fission products are released from the overheated core as it progressively degrades. A new computer module named ELSA is being developed to calculate fission product release. The authors approach is to model the key phenomena, as opposed to more complete mechanistic approaches. Here they present the main features of the module. Different release mechanisms have been identified and are modeled in ELSA, depending on fission product volatility: diffusion seems to govern the release of the highly volatile species if fuel oxidation is properly accounted for, whereas mass transport governs that of lower volatility fission products and fuel volatilization that of the practically involatile species

Simulation of fission products behavior in severe accidents for advanced passive PWR

International Nuclear Information System (INIS)

Tong, L.L.; Huang, G.F.; Cao, X.W.

2015-01-01

Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

Analysis of iodine chemical form noted from severe fuel damage experiments

International Nuclear Information System (INIS)

Cronenberg, A.W.; Osetek, D.J.

1986-01-01

Data from the TMI-2 accident has shown that only small amounts of iodine (I) escaped the plant. The postulated reason for such limited release is the formation of CsI (a salt) within fuel, which remains stable in a reducing high-temperature steam-H 2 environment. Upon cooldown CsI would dissolve in water condensate to form an ionic solution. However, recent data from fuel destruction experiments indicate different iodine release behavior that is tied to fuel burnup and oxidation conditions, as well as fission product concentration levels in the steam/H 2 effluent. Analysis of the data indicate that at low-burnup conditions, atomic I release from fuel is favored. Likewise, at low fission product concentration conditions HI is the favored chemical form in the steam/H 2 environment, not CsI. Results of thermochemical equilibria and chemical kinetics analysis support the data trends noted from the PBF-SFD tests. An a priori assumption of CsI for risk analysis of all accident sequences may therefore be inappropriate

Contribution to the study of the behaviour, in the urban environment, during the runoff of rainwater, of the fission products emitted during a nuclear accident; Contribution a l`etude du devenir, en milieu urbain, pendant le ruissellement des eaux pluviales, des produits de fission emis en cas d`accident nucleaire

Energy Technology Data Exchange (ETDEWEB)

Pioch, M

1993-05-24

In the context of research into the environmental consequences of a serious accident occurring on a pressurized water reactor, this paper concerns the experimental study of behaviour of five fission products (caesium, strontium, iodine, ruthenium and tellurium) in the urban environment under the action of rainwater. Stable or radioactive multiple-element aerosols were produced. Their physicochemical characteristics and their solubility in rainwater were studied. Caesium and rubidium forms solutions totally and quickly, while strontium is partially soluble (approximately 50 %) and iodine is only slightly soluble. The behaviour of fission products on five urban surfaces was then studied. Batch experiments showed that the retention of dissolved forms of radioelements varied according to the material. The reactions involved are ion exchange reactions. The presence of certain ions in water (in particular NH{sub 4}{sup +}) increase the desorption of radioelements. Using a laboratory rainfall simulator, the re-entrainment of fission products by rainwater was examined. Two modes of deposition and two intensities of rainfall were simulated. The desorption of radioelements is greater after wet deposition and remobilization is reduced by an increase in intensity of rainfall. An addition of NH{sub 4}{sup +} in water is especially effective in the case of wet depositions. Suggestions are made in order to improve experimental protocols and continue the research. (author). 75 refs., 51 figs., 69 tabs., 14 appends.

Separation of fission Molybdenum for production of technetium generator

International Nuclear Information System (INIS)

Bayat, L.; Shaham, V.; Davarkha, R.

2002-01-01

There are two basically different methods for Mo-99 productions: Activation of Mo-99 contained at about 24% in natural isotopic mixtures. Mo-98 enriched targets are irradiated in high-flux reactors in order to achieve the highest possible specific activity of the product. Idolisation of fission molybdenum from irradiated nuclear fuel targets which have undergone short-term cooling. Maximum fission yield can be attained by irradiation of uranium-235 with the highest possible enrichment. On account of its approximately 1000 times higher specific activity. Fission molybdenum has almost replaced worldwide the product fabricated by activation. However, fission molybdenum-99 production has as its prerequisite a suitably advanced technology by which the production process taking place under high activity conditions can be controlled. An integral part of the process consist in the retention of the fission gases the recycling of non-consumed fuel and the treatment of the waste streams arising. This publication will deal with the individual steps in the process

Characteristics of fission product release from a molten pool

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2001-01-01

The volatile fission products are released from the debris pool, while the less volatile fission products tend to remain as condensed phases because of their low vapor pressure. The release of noble gases and the volatile fission products is dominated by bubble dynamics. The release of the less volatile fission products from the pool can be analyzed based on mass transport through a liquid with the convection flow. The physico-numerical models were orchestrated from existing submodels in various disciplines of engineering to estimate the released fraction of fission products from a molten pool. It was assumed that the pool has partially filled hemispherical geometry. For the high pool pressure, the diameter of the bubbles at detachment was calculated utilizing the Cole and Shulman correlation with the effect of system pressure. Sensitivity analyses were performed and results of the numerical calculations were compared with analysis results for the TMI-2 accident. (author)

Nuclear fission and fission-product spectroscopy: 3. International workshop on nuclear fission and fission-product spectroscopy

International Nuclear Information System (INIS)

Goutte, Heloise; Fioni, Gabriele; Faust, Herbert; Goutte, Dominique

2005-01-01

The present book contains the proceedings of the third workshop in a series of workshops previously held in Seyssins in 1994 and 1998. The meeting was jointly organized by different divisions of CEA and two major international laboratories. In the opening address, Prof. B. Bigot, the French High Commissioner for Atomic Energy, outlined France's energy policy for the next few decades. He emphasized the continuing progress of nuclear fission in both technical and economic terms, allowing it to contribute to the energy needs of the planet even more in the future than it does today. Such progress implies a very strong link between fundamental and applied research based on experimental and theoretical approaches. The workshop gathered the different nuclear communities studying the fission process, including topics as the following: - nuclear fission experiments, - spectroscopy of neutron rich nuclei, - fission data evaluation, - theoretical aspects of nuclear fission, - and innovative nuclear systems and new facilities. The scientific program was suggested by an International Advisory Committee. About 100 scientists from 13 different countries attended the conference in the friendly working atmosphere of the Castle of Cadarache in the heart of the Provence. The proceedings of the workshop were divided into 11 sections addressing the following subject matters: 1. Cross sections and resonances (5 papers); 2. Fission at higher energies - I (5 papers); 3. Fission: mass and charge yields (4 papers); 4. Light particles and cluster emission (4 papers); 5. Spectroscopy of neutron rich nuclei (5 papers); 6. Resonances, barriers, and fission times (5 papers); 7. Fragment excitation and neutron emission (4 papers); 8. Mass and energy distributions (4 papers); 9. Needs for nuclear data and new facilities - I (4 papers); 10. Angular momenta and fission at higher Energies - II (3 papers); 11. New facilities - II (2 papers). A poster session of 8 presentations completed the workshop

Studies of iodine concentration in steel and transition metals

International Nuclear Information System (INIS)

Kormann, C.; Kozlowski, W.; Oleksi-Frenzel, J.; Nachtigall, K.; Neste, A. van; Welsh, M.; Titze-Zaeske, B.; Plieth, W.

1990-01-01

Radioactive iodine which originates from nuclear fuel reprocessing plants as a fission product and consists mainly of the radioactive isotopes 129 I, 131 I, and of the inactive 127 I, must be quantitatively adsorbed at the end of the process for the purpose of safe waste disposal. Inspite of using high-alloy austenitic chrome nickel steels, major corrosion effects were observed at tubes and containers of the PASSAT dissolver waste gas purifier. This research project serves to clarify the question in which parts of the installation iodine concentration is to be expected under certain conditions. Furthermore it serves to identify the redox state of iodine in the various installation components. For this purpose steel (1.4306esu and 1.4563N), zirconium and titanium were studied in nitric iodic media. With the objective of calculating iodine adsorption isotherms, the following methods were used to obtain qualitative and quantitative data: FTIR, ellipsometry, 131 I radiotracer method, cyclo-voltametry, closed circuit potential measurements, ion chromatography. (orig./DG) [de

Determination of the fission products yields, lanthanide and yttrium, in the fission of 238U with neutrons of fission spectra

International Nuclear Information System (INIS)

Nicoli, I.G.

1981-06-01

A radiochemical investigation is performed to measure the cumulative fission product yields of several lantanides and yttrium nuclides in the 238 U by fission neutron spectra. Natural and depleted uranium are irradiated under the same experimental conditions in order to find a way to subtract the contribution of the 235 U fission. 235 U percentage in the natural uranium was 3.5 times higher than in the depleted uranium. Uranium oxides samples are irradiated inside the core of the Argonaut Reactor, at the Instituto de Engenharia Nuclear, and the lantanides and yttrium are chemically separated. The fission products gamma activities were detected, counted and analysed in a system constituted by a high resolution Ge(Li) detector, 4096 multichannel analyser and a PDP-11 computer. Cumulative yields for fission products with half-lives between 1 to 33 hours are measured: 93 Y, 141 La, 142 La, 143 Ce and 149 Nd. The chain total yields are calculated. The cumulative fission yields measured for 93 Y, 141 La, 142 La, 143 Ce and 149 Nd are 4,49%, 4,54%, 4,95%, 4,16% and 1,37% respectively and they are in good agreement with the values found in the literature. (Author) [pt

Gas-phase transport of fission products

International Nuclear Information System (INIS)

Tang, I.N.; Munkelwitz, H.R.

1982-01-01

The paper presents the results of an experimental investigation to show the importance of nuclear aerosol formation as a mechanism for semi-volatile fission product transport under certain postulated HTGR accident conditions. Simulated fission product Sr and Ba as oxides are impregnated in H451 graphite and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperatures. Increasing carrier-gas flow rate greatly enhances the extent of particulate transport. The release and transport of simulated fission product Ag as metal are also investigated. Electron microscopic examinations of the collected Sr and Ag aerosols show large agglomerates composed of primary particles roughly 0.06 to 0.08 μm in diameter

Pool scrubbing models for iodine components

Energy Technology Data Exchange (ETDEWEB)

Fischer, K [Battelle Ingenieurtechnik GmbH, Eschborn (Germany)

1996-12-01

Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs.

Pool scrubbing models for iodine components

International Nuclear Information System (INIS)

Fischer, K.

1996-01-01

Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs

Continuous radiochemical analysis of fission products in a nuclear reactor water coolant

International Nuclear Information System (INIS)

Moskvin, L.N.; Zakharov, L.K.; Leont'ev, G.G.; Mel'nikov, V.A.; Orlenkov, I.S.; Slutskij, G.K.

1975-01-01

Method for continuous radiochemical analysis of I, Cs, Ba, Sr and Ce isotopes in a reactor water heat-transfer agent was developed. A continuous two-dimensional chromatographic process of complex mixtures separation of substances proved to be feasible on several parallel sorbent layers, which moved at constant velocities and separated by stationary intermediate collectors. Tests on model solutions containing I, Ce, Cs and Ba isotopes and on heat-carrier samples showed quantitative separation of elements. The results were indicative of a basic possibility of using multisorbent chromatographs for continuous control of multicomponent mixtures, particularly for control of radioactive fission product compositions in water heat-transfer agents in nuclear power plants. A diagram is shown for a two-dimensional chromatographic separation of a multicomponent mixture. Also shown is a flow chart of an installation for continuous control of iodine and cesium isotope activities

Transport of fission products in matrix and graphite

International Nuclear Information System (INIS)

Hoinkis, E.

1983-06-01

In the past years new experimental methods were applied to or developed for the investigation of fission product transport in graphitic materials and to characterization of the materials. Models for fission product transport and computer codes for the calculation of core release rates were improved. Many data became available from analysis of concentration profiles in HTR-fuel elements. New work on the effect on diffusion of graphite corrosion, fast neutron flux and fluence, heat treatment, chemical interactions and helium pressure was reported on recently or was in progress in several laboratories. It seemed to be the right time to discuss the status of transport of metallic fission products in general, and in particular the relationship between structural and transport properties. Following a suggestion a Colloquium was organized at the HMI Berlin. Interdisciplinary discussions were stimulated by only inviting a limited number of participants who work in different fields of graphite and fission product transport research. (orig./RW)

Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

International Nuclear Information System (INIS)

Appelhans, A.D.

1983-01-01

A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

Fission product release from defected nuclear reactor fuel elements

International Nuclear Information System (INIS)

Lewis, B.J.

1983-01-01

The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO 2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO 2 (1.56 x 10 -10 to 7.30 x 10 -9 s -1 ), as well as escape rate constants (7.85 x 10 -7 to 3.44 x 10 -5 s -1 ) and diffusion coefficients (3.39 x 10 -5 to 4.88 x 10 -2 cm 2 /s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

Iodine behaviour under LWR accident conditions: Lessons learnt from analyses of the first two Phebus FP tests

International Nuclear Information System (INIS)

Girault, N.; Dickinson, S.; Funke, F.; Auvinen, A.; Herranz, L.; Krausmann, E.

2006-01-01

The International Phebus Fission Product programme, initiated in 1988 and performed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN), investigates through a series of in-pile integral experiments, key phenomena involved in light water reactor (LWR) severe accidents. The tests cover fuel rod degradation and the behaviour of fission products released via the primary coolant circuit into the containment building. The results of the first two tests, called FPT0 and Ftp, carried out under low pressure, in a steam rich atmosphere and using fresh fuel for Ftp and fuel burned in a reactor at 23 GWdt -1 for Ftp, were immensely challenging, especially with regard to the iodine radiochemistry. Some of the most important observed phenomena with regard to the chemistry of iodine were indeed neither predicted nor pre-calculated, which clearly shows the interest and the need for carrying out integral experiments to study the complex phenomena governing fission product behaviour in a PWR in accident conditions. The three most unexpected results in the iodine behaviour related to early detection during fuel degradation of a weak but significant fraction of volatile iodine in the containment, the key role played by silver rapidly binding iodine to form insoluble AgI in the containment sump and the importance of painted surfaces in the containment atmosphere for the formation of a large quantity of volatile organic iodides. To support the Phebus test interpretation small-scale analytical experiments and computer code analyses were carried out. The former, helping towards a better understanding of overall iodine behaviour, were used to develop or improve models while the latter mainly aimed at identifying relevant key phenomena and at modelling weaknesses. Specific efforts were devoted to exploring the potential origins of the early-detected volatile iodine in the containment building. If a clear explanation has not yet been found, the non-equilibrium chemical

Regulatory simplification of fission product chemistry

International Nuclear Information System (INIS)

Read, J.B.J.; Soffer, L.

1986-01-01

The requirements for design provisions intended to limit fission product escape during reactor accidents have been based since 1962 upon a small number of simply-stated assumptions. These assumptions permeate current reactor regulation, but are too simple to deal with the complex processes that can reasonably be expected to occur during real accidents. Potential chemical processes of fission products in severe accidents are compared with existing plant safety features designed to minimize off-site consequences, and the possibility of a new set of simply-stated assumptions to replace the 1982 set is discussed

Study on the calculation method of source term from fission products

International Nuclear Information System (INIS)

Zhou Jing; Gong Quan; Qiu Haifeng

2014-01-01

As a major part of radioactive nuclides, fission products play an important role in nuclear power plant design. The paper analyzes the calculation model of core activity inventory, the model of fission products releasing from the pellets to RCS, the balance model of fission products in RCS, and then proves them by calculation of the typical pressurized water reactor. The model is proved applicable for calculating fission products of pressurized water reactors. (authors)

Fission product yield measurements using monoenergetic photon beams

Science.gov (United States)

Krishichayan; Bhike, M.; Tonchev, A. P.; Tornow, W.

2017-09-01

Measurements of fission products yields (FPYs) are an important source of information on the fission process. During the past couple of years, a TUNL-LANL-LLNL collaboration has provided data on the FPYs from quasi monoenergetic neutron-induced fission on 235U, 238U, and 239Pu and has revealed an unexpected energy dependence of both asymmetric fission fragments at energies below 4 MeV. This peculiar FPY energy dependence was more pronounced in neutron-induced fission of 239Pu. In an effort to understand and compare the effect of the incoming probe on the FPY distribution, we have carried out monoenergetic photon-induced fission experiments on the same 235U, 238U, and 239Pu targets. Monoenergetic photon beams of Eγ = 13.0 MeV were provided by the HIγS facility, the world's most intense γ-ray source. In order to determine the total number of fission events, a dual-fission chamber was used during the irradiation. These irradiated samples were counted at the TUNL's low-background γ-ray counting facility using high efficient HPGe detectors over a period of 10 weeks. Here we report on our first ever photofission product yield measurements obtained with monoenegetic photon beams. These results are compared with neutron-induced FPY data.

Fission product yield measurements using monoenergetic photon beams

Directory of Open Access Journals (Sweden)

Krishichayan

2017-01-01

Full Text Available Measurements of fission products yields (FPYs are an important source of information on the fission process. During the past couple of years, a TUNL-LANL-LLNL collaboration has provided data on the FPYs from quasi monoenergetic neutron-induced fission on 235U, 238U, and 239Pu and has revealed an unexpected energy dependence of both asymmetric fission fragments at energies below 4 MeV. This peculiar FPY energy dependence was more pronounced in neutron-induced fission of 239Pu. In an effort to understand and compare the effect of the incoming probe on the FPY distribution, we have carried out monoenergetic photon-induced fission experiments on the same 235U, 238U, and 239Pu targets. Monoenergetic photon beams of Eγ = 13.0 MeV were provided by the HIγS facility, the world's most intense γ-ray source. In order to determine the total number of fission events, a dual-fission chamber was used during the irradiation. These irradiated samples were counted at the TUNL's low-background γ-ray counting facility using high efficient HPGe detectors over a period of 10 weeks. Here we report on our first ever photofission product yield measurements obtained with monoenegetic photon beams. These results are compared with neutron-induced FPY data.

Special syntheses of certain organic iodine compounds

International Nuclear Information System (INIS)

Henry, R.; Debuchy, D.; Junod, E.

1960-01-01

The technical difficulties encountered in working on radioactive products force us to choose the simplest methods of chemical synthesis possible. For iodine compounds, two special methods have been chosen: - by using fission recoil, we can prepare simple iodine compounds such as iodobenzene or methyl iodide in high yields and having a good degree of purity. The method consists in the irradiation of mixtures of uranium oxide and benzoic acid or ammonium acetate. The iodised product is separated by distillation, after dissolving the recoil medium in a solvent. - by isotopic exchange between the inorganic iodine of different valencies and complex molecules such as Bengal pink, and diodone, it is also possible to obtain satisfactory labelling yields. These reactions have been adapted so as to give a minimum time for isotopic exchange. In the case of Bengal pink, we have found a yield of 90 per cent after 60 minutes by exchange between Nal and the organic molecule in aqueous solution in presence of hydrogen peroxide. For diodone the method proposed by Liebster has been modified so as to reduce losses during purification. The analytical methods adopted for these different compounds are described. (author) [fr

Fission-product releases from a PHWR terminal debris bed

Energy Technology Data Exchange (ETDEWEB)

Brown, M.J.; Bailey, D.G., E-mail: morgan.brown@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

ENDF/B-5 Fission Products Library 1979

International Nuclear Information System (INIS)

Schwerer, O.; Lemmel, H.D.

1981-10-01

This document summarizes contents and documentation of the 1979 version of the Fission Products File of the ENDF/B Library maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

Radioactive iodine and environmental and sanitary effects - bibliographic study and quantification

International Nuclear Information System (INIS)

Guetat, Ph.; Armand, P.; Monfort, M.; Fritsch, P.; Flury Herard, A.; Menetrier, F.; Bion, L.; Schoech, C.; Masset, S.

2004-01-01

This document is intended to a large public. It reviews the different parameters needed to evaluate the potential act o radioactive releases from the emission to public. Its objectives are to evaluate the importance of different exposure pathways and to assess efficiency of the possible interventions for large public. The main conclusions are summarised hereafter: The radioactive decay chains have to be taken into account to evaluate the iodine source term in the nuclear plants in the case of fission accidents. The physico-chemical forms of iodine are important in order to determine the released activity and deposited activity on the soil. The isotopes to be taken into account are mainly iodine 131 for radiological assessments and also iodine 133 for the nuclear reactor accidents, and the chain Tellurium-Iodine 132 when no particulate filtration exists. Iodine 129 in French reprocessing plant cannot lead to significant accidents. The dominant exposure pathways are related to the consumption of contaminated food products (vegetable, milk) for the inorganic iodine. The iodine transfer to goat and sheep milk is greater than the one to cow milk. The meat production of herbivores at field is the most sensitive. The interest to remove rapidly herbivore from pasture appears relatively clearly. The banning of consumption of local contaminated food products (vegetables and meats) may reduce by about a factor of thirteen the impact due to iodine 131. The youngest the population is, the greatest are the thyroid radiosensitivity and variability within the population. Oral administration of stable iodine limits transfers to maternal milk and foetal thyroid. Ingestion of stable iodine is complementary to consumption banning of local contaminated food products. The earliest the ingestion is, the greatest is the efficiency. 0,1 TBq of 131 iodine released at a low height involves only limited and local actions whereas the release of 10 TBq involves direct and immediate protection

Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

International Nuclear Information System (INIS)

Selby, H.D.; Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C.

2010-01-01

We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99 Mo, 95 Zr, 137 Cs, 140 Ba, 141,143 Ce, and 147 Nd. Modest incident-energy dependence exists for the 147 Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ∼5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except

Microprobe study of fission product behavior in high-burnup HTR fuels

International Nuclear Information System (INIS)

Kleykamp, H.

Electron microprobe analysis of irradiated coated particles with high burnup (greater than 50 percent fima) gives detailed information on the chemical state and the transport behavior of the fission products in UO 2 and UC 2 kernels and in the coatings. In oxide fuel kernels, metallic inclusions and ceramic precipitations are observed. The solubility behavior of the fission products in the fuel matrix has been investigated. Fission product inclusions could not be detected in carbide fuel kernels; post irradiation annealed UC 2 kernels, however, give information on the element combinations of some fission product phases. Corresponding to the chemical state in the kernel, Cs, Sr, Ba, Pd, Te and the rare earths are released easily and diffuse through the entire pyrocarbon coating. These fission products can be retained by a silicon carbide layer. The initial stage of a corrosive attack of the SiC coating by the fission products is evidenced

Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

International Nuclear Information System (INIS)

Gauld, I.C.

2005-01-01

U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k eff ) to determine the net importance of cross sections to k eff . The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: 151 Sm, 103 Rh, 155 Eu, 150 Sm, 152 Sm, 153 Eu, 154 Eu, and 143 Nd

Attachment behavior of fission products to solution aerosol

Energy Technology Data Exchange (ETDEWEB)

Takamiya, Koichi; Tanaka, Toru; Nitta, Shinnosuke; Itosu, Satoshi; Sekimoto, Shun; Oki, Yuichi; Ohtsuki, Tsutomu [Research Reactor Institute, Kyoto University, Osaka (Japan)

2016-12-15

Various characteristics such as size distribution, chemical component and radioactivity have been analyzed for radioactive aerosols released from Fukushima Daiichi Nuclear Power Plant. Measured results for radioactive aerosols suggest that the potential transport medium for radioactive cesium was non-sea-salt sulfate. This result indicates that cesium isotopes would preferentially attach with sulfate compounds. In the present work the attachment behavior of fission products to aqueous solution aerosols of sodium salts has been studied using a generation system of solution aerosols and spontaneous fission source of {sup 248}Cm. Attachment ratios of fission products to the solution aerosols were compared among the aerosols generated by different solutions of sodium salt. A significant difference according as a solute of solution aerosols was found in the attachment behavior. The present results suggest the existence of chemical effects in the attachment behavior of fission products to solution aerosols.

Apparatus for measuring the release of fission gases and other fission products by degassing

Energy Technology Data Exchange (ETDEWEB)

Stradal, Karl Alfred

1970-10-15

In gas-cooled high-temperature reactors, the fuel is, in general, inserted in the fuel elements in the form of small particles, which are, for example, coated with pyrolytic carbon. The purpose of this coating is to keep the fission products separate from the coolant gas. The further development of these coated particles makes it necessary to check the retention capacity. One possible method of doing this is the degassing test after irradiation in the reactor. An apparatus is described below, which was developed and installed in order to measure to a higher degree of sensitivity and in serial measurements the release of fission gases and sparingly volatile fission products.

Sorption of radionuclides on geologic media - A literature survey. I: Fission Products

International Nuclear Information System (INIS)

Andersson, K.; Allard, B.

1983-01-01

The fission products investigated were cobalt, nickel, strontium, cesium, technetium and iodine. Parameters of importance to sorption have been identified and a tabulation of distribution coefficients for groundwater conditions (pH 7-9, low to medium ionic strength) is included in the report. For cobalt and nickel the sorption is related to hydrolysis. High sorption is observed at a pH where hydrolysis become important and the sorption is high as long as cationic hydrolysis products are formed. When pH is increased or negatively charged hydrolysis products may be formed and the sorption decreases. For strontium hydrolysis is of no importance at the normal pH of groundwater, but in groundwater above pH 9 carbonate complexation may occur. For most minerals, the sorption is low, ususally with a pronounced pH dependence. Other important parameters are ionic strength and CEC. A nonselective sorption due to electrostatic interactions between negatively charged mineral surfaces and Sr 2+ seems to occur. For cesium no hydrolysis may be expected and pH has less importance than for Sr. For most minerals, however, the sorption of Cs is higher than for Sr. Important parameters are nuclide concentration and ionic strength. A selective for Cs-sorption is found for some minerals, mostly sheet-silicates. For technetium sorption is due to a reduction of TcO 4 - TcO 2 (s) and as anions are poorly sorbed, the sorption is dependent on the redox potential. Iodine is also anionic and poorly sorbed. Minerals containing ions capable of forming iodides with low solubility (Ag, Mg, Pb etc) are, however, sorbing I - . (Author)

Yields of fission products produced by thermal-neutron fission of 245Cm

International Nuclear Information System (INIS)

Dickens, J.K.; McConnell, J.W.

1981-01-01

Absolute yields have been determined for 105 gamma rays emitted in the decay of 95 fission products representing 54 mass chains created during thermal-neutron fission of 245 Cm. These results include 17 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays between 30 sec and 0.3 yr after very short irradiations of thermal neutrons on a 1 μg sample of 245 Cm. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 84 and 156. The absolute overall normalization uncertainty is 239 Pu and for 252 Cf(s.f.); the influences of the closed shells Z=50, N=82 are not as marked as for thermal-neutron fission of 239 Pu but much more apparent than for 252 Cf(s.f.). Information on the charge distribution along several isobaric mass chains was obtained by determining fractional yields for 12 fission products. The charge distribution width parameter, based upon data for the heavy masses, A=128 to 140, is independent of mass to within the uncertainties of the measurements. Gamma-ray assignments were made for decay of short-lived fission products for which absolute gamma-ray transition probabilities are either not known or in doubt. Absolute gamma-ray transition probabilities were determined as (51 +- 8)% for the 374-keV gamma ray from decay of 110 Rh, (35 +- 7)% for the 1096-keV gamma ray from decay of 133 Sb, and (21.2 +- 1.2)% for the 255-keV gamma ray from decay of 142 Ba

Development of fission Mo-99 production technology

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2001-05-01

This R and D project is planed to supply domestic demands of Mo-99 through fission route, and consequently this project will be expected to rise up utilization of HANARO and KAERI's capability for marketing extension into domestic and oversea radiopharmaceutical market. HEU and LEU target types are decided and designed for fission Mo-99 production in domestic. Experimental study of target fabrication technology was performed and developed processing equipments. And conceptual design of target loading/unloading in/from HANARO device are performed. Tracer test of Mo-99 separation and purification process was performed, test results reach to Mo-99 recovery yield above 80% and decontamination factor above 1600. Combined Mo-99 separation and purification process was decided for hot test scheduled from next year, and performance test was performed. Conceptual design for modification of existing hot cell for fission Mo-99 production facility was performed and will be used for detail design. Assumption for the comparison of LEU and HEU target in fission Mo-99 production process were suggested and compared of merits and demerits in view of fabrication technology and economy feasibility.

Development of fission Mo-99 production technology

International Nuclear Information System (INIS)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2001-05-01

This R and D project is planed to supply domestic demands of Mo-99 through fission route, and consequently this project will be expected to rise up utilization of HANARO and KAERI's capability for marketing extension into domestic and oversea radiopharmaceutical market. HEU and LEU target types are decided and designed for fission Mo-99 production in domestic. Experimental study of target fabrication technology was performed and developed processing equipments. And conceptual design of target loading/unloading in/from HANARO device are performed. Tracer test of Mo-99 separation and purification process was performed, test results reach to Mo-99 recovery yield above 80% and decontamination factor above 1600. Combined Mo-99 separation and purification process was decided for hot test scheduled from next year, and performance test was performed. Conceptual design for modification of existing hot cell for fission Mo-99 production facility was performed and will be used for detail design. Assumption for the comparison of LEU and HEU target in fission Mo-99 production process were suggested and compared of merits and demerits in view of fabrication technology and economy feasibility

Fission product release mechanisms and groupings

International Nuclear Information System (INIS)

Iglesia, F.C.; Brito, A.C.; Liu, Y.

1995-01-01

During CANDU postulated accidents the reactor fuel is estimated to be exposed to a variety of conditions. These conditions are dynamic and, during the course of an accident, the fuel may experience a wide range of temperatures and conditions from highly oxidizing to mildly reducing environments. The exposure of the reactor fuel to these environments and temperatures may affect its stoichiometry and release performance. In this paper a review of the important fission product release mechanisms is presented, the results of three out-of-pile experimental programs are summarized, and fission product release groups, for both oxidizing and reducing conditions are proposed. (author)

Transmutation of fission products through accelerator

International Nuclear Information System (INIS)

Nakamura, H.; Tani, S.; Takahashi, T.; Yamamura, O.

1995-01-01

The transmutation of fission products through particle accelerators has been studied under the OMEGA program. The photonuclear reaction has also been investigated to be applied to transmuting long-lived fission products, such as Cesium and Strontium, which have difficulties on reaction with neutrons due to its so small cross section. It is applicable for the transmutation if the energy balance can be improved with a monochromatic gamma rays in the range of the Giant Dipole Resonance generated through an excellent high current electron linear accelerator. The feasibility studies are being conducted on the transmutation system using it through an electron accelerator. (authors)

Plutonium and surrogate fission products in a composite ceramic waste form

International Nuclear Information System (INIS)

Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

1999-01-01

Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known

REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

Science.gov (United States)

Chiotti, P.

1964-02-01

A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

Calculating the mass distribution of heavy nucleus fission product by neutrons

International Nuclear Information System (INIS)

Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.

1981-01-01

The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru

Production of 131 Iodine in research reactors from elementary tellurium

International Nuclear Information System (INIS)

Silva, Constancia Pagano Goncalves da

1970-01-01

Presents the history of the production of iodine-131 in the Institute of Atomic Energy (IEA), SP, Brazil), the preliminary works for the development of the method, which were done in small scale and it was not necessary protection for the operators, the evolution of these operations until the final assembling of the equipment in shielded cells, the efficiency of operations and product purity. The problems linked to the presence of iodate in the preparations and the changes made for elimination of that ion, harmful to many types and use of iodine-131 are examined. The quality of the product delivered today, an average of 140 departures per month, and the number of departures, per year since the beginning of iodine-131 production were indicated

Chemical immobilization of fission products reactive with nuclear reactor components

International Nuclear Information System (INIS)

Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

1975-01-01

This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

Fission Product Sorptivity in Graphite

Energy Technology Data Exchange (ETDEWEB)

Tompson, Jr., Robert V. [Univ. of Missouri, Columbia, MO (United States); Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States); Ghosh, Tushar [Univ. of Missouri, Columbia, MO (United States); Viswanath, Dabir [Univ. of Missouri, Columbia, MO (United States); Walton, Kyle [Univ. of Missouri, Columbia, MO (United States); Haffner, Robert [Univ. of Missouri, Columbia, MO (United States)

2015-04-01

Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one

Development of Commercial-scale Fission Mo-99 Production System

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung-Kon; Lee, Suseung; Hong, Soon-Bog; Jang, Kyung-Duk; Park, Ul Jael; Lee, Jun Sig [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

These days, worldwide {sup 99} Mo supply is not only insufficient but also unstable. Because, most of the main {sup 99}Mo production reactors are more than years old and suffered from frequent and unscheduled shutdown. Therefore, movement to replace old reactors to keep stable supply is now active. Under these conditions, KAERI (Korea Atomic Energy Research Institute) is developing LEU-based fission {sup 99}Mo production process which is connected to the new research reactor (Kijang New Research Reactor, KJRR), which is being constructed in Gijang, Busan, Korea. Historically, the most fission {sup 99}Mo producers have been used highly enriched uranium (HEU) targets so far. However, to reduce the use of HEU in private sector for non-proliferation, {sup 99}Mo producers are forced to convert their HEU-based process to use low enriched uranium (LEU) targets. Economic impact of a target conversion from HEU to LEU is significant. In this study, fission {sup 99}Mo process with non-irradiated LEU targets was presented except separation and purification steps. Pre- and post-irradiation tests of the fission {sup 99}Mo target will be done in 4th quarter of 2016. For the fission Mo production process development, hot experiments with irradiated LEU targets will be done in 4th quarter of 2016. Then, verification of the production process with quality control will be followed until the commercial production of fission {sup 99}Mo scheduled in 2019.

Fission product removal from molten salt using zeolite

International Nuclear Information System (INIS)

Pereira, C.; Babcock, B.D.

1996-01-01

Spent nuclear fuel (SNF) can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. The fuel is dissolved in molten chloride salt. Non-transuranic fission products in the molten salt are ion-exchanged into zeolite A, which is subsequently mixed with glass and consolidated. Zeolite was found to be effective in removing fission product cations from the molten salt. Breakthrough of cesium and the alkaline earths occurred more rapidly than was observed for the rare earths. The effluent composition as a function of time is presented, as well as results for the distribution of fission products along the length of the column. Effects of temperature and salt flow rate are also discussed

Fission Product Yield Study of 235U, 238U and 239Pu Using Dual-Fission Ionization Chambers

Science.gov (United States)

Bhatia, C.; Fallin, B.; Howell, C.; Tornow, W.; Gooden, M.; Kelley, J.; Arnold, C.; Bond, E.; Bredeweg, T.; Fowler, M.; Moody, W.; Rundberg, R.; Rusev, G.; Vieira, D.; Wilhelmy, J.; Becker, J.; Macri, R.; Ryan, C.; Sheets, S.; Stoyer, M.; Tonchev, A.

2014-05-01

To resolve long-standing differences between LANL and LLNL regarding the correct fission basis for analysis of nuclear test data [M.B. Chadwick et al., Nucl. Data Sheets 111, 2891 (2010); H. Selby et al., Nucl. Data Sheets 111, 2891 (2010)], a collaboration between TUNL/LANL/LLNL has been established to perform high-precision measurements of neutron induced fission product yields. The main goal is to make a definitive statement about the energy dependence of the fission yields to an accuracy better than 2-3% between 1 and 15 MeV, where experimental data are very scarce. At TUNL, we have completed the design, fabrication and testing of three dual-fission chambers dedicated to 235U, 238U, and 239Pu. The dual-fission chambers were used to make measurements of the fission product activity relative to the total fission rate, as well as for high-precision absolute fission yield measurements. The activation method was employed, utilizing the mono-energetic neutron beams available at TUNL. Neutrons of 4.6, 9.0, and 14.5 MeV were produced via the 2H(d,n)3He reaction, and for neutrons at 14.8 MeV, the 3H(d,n)4He reaction was used. After activation, the induced γ-ray activity of the fission products was measured for two months using high-resolution HPGe detectors in a low-background environment. Results for the yield of seven fission fragments of 235U, 238U, and 239Pu and a comparison to available data at other energies are reported. For the first time results are available for neutron energies between 2 and 14 MeV.

Trapping technology for gaseous fission products from voloxidation process

International Nuclear Information System (INIS)

Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S.

2005-05-01

The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, 14 C, Kr, Xe, I and 3 H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and 14 C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for 3 H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system

Preparation of a primary target for the production of fission products in a nuclear reactor

International Nuclear Information System (INIS)

Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

1976-01-01

A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

Estimation of penetration depth of fission products in cladding Hull

International Nuclear Information System (INIS)

Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

2005-01-01

A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

Development of fission Mo production technology

International Nuclear Information System (INIS)

Kim, B. K.; Park, K. B.; Jun, B. J.; Park, J. H.; Choung, W. M.; Lee, K. I.; Woo, M. S.; Whang, D. S.; Kim, Y. K.; Yoo, J. H.; Sohn, D. S.; Lee, Y. W.; Na, S. H.; Koo, Y. H.; Hwang, D. H.; Joo, P. K.

1997-08-01

The feasibility study is accomplished in this project for the development of fission moly production. The KAERI process proposed for development in KAERI is discussed together with those of the American Cintichem and Russian IPPE, each of which would be plausible for introduction whenever the indigenous development is not much feasible. For the conceptual design of the KAERI irradiation target, analysis method is set up and some preliminary analysis is performed accordingly for the candidate design. To establish chemical process concepts for the afore-mentioned three processes, characteristics, operation conditions, and the management of the generated wastes are investigated. Basic requirements of hotcell facilities for chemical processing and a possible way of utilizing the existing hotcells are discussed in parallel with the counter-measures for the construction of new hotcell facilities. Various conditions of target irradiation for fission moly production in Hanaro are analyzed. Plan for introduction of the relevant technology introduction and for procurement of highly enriched uranium are considered. On the basis of assuming some conditions, the economic feasibility study for fission moly production is also overviewed. (author). 22 refs., 28 tabs., 24 figs

Removal of volatile iodine from gas bubbles rising in water pools: review and assessment of pool scrubbing codes

Energy Technology Data Exchange (ETDEWEB)

Polo, J; Herranz, L E; Peyres, V; Escudero, M [CIEMAT, Nuclear Technology Institute, Madrid (Spain)

1996-12-01

During a hypothetical nuclear reactor accident with core damage the fission products released from the degrading fuel bundles often pass through aqueous beds before entering the containment, mitigating in part the source term. Several computer codes have been developed for predicting the fission product and aerosols removal in pool scrubbing scenarios. In addition to particle removal, these codes simulate the retention of some volatile iodine compounds. In this work a review of volatile iodine removal models included in SPARC and BUSCA codes is presented. Besides, the results and discussions of a validation of both codes against the available experimental data are summarized. SPARC and BUSCA codes model the diffusion of iodine toward the bubble interface by using the film penetration theory, which assumes a double layer gas-liquid at the interface. However, there are some differences between the two models, mainly related to the boundary conditions in the aqueous volume for the diffusion of molecular iodine. In SPARC, a set of fast reactions in the liquid phase control both the molecular iodine concentration in the pool and the partition coefficient of iodine at the interface. Thus, the aqueous chemistry plays an important role in the boundary conditions for the diffusion process. On the contrary, the BUSCA model has no chemical considerations at all, and assumes a null iodine concentration in the water bulk. Several sensitivity studies have been made in order to weight the effect of these differences. The variables examined in these studies were the pool temperature and the incoming iodine concentration in the pool. Additionally, sensitivity studies focused on the steam mass fraction of the injected gas were performed to study the effect of the different approach of both models for the condensation process. The results showed a different sensitivity of SPARC and BUSCA to the incoming concentration. (author) 5 tabs., 26 refs.

Interaction of noble-metal fission products with pyrolytic silicon carbide

International Nuclear Information System (INIS)

Lauf, R.J.; Braski, D.N.

1982-01-01

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain layers of pyrolytic carbon and silicon carbide, which act as a miniature pressure vessel and form the primary fission product barrier. Of the many fission products formed during irradiation, the noble metals are of particular interest because they interact significantly with the SiC layer and their concentrations are somewhat higher in the low-enriched uranium fuels currently under consideration. To study fission product-SiC interactions, particles of UO 2 or UC 2 are doped with fission product elements before coating and are then held in a thermal gradient up to several thousand hours. Examination of the SiC coatings by TEM-AEM after annealing shows that silver behaves differently from the palladium group

Fission product release mechanisms and groupings

Energy Technology Data Exchange (ETDEWEB)

Iglesia, F C; Brito, A C; Liu, Y [Ontario Hydro, Toronto, ON (Canada); and others

1996-12-31

During CANDU postulated accidents the reactor fuel is estimated to be exposed to a variety of conditions. These conditions are dynamic and, during the course of an accident, the fuel may experience a wide range of temperatures and conditions from highly oxidizing to mildly reducing environments. The exposure of the reactor fuel to these environments and temperatures may affect its stoichiometry and release performance. In this paper a review of the important fission product release mechanisms is presented, the results of three out-of-pile experimental programs are summarized, and fission product release groups, for both oxidizing and reducing conditions are proposed. (author) 92 refs., 6 tabs.

Chemistry of actinides and fission products

International Nuclear Information System (INIS)

Pruett, D.J.; Sherrow, S.A.; Toth, L.M.

1988-01-01

This task is concerned primarily with the fundamental chemistry of the actinide and fission product elements. Special efforts are made to develop research programs in collaboration with researchers at universities and in industry who have need of national laboratory facilities. Specific areas currently under investigation include: (1) spectroscopy and photochemistry of actinides in low-temperature matrices; (2) small-angle scattering studies of hydrous actinide and fission product polymers in aqueous and nonaqueous solvents; (3) kinetic and thermodynamic studies of complexation reactions in aqueous and nonaqueous solutions; and (4) the development of inorganic ion exchange materials for actinide and lanthanide separations. Recent results from work in these areas are summarized here

Treatment and separation of radioactive fission products tritium, rare gases and iodine in nuclear fuel reprocessing plants

Energy Technology Data Exchange (ETDEWEB)

Schnez, H.

1975-07-15

Rare gases must be separated from the process off-gases of the head-end of the Purex and Thorex processes. To achieve high decontamination factors, the quantity of off-gas should be kept as low as possible. For rare gas separation, there are two possible methods of routing the off-gas: (a) the open flushing gas circuit, in which the purified off-gas (generally air) is passed off via the stack and (b) the closed circuit in which the off-gas (nitrogen or rare gases) is recycled to the dissolver after purification. Tritium must not be entrained into the second extraction cycle or be emitted with off-gases in the form of water vapor (HTO) or HT, but must remain completely in the aqueous phase. Most of the process water is recycled, as a result of which the tritium becomes concentrated in it. This tritiated water is then subjected to tritium rectification at a suitable point in the process. Iodine is very difficult to isolate to a small number of process stages. Present aim is to release the iodine in the dissolver stage into the off-gas, so as to prevent it being entrained into the extraction part. By the injection of hot nitrogen or water vapor into the dissolver or into iodine-containing condensates, all of the iodine is passed into the gaseous phase. Scrubbers can also be used together with iodine-containing condensates to adjust the scrubbing solution. Capital cost of separation plants account for 1 to 10 percent of the total cost of the reprocessing installation, and even more if a sophisticated tritium separation system is required. (DLC)

Actinide and fission product separation and transmutation

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-07-01

The second international information exchange meeting on actinide and fission product separation and transmutation, took place in Argonne National Laboratory in Illinois United States, on 11-13 November 1992. The proceedings are presented in four sessions: Current strategic system of actinide and fission product separation and transmutation, progress in R and D on partitioning processes wet and dry, progress in R and D on transmutation and refinements of neutronic and other data, development of the fuel cycle processes fuel types and targets. (A.L.B.)

Standardization of iodine-129 by the TDCR liquid scintillation method and 4π β-γ coincidence counting

Science.gov (United States)

Cassette, P.; Bouchard, J.; Chauvenet, B.

1994-01-01

Iodine-129 is a long-lived fission product, with physical and chemical properties that make it a good candidate for evaluating the environmental impact of the nuclear energy fuel cycle. To avoid solid source preparation problems, liquid scintillation has been used to standardize this nuclide for a EUROMET intercomparison. Two methods were used to measure the iodine-129 activity: triple-to-double-coincidence ratio liquid scintillation counting and 4π β-γ coincidence counting; the results are in good agreement.

RSAC-6, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release

International Nuclear Information System (INIS)

Wenzel, Douglas R.; Schrader, Brad J.

2007-01-01

1 - Description of program or function: RSAC-6 is the latest version of the program RSAC (Radiological Safety Analysis Computer Program). It calculates the consequences of a release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory; decay and in-grow the inventory during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to downwind individuals. Internal dose from the inhalation and ingestion pathways is calculated. External dose from ground surface and plume gamma pathways is calculated. New and exciting updates to the program include the ability to evaluate a release to an enclosed room, resuspension of deposited activity and evaluation of a release up to 1 meter from the release point. Enhanced tools are included for dry deposition, building wake, occupancy factors, respirable fraction, AMAD adjustment, updated and enhanced radionuclide inventory and inclusion of the dose-conversion factors from FOR 11 and 12. 2 - Methods: RSAC6 calculates meteorological dispersion in the atmosphere using Gaussian plume diffusion for Pasquill-Gifford, Hilmeier-Gifford and Markee models. A unique capability is the ability to model Class F fumigation conditions, the meteorological condition that causes the highest ground level concentrations from an elevated release. Doses may be calculated for various pathways including inhalation, ingestion, ground surface, air immersion, water immersion pathways. Dose calculations may be made for either acute or chronic releases. Internal doses (inhalation and ingestion) are calculated using the ICRP-30 model with dose conversion factors from FOR 11. External factors are calculated using FOR 12. 3 - Unusual Features: RSAC6 calculates complete progeny in-growth and decay during all accident phases. The calculation of fission product inventories in particularly useful in the analysis of accidents where the

Properties of Fission-Product decay heat from Minor-Actinide fissioning systems

International Nuclear Information System (INIS)

Oyamatsu, Kazuhiro; Mori, Hideki

2000-01-01

The aggregate Fission-Product (FP) decay heat after a pulse fission is examined for Minor Actinide (MA) fissiles 237 Np, 241 Am, 243 Am, 242 Cm and 244 Cm. We find that the MA decay heat is comparable but smaller than that of 235 U except for cooling times at about 10 8 s (approx. = 3 y). At these cooling times, either the β or γ component of the FP decay heat for these MA's is substantially larger than the one for 235 U. This difference is found to originate from the cumulative fission yield of 106 Ru (T 1/2 = 3.2x10 7 s). This nuclide is the parent of 106 Rh (T 1/2 = 29.8 s) which is the dominant source of the decay heat at 10 8 s (approx. = 3 y). The fission yield is nearly an increasing function of the fissile mass number so that the FP decay heat is the largest for 244 Cm among the MA's at the cooling time. (author)

Analysis of the kinetic behaviour of iodine and caesium isotopes in the primary circuit of LWR's during severe fuel damage accidents

International Nuclear Information System (INIS)

Alonso, A.; Fernandez, S.; Buron, J.M.; Lopez, J.V.

1991-01-01

This State of the Art report deals with the chemical behaviour of caesium and iodine in the primary system, focusing particularly on kinetic chemical aspects. In case of a postulated severe accident in a nuclear reactor, cesium and iodine fission products are among the major contributors to health harm because of their high volatility and radiotoxicity. The extent of the release of such fission products to the environment depends on the effectiveness of transport through different structures in the reactor coolant system and within the reactor building. The release from fuel has been briefly studied; only those aspects concerning to iodine and caesium chemical forms when released have been reviewed; nevertheless the emphasis has been put on the transport of such elements and their species through the primary system. Some thermochemical equilibrium studies, applied to primary circuit conditions in LWR's, have been analyzed. The revision of the few kinetic studies existing on this matter has shown that kinetic behaviour of iodine and caesium isotopes in the primary circuit is an aspect poorly studied, despite the fact that kinetic aspects could have great importance on the chemical species formed under certain conditions. Other phenomena affecting iodine and caesium transport, besides chemical reactions, such as interactions with surfaces, aerosols or other chemical species have also been examined from available information on diverse experiments

A Research Program for Fission Product/Dust Transport in HTGR’s

Energy Technology Data Exchange (ETDEWEB)

Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States)

2016-02-01

High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuel was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.

Fission product behavior in the Molten Salt Reactor Experiment

International Nuclear Information System (INIS)

Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

1975-10-01

Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

BIG-10 fission product generation and reaction rates

International Nuclear Information System (INIS)

Rogers, J.W.

1976-01-01

Fission product generation rates for high quality fission foils and reaction rates of nonfission foils have been measured by gamma ray activation analyses. These foils were irradiated in the BIG-10 facility and the activities were measured by NaI counting techniques

User's manual for computer code RIBD-II, a fission product inventory code

International Nuclear Information System (INIS)

Marr, D.R.

1975-01-01

The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)

Trapping technology for gaseous fission products from voloxidation process

Energy Technology Data Exchange (ETDEWEB)

Shin, Jin Myeong; Park, J. J.; Park, G. I.; Jung, I. H.; Lee, H. H.; Kim, G. H.; Yang, M. S

2005-05-15

The objective of this report is to review the different technologies for trapping the gaseous wastes containing Cs, Ru, Tc, {sup 14}C, Kr, Xe, I and {sup 3}H from a voloxidation process. Based on literature reviews and KAERI's experimental results on the gaseous fission products trapping, appropriate trapping method for each fission product has been selected considering process reliability, simplicity, decontamination factor, availability, and disposal. Specifically, the most promising trapping method for each fission product has been proposed for the development of the INL off-gas trapping system. A fly ash filter is proposed as a trapping media for a cesium trapping unit. In addition, a calcium filter is proposed as a trapping media for ruthenium, technetium, and {sup 14}C trapping unit. In case of I trapping unit, AgX is proposed. For Kr and Xe, adsorption on solid is proposed. SDBC (Styrene Divinyl Benzene Copolymer) is also proposed as a conversion media to HTO for {sup 3}H. This report will be used as a useful means for analyzing the known trapping technologies and help selecting the appropriate trapping methods for trapping volatile and semi-volatile fission products, long-lived fission products, and major heat sources generated from a voloxidation process. It can also be used to design an off-gas treatment system.

An Evaluation of a Fission Product Inventory for CANDU Fuels

International Nuclear Information System (INIS)

Jung, Jong Yeob; Park, Joo Hwan

2007-01-01

Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

Fission product released experiment of coated fuel particles

Energy Technology Data Exchange (ETDEWEB)

Shijiang, Xu; Bing, Yang; Chunhe, Tang; Junguo, Zhu; Jintao, Huang; Binzhong, Zhang [Inst. of Nucl. Energy Technology, Tsinghua Univ., Beijing (China); Jinghan, Luo [Inst. of Atomic Energy, Beijing (China)

1992-01-15

Four samples of coated fuel particles were irradiated in the Heavy-Water Research Reactor of the Institute of Atomic Energy. Each of them was divided into two groups and irradiated to the burn up of 0.394% fima and 0.788% fima in two static capsules, respectively. After irradiation and cooling, post irradiation annealing experiment was carried out, the release ratios of the fission product {sup 133}Xe and {sup 131}I were measured, they are in the order of 10{sup -6}{approx}10{sup -7}. The fission product release ratio of naked kernel was also measured under the same conditions as for the coated fuel particles, the ratio of the fission product release of the coated fuel particles and of the naked kernel was in the order of 10{sup -5}{approx}10{sup -4}.

Repeated radiation injuries by fission products

International Nuclear Information System (INIS)

Vasilenko, I.Ya.

1986-01-01

Attention is given to repeated radiation injuries during internal irradiation of theoretical and practical interest, particularly in case of the intake into organism of young products of nuclear fission (PNF). The results of experiments with dogs with repeated radioactive iodine injury the isotopes of which (131-135sub(I)) constitute a considerable part of PNF activity are discussed. The blood reaction and protein metabolism state have been studied. Observations for dogs have been continued for about 4 years. The doses for thyroid, gastrointestinal tract and liver subjected to the most intensive irradiation consituted in the first series of experiments after the first intake about 3;0.3;0.05 Gy, after the second - 5;0.5;0.08 Gy and in the second series of experiments - 3;0.3;0.05 Gy and 0.6;0.06;0.01 Gy, respectively. Hematologic factors,thyroid function, changes in exchange and immunologic reactivity have been studied. The dogs have been under observation for 5 years. It is shown in case of repeated intake of Isup(131) PNF into animals organism in quantity which does not cause during the acute period a clinically outlined sickness, substantial differences in the organism reaction as compared with the first intake of radionuclides have not been found. The presence of residual radiation injuries did not cause charging action during the acute period during PNF and repeated intake which in the author's opinion testifies to perfection of compensator mechanisms in case of intake of such quantities of radioactive products. At the remote periods blastomogenic action manifested which is estimated as a result of general biological action of radionuclides administered to the organism. The necessity in subsequent investigations for obtaining the data on organism reactivity, clinic and pathogenesis with the aim of prophylaxis and treatment of such injuries is indicated

Fission product yield evaluation for the USA evaluated nuclear data files

International Nuclear Information System (INIS)

Rider, B.F.; England, T.R.

1994-01-01

An evaluated set of fission product yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

Detector for gaseous nuclear fission products

International Nuclear Information System (INIS)

Kobayashi, Yoshihiro; Kubo, Katsumi.

1979-01-01

Purpose: To facilitate the fabrication of a precipitator type detector, as well as improve the reliability. Constitution: Gas to be measured flown in an anode is stored in a gas processing system. By applying a voltage between the anode and the cathode, if positively charged Rb or Cs which is the daughter products of gaseous fission products are present in the gas to be measured, the daughter products are successively deposited electrostatically to the cathode. The daughter products issue beta-rays and gamma-rays to ionize the argon gas at the anode, whereby ionizing current flows between both of the electrodes. Pulses are generated from the ionizing current, and presence or absence, as well as the amount of the gaseous fission products are determined by the value recorded for the number of the pulses to thereby detect failures in the nuclear fuel elements. After the completion of the detection, the inside of the anode is evacuated and the cathode is heated to evaporate and discharge the daughter products externally. This eliminates the effects of the former detection to the succeeding detection. (Moriyama, K.)

Potential for large-scale uses for fission-product Xenon

International Nuclear Information System (INIS)

Rohrmann, C.A.

1983-03-01

Of all fission products in spent, low-enrichment-uranium power-reactor fuels, xenon is produced in the highest yield - nearly one cubic meter, STP, per metric ton. In aged fuels which may be considered for processing in the US, radioactive xenon isotopes approach the lowest limits of detection. The separation from accompanying radioactive 85 Kr is the essential problem; however, this is state-of-the-art technology which has been demonstrated on the pilot scale to yield xenon with pico-curie levels of 85 Kr contamination. If needed for special applications, such levels could be further reduced. Environmental considerations require the isolation of essentially all fission-product krypton during fuel processing. Economic restraints assure that the bulk of this krypton will need to be separated from the much-more-voluminous xenon fraction of the total amount of fission gas. Xenon may thus be discarded or made available for uses at probably very low cost. In contrast with many other fission products which have unique radioactive characteristics which make them useful as sources of heat, gamma and x-rays, and luminescence - as well as for medicinal diagnostics and therapeutics - fission-product xenon differs from naturally occurring xenon only in its isotopic composition which gives it a slightly hgiher atomic weight, because of the much higher concentrations of the 134 Xe and 136 Xe isotopes. Therefore, fission-product xenon can most likely find uses in applications which already exist but which can not be exploited most beneficially because of the high cost and scarcity of natural xenon. Unique uses would probably include applications in improved incandescent light illumination in place of krypton and in human anesthesia

ENDF/B-5 Fission Products Library. Rev. 2

International Nuclear Information System (INIS)

Schwerer, O.; Pronyaev, V.G.; Lemmel, H.D.

1984-07-01

This document summarizes contents and documentation of the 1984 version of the Fission Products Nuclear Data File of the ENDF/B-5 Library (Rev. 2) maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

Application of dynamic pseudo fission products and actinides for accurate burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

1996-09-01

The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)

Immobilization of fission products in phosphate ceramic waste forms

International Nuclear Information System (INIS)

Singh, D.

1996-01-01

The goal of this project is to develop and demonstrate the feasibility of a novel low-temperature solidification/stabilization (S/S) technology for immobilizing waste streams containing fission products such as cesium, strontium, and technetium in a chemically bonded phosphate ceramic. This technology can immobilize partitioned tank wastes and decontaminate waste streams containing volatile fission products

Synergistic effects of iodine and silver ions co-implanted in 6H–SiC

International Nuclear Information System (INIS)

Kuhudzai, R.J.; Malherbe, J.B.; Hlatshwayo, T.T.; Berg, N.G. van der; Devaraj, A.; Zhu, Z.; Nandasiri, M.

2015-01-01

Motivated by the aim of understanding the release of fission products through the SiC coating of fuel kernels in modern high temperature nuclear reactors, a fundamental investigation is conducted to understand the synergistic effects of implanted silver (Ag) and iodine (I) in 6H–SiC. The implantation of the individual species, as well as the co-implantation of 360 keV ions of I and Ag at room temperature in 6H–SiC and their subsequent annealing behaviour has been investigated by Secondary Ion Mass Spectrometry (SIMS), Atom Probe Tomography (APT) and X-ray Photoelectron Spectroscopy (XPS). SIMS and APT measurements indicated the presence of Ag in the co-implanted samples after annealing at 1500 °C for 30 h in sharp contrast to the samples implanted with Ag only. In samples implanted with Ag only, complete loss of the implanted Ag was observed. However, for I only implanted samples, some iodine was retained. APT of annealed co-implanted 6H–SiC showed clear spatial association of Ag and I clusters in SiC, which can be attributed to the observed I assisted retention of Ag after annealing. Such detailed studies will be necessary to identify the fundamental mechanism of fission products migration through SiC coatings. - Highlights: • Co-implantation of Ag and I ions in 6H–SiC was performed. • Clear spatial association of Ag and I clusters observed after annealing. • Complete loss of Ag after high temperature annealing of silver only sample. • Iodine was retained in iodine only sample after high temperature annealing. • Iodine was found to play a role in the retention of Ag in the co-implanted samples.

Utilization of fast reactor excess neutrons for burning long-lived fission products

International Nuclear Information System (INIS)

Kawashima, K.; Kobayashi, K.; Kaneto, K.

1995-01-01

An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

International Nuclear Information System (INIS)

Pierce, V.H.

2005-01-01

1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

Map of calculated radioactivity of fission product, (4)

International Nuclear Information System (INIS)

Takeda, Tsuneo

1978-07-01

The overall radioactivities of fission products depending on irradiation time and cooling time were calculated for 18 different neutron fluxes, which are presented in contour maps and tables. Irradiation condition etc. are the followings: neutron flux (n sub(th)) 1 x 10 12 - 6.8 x 10 14 n/cm 2 /sec, uranium quantity 1 mole (6 x 10 23 atoms, ca. 271 g UO 2 ), U-235 enrichment 2.7%, irradiation time 60. - 6 x 10 7 sec (1 min - 1.9 y), cooling time 0. and 60. - 6 x 10 7 sec (1 min - 1.9 y). The enrichment value represents those for LWRs. To calculate the overall radioactivities, 595 fission product nuclides were introduced. Overall radioactivities calculations were made for 68,000 combinations of irradiation time, cooling time and neutron flux. The many complex decay chains of fission products were treated with CODAC-No.6 computer code. (author)

Proposal to represent neutron absorption by fission products by a single pseudo-fragment

International Nuclear Information System (INIS)

Tsibulya, A.M.; Kochetkov, A.L.; Kravchenko, I.V.; Nikolaev, M.N.

1991-01-01

The concentration of fission products during reactor operation is analyzed. The dependence of a composite fission product capture cross-section as a function of time and on the nature of the A of the fissile nuclide are investigated, and the neutron radiative capture in fission products of a thermal reactor is evaluated. It is concluded that neutron absorption by fission products can be described by pseudo-fragments. (author). 18 refs, 2 figs, 3 tabs

Fission product model for lattice calculation of high conversion boiling water reactor

International Nuclear Information System (INIS)

Iijima, S.; Yoshida, T.; Yamamoto, T.

1988-01-01

A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

Yields of products from thermal-neutron induced fission of 235U

International Nuclear Information System (INIS)

Rudstam, G.; Aagaard, P.; Zwicky, H.U.

1985-01-01

Methods for fission yield determinations at an ISOL-system connected to a nuclear reactor have been developed. The present report contains detailed descriptions both of the experimental techniques and of the method used to correct the experimental yields for the decay of short-lived nuclear species in the delay between production and measurement. The methods have been applied to the determination of the fission yields of 40 fission products including 2 isometric pairs in the light mass region and those of 99 fission products including 25 isometric pairs or triplets in the heavy mass region. For 64 cases this is the first determination published. (author)

Methodology and application of the WIMS-D4M fission product data

International Nuclear Information System (INIS)

Mo, S.C.

1995-01-01

The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the 135 Xe and 149 Sm chains, and a lumped fission product to account for the absorption by fission products not explicitly treated. Burn-up calculations were performed for the ANS MEU core using WIMS and EPRI-CELL cross sections. The calculated eigenvalues and material loadings are in good agreements

Influence of iodine chemistry on source term assessment

International Nuclear Information System (INIS)

Herranz Puebla, L. E.; Lopez Diez, I.; Rodriguez Maroto, J. J.; Martinez Lopez-Alcorocho, A.

1991-01-01

The major goal of a phenomenology analysis of containment during a severe accident situation can be spitted into the following ones: to know the containment response to the different loads and to predict accurately the fission product and aerosol behavior. In this report, the main results coming from the study of a hypothetical accident scenario, based on LA-4 experiment of LACE project, are presented. In order to do it, several codes have been coupled: CONTEMPT4/MOD5 (thermohydraulics), NAUA/MOD5 (aerosol physics) and IODE (iodine chemistry). It has been demonstrated the impossibility of assessing with confidence the Source Term if the chemical conduct of some radionuclides is not taken into account. In particular, the influence on the iodine retention efficiency of the sump of variables such as pH has been proven. (Author)12 refs

Simulation of COMEDIE Fission Product Plateout Experiment Using GAMMA-FP

International Nuclear Information System (INIS)

Tak, Nam-il; Yoon, Churl

2014-01-01

FThis phenomenon is particularly important under a VHTR design with vented low pressure confinement (VLPC), because the vent allows the prompt release of fission products accumulated within the primary circuit to environment during an initial blow-down phase after pipe break accidents. In order to analyze the fission product plateout, an numerical model was developed by Yoo et al. and incorporated into the GAMMA-FP code in the past. The GAMMA-FP model was validated against two experiment data, i.e., VAMPYR-1 and OGL, during the development phase. One of the well-known experiments for fission product plateout is the COMEDIE experiment. In this work, the COMEDIE experiment has been simulated using the GAMMA-FP code to investigate the reliability and applicability of the plateout model of GAMMA-FP. The COMEDIE experiment for fission product plateout was simulated using the GAMMA-FP code in this work. A good agreement was achieved between the measured and predicted plateout activities. The existing solution scheme was modified to allow larger time step size for fission product analysis in order to speed-up the computational time. Nevertheless, the modification of the existing numerical model of GAMMA-FP is necessary when a simulation capability of a long duration of plateout period (e.g., 60 years) is targeted

The universal library of fission products and delayed neutron group yields

International Nuclear Information System (INIS)

Koldobskiy, A.B.; Zhivun, V.M.

1997-01-01

A new fission product yield library based on the Semiempirical method for the estimation of their mass and charge distribution is described. Contrary to other compilations, this library can be used with all possible excitation energies of fissionable actinides. The library of delayed neutron group yields, based on the fission product yield compilation, is described as well. (author). 15 refs, 4 tabs

Iodine speciation in the hot cell effluent gases

International Nuclear Information System (INIS)

Lee, B.S.; Jester, W.A.; Olynyk, J.M.

1990-01-01

The various species of airborne radioiodine can affect the iodine source term of a severe core damage accident because of the different transport and deposition properties. also, the radiobiological hazardness may vary according to their chemical form. The purpose of the work reported in this paper was to characterize the various chemical forms of airborne radioiodine in hot cell effluent gases of a radiopharmaceutical production facility that produces medical radioisotopes from separated fission products of irradiated uranium targets. It is concluded that the methyl iodide is the youngest chemical species in terms of effective decay time age, and the hot cell filter bank is least efficient in removing the methyl iodide

Characterization of wastes from fission 99 Mo production

International Nuclear Information System (INIS)

Endo, L.S.; Dellamano, J.C.

1992-07-01

This work is a preliminary study on waste-streams generated in a fission 99 Mo production plant, their characterization and quantification. The study is based on a plant whose 99 Mo production process is the alkaline dissolution of U-target. The target is made of 1 g of enriched 235 U, therefore most of radionuclides present in the waste-streams are fission products. All the radionuclides inventories were estimated based on ORIGEN-2 Code. The characterization was done as a primary stage for the establishment of waste management plan, which should be subject for further study. (author)

Fission product detection by means of photovoltaic cells

International Nuclear Information System (INIS)

Liatard, E.; Akrouf, S.; Bruandet, J.F.; Fontenille, A.; Glasser, F.; Stassi, P.; Tsan Ung Chan

1988-01-01

The response of photovoltaic cells to heavy ions and fission products have been tested in-beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12 ns (FWHM) has been measured between two cells. (orig.)

Recent developments in iodine chemistry in Canada. Present and future applications to reactor safety

International Nuclear Information System (INIS)

Fluke, R.J.; Weaver, K.R.; Kupferschmidt, W.C.H.; Wren, J.C.; Ball, J.M.; Evans, G.J.

1997-01-01

In Canada, there exists a mature research programme directed towards resolving the nuclear safety and licensing questions surrounding fission product iodine. This integrated programme has had the long term objectives of developing a mechanistic understanding of the important aspects of iodine behaviour in containment, and of producing a mechanistic computer code to predict iodine behaviour under conditions of interest. The research is funded by Canadian Utilities and AECL through the CANDU Owners Group. In parallel with the research, an applications effort is underway to put existing research results into service for the resolution of reactor safety and licensing issues, as well as to identify and develop solutions for more comprehensive problems, such as chemical mitigation strategies to control post-accident iodine release. The present paper describes (a) the present status and objectives of R and D on iodine, its recent evolution and its significant achievements, (b) the analytical tools that are emerging from this programme and (c) existing and planned future applications of the results of the iodine R and D programme. The author concludes with an assessment of the impact that the iodine programme has had on reactor safety and licensing questions in Canada

Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

Energy Technology Data Exchange (ETDEWEB)

Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

2011-07-01

In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

A Research Program for Fission Product/Dust Transport in HTGR's

International Nuclear Information System (INIS)

Loyalka, Sudarshan

2016-01-01

High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuel was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.

THAI test facility for experimental research on hydrogen and fission product behaviour in light water reactor containments

Energy Technology Data Exchange (ETDEWEB)

Gupta, S., E-mail: gupta@becker-technologies.com [Becker Technologies GmbH, Koelner Strasse 6, 65760 Eschborn (Germany); Schmidt, E.; Laufenberg, B. von; Freitag, M.; Poss, G. [Becker Technologies GmbH, Koelner Strasse 6, 65760 Eschborn (Germany); Funke, F. [AREVA GmbH, P.O. Box 1109, 91001 Erlangen (Germany); Weber, G. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Forschungszentrum, Boltzmannstraße 14, 85748 Garching (Germany)

2015-12-01

hydraulics, gas distribution) and ISP-49 (hydrogen combustion), EU-SARNET/SARNET2 code-benchmark exercises involving THAI data on iodine/surface interactions, iodine mass transfer, passive autocatalytic recombiner performance, iodine oxide behaviour and iodine transport in multi-compartment behaviour. The present paper provides an overview of the THAI experiments related to hydrogen and fission products issues performed in the frame of national and international projects. From the comprehensive THAI experimental database, a selection of typical results is presented to illustrate the multi-functionality of the THAI facility and the broad variety of the experimental investigations.

Transmutation of fission products and actinide waste at Hanford

Energy Technology Data Exchange (ETDEWEB)

Daemen, L.L.; Pitcher, E.J.; Russell, G.J. [Los Alamos National Laboratory, NM (United States)

1995-10-01

The authors studied the neutronics of an ATW system for the transmutation of the fission products ({sup 99}Tc in particular) and the type of actinide waste stored in several tanks at Hanford. The heart of the system is a highly-efficient neutron production target. It is surrounded by a blanket containing a moderator/reflector material, as well as the products to be transmuted. The fission products are injected into the blanket in the form of an aqueous solution in heavy water, whereas an aqueous actinides slurry is circulated in the outer part of the blanket. For the sake of definiteness, the authors focussed on {sup 99}Tc (the most difficult fission product to transmute), and {sup 239}Pu, {sup 237}Np, and {sup 241}Am. Because of the low thermal neutron absorption cross-section of {sup 99}Tc, considerable care and effort must be devoted to the design of a very efficient neutron source.

Use of dwell time concept in fission product inventory assessment for CANDU reactors

International Nuclear Information System (INIS)

Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

2003-01-01

A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

Chemistry of fission products for accident analysis

International Nuclear Information System (INIS)

Potter, P.E.

1985-01-01

Current knowledge concerning the chemical state of the fission product elements during the development of accidents in water reactor systems is reviewed in this paper. The fission product elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behavior of Mo, Ru and the lanthanides are also discussed. Some features of the reactions of the various species of these elements with other components of the reactor systems are described. The importance of having an adequate knowledge of thermodynamic data and phase equilibria of relatively simple systems in order to interpret experimental observations on complex multi-component systems is stressed

Heat and fission product transport in molten core material pool with crust

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2005-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

NEANDC specialists meeting on yields and decay data of fission product nuclides

International Nuclear Information System (INIS)

Chrien, R.E.; Burrows, T.W.

1983-01-01

Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information

Removal of volatile iodine from gas bubbles rising in water pools: review and assessment of pool scrubbing codes

International Nuclear Information System (INIS)

Polo, J.; Herranz, L.E.; Peyres, V.; Escudero, M.

1996-01-01

During a hypothetical nuclear reactor accident with core damage the fission products released from the degrading fuel bundles often pass through aqueous beds before entering the containment, mitigating in part the source term. Several computer codes have been developed for predicting the fission product and aerosols removal in pool scrubbing scenarios. In addition to particle removal, these codes simulate the retention of some volatile iodine compounds. Nonetheless, experimental data on the matter are rather scarce and further validation remains to be done. In this work a review of volatile iodine removal models included in SPARC and BUSCA codes is presented. Besides, the results and discussions of a validation of both codes against the available experimental data are summarized. SPARC and BUSCA codes model the diffusion of iodine toward the bubble interface by using the film penetration theory, which assumes a double layer gas-liquid at the interface. However, there are some differences between the two models, mainly related to the boundary conditions in the aqueous volume for the diffusion of molecular iodine. In SPARC, a set of fast reactions in the liquid phase control both the molecular iodine concentration in the pool and the partition coefficient of iodine at the interface. Thus, the aqueous chemistry plays an important role in the boundary conditions for the diffusion process. On the contrary, the BUSCA model has no chemical considerations at all, and assumes a null iodine concentration in the water bulk. Several sensitivity studies have been made in order to weight the effect of these differences. The variables examined in these studies were the pool temperature and the incoming iodine concentration in the pool. Additionally, sensitivity studies focused on the steam mass fraction of the injected gas were performed to study the effect of the different approach of both models for the condensation process. The results showed a different sensitivity of SPARC

Experimental investigations relevant for hydrogen and fission product issues raised by the Fukushima accident

Directory of Open Access Journals (Sweden)

Sanjeev Gupta

2015-02-01

with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine test facility (9.2 m high, 3.2 m in diameter, and 60 m3 volume are discussed in the light of the Fukushima accident.

A new technique to measure fission-product diffusion coefficients in UO2 fuel

International Nuclear Information System (INIS)

Hocking, W.H.; Verrall, R.A.; Bushby, S.J.

1999-01-01

This paper describes a new out-reactor technique for the measurement of fission-product diffusion rates in UO 2 . The technique accurately simulates in-reactor fission-fragment effects: a thermal diffusion that is due to localized mixing in the fission track, radiation-enhanced diffusion that is due to point-defect creation by fission fragments, and bubble resolution. The technique utilizes heavy-ion accelerators - low energy (40 keV to 1 MeV) for fission-product implantation, high energy (72 MeV) to create fission-fragment damage effects, and secondary ion mass spectrometry (SIMS) for measuring the depth profile of the implanted species. Preliminary results are presented from annealing tests (not in the 72 MeV ion flux) at 1465 deg. C and 1650 deg. C at low and high concentrations of fission products. (author)

Chemical reactions of fission products with ethylene using the gas jet technique

International Nuclear Information System (INIS)

Contis, E.T.; Rengan, Krish; Griffin, Henry C.

1994-01-01

An understanding of the nature of the chemical reactions taking place between fission products and their carrier gases, and the designing of a fast separation procedure were the purposes of this investigation. Chemical reactions of short-lived (less than one minute half-life) fission products with carrier gases lead to various chemical species which can be separated in the gas phase. The Gas Jet Facility at the Ford Nuclear Reactor was used to study the yields of volatile selenium and bromine fission products of 235 U using a semi-automatic batch solvent extraction technique. Heptane and water were used as organic and inorganic solvents. A carrier gas mixture of ethylene to pre-purified nitrogen (1 : 3) was used to sweep the fission products from the target to the chemistry area for analysis. The results indicated that the volatile selenium products generated by the interaction of selenium fission fragments with ethylene were predominantly organic in nature (84%), possibly organoselenides. The selenium values were used to resolve the fractions of the bromine nuclides, which come from two major sources, viz., directly from fission and from the beta-decay of selenium. The data showed that the fractions of independent bromine fission products in the organic phase were much lower compared to selenium; the bromine values range from 10 to 22% and varied with mass number. Results indicated that the bromine products were inorganic in nature, as possibly hydrogen chloride. ((orig.))

Development of glass ceramics for the incorporation of fission products

International Nuclear Information System (INIS)

De, A.K.; Luckscheiter, B.; Lutze, W.; Malow, G.; Schiewer, E.

1976-01-01

Spontaneous devitrification of fission-product-containing borosilicate glasses can be avoided by controlled crystallization after melting. Glass ceramics have been developed from a vitrified simulated waste and further improvement of product properties was achieved. In particular perovskite, h-celsian, diopside and eucryptite glass ceramics were prepared. These contained leach resistant host phases which exhibited considerable enrichment of long-lived fission products. All products showed increased impact resistance, but the thermal expansion was only slightly improved

NEANDC specialists meeting on yields and decay data of fission product nuclides

Energy Technology Data Exchange (ETDEWEB)

Chrien, R.E.; Burrows, T.W. (eds.)

1983-01-01

Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information. (WHK)

Reactivity effects of fission product decay in PWRs

International Nuclear Information System (INIS)

Aragones, J.M.; Ahnert, C.

1988-01-01

The purpose of the work reported in this paper is to analyze the effects of fission product chains with radioactive decay on the reactivity in pressurized water reactor (PWR) cores, calculating their accumulation and absorption rates along fuel burnup at continuous operation and after shutdown periods extending from 1 day to a few months. The authors PWR version of the WIMS-D4 code is first used to obtain the individual number densities, absorption rates, and averaged cross sections for every nuclide of the fission product chains with significant decay rates, as a function of fuel burnup at continuous irradiation. Next, by an auxiliary ad hoc code, these data, have been processed together with the required one for fissile nuclides and boron, also taken from WIMS at each burnup step, to calculate the average or effective values relevant for the analysis and the decay and change in overall absorption after several shutdown times. (1) The reactivity effect of fission product decay changes significantly with the shutdown time. The maximum absorption increase by decay is reached in ∼ 10 days' shutdown. (2) The dependence with fuel type, enrichment, and burnup is slight, but the change with previous power density is nearly linear, which might be significant after coast-down in previous cycles. (3) For long shutdown periods, the overall reactivity effect of decay in the three fission product chains considered is much less than if only the samarium peak due to 149 Nd is considered

Vitrification processes for fission product solutions

International Nuclear Information System (INIS)

Bonniaud, R.; Jouan, A.; Moncouyoux, J.P.; Sombret, C.

1982-10-01

The different processes for fission product vitrification in the world are reviewed. Continuous or discontinuous processes, induction or arc heating, in can melting or casting, tests with radioactive or simulated wastes and industrial realizations are described [fr

LOFC fission product release and circulating activity calculations for gas-cooled reactors

International Nuclear Information System (INIS)

Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

1977-01-01

The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

The potential for large scale uses for fission product xenon

International Nuclear Information System (INIS)

Rohrmann, C.A.

1983-01-01

Of all fission products in spent, low enrichment, uranium, power reactor fuels xenon is produced in the highest yield - nearly one cubic meter, STP, per metric ton. In aged fuels which may be considered for processing in the U.S. radioactive xenon isotopes approach the lowest limits of detection. The separation from accompanying radioactive 85 Kr is the essential problem; however, this is state of the art technology which has been demonstrated on the pilot scale to yield xenon with pico-curie levels of 85 Kr contamination. If needed for special applications, such levels could be further reduced. Environmental considerations require the isolation of essentially all fission product krypton during fuel processing. Economic restraints assure that the bulk of this krypton will need to be separated from the much more voluminous xenon fraction of the total amount of fission gas. Xenon may thus be discarded or made available for uses at probably very low cost. In contrast with many other fission products which have unique radioactive characteristics which make them useful as sources of heat, gamma and x-rays and luminescence as well as for medicinal diagnostics and therapeutics fission product xenon differs from naturally occurring xenon only in its isotopic composition which gives it a slightly higher atomic weight, because of the much higher concentrations of the 134 X and 136 Xe isotopes. Therefore, fission product xenon can most likely find uses in applications which already exist but which can not be exploited most beneficially because of the high cost and scarcity of natural xenon. Unique uses would probably include applications in improved incandescent light illumination in place of krypton and in human anesthesia

Fuel performance and fission product behaviour in gas cooled reactors

International Nuclear Information System (INIS)

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

Evaluation and compilation of fission product yields 1993

International Nuclear Information System (INIS)

England, T.R.; Rider, B.F.

1995-01-01

This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993

Evaluation and compilation of fission product yields 1993

Energy Technology Data Exchange (ETDEWEB)

England, T.R.; Rider, B.F.

1995-12-31

This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993.

Proton-fission for the accelerator production of Mo-99

International Nuclear Information System (INIS)

Lagunas-Solar, M.C.; Jungerman, J.A.; Castaneda, C.M.

1993-01-01

The production of Mo-99 (66.0 h) via de U-238(p,f) Mo-99 fission reaction is proposed as a non-reactor source of this essential precursor of 6.6-h Tc-99m, an isotope of wide use of diagnostic nuclear medicine applications. Measurements of the total excitation function for the U-238(p,f) reaction indicated a maximum and fairly constant cross section of 1.4 barns at > 30 MeV. Combining the advances of high-current (mA) H-accelerators with dual beam (dual target) operation, and assuming a 5% fission yield, estimates of Mo-99 reaches 5 to 14 Ci/h at 1 mA. The proton fission production of Mo-99 appears to more advantageous than the reactor produced via evaporation neutron-induced fission. An accelerator method could allow securing ample supply of Mo-99 independently of the current scarce reactor operation, while also simplifying the associated waste management problems as well as some of the environmental concerns

Measuring the noble metal and iodine composition of extracted noble metal phase from spent nuclear fuel using instrumental neutron activation analysis

International Nuclear Information System (INIS)

Palomares, R.I.; Dayman, K.J.; Landsberger, S.; Biegalski, S.R.; Soderquist, C.Z.; Casella, A.J.; Brady Raap, M.C.; Schwantes, J.M.

2015-01-01

Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO 2 fuel dissolved in nitric acid and UO 2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared. - Highlights: • The noble metal phase was chemically extracted from spent nuclear fuel and analyzed non-destructively. • Noble metal phase nuclides and long-lived iodine were identified and quantified using neutron activation analysis. • Activation to shorter-lived radionuclides allowed rapid analysis of long-lived fission products in spent fuel using gamma spectrometry

Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

International Nuclear Information System (INIS)

2009-01-01

This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations

Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

Mo-99 production by fission and future projections

International Nuclear Information System (INIS)

Carranza, E.C.; Novello, A.; Bronca, M.; Cestau, D.; Bavaro, R.; Centurion, R.; Bravo, C.; Bronca, P.; Gualda, E.; Fraguas, F.; Giomi, A.; Ivaldi, L.

2012-01-01

Description of the I-131 and Mo-99 production process: The process starts with the irradiation of uranium-aluminum mini plates in the RA-3, Argentinean Reactor No.3, Ezeiza Atomic Center. In a nuclear reactor there is a constant flow of neutrons and when a neutron with proper energy impacts on a nucleus of U-235, it is absorbed at the same time generate an unstable configuration nuclear. For this reason, the nucleus formed is fission, getting two different atoms. Approximately 6% of the fissions produce Mo-99 and 3% produce I-131; the percentage remaining corresponds to formation of atoms without interest for use in medicine. In conclusion, the objective of the process developed in the Fission Plant, is starting from uranium mini plates, separate the Mo-99 and I-131 generated, the remaining elements formed. - Evolution of Mo-99 Production in the last 10 years: The Fission Mo-99 Plant Production begins routine production of Mo-99 in 1985, using targets made of uranium enriched at 90% U-235. In the 1990s, global concern regarding the use of highly enriched uranium, due to non-proliferation issues, caused the interruption of supply of nuclear material (HEU enriched at 90% of U-235). Following this, Argentina developed target based on low-enriched uranium (less than 20% U-235), becoming in 2002 the first country in the world to produce Mo-99 with LEU targets. From 2002 to date, the activity produced of Mo-99 has been tripled annually (author)

Analysis of fission product release from HTGR core during transient temperature excursion

International Nuclear Information System (INIS)

Saito, Takao; Yamatoya, Naotoshi; Onuma, Mamoru

1978-01-01

The computer program ''FRANC'' was developed to calculate the release activity of fission products from a high-temperature gas cooled reactor (HTGR) core during transient temperature excursions such as a hypothetical loss of forced circulation combined with design basis depressurization. The program utilizes a segmented cylindrical core spatial model with the associated values of the prior fuel irradiation history and temperature conditions. The fission product transport and decay chain behavior is expressed by a set of differential equations. This set of equations describes the entire core inventory of fission products by means of calculated parameters based on the detailed spatial core conditions. The program determines the time-dependent amounts of fission product nuclides escaping from the core into the coolant. Coded in Continuous System Simulation Language (CSSL) with double precision, FRANC showed appropriate results for both short- and long-lived fission product nuclides. The sample calculation conducted by applying the program to a large HTGR indicated that it would take about one hour for noble gases and volatile nuclides to be released to the coolant, and several hours for metalic nuclides. (auth.)

The behavior of radioactive iodine at the time of reactor accident and its counterplan

International Nuclear Information System (INIS)

Murata, Toshifumi

1974-01-01

When an accident occurs in a reactor, very volatile radioactive iodine is most dangerous among fission products. Among the isotopes of radioactive iodine, 131 I which has longer half-life is harmful. Supposing one-tenth of the radioactivity of 10 8 Ci in a reactor of 10 6 Kw heat output is due to 131 I, it weights about 100g. The behavior of the radioactive iodine is greatly subjected to the influence of inside temperature and other conditions. Therefore, very prudent policies are adopted by installing emergency core cooling system, containment vessels, and activated carbon filters. For the emergency core cooling system, water spraying, flooding with low pressure water, and maintaining of water level by high pressure water injection are adopted, while in the containment vessels, measures are taken so as to lower the inside pressure and minimize leakage. (Kobatake, H.)

Ceramic Hosts for Fission Products Immobilization

Energy Technology Data Exchange (ETDEWEB)

Peter C Kong

2010-07-01

Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

Analytical evaluation of fission product sensitivities

International Nuclear Information System (INIS)

Sola, A.

1977-01-01

Evaluating the concentration of a fission product produced in a reactor requires the knowledge of a fairly large number of variables. Sensitivity studies were made to ascertain the important variables. Analytical formulae were developed sufficiently simple to allow numerical computations. Some simplified formulas are also given and they are applied to the following isotopes: 80 Se, 82 Se, 81 Br, 82 Br, 82 Kr, 83 Kr, 84 Kr, 85 Kr, 86 Kr. Their sensitivities to capture cross sections, fission yields, ratios of activation cross sections, half-lives (during and after irradiation), branching ratios, as well as to the neutron flux and to the time are considered

Fission product source term research at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Malinauskas, A.P.

1985-01-01

The purpose of this work is to describe some of the research being performed at ORNL in support of the effort to describe, as realistically as possible, fission product source terms for nuclear reactor accidents. In order to make this presentation manageable, only those studies directly concerned with fission product behavior, as opposed to thermal hydraulics, accident sequence progression, etc., will be discussed

Chemical factors affecting fission product transport in severe LMFBR accidents

International Nuclear Information System (INIS)

Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

1984-10-01

This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

Fuel performance and fission product behaviour in gas cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

Investigation of the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4

International Nuclear Information System (INIS)

Soares, M.I.

1981-12-01

To investigate the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4 tubes, deformation tests under pressure of samples hydrided in autoclave and of samples containing iodine were carried out, in order to simulate the fission product. The same tests were carried out in samples without hydride and iodine contents that were used as reference samples in the temperature range of 650 0 C-950 0 C. The hydrided samples and the samples containing iodine tested at 650 0 C and 750 0 C showed a higher ductility than the samples of reference. The hydrided samples tested at 850 0 C and 950 0 C showed a higher embritlement than the samples of reference and than the samples containing iodine tested at the same temperatures. A mechanical test has been developed to investigate the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4 tubes. The mechanical test were carried out at room temperature. At room temperature the hydrition decreased the ductility of zircaloy-4. At room temperature the sample containing iodine showed a higher ductility than the sample without iodine. The combined action of hydrogen and iodine at room temperature enhanced the embrittlment of the samples zircaloy-4. (Author) [pt

Tables and figures from JNDC Nuclear Data Library of fission products, version 2

International Nuclear Information System (INIS)

Ihara, Hitoshi

1989-11-01

The content of JNDC (Japanese Nuclear Data Committee) FP (Fission Product) Nuclear Data Library version 2 for 1227 fission products is presented in the form of tables and figures. The library is inclusive of evaluated decay data such as decay constant, Q-value, average energies of beta, gamma and internal conversion electron, spin-parity, branching ratio of each decay mode and fission yield. The neutron capture cross-sections are also contained for 166 nuclides. The mass number of the fission product nuclides ranges from A = 66 to A = 172. (author)

Considerations on the influence of fission products in whole core accidents

International Nuclear Information System (INIS)

Meyer Heine, A.; Pattoret, A.; Schmitz, F.

1977-01-01

If the hypothetical Whole Core Accidents which are taken into account in reactor safety analysis can change from one country to another, there is nevertheless a general agreement over their description and main phases. Furthermore the important parameters have also been identified by every laboratory. During the development of such core accidents the role of the fission products in essential. It is not the purpose of this paper to give an exhaustive description of the phases which can be influenced by the fission products, we will try however to focus this study on the most important ones. In a second step we will discuss the equation of state of irradiated fuels; here again one principal preoccupation being to quantify the influence of fission products on reactor accidents. It is not our purpose to enter on the fundamental aspects of the equation of state. The studies and the experimental program launched at the CEA will then be described. Special attention will be directed towards the eventual role of fission products in molten fuel-coolant interactions (MFCls) or the events leading to the initiation of whole core accidents. This paper will be limited to oxide fuels. Whether the whole core accident is initiated by a reactivity defect or a coolant coast-down, one has to deal with four great categories of phenomena. Loss of flow: the power is around the nominal value, while the coolant flow has been reduced by a factor of 5 to 10. This induces boiling and clad weakening. Will the plenum pressure lead to a clad rupture? In case of a rupture, what will be the effect on the voiding of the channel? Transient over power: influence of gases from gaseous and volatile fission products on the fuel movements? MFCIs: Influence of the fission products in the mode of contact between fuel and coolant? Influence on the fuel characteristics. Sodium vapour bubble expansion: influence of the fission products on the heat transfer and eventual condensation of the bubble?

Post-irradiation studies on knock-out and pseudo-recoil releases of fission products from fissioning UO2

International Nuclear Information System (INIS)

Yamagishi, S.; Tanifuji, T.

1976-01-01

By using post-irradiation techniques, in-pile releases of 133 Xe, sup(85m)Kr, 88 Kr, 87 Kr and 138 Xe from UO 2 fissioning at low temperatures below about 200 0 C are studied: these are analyzed into a time-dependent knock-out and time-independent pseudo-recoil releases. For the latter, a 'self knock-out' mechanism is proposed: when a fission fragment loses thoroughly its energy near the UO 2 surface and stops there, it will knock out the surface substances and accordingly the fragment (i.e. the fission product) will be released. The effective thickness of the layer where the self knock-out occurs is found to be approximately 7A. As for the knock-out release, the following is estimated from its dependence on various factors: the knock-out release of fission products occurs from the surface layer with the effective thickness of approximately 20A: the shape of UO 2 matrix knocked out by one fission fragment passing through the surface is equivalent to a cylinder approximately 32A diameter by approximately 27A thick, (i.e. the knock-out coefficient for UO 2 is approximately 660 uranium atoms per knock-out event). On the basis of the above estimations, the conclusions derived from the past in-pile studies of fission gas releases are evaluated. (Auth.)

Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

International Nuclear Information System (INIS)

Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

1997-01-01

Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)

Application of mercury cathode electrolysis to fission-product separation

International Nuclear Information System (INIS)

Besson, A.; Prigent, Y.; Van-Kote, F.

1969-01-01

A method involving controlled potential mercury cathode electrolysis has been developed to separate fission products. It allows the radiochemical determination of Ag, Cd, Pd, Rh, Ru, Sn, Te, Sb and Mo from solutions of fission products highly concentrated in mineral salts. The general procedure consists in three main steps: electrolytic amalgam generation, destruction of amalgams and ultimate purification of elements by other means. Electrolytic operations last about five hours. Chemical yields lie between 10 per cent and 70 per cent. (authors) [fr

SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

International Nuclear Information System (INIS)

Kodaira, Hideki

1990-01-01

1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

The main outcomes of the OECD Behaviour of Iodine (BIP) Project

International Nuclear Information System (INIS)

Glowa, Glenn A.; Moore, Chris J.; Ball, Joanne M.

2013-01-01

Highlights: • Moisture affects the rate of iodine adsorption on paint. • Irradiation of iodine-loaded paint yields methyl iodide. • The BIP project is complimentary to the EPICUR project. • BIP results help with the interpretation of Phébus FP results. - Abstract: An Organisation for Economic Co-operation and Development (OECD) status report on iodine behaviour published in February 2007 concluded that although the understanding of iodine behaviour in containment had advanced considerably over the past several decades, there were still areas where further investigation was warranted. The OECD initiated the Behaviour of Iodine Project (BIP) to investigate two of these areas: • The interaction of iodine with painted surfaces: adsorption of iodine from the gas phase to several containment surfaces was investigated under a variety of conditions, with focus on the role of water on the adsorption of iodine onto epoxy paints. The results show that the relative humidity is very important to the deposition velocity of iodine on paint; higher humidity caused faster deposition. Adsorption parameters are a critical input for containment codes. • The formation of organic iodide from painted surfaces: gas chromatography was used to monitor the evolution of CH 3 I during the irradiation of epoxy coupons pre-loaded with iodine. The results showed that organic iodide formation is greater, and has a different temporal behaviour, when the iodine is absorbed by the paint from the aqueous phase as opposed to from the gas phase. The results also highlighted the importance of destruction of organic iodides by gas phase radiolysis to the concentration of CH 3 I. The BIP organic iodide production experiments were different, but complementary, to the experiments performed in the EPICUR (Experimental Program for Iodine Chemistry Under Radiation) facility. However, both projects share an objective of supporting the explanation of iodine behaviour in the Phébus Fission Product

Solidification of residual fission-product solutions; laboratory studies; Solidification de solutions residuaires concentrees de produits de fission: etudes de laboratoire

Energy Technology Data Exchange (ETDEWEB)

Boniaud, R; Cohen, P; Sombret, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1961-07-01

This paper describes the results obtained, at laboratory scale, during the study of the incorporation of fission products into glasses and synthetic micas. The rate of leaching of fission products from the glass and their volatility during firing were measured. A hot cell was built to complete these results. (author) [French] Ce rapport resume le resultat des etudes faites en laboratoire (activite de l'ordre du millicurie) sur l'incorporation des produits de fission dans des verres et micas synthetiques. On a mesure le taux de lixiviation des produits de fission et leur volatilisation au cours de la cuisson. Une cellule chaude a ete installee pour completer ces resultats au moyen d'essais realises avec une activite superieure. (auteur)

Separation of caesium-137 from fission products using phosphotungstic acid

International Nuclear Information System (INIS)

Murthy, T.S.; Balasubramaniam, K.R.; Ananthakrishnan, M.; Varma, R.N.

1977-01-01

Separation of caesium 137 from fission products using phosphotungstic acid is reported. Phosphotungstate caesium is precipitated as caesium from fission product waste solution in acid medium and subsequently purified. Separation of phosphate and tungstate ions has been done using a typical hydrous oxide like alumina. The exchange capacity of alumina for phosphate and tungstate ions, and the purity of the product are determined. Results are discussed. Based on the findings a procedure is recommended for caesium 137 separation. (A.K.)

State of the art report on iodine chemistry

International Nuclear Information System (INIS)

Clement, B.; Cantrel, L.; Ducros, G.; Funke, F.; Herranz, L.; Rydl, A.; Weber, G.; Wren, C.

2007-01-01

An accident in which the normal core cooling is lost could lead to fuel elements melting and fission product release beyond the plant limits. Nuclear power plants are designed with engineering systems and associated operational procedures which provide an in-depth defence against such accidents. Iodine is a major contributor to the potential source term to the environment, thus a good understanding of its behaviour and validated calculation tools are required to perform meaningful risk analyses and make decisions in the field of accident management, mitigation measures and emergency procedures. A number of experimental programmes, involving separate-effect and integral tests have been carried out during the last decade, providing new and valuable results that have improved our understanding of iodine phenomena. A modelling effort has also been pursued in order to encapsulate the acquired knowledge in the calculation tools prepared for predicting the iodine behaviour under severe accident conditions. In view of the progress made, the Working Group on Analysis and Management of Accidents (GAMA) considered the necessity of producing a status paper on iodine chemistry, with the following objectives: - to review insights gained and evaluate the progress made during the last 10 years on the understanding of phenomena governing iodine chemistry and release in the case of a reactor severe accident, - to evaluate the current status of iodine chemistry knowledge and tools used for source term prediction in connection with accident management and emergency planning, under various reactor conditions, to identify the remaining weaknesses, discuss the reactor safety relevance of these issues and make recommendations as necessary. This paper aims at shedding light on the present situation, helping end-users and decision makers to adequately address questions related to iodine behaviour under severe accident conditions, and to essential programmes of work in this area

Characteristic relation for the mass and energy distribution of the nuclear fission products

International Nuclear Information System (INIS)

Alexandru, G.

1977-01-01

The dispersion relation for nuclear fission is written in the two part fragmentation approach which allows to obtain the characteristic relation for the mass and energy distribution of the nuclear fission products. One explains the resonance approximation in the mass distribution of the fission products taking into account the high order resonances too. (author)

Synergistic effects of iodine and silver ions co-implanted in 6H-SiC

Science.gov (United States)

Kuhudzai, R. J.; Malherbe, J. B.; Hlatshwayo, T. T.; van der Berg, N. G.; Devaraj, A.; Zhu, Z.; Nandasiri, M.

2015-12-01

Motivated by the aim of understanding the release of fission products through the SiC coating of fuel kernels in modern high temperature nuclear reactors, a fundamental investigation is conducted to understand the synergistic effects of implanted silver (Ag) and iodine (I) in 6H-SiC. The implantation of the individual species, as well as the co-implantation of 360 keV ions of I and Ag at room temperature in 6H-SiC and their subsequent annealing behaviour has been investigated by Secondary Ion Mass Spectrometry (SIMS), Atom Probe Tomography (APT) and X-ray Photoelectron Spectroscopy (XPS). SIMS and APT measurements indicated the presence of Ag in the co-implanted samples after annealing at 1500 °C for 30 h in sharp contrast to the samples implanted with Ag only. In samples implanted with Ag only, complete loss of the implanted Ag was observed. However, for I only implanted samples, some iodine was retained. APT of annealed co-implanted 6H-SiC showed clear spatial association of Ag and I clusters in SiC, which can be attributed to the observed I assisted retention of Ag after annealing. Such detailed studies will be necessary to identify the fundamental mechanism of fission products migration through SiC coatings.

Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

International Nuclear Information System (INIS)

Kramer, S.J.; Kotzer, T.G.; Chant, L.A.

2001-06-01

In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope 131 I (t 1/2 = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product 129 I has a long half-life (t 1/2 = 1.57x10 7 y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ( 127 I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ( 127 I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine concentrations. The concentrations of

Convective-diffusive transport of fission products in the gap of a failed fuel element

International Nuclear Information System (INIS)

Lian, Z.W.; Carlucci, L.N.; Arimescu, V.I.

1995-03-01

A model is presented to describe the transport behaviour of gaseous fission products along the axial fuel-to-sheathe gap of a failed fuel element to the coolant system. The model is applicable to an element having failed under normal operating conditions or loss-of coolant-accident conditions. Because of the large differences in operating parameters, the transport characteristics of gaseous fission products in a failed element under these two operating conditions are significantly different. However, in both cases the transport process can be described by convection-diffusion caused by the continuous release of fission products from the fuel to the gap. Under normal operating conditions, the bulk-flow velocity is found to be negligible, due to the low release rate of fission products from fuel. The process can be well approximated by the diffusion of fission products in a stagnant gas-steam mixture. The effect of convection on the fission product transport, however, becomes significant under loss-of-coolant-accident conditions, where the release rates of fission products from fuel can be several orders of magnitude higher that that under normal operating conditions. The convection of the mixture in the gap not only contributes an additional flux to the gas-mixture transport, but also increases the gradient of fission products concentration across the opening, and therefore increases the diffusion flux to the coolant. As a result of the bulk flow, the transport of fission products along the gap is accelerated and the hold-up of short-lived isotopes in the gap is significantly reduced. Steam ingress through the opening into the gap is obstructed by the bulk flow, resulting in low steam concentrations in the gap under loss-of-coolant-accident conditions. (author). 6 refs., 8 figs

The chemistry of fission products for accident analysis

International Nuclear Information System (INIS)

Potter, P.E.

1985-01-01

Current knowledge concerning the chemical state of the fission product elements during the development of accidents in water reactor systems is reviewed in this paper. The fission products elements which have been considered are Cs, I, Te, Sr and Ba but aspects of the behaviour of Mo, Ru and the lanthanides are also discussed. Some features of the reactions of the various species of these elements with other components of the reactor systems are described. The importance of having an adequate knowledge of thermodynamic data and phase equilibria of relatively simple systems in order to interpret experimental observations on complex multi-component systems is stressed

Automatic production of Iodine-123 with PLC 135/U

International Nuclear Information System (INIS)

Moghaddam-Banaem, L.; Afarideh, H.

2004-01-01

In this project, the automatic system for production of Iodine-123 with PLC/135μ Siemens, which is designed and installed for the first time in Iran, is discussed. The PLC (Programmable Logic Controller) is used to control industrial processing, which is similar to a computer and consists of central processing unit and memory and Input/Output units. PLC receives input information from auxiliary units such as sensors, switches, etc. and software processes data in memory and then sends commands to output units such as relays, motors, etc.The target section in Iodine production consists of 8 stages. In order to be sure automation works properly the system can be operated both manually and automatically. First PLC checks Manual/Automatic switch and in the case of automatic mode, PLC runs the program in memory and processing is done automatically. For this purpose, PLC takes the value of pressures and temperatures from analog inputs and after processing them it sends commands to digital output to activate valves or vacuum pumps or heaters. In this paper the following subjects are discussed: 1) Production of Iodine 123 2) PLC structure and auxiliary boards 3) Sensors and actuators and their connection to PLC 4) Software flowchart

Preliminary results utilizing high-energy fission product γ-rays to detect fissionable material in cargo

Science.gov (United States)

Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Church, J. A.; Descalle, M. A.; Gosnell, T. B.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Mauger, G. J.; Moore, T. L.; Norman, E. B.; Pohl, B. A.; Pruet, J. A.; Petersen, D. C.; Walling, R. S.; Weirup, D. L.; Prussin, S. G.; McDowell, M.

2005-12-01

A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their β-delayed neutron emission or β-delayed high-energy γ radiation between beam pulses provide the detection signature. Fission product β-delayed γ-rays above 3 MeV are nearly 10 times more abundant than β-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified.

Cea assessment of the sulphur-iodine cycle for hydrogen production

International Nuclear Information System (INIS)

Caries, Ph.; Vitart, X.; Yvon, P.

2010-01-01

The sulphur-iodine cycle is a promising process for hydrogen production using nuclear heat: - it is a purely thermochemical cycle, implying that hydrogen production will scale with volume rather than surface; - it only involves fluids, thus avoiding the often difficult handling of solids; - its heat requirements are well matched to the temperatures available from a Generation IV very/high temperature reactor. These characteristics seem very attractive for high efficiency and low cost massive hydrogen production. On the other hand, the efficiency of the cycle may suffer from the large over-stoichiometries of water and iodine and the very important heat exchanges it involves; furthermore, due to lack of adequate thermodynamic models, its efficiency is difficult to assess with confidence. Besides, the large quantities of chemicals that need to be handled, and the corrosiveness of these chemicals, are factors not to be overlooked in terms of investment and operation costs. In order to assess the actual potential of the sulphur-iodine cycle for massive hydrogen production at a competitive cost, CEA has been conducting an important programme on this cycle, ranging from thermodynamic measurements to hydrogen production cost evaluation, with flow sheet optimisation, component sizing and investment cost estimation as intermediate steps. The paper will present the method used, the status of both efficiency and production cost estimations, and discuss perspectives for improvement. (authors)

Preparative electrophoresis of industrial fission product solutions

International Nuclear Information System (INIS)

Tret, Joel

1971-07-01

The aim of this work is to contribute to the development of the continuous electrophoresis technique while studying its application in the preparative electrophoresis of industrial fission product solutions. The apparatus described is original. It was built for the purposes of the investigation and proved very reliable in operation. The experimental conditions necessary to maintain and supervise the apparatus in a state of equilibrium are examined in detail; their stability is an important factor, indispensable to the correct performance of an experiment. By subjecting an industrial solution of fission products to preparative electrophoresis it is possible, according to the experimental conditions, to prepare carrier-free radioelements of radiochemical purity (from 5 to 7 radioelements): 137 Cs, 90 Sr, 141+144 Ce, 91 Y, 95 Nb, 95 Zr, 103+106 Ru. (author) [fr

Behaviour of short-lived iodines in operating UO2 fuel elements

International Nuclear Information System (INIS)

Lipsett, J.J.; Hastings, I.J.; Hunt, C.E.L.

1984-11-01

Sweep gas experiments have been done to determine the behaviour of short-lived fission products within operating UO 2 fuel elements at linear powers of 45, 54, and 60 KW/m, and to burnups of 70, 80, and 50 MWh/kgU respectively. Although radioiodine transport was not observed directly during normal operation, equilibrium gap inventories for I-131 were deduced from the shutdown decay behaviour of the fission gases. These inventories were a strong function of fuel power and ranged from 10 GBq (0.27 Ci) to 100 GBq (2.7 Ci) over the range tested. We conclude that the iodine inventory was adsorbed onto the fuel and/or sheath surfaces with a volatile fraction of less than 10 -2 and a charcoal-filter-penetrating fraction of less than 2x10 -4

FISPRO: a simplified computer program for general fission product formation and decay calculations

International Nuclear Information System (INIS)

Jiacoletti, R.J.; Bailey, P.G.

1979-08-01

This report describes a computer program that solves a general form of the fission product formation and decay equations over given time steps for arbitrary decay chains composed of up to three nuclides. All fission product data and operational history data are input through user-defined input files. The program is very useful in the calculation of fission product activities of specific nuclides for various reactor operational histories and accident consequence calculations

Iodine Deficiency

NARCIS (Netherlands)

Zimmermann, M.B.

2009-01-01

Iodine deficiency has multiple adverse effects in humans, termed iodine deficiency disorders, due to inadequate thyroid hormone production. Globally, it is estimated that 2 billion individuals have an insufficient iodine intake, and South Asia and sub-Saharan Africa are particularly affected.

Background and derivation of ANS-5.4 standard fission product release model. Technical report

International Nuclear Information System (INIS)

1982-01-01

ANS Working Group 5.4 was established in 1974 to examine fission product releases from UO2 fuel. The scope of ANS-5.4 was narrowly defined to include the following: (1) Review available experimental data on release of volatile fission products from UO2 and mixed-oxide fuel; (2) Survey existing analytical models currently being applied to lightwater reactors; and (3) Develop a standard analytical model for volatile fission product release to the fuel rod void space. Place emphasis on obtaining a model for radioactive fission product releases to be used in assessing radiological consequences of postulated accidents

Fission product vapour - aerosol interactions in the containment: simulant fuel studies

International Nuclear Information System (INIS)

Beard, A.M.; Benson, C.G.; Bowsher, B.R.

1988-12-01

Experiments have been conducted in the Falcon facility to study the interaction of fission product vapours released from simulant fuel samples with control rod aerosols. The aerosols generated from both the control rod and fuel sample were chemically distinct and had different deposition characteristics. Extensive interaction was observed between the fission product vapours and the control rod aerosol. The two dominant mechanisms were condensation of the vapours onto the aerosol, and chemical reactions between the two components; sorption phenomena were believed to be only of secondary importance. The interaction of fission product vapours and reactor materials aerosols could have a major impact on the transport characteristics of the radioactive emission from a degrading core. (author)

Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

Science.gov (United States)

Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

2005-12-01

Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

Fission-product yields for thermal-neutron fission of 243Cm determined from measurements with a high-resolution low-energy germanium gamma-ray detector

International Nuclear Information System (INIS)

Merriman, L.D.

1984-04-01

Cumulative fission-product yields have been determined for 13 gamma rays emitted during the decay of 12 fission products created by thermal-neutron fission of 243 Cm. A high-resolution low-energy germanium detector was used to measure the pulse-height spectra of gamma rays emitted from a 77-nanogram sample of 243 Cm after the sample had been irradiated by thermal neutrons. Analysis of the data resulted in the identification and matching of gamma-ray energies and half-lives to individual radioisotopes. From these results, 12 cumulative fission product yields were deduced for radionuclides with half-lives between 4.2 min and 84.2 min. 7 references

Absorption behavior of iodine from molten salt mixture to zeolite

International Nuclear Information System (INIS)

Sugihara, Kei; Terai, Takayuki; Suzuki, Akihiro; Uozumi, Koichi; Tsukada, Takeshi; Koyama, Tadafumi

2011-01-01

Behavior of zeolite to absorb anion fission product (FP) elements in molten LiCl-KCl eutectic salt was studied using iodine. At first, zeolite-A was selected as the suitable type of zeolite among zeolite-A (powder), zeolite-X (powder and granule), and zeolite-Y (powder) through experiments to heat the zeolite together with LiCl-KCl-KI salt, respectively. As the next step, similar experiments to immerse zeolite-A in molten LiCl-KCl-KI salt containing various concentrations of iodine were performed. The affinity of iodine to zeolite was evaluated using the separation factor (SF) value, which is defined as [I/(I+Cl) mol ratio in zeolite after immersion]/[I/(I+Cl) mol ratio in salt after immersion]. Since the SF values ranged between 4.3 and 9.1, stronger affinity of iodine than chlorine to zeolite-A was revealed. Finally, influence of co-existing cation FPs was studied by similar absorption experiments in LiCl-KCl-KI salt containing CsCl, SrCl 2 , or NdCl 3 . The SF values were less than those obtained in the LiCl-KCl-KI salt and this can be ascribed to the sharing of inner space of zeolite cage among absorbed cations and anions. (author)

Modification of the fission product inventory program FISPIN

International Nuclear Information System (INIS)

Thomas, R.B.

1977-05-01

The fission product inventory program FISPIN calculates inventories of fission products, actinides and activation products, during and after irradiation in a nuclear reactor, estimates also being given for heat output and radioactive activity of the isotopes. The program has been developed further by making provision for the simulation of fuel reprocessing and in providing subroutines to make the program compatible with nuclear data in a slightly modified ENDF/B4 format. Continuous development of FISPIN over the years has however involved many program alterations and additions, and this has resulted in a generally untidy and cumbersome program. An attempt has therefore been made to improve the basic structure of the program. The subject is dealt with under the following headings: modularisation, direct access data, override facility, selective output, flowcharts, summary. (U.K.)

Measurements of fission product yield in the neutron-induced fission of 238U with average energies of 9.35 MeV and 12.52 MeV

Science.gov (United States)

Mukerji, Sadhana; Krishnani, Pritam Das; Shivashankar, Byrapura Siddaramaiah; Mulik, Vikas Kaluram; Suryanarayana, Saraswatula Venkat; Naik, Haladhara; Goswami, Ashok

2014-07-01

The yields of various fission products in the neutron-induced fission of 238U with the flux-weightedaveraged neutron energies of 9.35 MeV and 12.52 MeV were determined by using an off-line gammaray spectroscopic technique. The neutrons were generated using the 7Li(p, n) reaction at Bhabha Atomic Research Centre-Tata Institute of Fundamental Research Pelletron facility, Mumbai. The gamma- ray activities of the fission products were counted in a highly-shielded HPGe detector over a period of several weeks to identify the decaying fission products. At both the neutron energies, the fission-yield values are reported for twelve fission product. The results obtained from the present work have been compared with the similar data for mono-energetic neutrons of comparable energy from the literature and are found to be in good agreement. The peak-to-valley (P/V) ratios were calculated from the fission-yield data and were found to decreases for neutron energy from 9.35 to 12.52 MeV, which indicates the role of excitation energy. The effect of the nuclear structure on the fission product-yield is discussed.

Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

International Nuclear Information System (INIS)

Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

1984-01-01

The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

Atomic iodine production in a gas flow by decomposing methyl iodide in a dc glow discharge

International Nuclear Information System (INIS)

Mikheyev, P A; Shepelenko, A A; Voronov, A I; Kupryaev, Nikolai V

2002-01-01

The production of atomic iodine for an oxygen - iodine laser is studied by decomposing methyl iodide in a dc glow discharge in a vortex gas flow. The concentration of iodine atoms in discharge products was measured from the atomic iodine absorption of the radiation of a single-frequency tunable diode laser at a wavelength of 1.315 μm. Atomic iodine concentrations sufficient for the operation of an oxygen - iodine laser were obtained. The concentration of atomic iodine amounted to 3.6 x 10 15 cm -3 for a pressure of the carrying argon gas of 15 Torr. The discharge stabilisation by a vortex gas flow allowed the glow discharge to be sustained in a strongly electronegative halogen-containing gas mixture for pressures up to 20 Torr. (active media)

Heat and Fission Product Transport in a Molten U-Zr-O Pool With Crust

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2002-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry and the change of pool geometry during the numerical calculation was neglected. Results of the numerical calculation revealed that the peak temperature of the molten pool significantly decreased and most of the volatile fission products were released from the molten pool during the accident. (authors)

Integral measurement of fission products capture in fast breeder reactors

International Nuclear Information System (INIS)

Martin Deidier, Loick.

1979-12-01

For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

Experimental and analytical studies of iodine mass transfer from xenon-iodine mixed gas bubble to liquid sodium pool

International Nuclear Information System (INIS)

Miyahara, S.; Sagawa, N.; Shimoyama, K.

1996-01-01

In the fuel pin failure accident of a liquid metal fast reactor, volatile fission products play an important role in the assessment of radiological consequences. Especially the radioisotopes of elemental iodine are important because of their high volatility and of the low permissible dose to human thyroid. The released iodines are known to be retained in the coolant sodium as sodium iodide due to the chemical affinity between alkali metals and halogens. However, the xenon and krypton released with iodines into the sodium pool as bubbles may influence the reaction rate of iodine with sodium during the bubble rising. So far, the only few experimental results have been available concerning the decontamination factor (DF: the ratio of the initial iodine mass in the mixed gas bubble to the released mass into the cover gas) of iodine in this phenomenon. Therefore, experimental and analytical studies were carried out to study the mass transfer of iodine from a xenon-iodine mixed gas bubble to the liquid sodium pool. In the experiments, the bubble was generated in the sodium pool by cracking a quartz ball which contains the xenon-iodine mixed gas and then, the mixed gas released into the argon cover gas was collected to determine the transferred iodine mass into the pool. A rising velocity of the bubble was measured by Chen-type void sensors arranged vertically in the pool. From the measured rising velocity and another observation of bubble behavior in simulated water experiments, it is found that the generated bubble breaks up into several smaller bubbles of spherical cap type during the rising period. Transferred iodine mass per unit initial bubble volume from the bubble to the sodium pool shows increases with increasing time and the initial iodine concentration. A mass transfer rate obtained by differentiating the transferred iodine mass with respect to the time indicates a rapid decrease just after the bubble generation and a slow decrease for the successive period

Fission product release from TRIGA-LEU reactor fuels

International Nuclear Information System (INIS)

Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

1980-01-01

Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

Recoil release of fission products from nuclear fuel

International Nuclear Information System (INIS)

Wise, C.

1985-01-01

An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO 2 . The calculations presented here are one way of allowing for this, other methods are suggested. (orig.)

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

Report on the Behavior of Fission Products in the Co-decontamination Process

International Nuclear Information System (INIS)

Martin, Leigh Robert; Riddle, Catherine Lynn

2015-01-01

This document was prepared to meet FCT level 3 milestone M3FT-15IN0302042, 'Generate Zr, Ru, Mo and Tc data for the Co-decontamination Process.' This work was carried out under the auspices of the Lab-Scale Testing of Reference Processes FCT work package. This document reports preliminary work in identifying the behavior of important fission products in a Co-decontamination flowsheet. Current results show that Tc, in the presence of Zr alone, does not behave as the Argonne Model for Universal Solvent Extraction (AMUSE) code would predict. The Tc distribution is reproducibly lower than predicted, with Zr distributions remaining close to the AMUSE code prediction. In addition, it appears there may be an intricate relationship between multiple fission product metals, in different combinations, that will have a direct impact on U, Tc and other important fission products such as Zr, Mo, and Rh. More extensive testing is required to adequately predict flowsheet behavior for these variances within the fission products.

The ASIND-MEPhI library of independent actinide fission product yields

International Nuclear Information System (INIS)

Bogomolova, E.S.; Grashin, A.F.; Efimenko, A.D.; Lukasevich, I.B.

1997-01-01

This data base of independent fission product yields has been set up at the Moscow Engineering Physics Institute on the basis of theoretical calculations within the framework of the super-nonequilibrium thermodynamic model. The database consists of independent yield sets for 1163 fission products in the wide range of fissile nuclides from thorium-229 to fermium-257 with excitation energies up to 20 MeV. The use of the theoretical model made it possible to raise the accuracy of prediction for poorly explored fission reactions. The number of yield sets is larger than in the ENDF/B. For example, photofission product yields are included in the ASIND-MEPhI database as virtual sets. (author). 14 refs, 17 figs, 2 tabs

Highlights from the IAEA coordinated research programme on fuel performance and fission product data

International Nuclear Information System (INIS)

Nabielek, H.; Schenk, W.; Verfondern, K.

1996-01-01

Seven countries are cooperating with the objectives (i) to document the status of the experimental data base and of the predictive methods for Gas-Cooled Reactor fuel performance and fission product behaviour; (ii) to verify and validate methods in fuel performance and fission product retention prediction. These countries are China, France, Germany, Japan, Russia, USA and the UK. Duration of the programme is 1993-96. The technology areas addressed in this IAEA Coordinated Research Programme are: Fuel design and manufacture, Normal operation fuel performance and fission product behaviour, Accident condition fuel performance and fission product behaviour, -core heatup, -fast transients, -oxidising conditions (water and air ingress), Plateout, re-entrainment of plateout, fission product behaviour in the reactor building, and Performance of advanced fuels. Work performed so far has generated a 300-page draft document with important information for normal operations (Germany, Japan, China, Russia) and accident conditions (USA, Japan, Germany, Russia) and, additionally, a special chapter on advanced fuels (Japan). (author)

Fission product behaviour in the primary circuit of an HTR

International Nuclear Information System (INIS)

Decken, C.B. von der; Iniotakis, N.

1981-01-01

The knowledge of fission product behaviour in the primary circuit of a High Temperature Reactor (HTR) is an essential requirement for the estimations of the availability of the reactor plant in normal operation, of the hazards to personnel during inspection and repair and of the potential danger to the environment from severe accidents. On the basis of the theoretical and experimental results obtained at the ''Institute for Reactor Components'' of the KFA Juelich /1/,/2/ the transport- and deposition behaviour of the fission- and activation products in the primary circuit of the PNP-500 reference plant has been investigated thoroughly. Special work had been done to quantify the uncertainties of the investigations and to calculate or estimate the dose rate level at different components of the primary cooling circuit. The contamination and the dose rate level in the inspection gap in the reactor pressure vessel is discussed in detail. For these investigations in particular the surface structure and the composition of the material, the chemical state of the fission products in the cooling gas, the composition of the cooling gas and the influence of dust on the transport- and deposition behaviour of the fission products have been taken into account. The investigations have been limited to the nuclides Ag-110m; Cs-134 and Cs-137

Measurements of fission product yield in the neutron-induced fission of {sup 238}U with average energies of 9.35 MeV and 12.52 MeV

Energy Technology Data Exchange (ETDEWEB)

Mukerji, Sadhana; Krishnani, Pritam Das; Suryanarayana, Saraswatula Venkat; Naik, Haladhara; Goswami, Ashok [Bhabha Atomic Research Centre, Mumbai (India); Shivashankar, Byrapura Siddaramaiah [Manipal University, Manipal (India); Mulik, Vikas Kaluram [University of Pune, Pune (India)

2014-07-15

The yields of various fission products in the neutron-induced fission of {sup 238}U with the flux-weighted averaged neutron energies of 9.35 MeV and 12.52 MeV were determined by using an off-line gamma ray spectroscopic technique. The neutrons were generated using the {sup 7}Li(p, n) reaction at Bhabha Atomic Research Centre-Tata Institute of Fundamental Research Pelletron facility, Mumbai. The gamma- ray activities of the fission products were counted in a highly-shielded HPGe detector over a period of several weeks to identify the decaying fission products. At both the neutron energies, the fission-yield values are reported for twelve fission product. The results obtained from the present work have been compared with the similar data for mono-energetic neutrons of comparable energy from the literature and are found to be in good agreement. The peak-to-valley (P/V) ratios were calculated from the fission-yield data and were found to decreases for neutron energy from 9.35 to 12.52 MeV, which indicates the role of excitation energy. The effect of the nuclear structure on the fission product-yield is discussed.

Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

International Nuclear Information System (INIS)

Lewis, B.J.; Corse, B.J.; Thompson, W.T.; Kaye, M.H.; Iglesias, F.C.; Elder, P.; Dickson, R.; Liu, Z.

1997-01-01

An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

Cross sections of the lumped fission products for the AMZ library

International Nuclear Information System (INIS)

Ono, S.; Corcueca, R.P.; Nascimento, J.A.

1985-01-01

The preparation of the lumped fission product cross section for the AMZ library is described. For this purpose 100 nuclides were selected. The cross sections for each nuclide were generated by the NJOY code with evaluated nuclear data from ENDF/B-V, complemented with ENDF/B-IV data. A comparison is performed between the data obtained and the lumped fission product cross section of JFS-II [pt

Release of fission products from contaminated sodium fires

International Nuclear Information System (INIS)

Jordan, S.

1976-01-01

Leaks in the primary coolant system of a LMFBR and also serious incidents with tank rupture may entail the escape of fission products into the containment of the reactor. For incident analysis it is important to know the retention capability of sodium for the different fission products. The release of cesium and strontium from pools contaminated with 100 to 1000 ppM was investigated by experiments. The cesium content of airborne aerosols depends on oxygen concentration: at 21 percent oxygen concentration the Cs content of sodium-oxide aerosols is 3 times and at 0.5 percent 15 times as high as the initial Cs concentration in the pool. Strontium content of aerosols over burning contaminated sodium pools is 10 3 times smaller than the strontium pool concentration

Tracking of fission products release during refueling operations

International Nuclear Information System (INIS)

Agarwal, Sharad; Prajapat, M.K.; Vyas, Shyam; Hussain, S.A.

2001-01-01

It has been always observed that the release of fission products increase during refueling operations. At RAPP-3 and 4 an attempt has been made to follow-up the change in fission products activity release at each stage of refueling operation and quantification of concentrations of various radionuclides. This exercise was also extended to refueling operation of the channels containing suspected failed fuel. A level of FPNG ( 133 Xe) was observed to increase by a factor of about 10-40 during refueling of failed channel as compared to healthy channel. It can be concluded that by monitoring FPNG levels in exhaust status of the healthiness of spent fuel can be found out. This report discusses in detail the experiment conducted for this purpose. (author)

HAMCIND, Cell Burnup with Fission Products Poisoning

International Nuclear Information System (INIS)

Abe, Alfredo Y.; Dos Santos, Adimir

2002-01-01

1 - Description of program or function: HAMCIND is a cell burnup code based in a coupling between HAMMER-TECHNION and CINDER. The fission product poisoning is taken into account in an explicit fashion. 2 - Method of solution: The nonlinear coupled set of equations for the neutron transport and nuclide transmutation equations and nuclide transmutation equations in a unit cell is solved by HAMCIND in a quasi-static approach. The spectral transport equation is solved by HAMMER-TECHNION at the beginning of each time-step while the nuclide transmutation equations are solved by CINDER for every time-step. The HAMMER-TECHNION spectral calculations are performed taking into account the fission product contribution to the macroscopic cross sections (fast and thermal), in the inelastic scattering matrix and even in the thermal scattering matrices. 3 - Restrictions on the complexity of the problem: Restrictions and/or limitations for HAMCIND depend upon the local operating system

Simulating fission product transients via the history-based local-parameter methodology

International Nuclear Information System (INIS)

Jenkins, D.A.; Rouben, B.; Salvatore, M.

1993-01-01

This paper describes the fission-product-calculation capacity of the history-based local-parameter methodology for evaluating lattice properties for use in core-tracking calculations in CANDU reactors. In addition to taking into account the individual past history of each bundles flux/power level, fuel temperature, and coolant density and temperature that the bundle has seen during its stay in the core, the latest refinement of the history-based method provides the capability of fission-product-drivers. It allows the bundle-specific concentrations of the three basic groups of saturating fission products to be calculated in steady state or following a power transient, including long shutdowns. The new capability is illustrated by simulating the startup period following a typical long-shutdown, starting from a snapshot in the Point Lepreau operating history. 9 refs., 7 tabs

Mass-yield distributions of fission products from 20, 32, and 45 MeV proton-induced fission of 232Th

Science.gov (United States)

Naik, H.; Goswami, A.; Kim, G. N.; Kim, K.; Suryanarayana, S. V.

2013-10-01

The yields of various fission products in the 19.6, 32.2, and 44.8 MeV proton-induced fission of 232Th have been determined by recoil catcher and an off-line γ-ray spectrometric technique using the BARC-TIFR Pelletron in India and MC-50 cyclotron in Korea. The mass-yield distributions were obtained from the fission product yield using the charge distribution corrections. The peak-to-valley (P/V) ratio of the present work and that of literature data for 232Th(p,f) and 238U(p,f) were obtained from the mass yield distribution. The present and the existing literature data for 232Th(p,f), 232Th(n,f), and 232Th( γ,f) at various energies were compared with those for 238U(p,f), 238U(n,f), and 238U( γ,f) to examine the probable nuclear structure effect. The role of Th-anomaly on the peak-to-valley ratio in proton-, neutron-, and photon-induced fission of 232Th was discussed with the similar data in 238U. On the other hand, the fine structure in the mass yield distributions of the fissioning systems at various excitation energies has been explained from the point of standard I and II asymmetric mode of fission besides the probable role of even-odd effect, A/ Z ratio, and fissility parameter.

Separation and utilization of fission products considering economic aspects

International Nuclear Information System (INIS)

Beer, M.; Gorski, B.; Hennrich, M.; Pfrepper, G.; Richter, M.

1982-01-01

The quantity of usable fission products which will be obtained by nuclear fission till the year 2000 is estimated on the basis of prognostics for the development of nuclear energy in the world considering especially the development in the U.S.S.R. and the CMEA. The possibilities of utilization of cesium as gamma-ray source are discussed, and the present fields of application of palladium and the development of its price on the world market are shown. The fields of application of technetium, which wasn't available as artificial element in a greater quantity till now, have to be developed. The economic estimations base on data of a project for the separation of fission products in connection with a reprocessing plant, which was developed in the U.S.A. in 1978. The data show, that it is possible to produce the platinum metals and cesium with profit, the same can be expected for technetium. (author)

Measurements of isomeric yield ratios of fission products from proton-induced fission on natU and 232Th via direct ion counting

Directory of Open Access Journals (Sweden)

Rakopoulos Vasileios

2017-01-01

Full Text Available Independent isomeric yield ratios (IYR of 81Ge, 96Y, 97Y, 97Nb, 128Sn and 130Sn have been determined in the 25 MeV proton-induced fission of natU and 232Th. The measurements were performed at the Ion Guide Isotope Separator On-Line (IGISOL facility at the University of Jyväskylä. A direct ion counting measurement of the isomeric fission yield ratios was accomplished for the first time, registering the fission products in less than a second after their production. In addition, the IYRs of natU were measured by means of γ-spectroscopy in order to verify the consistency of the recently upgraded experimental setup. From the obtained results, indications of a dependence of the production rate on the fissioning system can be noticed. These data were compared with data available in the literature, whenever possible. Using the TALYS code and the experimentally obtained IYRs, we also deduced the average angular momentum of the fission fragments after scission.

European commission research activities on iodine

Energy Technology Data Exchange (ETDEWEB)

Loggia, E della [European Commission, Brussels (Belgium)

1996-12-01

The research on iodine, as on other important fission products which would be released during a severe accident, carried out directly or organized by the European Commission stems from the Euratom Treaty, namely from Chapter III of the treaty which deals with the protection of the health of the population against radiations and from Chapter I which deals with research. In this paper we do not consider the Commission radiological protection programme: we limit ourselves to the presentation of the research carried out on Iodine as part of the most recent source term studies within the framework Programmes as are called the research programme of the European Commission, usually valid for a 4 year periods. The research activities are carried out by the European Commission either directly through the Joint Research Centres (JRC) or indirectly through collaboration with research organizations of Member States. Concerning the iodine research carried out as Direct Action in the Joint Research Centres, are mentioned here the most relevant activities carried out in this field at the JRC of Ispra and Karlsruhe (TUI). As Indirect Action, we present here the results of some studies allocated by the European Commission to experts of research organizations of Member Countries, followed by a short description of the main results achieved by the Reinforced Concerted Action, within the III Framework Programme (1992-1995). At the end of the paper are described the research on iodine being carried out or proposed within the IV Framework Programme (1995-1998). Mention is also done of the Commission participation, relevant in terms of financial and human efforts, to the PHEBUS FP Project. (author) refs.

European commission research activities on iodine

International Nuclear Information System (INIS)

Loggia, E. della

1996-01-01

The research on iodine, as on other important fission products which would be released during a severe accident, carried out directly or organized by the European Commission stems from the Euratom Treaty, namely from Chapter III of the treaty which deals with the protection of the health of the population against radiations and from Chapter I which deals with research. In this paper we do not consider the Commission radiological protection programme: we limit ourselves to the presentation of the research carried out on Iodine as part of the most recent source term studies within the framework Programmes as are called the research programme of the European Commission, usually valid for a 4 year periods. The research activities are carried out by the European Commission either directly through the Joint Research Centres (JRC) or indirectly through collaboration with research organizations of Member States. Concerning the iodine research carried out as Direct Action in the Joint Research Centres, are mentioned here the most relevant activities carried out in this field at the JRC of Ispra and Karlsruhe (TUI). As Indirect Action, we present here the results of some studies allocated by the European Commission to experts of research organizations of Member Countries, followed by a short description of the main results achieved by the Reinforced Concerted Action, within the III Framework Programme (1992-1995). At the end of the paper are described the research on iodine being carried out or proposed within the IV Framework Programme (1995-1998). Mention is also done of the Commission participation, relevant in terms of financial and human efforts, to the PHEBUS FP Project. (author) refs

Energy production using fission fragment rockets

International Nuclear Information System (INIS)

Chapline, G.; Matsuda, Y.

1991-08-01

Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs

Radionuclides measurements for the Phebus FP project - preliminary study

International Nuclear Information System (INIS)

Schuster, E.; Nopitsch, K.

1989-09-01

This report gives a review of proposed measurement techniques to measure quantitatively release and transport of fission products during and after fission product release tests. The measurement devices provided are for example gamma-spectrometers, scintillation counters and sampling vessels. The second part reviews the behaviour of fission product iodine under severe accident conditions and describes the iodine sampling system and the determination of iodine species in aqueous solutions

Iodine and Pregnancy

OpenAIRE

Yarrington, Christina; Pearce, Elizabeth N.

2011-01-01

Iodine is a necessary element for the production of thyroid hormone. We will review the impact of dietary iodine status on thyroid function in pregnancy. We will discuss iodine metabolism, homeostasis, and nutritional recommendations for pregnancy. We will also discuss the possible effects of environmental contaminants on iodine utilization in pregnant women.

Yields of fission products produced by thermal-neutron fission of 249Cf

International Nuclear Information System (INIS)

Dickens, J.K.; McConnell, J.W.

1981-01-01

Absolute yields have been determined for 107 gamma rays emitted in the decay of 97 fission products representing 54 mass chains created during thermal-neutron fission of 249 Cf. These results include 14 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays emanating from a 0.4 μg sample of 249 Cf between 45 s and 0.4 yr after very short irradiations of the 249 Cf by thermal neutrons. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 89 and 156. The absolute overall normalization uncertainty is approx.8%. The measured A-chain cumulative yields make up 77% of the total light mass (A 249 Cf

Decommissioning of the Fission Product Development Laboratory at Holifield National Laboratory

International Nuclear Information System (INIS)

Schaich, R.W.

1975-01-01

The decontamination of the Fission Product Development Laboratory was initiated in FY 1975 after 17 years of processing fission product waste streams to produce commercial quantities of 90 Sr, 137 Cs, 144 Ce, and 147 Pm. The objective of the decommissioning program is the removal of all radiation and contamination areas in the facility to a level which will be compatible with the environment in the foreseeable future

Role of fission product in whole core accidents: research in the USA

International Nuclear Information System (INIS)

Jackson, J.F.; Deitrich, L.W.

1977-01-01

The techniques being developed in the United States for analyzing postulated whole-core accidents in LMFBRs are briefly reviewed. The key mechanistic analysis methods are discussed in detail. Important research projects in the area of fission product effects are examined. Some typical results on the role of fission products in whole-core accidents are presented

Release and transport of fission product cesium in the TMI-2 accident

International Nuclear Information System (INIS)

Lorenz, R.A.; Collins, J.L.

1986-01-01

Approximately 50% of the fission product cesium was released from the overheated UO 2 fuel in the TMI-2 accident. Steam that boiled away from a water pool in the bottom of the reactor vessel transported the released fission products throughout the reactor coolant system (RCS). Some fission products passed directly through a leaking valve with steam and water into the containment structure, but most deposited on dry surfaces inside of the RCS before being dissolved or resuspended when the RCS was refilled with water. A cesium transport model was developed that extended measured cesium in the RCS back to the first day of the accident. The model revealed that ∼62% of the released 137 Cs deposited on dry surfaces inside of the RCS before being slowly leached and transported out of the RCS in leaked or letdown water. The leach rates from the model agreed reasonably well with those measured in the laboratory. The chemical behavior of cesium in the TMI-2 accident agreed with that observed in fission product release tests at Oak Ridge National Laboratory (ORNL)

Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

Energy Technology Data Exchange (ETDEWEB)

Kramer, S.J.; Kotzer, T.G.; Chant, L.A

2001-06-01

In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope {sup 131}I (t{sub 1/2} = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product {sup 129}I has a long half-life (t{sub 1/2} = 1.57x10{sup 7} y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ({sup 127}I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ({sup 127}I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine

Chemical and physical factors which control the substitution reactions of direct fission-produced iodine with gaseous methane and the methyl halides

International Nuclear Information System (INIS)

Kikuchi, M.; Church, L.B.

1976-01-01

The factors controlling the hydrogen and halogen substitution reactions resulting from direct fission-produced iodine isotopes (*I) with gaseous methane and methyl halides were investigated. The chemical reaction probabilities, corrected for any secondary decomposition, were compared as a function of various chemical and physical parameters. These include carbon-halogen bond strength, halogen electronegativity, volume and cross sectional area of the substituted and neighboring atoms. On the basis of this analysis, it is concluded that *I-for-X (where X = H, F, Cl, Br and I) substitution reactions are controlled by the cross sectional area of the X atom. The *I-for-H substitution probability is reduced in proportion to the volume of X, suggesting that steric interference is the dominant factor influencing the reaction probability. (orig.) [de

Status report on actinide and fission product transmutation studies

International Nuclear Information System (INIS)

1997-06-01

The management of radioactive waste is one of the key issues in today's political and public discussions on nuclear energy. One of the fields that looks into the future possibilities of nuclear technology is the neutronic transmutation of actinides and of some most important fission products. Studies on transmutation of actinides are carried out in various countries and at an international level. This status report which gives an up-to-date general overview of current and planned research on transmutation of actinides and fission products in non-OECD countries, has been prepared by a Technical Committee meeting organized by the IAEA in September 1995. 168 refs, 16 figs, 34 tabs

Irradiation effects and behaviour of fission products in zirconia and spinel

International Nuclear Information System (INIS)

Gentils, A.

2003-10-01

Crystalline oxides, such as zirconia (ZrO 2 ) and spinel (MgAl 2 O 4 ), are promising inert matrices for the transmutation of plutonium and minor actinides. This work deals with the study of the physico-chemical properties of these matrices, more specifically their behaviour under irradiation and their capacity to retain fission products. Irradiations at low energy and incorporation of stable analogs of fission products (Cs, I, Xe) into yttria-stabilized zirconia and magnesium-aluminate spinel single crystals were performed by using the ion implanter IRMA (CSNSM-Orsay). Irradiations at high energy were made on several heavy ion accelerators (GANIL-Caen, ISL-Berlin, HIL-Warsaw). The damage induced by irradiation and the release of fission products were monitored by in situ Rutherford Backscattering Spectrometry experiments. Transmission electron microscopy was also used in order to determine the nature of the damage induced by irradiation. The results show that irradiation of ZrO 2 and MgAl 2 O 4 with heavy ions (about hundred keV and about hundred MeV) induces a huge structural damage in crystalline matrices. Total disorder (amorphization) is however never reached in zirconia, contrary to what is observed in the case of spinel. The results also emphasize the essential role played by the concentration of implanted species on their retention capacity. A dramatic release of fission products was observed when the concentration exceeds a threshold of a few atomic percent. Irradiation of implanted samples with medium-energy noble-gas ions leads to an enhancement of the fission product release. The exfoliation of spinel crystals implanted at high concentration of Cs ions is observed after a thermal treatment at high temperature. (author)

Research on in-pile release of fission products from coated particle fuels

International Nuclear Information System (INIS)

Fukuda, K.; Iwamoto, K.

1985-01-01

Coated particle fuels fabricated in accordance with VHTR (Very High Temperature gas-cooled Reactor) fuel design have been irradiated by both capsules and an in-pile gas loop (OGL-1), and data on the fission products release under irradiation were obtained for loose coated particles, fuel compacts and fuel rods in the temperature range between 800 deg. C and 1600 deg. C. For the fission gases, temperature- and time dependences of the fractional release(R/B) were measured. Relation between release and failure fraction of the coated particles was elucidated on the VHTR reference fuels. Also measured was tritium concentration in the helium coolant of OGL-1. In-pile release behavior of the metallic fission products was studied by measuring the activities of the fission products adsorbed in the graphite sleeves of the OGL-1 fuel rods and the graphite fuel container of the sweep gas capsules in the PIE. Investigation on palladium interaction with SiC coating layer was included. (author)

Development of an automation system for iodine-125 brachytherapy seed production by ND:YAG laser welding

International Nuclear Information System (INIS)

Somessari, Samir L.; Feher, Anselmo; Sprenger, Francisco E.; Rostellato, Maria Elisa C.M.; Costa, Fabio E.; Calvo, Wilson A.P.

2009-01-01

The aim of this work is to develop an automation system for iodine-125 radioactive seed production by Nd:YAG laser welding, which has been used successfully in low dose rate brachytherapy treatment. This small seed consists of a welded titanium capsule, with 0.8 mm in diameter and 4.5 mm in length, containing iodine-125 adsorbed onto a silver rod. The iodine-125 seeds are implanted into the human prostate to irradiate the tumor for cancer treatment. Nowadays, the Radiation Technology Center, at IPEN-CNEN/SP imports and distributes 36,000 iodine-125 seeds per year, for the clinics and hospitals in the country. However, the Brazilian market potential is now over 8,000 iodine-125 seeds per month. The local production of these iodine-125 radioactive sources became a priority for the Institute, in order to reduce the price and the problems of prostate cancer management. It will permit to spread their use to a larger number of patients in Brazil. On the other hand, the industrial automation plays an important role for iodine-125 seeds in order to increase the productivity, with high quality and assurance, avoiding human factors, implementing and operating with good manufacturing practices. The technology consists of appliance electronic and electro-mechanical parts and components to control machines and processes. The automation system technology for iodine-125 seed production developed in this work was mainly assembled employing a Programmable Logic Controller, a stepper motor, an Nd:YAG laser welding machine and a supervisory. (author)

Development of an automation system for iodine-125 brachytherapy seed production by (Nd:YAG) laser welding

International Nuclear Information System (INIS)

Somessari, Samir Luiz

2010-01-01

The aim of this work is to develop an automation system for iodine-125 radioactive seed production by (Nd:YAG) laser welding, which has been used successfully in Low Dose Rate (LDR) brachytherapy treatment. This small seed consists of a welded titanium capsule, with 0.8mm in diameter and 4.5mm in length, containing iodine-125 adsorbed onto a silver rod. The iodine-125 seeds are implanted into the human prostate to irradiate the tumor for cancer treatment. Nowadays, the Radiation Technology Center, at IPEN-CNEN/SP imports and distributes 36,000 iodine-125 seeds per year, for the clinics and hospitals in the country. However, the Brazilian market potential is now over 8,000 iodine-125 seeds per month. The local production of these iodine-125 radioactive sources becomes a priority for the Institute, in order to reduce the price and the problems of prostate cancer management. It will permit to spread their use to a largest number of patients in Brazil. On the other hand, the industrial automation plays an important role for iodine-125 seeds in order to increase the productivity, with high quality and assurance, avoiding human factors, implementing and operating with Good Manufacturing Practices (GMP). The technology consists of appliance electronic and electro-mechanical parts and components to control machines and processes. The automation system technology for iodine-125 seed production developed in this work was mainly assembled employing Programmable Logic Controller (PLC), stepper motors, drivers, (Nd:YAG) laser welding machine, photoelectric sensors and supervisory. (author)

Transmutation analysis considering and explicit fission product treatment based on a coupled Hammer-Technion and Cinder-2 system

International Nuclear Information System (INIS)

Abe, A.Y.

1989-01-01

This work presents a study about neutron absorption in a typical PWR cell by considering an explicit treatment for the fission products. The proposed methodology to treat fission product neutron absorption in a lattice calculation combines the HAMMER-TECHNION and CINDER-2 codes. The fission product chain treatment considers nearly 99% of all original CINDER-2 neutron absorption chain treatment. Parallel to the explicit treatment, a cross section library in the HAMMER-TECHNION code multigroup structure for the fission products was generated using the ENDF/B-V fission product library and processed by NJOY and AMPX-II processing codes. The methodology validation was investigated against two available benchmarks and it was obtained excellent results for the K-Infinity (IAEA-TECDOC-233) as function of burnup and enrichment and for the aggregate quantity sup(σ)2200 in units of barns/fission cross sections (OKAZAKI and SOKOLOWSKI). This work contributed for a better understanding of the fission product neutron absorption in a typical PWR cell and showed that the explicit fission product treatment can be successfully achieved. Besides that the performance of the ENDF/B-V fission product library was accessed. (author)

Study on the behavior of unirradiated light water reactor fuel with iodine-127 under the reactivity initiated accident (RIA) conditions

International Nuclear Information System (INIS)

Sasajima, Hideo; Yanagisawa, Kazuaki; Kanazawa, Hiroyuki

1988-07-01

In a light water reactor fuel, a stress arised from pellet-cladding interaction (PCI) will have possibility to cause stress corrosion claddling (PCI failure) under an environment of corrosive fission product ; line iodine. A pulse irradiation experiment by NSRR was carried out to examine whether or not an unirradiated NSRR standard fuel rod in which 1.3 mg (33 x 10 -6 g/cm 2 ) of iodine was artificially filled could cause the PCI failure. Obtained results are: (1) The fuel rods with iodine did not fail both at deposited energy levels of 268 and 280 cal/g · UO 2 . On the other hand, the fuel rods without iodine failed at the same energy levels due to thinning of the cladding wall thickness. Within this experimental scope, PCI-failure did not occur on iodine filled fuel rods. (2) At a periphery of the fuel pellet of iodine filled rod, an uniform torus ring was formed. The torus ring consisted of an equi-axed large grains at 268 cal/g · UO 2 and a columnar ones at 280 cal/g · UO 2 . The torus ring was not formed in the fuel without iodine. (author)

Design considerations of fission and corrosion product in primary system of MONJU

International Nuclear Information System (INIS)

Yanagisawa, T.; Akagane, K.; Yamamoto, K.; Kawashima, K.

1976-01-01

General influence of fission and corrosion products in primary system on MONJU plant design is reviewed. Various research and development works are now in progress to decrease the generation rate, to remove the products more effectively and to develop the methods of evaluation the behaviour of radioactive products. The inventory and distribution of fission and corrosion products in the primary circuit of MONJU are given. The radiation levels on the primary components are estimated to be several roentgens per hour. (author)

Investigations concerning the exchange of iodine from non-volatile organic iodine compounds

International Nuclear Information System (INIS)

Psarros, N.; Duschner, H.; Molzahn, D.; Schmidt, L.; Heise, S.; Jungclas, H.; Brandt, R.; Patzelt, P.

1990-10-01

The iodine produced by nuclear fission is removed during the reprocessing of exhausted nuclear fuel elements by desorption achieving good decontamination factors. Nevertheless the further optimization of the process requires detailed information about the iodine speciation during fuel reprocessing, and about possible reactions. For the study of decomposition reactions of iodo-alcanes, which are built up during the fuel recycling process, we developed a method for the synthesis of labelled iodo-dodecane, which was used as tracer. In order to identify the iodo species in the organic phase of the reprocessing cycle we applied plasma desorption time-of-flight mass spectroscopy. The problem of the volatility of the iodo-compounds in the ultra vacuum of the mass spectrometer was overcome by derivatization of the iodo-alcanes with dithizon, which yielded non-volatile ionic alcyltetrazolium iodides. Beta-spectrometric analysis of the exhaust condensates collected from the organic phase of the WAK reprocessing cycle revealed beside iodine-129 the existence of a low-energetic beta emitter, which has yet to be identified. A literature survey on the topic was also performed. (orig.) With 42 refs., 9 figs [de

Radioiodine source term and its potential impact on the use of potassium iodide

International Nuclear Information System (INIS)

Malinauskas, A.P.

1982-01-01

Information is presented concerning chemical forms of fission product iodine in the primary circuit; chemical forms of fission product iodine in the containment building; summary of iodine chemistry in light water reactor accidents; and impact of the radiodine source term on the potassium iodide issue

Production of high specific activity 123I for protein iodination for medical use

International Nuclear Information System (INIS)

Legoux, Y.; Cieur, M.; Crouzel, C.; Syrota, A.

1985-01-01

Iodine-123 is produced via xenon-133 by irradiation of a sodium iodide target with 108 MeV deuterons from the synchrocyclotron of IPN. The on-line production method is described. The specific activity of the iodine is determined by neutron activation analysis and by a radioimmunological method. The conditions labelling different proteins (insulin, angiotensin) are given and also the purification method to obtain a product ready for injection to patients. (author)

Adsorption of fission products on mediterranean mud; Adsorption des produits de fission sur des vases de mediterranee

Energy Technology Data Exchange (ETDEWEB)

Cohen, P; Gailledreau, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

Partition coefficients of some fission products have been measured in sea water on mud taken from the bottom of the Mediterranean sea. A discussion follows on the behaviour of these radioisotopes. (author) [French] On a mesure les coefficients de partage dans l'eau de mer de quelques produits de fission a longue periode sur des echantillons de vase preleves en Mediterranee. Les valeurs trouvees sont elevees. Le comportement de ces isotopes radioactifs est discutee. (auteur)

Fission product yield data for the transmutation of minor actinide nuclear waste

International Nuclear Information System (INIS)

2008-04-01

A report issued by an international study group for the transmutation of nuclear waste using accelerator driven systems has highlighted the need for specific sets of nuclear data. These authoritative requirements include fission product yields at an intermediate incident neutron energy of up to 150 MeV. Before the start of the present CRP on fission product yield data for the transmutation of nuclear waste, only four types of evaluated fission yield data sets existed, namely for spontaneous fission, and for fission induced by thermal, fast (or fission) spectrum, and by 'high energy' (14-15 MeV) neutrons. A new type of evaluation for energy dependent neutron induced fission yields was required for this project. In view of the scarcity of experimental data, such an evaluation has to be based on systematics and theoretical model calculations. Unlike fission cross-sections, where nuclear models are being used successfully for the calculation of unmeasured cross-section ranges, such models or theories existed only for low energy fission yields. Hence the CRP participants entered a completely new field of research for which the progress and outcome were unpredictable. Clearly the ultimate goal of such an effort, namely an evaluation of energy dependent fission yields, could not be realized within the perceived lifetime of a CRP. The main emphasis of the CRP was on the development of adequate systematics and models for the calculation of energy dependent fission yields up to 150 MeV incident neutron energy. Several problems had to be solved, such as the correct choice of model parameters and multiplicity distributions of emitted neutrons, and the effect of multi-chance fission. Models and systematics have been tested for lower energy yields, but they failed to reproduce recent experimental data, particularly at higher energies, and the parameters had to be modified. Other models have been developed from the analysis of experimental data in order to derive systematic

Iodine deficiency and nutrition in Scandinavia

DEFF Research Database (Denmark)

Manousou, Sofia; Dahl, Lisbeth; Heinsbaek Thuesen, Betina

2017-01-01

Iodine nutrition is a result of geological conditions, iodine fortification and monitoring strategies within a country together with the dietary habits of the population. This review summarizes the basis for the current iodine situation in the Scandinavian countries in order to identify gaps...... strategies have been used in Scandinavia to improve iodine nutrition. The major source of iodine is iodized salt in Sweden and from milk and dairy products in Norway. In Denmark, drinking water, milk, dairy products and iodized salt used in commercial production of bread are the important sources of iodine....... The current iodine status in Scandinavia is not optimal and action is ongoing to increase iodination in Denmark, where there is mild iodine deficiency in the general population. Data from all three countries indicate insufficient iodine nutrition during pregnancy and there is a need for data from children...

Sensitivity and uncertainty analysis for fission product decay heat calculations

International Nuclear Information System (INIS)

Rebah, J.; Lee, Y.K.; Nimal, J.C.; Nimal, B.; Luneville, L.; Duchemin, B.

1994-01-01

The calculated uncertainty in decay heat due to the uncertainty in basic nuclear data given in the CEA86 Library, is presented. Uncertainties in summation calculation arise from several sources: fission product yields, half-lives and average decay energies. The correlation between basic data is taken into account. The uncertainty analysis were obtained for thermal-neutron-induced fission of U235 and Pu239 in the case of burst fission and irradiation time. The calculated decay heat in this study is compared with experimental results and with new calculation using the JEF2 Library. (from authors) 6 figs., 19 refs

Putting to point the production process of iodine-131 by dry distillation (Preoperational tests)

International Nuclear Information System (INIS)

Alanis M, J.

2002-12-01

With the purpose of putting to point the process of production of 131 I, one of the objectives of carrying out the realization of operational tests of the production process of iodine-131, it was of verifying the operation of each one of the following components: heating systems, vacuum system, mechanical system and peripheral equipment that are part of the production process of iodine-131, another of the objectives, was settling down the optimal parameters that were applied in each process during the obtaining of iodine-131, it is necessary to point out that this objective is very important, since the components of the equipment are new and its behavior during the process is different to the equipment where its were carried out the experimental studies. (Author)

Identification of Organic Iodine Compounds and Their Transformation Products in Edible Iodized Salt Using Liquid Chromatography-High Resolution Mass Spectrometry.

Science.gov (United States)

Yun, Lifen; Peng, Yue'e; Chang, Qing; Zhu, Qingxin; Guo, Wei; Wang, Yanxin

2017-07-05

The consumption of edible iodized salt is a key strategy to control and eliminate iodine deficiency disorders worldwide. We herein report the identification of the organic iodine compounds present in different edible iodized salt products using liquid chromatography combined with high resolution mass spectrometry. A total of 38 organic iodine compounds and their transformation products (TPs) were identified in seaweed iodine salt from China. Our experiments confirmed that the TPs were generated by the replacement of I atoms from organic iodine compounds with Cl atoms. Furthermore, the organic iodine compound contents in 4 seaweed iodine salt samples obtained from different manufacturers were measured, with significant differences in content being observed. We expect that the identification of organic iodine compounds in salt will be important for estimating the validity and safety of edible iodized salt products.

Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

International Nuclear Information System (INIS)

Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

2003-01-01

During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

HAC and fission reactors

International Nuclear Information System (INIS)

Fujiwara, I.; Moriyama, H.; Tachikawa, E.

1984-01-01

In the fission process, newly formed fission products undergo hot atom reactions due to their energetic recoil and abnormal positive charge. The hot atom reactions of the fission products are usually accompanied by secondary effects such as radiation damage, especially in condensed phase. For reactor safety it is valuable to know the chemical behaviour and the release behaviour of these radioactive fission products. Here, the authors study the chemical behaviour and the release behaviour of the fission products from the viewpoint of hot atom chemistry (HAC). They analyze the experimental results concerning fission product behaviour with the help of the theories in HAC and other neighboring fields such as radiation chemistry. (Auth.)

Separation of fission products by the use of recoil; Separation des produits de fission par utilisation du recul

Energy Technology Data Exchange (ETDEWEB)

Henry, R; Beydon, J; Bardy, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

We have studied fission recoil in U{sub 3}O{sub 8} organic solvent mixtures. The organic phase chosen was first naphtalene then terphenyl. Graphite and activated carbon were also tried out as recoil media. We first verified that the fission fragments are ejected from the uranium oxide particles under our experimental conditions. The retention phenomenon observed is due to an adsorption occurring either during irradiation or during the chemical treatment. Using naphthalene or terphenyl, the individual separation of the fission products has made it possible to show the influence of the chemical nature of the recoil medium on the retention of each fission product. We put forward a hypothesis concerning this phenomenon: experiments carried out using 'scavengers', together with kinetic studies make it possible to explain the retention phenomenon and to choose the most favorable conditions for reducing this retention to a low value. The thermal recombination kinetics demonstrate the influence of the fission ion charge on the final value of the retention for a given temperature. The origins of this thermal recombination are discussed. (author) [French] On a etudie le recul de fission dans les melanges U{sub 3}0{sub 8}, phase organique. La phase organique choisie a ete le naphtalene puis le terphenyle. Le graphite et le charbon actif ont egalement ete essayes comme milieux de recul. On a d'abord determine que les fragments de fission sortent des particules d'oxyde d'uranium avec un rendement de 100 pour cent dans nos conditions experimentales. Le phenomene de retention observe est du a une adsorption ayant lieu pendant l'irradiation ou pendant le traitement chimique. Dans le naphtalene et le terphenyle, la separation individuelle des produits de fission a permis de mettre en evidence l'influence de la nature chimique du milieu de recul sur la retention de chaque produit de fission. On avance une hypothese sur ce phenomene: des experiences effectuees avec des 'scavengers

Radiation research contracts: Distribution of fission products in the biosphere

Energy Technology Data Exchange (ETDEWEB)

Schoenfeld, T [Vienna University, First Chemical Institute, Vienna (Austria)

1959-04-15

Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. To make the routine release of small amounts of fission products safe and to be able to estimate the possible effect of larger releases in accidents, a considerable amount of information is required

Radiation research contracts: Distribution of fission products in the biosphere

International Nuclear Information System (INIS)

Schoenfeld, T.

1959-01-01

Protection against ionizing radiation given off in nuclear transformations is one of the foremost safety problems in all atomic energy operations. While every effort is being made to prevent reactors, processing plants and all other installations from releasing radioactive materials into the biosphere - air, water and earth - under any foreseeable conditions, small amounts of it are actually released into man's living space. Undoubtedly, this will continue to be so, at least for the time being. For example, low activity liquid wastes from some chemical processing plants are decontaminated in special processes, but traces of fission products remain in the liquids finally discharged on the ground or to nearby waterways. In some installations low and medium activity liquid wastes are even released on the ground or into swamps without prior decontamination. It is also to be expected that in accidents larger amounts of fission products may occasionally be released. To make the routine release of small amounts of fission products safe and to be able to estimate the possible effect of larger releases in accidents, a considerable amount of information is required

ENDF/B-6 fission-product yield sublibraries

International Nuclear Information System (INIS)

Lemmel, H.D.

1994-01-01

The contents and the documentation of the ENDF/B-6 fission-product yield sublibraries which were released in 1991 and updated in 1993, are summarized. Copies of the data libraries are available on magnetic tape of PC diskettes from the IAEA Nuclear Data Section, costfree upon request. (author). 1 tab

The use of recoil for the separation of uranium fission products; Utilisation du recul pour la separation des produits de fission de l'uranium

Energy Technology Data Exchange (ETDEWEB)

Henry, R; Herczec, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

The recoil distance of fission fragments in U{sub 3}O{sub 8} is about 8 microns. By using highly diluted suspensions of uranium oxide particles having dimension much smaller than this figure (mean diameter 0,5 micron), we were able to study the re-adsorption of fission products on uranium oxide. Separation results have been studied as a function of the nature of the irradiation medium (solid or liquid) and the separation medium, of particle size and of concentration of particles in the dispersing medium. Decay curves can be used to discriminate between {sup 239}Np and mixed fission products. Most of the {sup 239}Np is found in the U{sub 3}O{sub 8} particles. The location of fission products in solid dispersing media has been determined, fission products being found always inside the dispersing medium particles. The results obtained can be applied to the rapid separation of short-lived fission products from a uranium-free starting material. (author) [French] Le parcours de recul des fragments de fission est en moyenne de 8 microns dans l'U{sub 3}O{sub 8}. En prenant des suspensions d'oxyde d'uranium dont les particules, tres diluees, ont des dimensions nettement inferieures a cette valeur (diametre moyen 0,5 micron), on a pu etudier directement la readsorption des produits de fission sur l'oxyde d'uranium. Les resultats de separation ont ete etudies en fonction de la nature du milieu d'irradiation (solide ou liquide) et du milieu de separation, de la taille des particules d'oxyde et de leur concentration dans le milieu dispersant. Les courbes de decroissance permettent de determiner la perturbation apportee dans les mesures par le {sup 239}Np qui reste en majorite dans les grains d'U{sub 3}O{sub 8}. On a determine enfin l'emplacement des produits de fission dans le cas des melanges solides; ils se trouvent toujours a l'interieur des grains du milieu recepteur. Les resultats obtenus permettent d'envisager la separation rapide de produits de fission a periode courte a

A separate effect study of the influence of metallic fission products on CsI radioactive release from nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Di Lemma, F.G., E-mail: fidelma.dilemma@gmail.com [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands); Colle, J.Y., E-mail: jean-yves.colle@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Beneš, O. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Konings, R.J.M. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands)

2015-10-15

The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas–gas and gas–aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

International Nuclear Information System (INIS)

Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

1999-01-01

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

Directory of Open Access Journals (Sweden)

Salahuddin Asif

2013-01-01

Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

Quantitative analysis of fission products by γ spectrography

International Nuclear Information System (INIS)

Malet, G.

1962-01-01

The activity of the fission products present in treated solutions of irradiated fuels is given as a function of the time of cooling and of the irradiation time. The variation of the ratio ( 144 Ce + 144 Pr activity)/ 137 Cs activity) as a function of these same parameters is also given. From these results a method is deduced giving the 'age' of the solution analyzed. By γ-scintillation spectrography it was possible to estimate the following elements individually: 141 Ce, 144 Ce + 144 Pr, 103 Ru, 106 Ru + 106 Rh, 137 Cs, 95 Zr + 95 Nb. Yield curves are given for the case of a single emitter. Of the various existing methods, that of the least squares was used for the quantitative analysis of the afore-mentioned fission products. The accuracy attained varies from 3 to 10%. (author) [fr

Low-level seaweed supplementation improves iodine status in iodine-insufficient women.

Science.gov (United States)

Combet, Emilie; Ma, Zheng Feei; Cousins, Frances; Thompson, Brett; Lean, Michael E J

2014-09-14

Iodine insufficiency is now a prominent issue in the UK and other European countries due to low intakes of dairy products and seafood (especially where iodine fortification is not in place). In the present study, we tested a commercially available encapsulated edible seaweed (Napiers Hebridean Seagreens® Ascophyllum nodosum species) for its acceptability to consumers and iodine bioavailability and investigated the impact of a 2-week daily seaweed supplementation on iodine concentrations and thyroid function. Healthy non-pregnant women of childbearing age, self-reporting low dairy product and seafood consumption, with no history of thyroid or gastrointestinal disease were recruited. Seaweed iodine (712 μg, in 1 g seaweed) was modestly bioavailable at 33 (interquartile range (IQR) 28-46) % of the ingested iodine dose compared with 59 (IQR 46-74) % of iodine from the KI supplement (n 22). After supplement ingestion (2 weeks, 0·5 g seaweed daily, n 42), urinary iodine excretion increased from 78 (IQR 39-114) to 140 (IQR 103-195) μg/l (Pseaweed was palatable and acceptable to consumers as a whole food or as a food ingredient and effective as a source of iodine in an iodine-insufficient population. In conclusion, seaweed inclusion in staple foods would serve as an alternative to fortification of salt or other foods with KI.

Radioactive iodine and environmental and sanitary effects - bibliographic study and quantification; Iodes radioactifs et impacts environnemental et sanitaire - etude bibliographique et quantification

Energy Technology Data Exchange (ETDEWEB)

Guetat, Ph.; Armand, P.; Monfort, M.; Fritsch, P. [CEA Bruyeres-le-Chatel, 91 (France); Flury Herard, A. [CEA, Dir. des Sciences du Vivant, 75 - Paris (France); Menetrier, F. [CEA Fontenay-aux-Roses, Dir. des Sciences du Vivant, 92 (France); Bion, L. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France); Schoech, C.; Masset, S. [Societe EX-IN - Expertise et Ingenierie, 92 - Le Plessis-Robinson (France)

2004-07-01

This document is intended to a large public. It reviews the different parameters needed to evaluate the potential act o radioactive releases from the emission to public. Its objectives are to evaluate the importance of different exposure pathways and to assess efficiency of the possible interventions for large public. The main conclusions are summarised hereafter: The radioactive decay chains have to be taken into account to evaluate the iodine source term in the nuclear plants in the case of fission accidents. The physico-chemical forms of iodine are important in order to determine the released activity and deposited activity on the soil. The isotopes to be taken into account are mainly iodine 131 for radiological assessments and also iodine 133 for the nuclear reactor accidents, and the chain Tellurium-Iodine 132 when no particulate filtration exists. Iodine 129 in French reprocessing plant cannot lead to significant accidents. The dominant exposure pathways are related to the consumption of contaminated food products (vegetable, milk) for the inorganic iodine. The iodine transfer to goat and sheep milk is greater than the one to cow milk. The meat production of herbivores at field is the most sensitive. The interest to remove rapidly herbivore from pasture appears relatively clearly. The banning of consumption of local contaminated food products (vegetables and meats) may reduce by about a factor of thirteen the impact due to iodine 131. The youngest the population is, the greatest are the thyroid radiosensitivity and variability within the population. Oral administration of stable iodine limits transfers to maternal milk and foetal thyroid. Ingestion of stable iodine is complementary to consumption banning of local contaminated food products. The earliest the ingestion is, the greatest is the efficiency. 0,1 TBq of 131 iodine released at a low height involves only limited and local actions whereas the release of 10 TBq involves direct and immediate protection

Direct and preequilibrium effects in the fission-product mass range

International Nuclear Information System (INIS)

Gruppelaar, H.; Hogenbirk, A.

1992-07-01

Until recently inelastic scattering did not gain the proper attention in fission-product cross section evaluations. In many existing evaluations global spherical optical models have been used, neglecting direct and pre-equilibrium effects. There are also few experimental data relevant to inelastic scattering in fission products. This paper is focussed on the anomalously high inelastic scattering cross sections observed in even-mass nuclei near mass A=100 at low energies. Both more data and more refined theoretical analyses are required. A number of suggestions for relevant coupled-channel calculations is made. (author). 29 refs., 5 figs., 1 tab

Transmutation of fission products in reactors and accelerator-driven systems