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Sample records for chemical burnup determination

  1. Chemical separation for the burnup determination of the U3Si/Al spent fuels

    International Nuclear Information System (INIS)

    The separation of U, Pu, and Nd for the burnup determination of the U3Si/Al spent fuel samples has been studied. The preliminary experiments were carried out with the simulated spent fuel solution. The solutions were prepared by adding of fission product elements to unirradiated U3Si/Al fuel samples. The fuel samples were dissolved in 6 M HNO3, 6 M HNO3 using mercury catalyst, or applying a mixture of HCl and HNO3 without any catalyst. All dissolved fuel solutions contained a small amount of a residue(silica). The trace silica reprecipitated from the fuel solutions taken for the separation was dissolved in HF and removed by subsequent evaporating to dryness. The separation of U and fission product elements from the various sample solutions was achieved by two sequential anion exchange resin separation procedures. The U, Pu and Nd can be purely isolated from the sample solutions with a large excess of Al by this chromatographic procedures. The dissolution and separation procedure used in this experiment were applied for burnup determination of real U3Si/Al spent fuels from HANARO reactor

  2. Chemical analytical considerations on the determination of burnup in irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Burnup in an irradiated nuclear fuel may be defined as the energy produced per mass unit, from the time the fuel is introduced into the reactor and until a given moment. It is usually shown in megawatt/day or megawatt/hour generated per ton or kilo of fuel. It is also indicated as the number of fission produced per volume unit (cm3) or per every 100 initial fissionable atoms. The yield of a power plant is directly related to the burnup of its fuel load and knowing the latter contributes to optimizing the economy in reactor operation and the related technologies. The development of nuclear fuels and the operation of reactors require doing with exact and accurate methods allowing to know the burnup. Errors in this measurement have an incidence upon the fuel design, the physical and nuclear calculations, the shielding requirements, the design of vehicles for the transportation of irradiated fuels, the engineering of processing plants, etc. All these factors, in turn, have an incidence upon the cost of nuclear power generation. (Author)

  3. Quantitative burnup determination: A comparison of different experimental methods

    International Nuclear Information System (INIS)

    The burn-up of nuclear fuel is defined as the energy produced per mass of fuel and, hence, is related to the inventory of fission products formed in the matrix of the fuel. It affects both neutron-physical and material properties. Therefore, it is essential to have methods available that allow a reliable determination of this important parameter. The burn-up is usually determined by measuring the content of an element that results from the fission process. The isotope 148Nd has proven to be an ideal monitor due to its chemical and neutron physical properties. On the other hand, 148Nd can only be determined by wet-chemistry methods, which means a rather costly and time consuming chemistry process. Another method using the sum of 145Nd and 146Nd is proposed. In case of very high burn-ups of U02 fuel and, especially, MOX fuel this method needs weighed yields for U and Pu to obtain a sufficient accuracy. Among the non-destructive spectrometric methods, the burn-up determination with 137Cs provides adequate results provided the gamma radiation detector is calibrated and self-attenuation effects of Cs together with measurement geometries are considered. (Author)

  4. Determination of research reactor fuel burnup

    International Nuclear Information System (INIS)

    This report was prepared by a Consultants Group which met during 12-15 June 1989 at the Jozef Stefan Institute, Yugoslavia, and during 11-13 July 1990 at the IAEA Headquarters in Vienna, Austria, with subsequent contributions from the Consultants. The report is intended to provide information to research reactor operators and managers on the different, most commonly used methods of determining research reactor fuel burnup: 1) reactor physics calculations, 2) measurement of reactivity effects, and 3) gamma ray spectrometry. The advantages and disadvantages of each method are discussed. References are provided to assist the reactor operator planning to establish a programme for burnup determination of fuel. Destructive techniques are not included since such techniques are expensive, time consuming, and not normally performed by the reactor operators. In this report, TRIGA fuel elements are used in most examples to describe the methods. The same techniques however can be used for research reactors which use different types of fuel elements. 22 refs, 13 figs, 2 tabs

  5. Burnup determination of water reactor fuel

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency in consultation with the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The meeting was hosted by the Commission of the European Communities, at the Transuranium Research Laboratory, Joint Research Centre Karlsruhe, in the Federal Republic of Germany. This subject was dealt with for the first time by the IAEA. It was found to correspond adequately to this type of Specialist Meeting and to be suitable at a moment when the extension of burnup constitutes a major technical and economical issue in fuel technology. It was stressed that analysis of highly burnt fuels, mixed oxides and burnable absorber bearing fuels required extension of the experimental data base, to comply with the increasing demand for an improved fuel management, including better qualification of reactor physics codes. Twenty-seven participants from eleven countries plus two international organizations attended the Meeting. Twelve papers were given during three technical sessions, followed by a panel discussion which allowed to formulate the conclusions of the meeting and recommendations to the Agency. In addition, participants were invited to give an outline of their national programmes, related to Burnup Determination of Water Reactor Fuel. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

  6. Automated system for determining the burnup of spent nuclear fuel

    Directory of Open Access Journals (Sweden)

    Mokritskii V. A.

    2014-12-01

    Full Text Available The authors analyze their experience in application of semi-conductor detectors and development of a breadboard model of the monitoring system for spent nuclear fuel (SNF. Such system should use CdZnTe-detectors in which one-charging gathering conditions are realized. The proposed technique of real time SNF control during reloading technological operations is based on the obtained research results. Methods for determining the burnup of spent nuclear fuel based on measuring the characteristics of intrinsic radiation are covered in many papers, but those metods do not usually take into account that the nuclear fuel used during the operation has varying degrees of initial enrichment, or a new kind of fuel may be used. Besides, the known methods often do not fit well into the existing technology of fuel loading operations and are not suitable for operational control. Nuclear fuel monitoring (including burnup determination system in this research is based on the measurement of the spectrum of natural gamma-radiation of irradiated fuel assemblies (IFA, as from the point of view of minimizing the time spent, the measurement of IFA gamma spectra directly during fuel loading is optimal. It is the overload time that is regulated rather strictly, and burnup control operations should be coordinated with the schedule of the fuel loading. Therefore, the real time working capacity of the system should be chosen as the basic criterion when constructing the structure of such burnup control systems.

  7. Chemical form of fission products in high burnup fuels

    International Nuclear Information System (INIS)

    In order to make a proper assessment of candidate materials for advanced high-burnup fuels, thermochemical studies of fuel materials have been performed. Using data from the ECN thermochemical database (TBASE), which has been updated and extended for the present work, the suitability of various advanced fuel materials and inert matrices is studied. Detailed thermodynamic equilibrium calculations are performed for Pu0.42U0.58O2 and Pu0.40U0.60N for values of the burnup up to 200 MWd/kgHM. The formation of metallic phases, the pressure buildup and the stability of nitride or oxide phases is studied for each fuel type. The results for the chemical form of the solid fission products are given. The chemical aspects of the use of the inert matrix spinel (MgAl2O4) in combination with oxide fuel will be discussed. Experimental research on the compatibility of various types of inert matrices (nitrides, spinel) is in progress at ECN. (author)

  8. Destructive radiochemical analysis of uraniumsilicide fuel for burnup determination

    Energy Technology Data Exchange (ETDEWEB)

    Gysemans, M.; Bocxstaele, M. van; Bree, P. van; Vandevelde, L.; Koonen, E.; Sannen, L. [SCK-CEN, Boeretang, Mol (Belgium); Guigon, B. [CEA, Centre de Cadarache, Saint Paul lez Durance (France)

    2004-07-01

    During the design phase of the French research reactor Jules Horowitz (RJH) several types of low enriched uranium fuels (LEU), i.e. <20% {sup 235}U enrichment, are studied as possible candidate fuel elements for the reactor core. One of the LEU fuels that is taken into consideration is an uraniumsilicide based fuel with U{sub 3}Si{sub 2} dispersed in an aluminium matrix. The development and evaluation of such a new fuel for a research reactor requires an extensive testing and qualification program, which includes destructive radiochemical analysis to determine the burnup of irradiated fuel with a high accuracy. In radiochemistry burnup is expressed as atom percent burnup and is a measure for the number of fissions that have occurred per initial 100 heavy element atoms (%FIMA). It is determined by measuring the number of heavy element atoms in the fuel and the number of atoms of selected key fission products that are proportional to the number of fissions that occurred during irradiation. From the few fission products that are suitable as fission product monitor, the stable Nd-isotopes {sup 143}Nd, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148Nd}, {sup 150}Nd and the gamma-emitters {sup 137}Cs and {sup 144}Ce are selected for analysis. Samples form two curved U{sub 3}Si{sub 2} plates, with a fuel core density of 5.1 and 6.1 g U/cm{sup 3} (35% {sup 235}U) and being irradiated in the BR2 reactor of SCK x CEN{sup [1]}, were analyzed. (orig.)

  9. Burnup determination and age dating of spent nuclear fuel using noble gas isotopic analysis

    International Nuclear Information System (INIS)

    During the chopping and dissolving phases of reprocessing, gases (such as tritium, krypton, xenon, iodine, carbon dioxide, nitrogen oxide, and steam) are released. These gases are traditionally transferred to a gas-treatment system for treatment, release, and/or recycle. Because of their chemically inert nature, the xenon and krypton noble gases are generally released directly into the loser atmosphere through the facility's stack. These gases (being fission products) contain information about the fuel being reprocessed and may prove a valuable monitor of reprocessing activities. Two properties of the fuel that may prove valuable from a safeguards standpoint are the fuel burnup and the fuel age (or time since discharge from the reactor). Both can be used to aid in confirming declared activities, and the burnup is generally indicative of the usability of the fuel for fabricating nuclear explosives. A study has been ongoing at Los Alamos National Laboratory to develop a methodology to determine spent-fuel parameters from measured xenon and/or krypton isotopic ratios on-stack at reprocessing facilities. This study has resulted in the generation of the NOVA data analysis code, which links to a comprehensive database of reactor physics parameters (calculated using the Monteburns 3.01 code system). NOVA has been satisfactorily tested for burnup determination of weapons-grade fuel from a US production reactor. Less effort has been spent quantifying NOVA's ability to predict burnup and fuel age for power reactor fuel. The authors describe the results predicted by NOVA for xenon and krypton isotopic ratios measured after the dissolution of spent-fuel samples from the Borssele reactor. The Borssele reactor is a 450-MW(electric) pressurized water reactor (PWR) consisting of 15 x 15 KWU assemblies. The spent-fuel samples analyzed were single fuel rods removed from one assembly and dissolved at the La Hague reprocessing facility. The assembly average burnup was estimated at 32

  10. Determination of the accuracy of utility spent fuel burnup records. Interim report

    International Nuclear Information System (INIS)

    In order to develop a NRC-licensable burnup credit methodology, the pedigree and uncertainty of commercial spent nuclear fuel assembly burnup records needs to be established. Typically the assembly average burnup for each assembly is maintained in the plant records. It is anticipated that the repository for the disposal of spent fuel will utilize burnup credit and will require knowledge of the uncertainty of reactor burnup records. The uncertainty of the assembly average burnup record depends on the uncertainty of the method used to develop the record. Such records are generally based on core neutronic analysis coupled with analysis of in-core power detector data. This report evaluates the uncertainties in the burnup of fuel assemblies utilizing in-core measurements and core neutronic calculations for a Westinghouse PWR. To quantify the uncertainty, three cycles of in-core movable detector data were used. The data represents a first cycle of operation, a transition cycle and a low leakage cycle. These three cycles of data provide a true test of the uncertainty methodology. Three separate sets of results were used to characterize the burnup uncertainty of the fuel assemblies. The first set of results compared the measured and calculated reaction rates in instrumented assemblies and determined the uncertainty in the reaction rates. The second set of results determined the uncertainty in relative assembly power for both the instrumented and un-instrumented assemblies. The third set of results determined the burnup uncertainty of the discharged fuel in each cycle

  11. Methods used in burn-up determination of the irradiated fuel rods at TRIGA reactor

    International Nuclear Information System (INIS)

    A short presentation of the methods used at INR TRIGA reactor for the burn-up determination is given together with some considerations on ORIGEN 2 computer code used for calculating fission products activities and nuclide concentration. Burn-up is determined by gamma spectroscopy and thermal power monitoring. (Author)

  12. Burnup determination in irradiated fuel by means of isotopic analysis and comparison to CASMO calculations

    International Nuclear Information System (INIS)

    One of the traditional methods for determining the burnup of irradiated Light Water Reactor (LWR) fuel is the 148Nd method according to ASTM E-321. Probably one of the largest sources for systematic errors in this method is the assumed fission yield, requiring knowledge of the fraction of fissions occurring in different fissile nuclides. Another traditional method for burnup determination is based on the uranium and plutonium isotopic composition; however, this method is rarely used for LWR fuel due to its rather simplified and rough assumptions regarding the neutron spectrum and fission fractions. However, modern physics codes like CASMO and HELIOS are instead able to calculate the amount of fission products and actinides formed or consumed during reactor operation in a much more sophisticated way. Isotopic Dilution Analysis with chemical separation of elements of interest, followed by isotopic analysis with a Thermal Ionization Mass Spectrometer (TIMS) is a well established method for determining the content of selected isotopes in samples of dissolved irradiated fuel. This method normally provides very accurate and precise results. High Performance Liquid Chromatography (HPLC) for elemental separations, combined with Inductively Coupled Plasma Mass Spectrometry (ICP-MS) has become a much faster alternative. In general, this method is somewhat less precise. This disadvantage is at least partly compensated by the possibility of analyzing a larger number of nuclides and samples. The local pellet burnup of a well characterised fuel sample irradiated in the Swedish Boiling Water Reactor Forsmark 3 to about 60 MWd/kgU was determined. Weight ratios of neodymium isotopes relative to 238U, analysed by Isotope Dilution Analysis applying HPLC-ICP-MS as well as 235U and 239Pu abundance values were compared to corresponding values calculated by a single-assembly CASMO-4 simulation. Input data were generated by CASMO-4/POLCA7 core tracking calculations. The overall result

  13. Determination of the burn-up of TRIGA fuel elements by calculation and reactivity experiments

    International Nuclear Information System (INIS)

    The burnup of 17 fuel elements of the TRIGA Mark-II reactor in Vienna was measured. Different types of fuel elements had been simultaneously used for several years. The measured burnup values are compared with those calculated on the basis of core configuration and reactor operation history records since the beginning of operation. A one-dimensional, two-group diffusion computer code TRIGAP was used for the calculations. Comparison with burnup values determined by γ-scanning is also made. (orig./HP)

  14. Radiochemical burnup determination and isotope analysis of four IFA 148 samples: Ris/o/ Fission Gas Project

    Energy Technology Data Exchange (ETDEWEB)

    Mogensen, M.; Larsen, E.; Funck, J.; Strauss, T.R.

    1981-06-01

    In the frame of the Ris/o/ Fission Gas Project, radiochemical burnup determinations (Nd-148) and heavy isotope analyses have been carried out on four IFA 148 samples. The Nd-148 burnup determinations serve as a calibration of the Cs-137 gamma scans so that the absolute burnup can be determined in all axial positions from the Cs-137 curves and the Nd-148 burnup results. The analysis of uranium and transuranium isotopes have the objective of establishing the distribution of the total burnup on U-235, Pu-239 and Pu-241 fissions. 6 refs., 9 figs., 3 tabs.

  15. Burn-up determination of irradiated thoria samples by isotope dilution-thermal ionisation mass spectrometry

    International Nuclear Information System (INIS)

    Burn-up was determined experimentally using thermal ionization mass spectrometry for two samples from ThO2 bundles irradiated in KAPS-2. This involved quantitative dissolution of the irradiated fuel samples followed by separation and determination of Th, U and a stable fission product burn-up monitor in the dissolved fuel solution. Stable fission product 148Nd was used as a burn-up monitor for determining the number of fissions. Isotope Dilution-Thermal Ionisation Mass Spectrometry (ID-TIMS) using natural U, 229Th and enriched 142Nd as spikes was employed for the determination of U, Th and Nd, respectively. Atom % fission values of 1.25 ± 0.03 were obtained for both the samples. 232U content in 233U determined by alpha spectrometry was about 500 ppm and this was higher by a factor of 5 compared to the theoretically predicted value by ORIGEN-2 code. (author)

  16. Determination of burn-up of irradiated PHWR fuel samples from KAPS-1 by mass spectrometry

    International Nuclear Information System (INIS)

    Burn-up was determined experimentally using thermal ionization mass spectrometry for three spent UO2 fuel samples, which had undergone extended irradiation in Kakrapar Atomic Power Station Unit 1 (KAPS-1). The method involves dissolution of the irradiated fuel sample, separation and determination of burn-up monitor, uranium and plutonium. Isotope Dilution-Thermal Ionisation Mass Spectrometry (ID-TIMS) using Triple Spike Mixture consisting of (142Nd+233U+242Pu) was employed for the concentration determination of Nd, U and Pu in the dissolved fuel samples. The atom percent fission was calculated based on 148Nd as a burn-up monitor and also from the changes in the abundances of heavy element isotopes. Fractional fission contributions from the major fissile nuclides were calculated from heavy elemental data and also from the Nd isotopic ratios. (author)

  17. A computer program for nuclear fuel burnup determination using gamma spectrometric methods

    International Nuclear Information System (INIS)

    In the end of its service life in the reactor, the fuel needs to be characterized for reasons relating both to safety and economy. The main investigations carried out are oriented towards verifying the fuel cladding integrity and determining the fissile content and the fuel burnup. A computer program for fast burnup evaluation was developed at the Post-Irradiation Examination Laboratory (PIEL) from INR Pitesti, the only laboratory of this kind in Romania. The input data consists, on one hand, of axial and radial gamma-scanning profiles (for the experimental evaluation of the number of nuclei of a given fission product - selected as burnup monitor - in the end of irradiation) and, on the other hand, of the history of irradiation (the time length and relative value of the neutron flux for each step of irradiation). Using the equation for the build-up and decay of the burnup monitor during irradiation the flux value is iteratively adjusted until the calculated number of nucleus is equal to the experimental one. Then the flux value is used in the equations of evolution of the fissile and fertile nuclei to determine the number of fissions and consequently the fuel burnup. The program was successfully used in the analysis of more then one hundred of TRIGA and CANDU-type fuel rods. An experimental result is reported in some details. (authors)

  18. Burnup determination of power reactor fuel elements by gamma spectrometry

    International Nuclear Information System (INIS)

    This report describes a method for determining by γ spectrometry the burn up and the specific power of fuel elements irradiated in power reactors. The energy spectrum of γ rays emitted by fission products is measured by means of a simple equipment using a sodium iodide detector and a multichannel analyzer. In order to extract from the spectrum a quantity proportional to the burn up, it is necessary to: - isolate an activity specific of one emitter,- give the same importance to fissions in uranium and plutonium - take into account the radioactive decay during and after irradiation. One hundred fuel elements were studied and burn up values obtained by γ spectrometry are compared to results given by chemical analyses. Preliminary measurements show that the accuracy of the results is greatly increased by the use of a germanium detector, due to its good resolution. (authors)

  19. Non-destructive burnup determination of PWR spent fuel using Cs-134/Cs-137 and Eu-154/Cs-137

    International Nuclear Information System (INIS)

    Burnups for 36 points of five rods in the G23 assembly of Kori unit 1 have been determined on the basis of gamma-ray spectrometric measurement of two isotopic ratios, Cs-134/Cs-137 and Eu-154/Cs-137 in combination with the results calculated by the SCALE4.4 SAS2H module. Benchmarking of the SAS2H module has been done for the existing experimental data of Cs-13134, Cs-137 and Eu-154 isotopic compositions in PWR spent fuel. The gamma ray counts of two isotopic ratios have been corrected with their branching ratios, decay rates and energy dependent counting efficiencies in order to get true ratios. The energy dependent counting efficiencies have been determined as a quadratic equation based on the gamma ray counts for Cs-134 and Eu-154 at fourth energy points. Finally, burnups have been determined by putting true ratios of two isotopic ratios to their burnup-to-ratio fitting functions, respectively. Then the measured burnups have been compared with the declared burnup by the nuclear power plant. It is revealed that burnups determined from Cs-134/Cs-137 are agreeable with the declared burnups in most cases within about 12% error except a measuring point of C13, one of G23 fuel rods. In the case of Eu-154/Cs-137, the measured burnup is much lower than the declared burnup, which seems to be derived from system errors. (author)

  20. Development of high performance liquid chromatography for rapid determination of burn-up of nuclear fuels

    International Nuclear Information System (INIS)

    Burn-up an important parameter during evaluation of the performance of any nuclear fuel. Among the various techniques available, the preferred one for its determination is based on accurate measurement of a suitable fission product monitor and the residual heavy elements. Since isotopes of rare earth elements are generally used as burn-up monitors, conditions were standardized for rapid separation (within 15 minutes) of light rare earths using high performance liquid chromatography based on either anion exchange (Partisil 10 SAX) in methanol-nitric acid medium or by cation exchange on a reverse phase column (Spherisorb 5-ODS-2 or Supelcosil LC-18) dynamically modified with 1-octane sulfonate or camphor-10-sulfonic acid (β). Both these methods were assessed for separation of individual fission product rare earths from their mixtures. A new approach has been examined in detail for rapid assay of neodymium, which appears promising for faster and accurate measurement of burn-up. (author)

  1. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    Energy Technology Data Exchange (ETDEWEB)

    Ozdemir, Levent, E-mail: levent.ozdemir@taek.gov.tr [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey); Acar, Banu Bulut; Zabunoglu, Okan H. [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey)

    2011-02-15

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of {sup 239}Pu and {sup 241}Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  2. Determination of burnup balance for nuclear reactor fuel on the basis of γ-spectrometric determination of fission products

    International Nuclear Information System (INIS)

    Results are given of experimental investigations in one of the versions of the method for determination of the balance of nuclear fuel burnup process by means of the γ-spectrometry of fission products. In the version being considered a balance of the burnup process was determined on the base of 106Ru, 134Cs.Activity was measured by means of a γ-spectrometer with Ge counter. Investigations were done on the natural uranium metal fuel from the heavy-water moderated reactor of the first Czechoslovakian nuclear power plant A1 in Yaslovske Bohunice. Possibility was checked of determination of the fuel burnup depth as well as of the isotope ratio and content of plutonium. Results were compared with the control data which had been obtained on the base of the mass-spectrometry of U, Pu and Nd. The reasors for deviations were estimated in the cases when they were greater tan error in the control data

  3. ORIGEN computer code use in non-destructive analysis of irradiated fuel elements for burn-up determination

    International Nuclear Information System (INIS)

    An iterative method for burn-up determination in the non-destructive analysis of irradiated fuel elements using the ORIGEN computer code is presented. On the bases of data obtained from ORIGEN code the calibration coefficient for the neutron flux is determined as a function of one fission product activity while the burn-up is determined as a function of the calibration coefficient for a given irradiation history. These functions are used for determining the burn-up of nuclear fuel elements measured by gamma-scanning. The method is tested for fuel elements irradiated in a TRIGA reactor facility. (Author)

  4. Determination of deuterium–tritium critical burn-up parameter by four temperature theory

    Energy Technology Data Exchange (ETDEWEB)

    Nazirzadeh, M.; Ghasemizad, A. [Department of Physics, University of Guilan, 41335-1914 Rasht (Iran, Islamic Republic of); Khanbabei, B. [School of Physics, Damghan University, 36716-41167 Damghan (Iran, Islamic Republic of)

    2015-12-15

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  5. Determination of deuterium-tritium critical burn-up parameter by four temperature theory

    Science.gov (United States)

    Nazirzadeh, M.; Ghasemizad, A.; Khanbabei, B.

    2015-12-01

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  6. Determination of the fuel element burn-up for mixed TRIGA core by measurement and calculation with new TRIGLAV code

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)

  7. Burnup calculations and chemical analysis of irradiated fuel samples studied in LWR-PROTEUS phase II

    International Nuclear Information System (INIS)

    The isotopic compositions of 5 UO2 samples irradiated in a Swiss PWR power plant, which were investigated in the LWR-PROTEUS Phase II programme, were calculated using the CASMO-4 and BOXER assembly codes. The burnups of the samples range from 50 to 90 MWd/kg. The results for a large number of actinide and fission product nuclides were compared to those of chemical analyses performed using a combination of chromatographic separation and mass spectrometry. A good agreement of calculated and measured concentrations is found for many of the nuclides investigated with both codes. The concentrations of the Pu isotopes are mostly predicted within ±10%, the two codes giving quite different results, except for 242Pu. Relatively significant deviations are found for some isotopes of Cs and Sm, and large discrepancies are observed for Eu and Gd. The overall quality of the predictions by the two codes is comparable, and the deviations from the experimental data do not generally increase with burnup. (authors)

  8. Prototype studies on the nondestructive online burnup determination for the modular pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: • Prototype study of online burnup measurement for HTR proves its feasibility. • Calibration and its correction of burnup assay device is discussed and verified. • Analysis of simulated gamma spectra shows good performance of spectra-unfolding method. - Abstract: The online fuel pebble burnup determination in future modular pebble bed reactor is implemented by measuring nondestructively the activity of the monitoring nuclide Cs-137 with HPGe detector on a pebble-by-pebble basis. Based on a full size prototype the feasibility is investigated. The prototype was first tested by using double sources to show that a precision of 2.8% (1σ) can be achieved in the determination of the Cs-137 net counting rate. Then, the relationship between the Cs-137 activity and the net counting rate recorded in the HPGe detector is calibrated with a standard Cs-137 source contained in the center of a graphite sphere with the same dimension as a real fuel pebble. Because the self attenuation of the calibration source differs with a fuel pebble, a correction factor of 1.07 ± 0.02 (p = 0.95) to the calibration is derived by using the efficiency transfer method. Last, by analyzing the spectra generated with KORIGEN software followed by Monte Carlo simulation, it is predicted that the relative standard deviation of the Cs-137 net counting rate can be still controlled below 3.5% despite of the presence of all the interfering peaks. The results demonstrate the feasibility of utilizing HPGe gamma spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors

  9. Uranium and plutonium determinations for evaluation of high burnup fuel performance

    International Nuclear Information System (INIS)

    Purpose of this work is to experimentally test computational methods being developed for reactor fuel operation. Described are the analytical techniques used in the determination of uranium and plutonium compositions on PWR fuel that has spanned five power cycles, culminating in 55,000 to 57,000 MWd/T burnup. Analyses have been performed on ten samples excised from selected sections of the fuel rods. Hot cell operations required the separation of fuel from cladding and the comminution of the fuel. These tasks were successfully accomplished using a SpectroMil, a ball pestle impact grinding and blending instrument manufactured by Chemplex Industries, Inc., Eastchester, New York. The fuel was dissolved using strong mineral acids and bomb dissolution techniques. Separation of the fuel from fission products was done by solvent (hexone) extraction. Fuel isotopic compositions and assays were determined by the mass spectrometric isotope dilution (MSID) method using NBS standards SRM-993 and SRM-996. Alpha spectrometry was used to determine the 238Pu composition. Relative correlations of composition with burnup were obtained by gamma-ray spectrometry of selected fission products in the dissolved fuel

  10. Analysis of neodymium 148 in order to determin of nuclear fuel burnup

    International Nuclear Information System (INIS)

    To determine the degree of the nuclear fuel burnup experiments were conducted to introduce improvements in the mass-spectrometric study of neodymium-148 by the method of isotopic dilution with Nd-150 taken as a diluent. The separation of neodymium out of the mixture of the fission products and uranium was carried out in two stages. In the first stage a group of rare earth elements was isolated on the Vofatit SBV anionite in the mixture of nitric acid and methanol. The second stage involved the separation of the rare earth group on the Vofatite KPS cationite with the aid of the complexing agent of α-hydroxy-isobutyric acid. To identify the neodymium fraction, the traces of americium-241 were added at elution. The possibilities of the above analytical method are examplified by the isolation of neodymium out of the burned-up fuel of type EK-10. The isotopic ratios were determined by the spectroscopic method to the accuracy of +-1.2%. A highly enriched compound of neodymium-150 was used as a diluent. The factors are discussed affecting the degree of the burnup obtained by this method

  11. Development of destructive methods of burn-up determination and their application on WWER type nuclear fuels

    International Nuclear Information System (INIS)

    Results are described of a cooperation between the Central Institute of Nuclear Research Rossendorf and the Radium Institute 'V.G. Chlopin' Leningrad in the field of destructive burn-up determination. Laboratory methods of burn-up determination using the classical monitors 137Cs, 106Ru, 148Nd and isotopes of heavy metals (U, Pu) as well as the usefulness of 90Sr, stable isotopes of Ru and Mo as monitors are dealt with. The analysis of the fuel components uranium (spectrophotometry, potentiometric titration, mass-spectrometric isotope dilution) and plutonium (spectrophotometry, coulometric titration, mass- and alpha-spectrometric isotope dilution) is fully described. Possibilities of increasing the reproducibility (automatic adjusting of measurement conditions) and the sensibility (ion impuls counting) of mass-spectrometric measurements are proposed and applied to a precise determination of Am and Cm isotopic composition. The methods have been used for burn-up analysis of spent WWER (especially WWER-440) fuel. (author)

  12. Determination of nuclear fuel burnup by non-destructive gamma spectroscopy

    International Nuclear Information System (INIS)

    The determination of nuclear fuel burnup by the non-destructive gamma spectroscopy method is studied. A MTR (Materials Testing Reactor) -type fuel element is used in the measurement. The fuel element was removed from the reactor core in 1958 and, because of the long decay time, show only one peak in is gamma spectrum at 661.6 Kev. Corresponding to 137Cs. Measurements are made at 330 points of the element using a Nal detector and the final result revealed that the quantity of 235U consumed was 3.3 +- 0,8 milligram in the entire element. The effect of the migration of 137Cs in the element is neglected in view of the fact that it occurs only when the temperature is above 10000C, which is not the case in IEAR-1. (Author)

  13. Burn-Up Determination by High Resolution Gamma Spectrometry: Axial and Diametral Scanning Experiments

    International Nuclear Information System (INIS)

    In the gamma spectrometric determination of burn-up the use of a single fission product as a monitor of the specimen fission rate is subject to errors caused by activity saturation or, in certain cases, fission product migration. Results are presented of experiments in which all the resolvable gamma peaks in the fission product spectrum have been used to calculate the fission rate; these results form a pattern which reflect errors in the literature values of the gamma branching ratios, fission yields etc., and also represent a series of empirical correction factors. Axial and diametral scanning experiments on a long-irradiated low-enrichment fuel element are also described and demonstrate that it is possible to differentiate between fissions in U-235 and in Pu-239 respectively by means of the ratios of the Ru-106 activity to the activities of the other fission products

  14. Determination of axial profit performed burnup credit by SCALE 4.3-system

    International Nuclear Information System (INIS)

    SCALE 4.3 is a modular code system designed for realizing standard computational analysis for licensing evaluation. Since now, spent fuel storage pools criticality analysis have been done considering this fuel as fresh, with its maximum enrichment. With burnup credit we can obtain cheaper and compact configurations. The procedure for calculating a spent fuel storage consists of a burnup calculation plus a criticality calculation. We can perform a conservative approximation for the burnup calculations using 1-D results, but, besides the geometry configurations for the 3-D criticality calculation. we need an appropriate approximation to model the burnup axial variation. We assume that for a burnup profile set, the most conservative profile is between the lower and the upper range of this profile, set. We consider only combinations of the maximum and minimum burnup in each axial region, for each burnup range. This gives an estimation of the different burnup shapes effect and the general characteristics of the most conservative shape. (Author) 6 refs

  15. Chemical analyses and calculation of isotopic compositions of high-burnup UO2 fuels and MOX fuels

    International Nuclear Information System (INIS)

    Chemical analysis activities of isotopic compositions of high-burnup UO2 fuels and MOX fuels in CRIEPI and calculation evaluation are reviewed briefly. C/E values of ORIGEN2, in which original libraries and JENDL-3.2 libraries are used, and other codes with chemical analysis data are reviewed and evaluated. Isotopic compositions of main U and Pu in fuels can be evaluated within 10% relative errors by suitable libraries and codes. Void ratio is effective parameter for C/E values in BWR fuels. JENDL-3.2 library shows remarkable improvement compared with original libraries in isotopic composition evaluations of FP nuclides. (author)

  16. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

    International Nuclear Information System (INIS)

    This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ''end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified

  17. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.; DeHart, M.D.

    2000-03-01

    This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.

  18. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Wang Tienko E-mail: tkw@faculty.nthu.edu.tw; Peir Jinnjer

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products {sup 97}Zr/{sup 97}Nb, {sup 132}I, and {sup 140}La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, {sup 235}U burn-up values can be deduced by iterative calculations. The complication caused by {sup 239}Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products {sup 137}Cs, {sup 134}Cs/{sup 137}Cs ratio and {sup 106}Ru/{sup 137}Cs ratio.

  19. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    International Nuclear Information System (INIS)

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio

  20. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry.

    Science.gov (United States)

    Wang, T K; Peir, J J

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

  1. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  2. Burn-up determination of irradiated uranium oxide by means of direct gama spectrometry and by radiochemical method

    International Nuclear Information System (INIS)

    The burn-up of thermal neutrons irradiated U3O8 (natural uranium) samples has been determined by using both direct gamma spectrometry and radiochemical methods and the results obtained were compared. The fission products 144Ce, 103Ru, 106Ru, 137Cs and 95Zr were chosen as burn-up monitors. In order to isolate the radioisotopes chosen as monitors, a radiochemical separation procedure has been established, in which the solvent extraction technique was used to separate cerium, cesium and ruthenium one from the other and all of them from uranium. The separation between zirconium and niobium and of both elements from the other radioisotopes and uranium was accomplished by means of adsorption on a silica-gel column, followed by selective elution of zirconium and of niobium. When use was made of the direct gamma-ray spectrometry method, the radioactivity of each nuclide of interest was measured in presence of all others. For this purpose use was made of gamma-ray spectrometry and of a Ge-Li detector. Comparison of burn-up values obtained by both methods was made by means of Student's 't' test, and this showed that results obtained in each case are statistically equal. (Author)

  3. Determination of Plutonium Contribution to the Total Burnup of a Spent Nuclear Fuel by Mass Spectrometric Measurements of Uranium and Ruthenium

    International Nuclear Information System (INIS)

    The U and Ru isotope patterns provide information on the real irradiation characteristics which are necessary for evaluating a fuel's performance in a reactor. A comparison of the Pu contribution values determined independently provides a promising way to check on the validity of the results. In order to check the consistency of the post-irradiation analysis results, correlations between the parameters of the irradiated nuclear fuels such as the concentration of the heavy elements and fission products, ratios of their isotopes and burnup were established. These correlations can be used to identify the reactor fuels and to estimate the burnup and Pu production. A new approach was carried out with Ru isotopic ratio for the determination of Pu contribution to the total burnup of a spent nuclear fuel from a power reactor. The principle of this approach was based on the use of the difference in the fission yield ratios of the Ru fission products involved for the three main fissionable nuclides such as 235U, 239Pu, and 241Pu. In this work, to determine the contribution of Pu to the total burnup of the fuel, the following two independent methods have been applied: by measuring the isotope ratios of the stable Ru fission products 101Ru/104Ru, and by determining the total burnup by Nd-148 method and subtracting partial burnup, which is determined form the measured values of U isotope ratios

  4. The fuel burn-up determination by combined passive neutron and gamma - spectrometric ND - measurements. Final report for the period 1 July 1980 - 30 April 1988

    International Nuclear Information System (INIS)

    The non-destructive gamma-spectrometric method (HRGS) and the passive neutron technique (PNT) were applied to the determination of WWER 440 reactor spent fuel assemblies burn-up for safeguard purposes. Rapid codes FISPR-2 and BUNECO were compiled on HP-85 and OLIVETTI M24 computers for the IAEA inspectors. An improved equipment was constructed and tested for the fixation of the ''fork detector''. Measurements were carried-out at 1st unit of NPP Bohunice. The correlation between burn-up and fission product ratio 134Cs/137Cs or neutron count rate were analysed. Correlations between concentrations 235U, Pu and burn-up were analysed as well. A simplified procedure for assemblies burn-up check by PNT was proposed for the inspectors. Refs, figs and tabs

  5. Measurement of gamma attenuation coefficients in UO2 and zirconium for self-absorption corrections of burn-up determination

    International Nuclear Information System (INIS)

    UO2 pellets from ALUOX fuel elements were used in measuring the absorption coefficient of gamma radiation in UO2. The results of measurements of the energy dependence of the linear absorption coefficient (within 622 to 796 keV) and of the dependence on pellet density showed that in the given density interval the absorption coefficient was almost constant. The density interval was chosen to be typical for pellet fuel used in water cooled and water moderated power reactors. The results are also shown of the dependence of the mass absorption coefficient of gamma radiation in Zr on radiation energy and compared with the mass absorption coefficient of Mo; these also showed the independence of the absorption coefficient on density. The linear and mass absorption coefficients of UO2 are considerably high and correspond approximately to the absorption coefficient of lead. For the measured energy range the variation of absorption coefficient is about 40%, which causes errors in burnup determination. The efficiency was also determined of Ge(Li) detectors for the energy range 0.5 to 1.2 MeV. The determination of the above coefficients was used for improving the gamma fuel scanning technique in determining the activity and burnup of spent fuel elements. (J.P.)

  6. Nondestructive methods of determination of isotope composition and burnup of spent fuel from WWER-type reeactor

    International Nuclear Information System (INIS)

    Application of the nondestructive methods of analysis (NDA) is discussed being used for determination of burnup and isotope ratio as well as composition of spent fuel elements in the fuel assemblies of WWER-type reactor. Results are discussed which have been obtained by means of the γ-spectrometric method. Prospects are noted of a semiempiric method of determination of the burnup and isotope ratio of the fuel in WWER-type reactors. This method is based on the combination of the data which have been obtained by the γ-spectroscopy and of calculation. NDA has been considered which is based on the registration of self radiation of neutrons from spent fuel elements and assemblies. This method has some advantages as to compare to the γ-spectrometric one which permits to hope for successful, in the case of it's firther elaboration, appliaction of the neutron passive method both in the fuel cycle and in the safeguard system. In the conclusion, prospects are discussed of development and application of the NDA for spent fuels from nuclear power plants

  7. Application of radiochemical-and direct gamma ray spectrometry methods for the determination of the burnup of irradiated uranium oxide

    International Nuclear Information System (INIS)

    The burn-up of U3O8 (natural uranium) samples was determined by using both destructive and non-destructive methods, and comparing the results obtained. The radioisotopes 144Ce, 103Ru, 106Ru, 137Cs and 95Zr were chosen as monitors. In order to isolate the radioisotopes chosen as monitors, a separation scheme has been established in which the solvent extraction technic is used to separate cerium, cesium, and ruthenium one from the other and from uranium. The separation between zirconium and niobium and of both from the others was accomplished by means of adsorption on a silica-gel column. When the non-destructive method was used, the radioactivity of each nuclide of interest was measured in the presence of all others. For this purpose, use was made of gamma-ray spectrometry and a Ge-Li detector. The comparison of burn-up values obtained by both destructive and non-destructive methods was made by means of Student's 't' test, and it has shown that the averages of results obtained in each case are equal. (Author)

  8. Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

    International Nuclear Information System (INIS)

    Highlights: ► Elastic constants of nuclear fuel were measured by acoustic microscopy. ► Our approach was in line with existing literature on non-irradiated material. ► Measurements on several samples of irradiated fuel (HBRP and N118) were performed. ► A decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. ► This trend is in good agreement with measurements conducted with indentation method. - Abstract: We report the measurement of elastic constants of non-irradiated UO2, SIMFUEL (simulated spent fuel: UO2 with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young’s modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.

  9. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However

  10. Experimental Fission Gas Release Determination at High Burnup by Means of Gamma Measurements on Fuel Rods in OL2

    International Nuclear Information System (INIS)

    off between counting statistic and number of fuel rods measured must be exercised. For these measurements, a statistical counting error of 1% or less was achieved for each rod. For the evaluation of the results, a specially designed computer code, LADAKH, was used. The LADAKH program was created in-house at Westinghouse for the sole purpose of determining fission gas release in gamma measured fuel rods. Specifically, LADAKH uses the raw spectrum data along with other inputs such as, for instance, the mechanical characteristics of the fuel rod and individual measurement times to finally determine the percentage of 85Kr released from the fuel matrix. The results showed that the experimentally determined fission gas release agreed well with those values calculated by a fuel performance code in all cases but one, this one case being affected most probably by a relatively large channel bow in that particular assembly. Some efforts were made to evaluate the effect of channel bow on the bundle power distribution and on the rod fission gas release by computer analyses. Another noteworthy point in the fuel performance analyses was that the fission gas release in the rods of a larger diameter were over-predicted by the code, and that this observation was more pronounced when going from four to five cycles of assembly irradiation. Additionally, an estimation based on the amount of fission gas release was done to predict the internal pressure of the fuel rod which, in principle, scales linearly with the fission gas release. In conclusion, all rods were successfully measured for fission gas release and the rod internal pressure was estimated for all rods based upon these measurements. Overall, a successful measurement campaign was conducted adding both valuable data, which will support TVO's burnup increase endeavours as well as additional data for Westinghouse's large data base of measured fuel rods. (authors)

  11. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel

    International Nuclear Information System (INIS)

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However, relevant Xe

  12. Changes of the inventory of radioactive materials in reactor fuel from uranium in changing to higher burn-up and determining the important effects of this

    International Nuclear Information System (INIS)

    The knowledge of the nuclide composition during and after use in the reactor is an essential, in order to be able to determine the effects associated with the operation of nuclear plants. The missing reliable data on the inventory of radioactive materials resulting from the expected change to higher burn-ups of uranium fuels in West Germany are calculated. The reliability of the program system used for this, which permits a one-dimensional account taken of the fuel rod cell and measurement of the changes of specific sets of nuclear data depending on burn-up, is confirmed by the comparison with experimentally found concentrations of important nuclides in fuel samples at Obrigheim nuclear power station. Realistic conditions of use are defined for a range of burn-up of 33 GWd/t to 55 GWd/t and the effects of changes of the number of cycles and the use of types of fuel elements being developed on the composition of the inventory are determined. The plutonium compositions during use in the reactor are given and are tabulated with the inventory for decay times up to 30 years. Effects during change to higher burn-ups are examined and discussed for the maximum inventories during use of fuel and for heat generation during final storage. (orig./HP)

  13. Reactivity effects of nonuniform axial burnup distributions on spent fuel

    International Nuclear Information System (INIS)

    When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup

  14. Reactivity effects of nonuniform axial burnup distributions on spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Leary, R.W. II; Parish, T.A. [Texas A & M Univ., College Station, TX (United States)

    1995-12-01

    When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup.

  15. Burnup determination of silicide MTR fuel elements (20% 235U) in the LFR laboratory

    International Nuclear Information System (INIS)

    The LFR facility is a radiochemical laboratory designed and constructed with a hot-cells line, a glove-box and a fume hood, all of them suited to work radioactive materials. At the beginning of the LFR operation a series of dissolutions of MTR irradiated silicide fuel elements was performed, and determined its isotopic composition of 235U, 239Pu and 148Nd (the last one as burn up monitor), by the thermal ionization mass spectrometry (TIMS). These assays are linked to the IAEA RLA/4/018 Regional Project 'Management of Spent Fuel from Research Reactors'. It is concluded that this technique of burn up measurement is powerful and accurate when properly applied, and permit to validate the calculation codes when isotopic dilution is performed. It is worth noticed the LFR capacity to carry on different research and development programs in the nuclear fuel cycle field, such as the previously mentioned absolute burn up measurements, or the evaluation of radioactive waste immobilization processes and researches on burnable poisons. (author)

  16. Determination of the burn-up in fuels of the MTR type by means of gamma spectroscopy with crystal of INa(Tl)

    International Nuclear Information System (INIS)

    One of the responsibilities of the Laboratory of Analysis by Neutronic Activation of the RA-6 reactor is to determine the burn-up in fuels of the MTR type. In order to gain experience, up to the arrival of the hyperpure Germanium detector (HPGe) to be used in normal operation, preliminary measurements with a crystal of INa(Tl) were made. The fuel elements used are originated in the RA-3 reactor, with a decay superior to the thirteen years. For this reason, the unique visible photoelectric peak is the one of Cs-137, owing to the low resolution of the INa(Tl). After preliminary measurements, the profiles of burn-up, rectified by attenuation, were measured. Once the efficiency of the detector was determined, the calculation of the burn-up was made; for the element No. 144, a value of 21.6 ± 2.9 g was obtained to be compared with the value 21.9 g which was the evaluation made by the operators. (Author)

  17. Burn-up measurement of irradiated rock-like fuels

    International Nuclear Information System (INIS)

    In order to obtain burn-up data of plutonium rock-like (ROX) fuels irradiated at JRR-3M in JAERI, destructive chemical analysis of zirconia or thoria system ROX fuels was performed after development of a new dissolution method. The dissolution method and procedure have been established using simulated ROX fuel, which is applicable to the hot-cell handling. Specimens for destructive chemical analysis were obtained by applying the present method to irradiated ROX fuels in a hot-cell. Isotopic ratios of neodymium and plutonium were determined by mass-spectrometry using the isotope dilution procedure. Burn-up of the irradiated ROX fuels was calculated by the 148Nd procedure using measured data. The burn-ups of thoria and zirconia system fuels that irradiated same location in the capsule showed almost same values. For the ROX fuel containing thorium, 233U was also determined by the same techniques in order to evaluate the effect of burn-up of thorium. As the result, it was found that the fission of 233U was below 1% of total fission number and could be negligible. In addition, americium and curium were determined by alpha-spectrometry. These data, together with isotopic ratio of plutonium, are important data to analyze the irradiation behavior of plutonium. (author)

  18. Determination of dependence of fissile fraction in MOX fuels on spent fuel storage period for different burnup values

    International Nuclear Information System (INIS)

    Highlights: ► In a previous study, an expression to calculate fissile fraction of MOX for various burnups was obtained for 5-year cooled SF. ► In this follow-up study, a correction factor for spent fuel storage periods other than 5 years is derived. ► Thus, one major restriction on use of the expression derived in the initial study is eliminated. - Abstract: The purpose of this technical note is to remove one of the limitations of a derived expression in a previously published article (Özdemir et al., 2011). The original article focused on deriving (computationally) an expression for calculating total fissile fraction of mixed oxid (MOX) fuels depending on discharge burnup of spent fuel and desired burnup of MOX fuel; consequently, such an expression was obtained and put forward, together with its limitations. One of the limitations has been that all the computations and therefore the resulting expression are based on the assumption of a spent fuel storage period of 5 years. This follow-up study simply aims to obtain a correction factor for spent fuel storage periods other than 5 years; thus to remove one major restriction on use of the expression derived in the original article

  19. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    International Nuclear Information System (INIS)

    Highlights: • The burnup of irradiated AGR-1 TRISO fuel was analyzed using gamma spectrometry. • The burnup of irradiated AGR-1 TRISO fuel was also analyzed using mass spectrometry. • Agreement between experimental results and neutron physics simulations was excellent. - Abstract: AGR-1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR-1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non-destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR-1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs-137 activity and the other based on the ratio of Cs-134 and Cs-137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA (fissions per initial heavy metal atom) for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can be determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP-MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma

  20. Determination of Fission Gas Inclusion Pressures in High Burnup Nuclear Fuel using Laser Ablation ICP-MS combined with SEM/EPMA and Optical Microscopy

    International Nuclear Information System (INIS)

    In approximately 20% of all fissions at least one of the fission products is gaseous. These are mainly xenon and krypton isotopes contributing up to 90% by the xenon isotopes. Upon reaching a burn-up of 60 - 75 GWd/tHM a so called High Burnup Structure (HBS) is formed in the cooler rim of the fuel. In this region a depletion of the noble fission gases (FG) in the matrix and an enrichment of FG in μm-sized pores can be observed. Recent calculations show that in these pores the pressure at room temperature can be as large as 30 MPa. The knowledge of the FG pressure in pores is important to understand the high burn-up fuel behavior under accident conditions (i.e. RIA or LOCA). With analytical methods routinely used for the characterization of solid samples, i.e. Electron Probe Micro Analysis (EPMA), Secondary Ion Mass Spectrometry (SIMS), the quantification of gaseous inclusions is very difficult to almost impossible. The combination of a laser ablation system (LA) with an inductively coupled plasma mass spectrometer (ICP-MS) offers a powerful tool for quantification of the gaseous pore inventory. This method offers the advantages of high spatial resolution with laser spot sizes down to 10 μm and low detection limits. By coupling with scanning electron microscopy (SEM) for the pore size distribution, EPMA for the FG inventory in the fuel matrix and optical microscopy for the LA-crater sizes, the pressures in the pores and porosity was calculated. As a first application of this calibration technique for gases, measurements were performed on pressurized water reactor (PWR) fuel with a rod average of 105 GWd/tHM to determine the local FG pressure distribution. (authors)

  1. Experimental control of burn-up calculations for high temperature reactor fuel by introduction of a special alpha spectrometric method for the determination of transuranium content. An attempt to establish isotopic correlations

    International Nuclear Information System (INIS)

    In the field of high-temperature-reactor (HTR) fuel investigation there is a great interest in the experimental and calculational determination of heavy metal content under the aspects of burn-up physics and for the prediction of reliable data for reprocessing and waste management. Using a laser-micro-boring preparation method, high resolution alpha-spectroscopy and sophisticated computer decomposition programs we identify qualitatively and quantitatively most of the important actinide isotopes in irradiated HTR-fuel. Additionally we use data, delivered by gamma- and mass-spectroscopy of the same fuel samples. The evaluated results are compared with calculational results from the burn-up code ORIGEN, using a special generated HTR-neutron-cross-section library. In a first step we determine new cross sections for the uranium and plutonium isotopes depending on the irradiation conditions. In a second step we calculate correlations between the heavy metal isotopes and the burn-up or the fission products

  2. Chemical Analysis of High Burn-up PuO2 Fuel. II Results on Dragon-Fuel. RCN Report

    International Nuclear Information System (INIS)

    The results of a chemical analysis with respect to isotopic composition and total content of the elements Zr, Mo, Ru, Cs, Nd, Pm, Sm, Eu and Pu in a batch of irradiated pyro-carbon / silicon-carbide-coated PuO2 fuel particles are reported and discussed. (author)

  3. FTR tag burnup

    International Nuclear Information System (INIS)

    The gas tag burnup changes investigated were limited to the three tags (Kr-78/Kr-80, Xe-126/Xe-129 and Kr-82/Kr-80) currently accepted as being the most desirable. Control rod tag burnup was significantly greater than fuel rod tag burnup. This occurs because control rods stay in the reactor longer and occupy positions of greater low-energy flux. Thus, minimum tag spacings were set by the control rods as 1.079 for Kr-78/Kr-80, 1.189 for Xe-126/Xe-129 and 1.134 for Kr-82/Kr-80

  4. Burnup performances of boron nitride and boron coated nuclear fuels

    International Nuclear Information System (INIS)

    The nuclear fuels of urania (UOV) and 5% and 10% gadolinia (Gd2O3) containing UO2 previously produced by sol-gel technique were coated with first boron nitride (BN) then boron (B) thin layer by chemical vapor deposition (CVD) and also by plasma enhanced chemical vapor deposition (PECVD) techniques to increase the fuel cycle length and to improve the physical properties. From the cross-sectional view of BN and B layers taken from scanning electron microscope (SEM), the excellent adherence of BN onto fuel and B onto BN layer was observed in both cases. The behavior of fuel burnup, depletion of BN and B, the effect of coating thickness and also Gd2O3 content on the burnup performances of the fuels were identified by using the code WIMS-D/4 for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) cores. The optimum thickness ratio of B to BN was found as 4 and their thicknesses were chosen as 40 mm and 10 mm respectively in both reactor types to get extended cycle length. The assemblies consisting of fuels with 5% Gd2O3 and also coated with 10 mm BN and 40 mm B layers were determined as candidates for getting higher burnup in both types of reactors

  5. Burnup credit in Spain

    International Nuclear Information System (INIS)

    The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analytical activities in the country are described, as well as international collaborations in this field. Perspectives for future research and development of new applications are finally resumed. (author)

  6. Application of burnup credit with partial boron credit to PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    The outcome of performing a burnup credit criticality safety analysis of a PWR spent fuel storage pool is the determination of burnup credit loading curves BLC=BLC(e) for the spent fuel storage racks designed for burnup credit, cp. Reference. A burnup credit loading curve BLC=BLC(e) specifies the loading criterion by indicating the minimum burnup BLC(e) necessary for the fuel assembly with a specific initial enrichment e to be placed in storage racks designed for burnup credit. (orig.)

  7. Experimental verification of the depletion code (ORIGEN-S) by chemical assay method

    International Nuclear Information System (INIS)

    Burnup of the fuel rod taken from spent fuel assemblies C15, G23 and J14 discharged from Kori-1 was examined by destructive chemical assay and determined by Nd-148 mass spectrometry method. And then they were compared with the calculated burnup of ORIGEN-S for the code verification by using activity ratio Eu-154/ Cs-137. As a result of the comparison, burnup of Nd-148 chemical method has a good agreement with the calculated burnup within 1% error in C-15 and G23 rod, and within 8% in the J-14 rod

  8. Application of the radiochemical - and the direct gamma ray spectrometry method to the burnup determination of irradiated uranium oxide

    International Nuclear Information System (INIS)

    The burn up of natural U3O8 that occurs by the action of thermal neutrons was determined, using the radioisotopes 144Ce, 137Cs, 103Ru, 106Ru and 95Zr as monitors. The determination of the burn up was made using both destructive and non-destructive methods. In the non-destructive method, the technique of direct gamma-ray spectrometry was used and the radioisotopes mentioned were simultaneously counted in a Ge-Li detector. In the radiochemical method the same radioisotopes were isolated one from the other and from all other fission products before counting. The solvent extraction technique was used for the radiochemical separation of uranium, cerium, cesium and ruthenium. To separate zirconium and niobium, adsorption in silica-gel was used. The extraction agent employed to isolate cesium was dipycrilamine and for the separation of the other radioisotopes Di-(2-Ethyl Hexyl) Phosphoric acid (HDEHP) was used. (Author)

  9. The method of correction of irradiation history in burn-up determination using fission product cesium-137, cerium-144, and neodymium-148 as monitors

    International Nuclear Information System (INIS)

    In this paper, for cesium-137, cerium-144 and neodymium-148 nuclids the average yield, the quantity of correction for (n, γ) reaction, the quantity of correction for radioactive decay in reactor and the average fission energy of fissionable nuclide were calculated. The result improved precision of parameter and gave quite well value of burn-up

  10. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    This paper presents the development of an automated generation of a new burnup chain for reactor analysis applications. The JENDL FP Decay Data File 2011 and Fission Yields Data File 2011 were used as the data sources. The nuclides in the new chain are determined by restrictions of the half-life and cumulative yield of fission products or from a given list. Then, decay modes, branching ratios and fission yields are recalculated taking into account intermediate reactions. The new burnup chain is output according to the format for the SRAC code system. Verification was performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Further development and applications are being planned with the burnup chain code. (author)

  11. Thermonuclear burn-up in deuterated methane $CD_4$

    CERN Document Server

    Frolov, Alexei M

    2010-01-01

    The thermonuclear burn-up of highly compressed deuterated methane CD$_4$ is considered in the spherical geometry. The minimal required values of the burn-up parameter $x = \\rho_0 \\cdot r_f$ are determined for various temperatures $T$ and densities $\\rho_0$. It is shown that thermonuclear burn-up in $CD_4$ becomes possible in practice if its initial density $\\rho_0$ exceeds $\\approx 5 \\cdot 10^3$ $g \\cdot cm^{-3}$. Burn-up in CD$_2$T$_2$ methane requires significantly ($\\approx$ 100 times) lower compressions. The developed approach can be used in order to compute the critical burn-up parameters in an arbitrary deuterium containing fuel.

  12. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

  13. Burn-up measurements coupling gamma spectrometry and neutron measurement

    International Nuclear Information System (INIS)

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  14. 2005 status and future of burnup credit in the USA

    International Nuclear Information System (INIS)

    At the beginning of 2005 in the USA burnup credit is licensed for PWR and BWR spent fuel pools, is under license review for a transport cask, is under discussion for disposal criticality. Two basic approaches exist for burnup credit. The first approach, which is licensed for spent fuel pools, utilizes criticality experience with spent fuel that has not been chemically assayed. The second approach to burnup credit comes from utilizing chemical assay data to validate the depletion calculations and then clean critical experiments to validate the criticality calculation. A burnup credit standard (ANS/ANSI-8.27) is under development where the two approaches are actively discussed. Issues related to the two approaches are presented as well as possible ways of resolving the issues. (author)

  15. Radiochemical analysis of nuclear fuel burn-up and spent fuel key nuclides

    International Nuclear Information System (INIS)

    Destructive radiochemical analysis of spent nuclear fuels is an important tool to determine burn-up with high accuracy and to better understand the process of depletion and formation of actinides and fission products during irradiation as a result of fission and successive neutron capture. The resulting isotope inventories and nuclear databases that are created, are of high importance to evaluate the performance of nuclear fuels in a reactor, to evaluate computer codes applied for a safe transport, storage and disposal/reprocessing of spent fuels and to safeguard fissile material. The objective is to provide chemical and radiochemical analyses procedures for an accurate determination of isotopic compositions and concentrations of actinides and fission products in different types of industrial (UO2, MOX) and experimental nuclear fuels (UAlx, U3Si2, UMo, ...). For a burn-up determination program typically 21 actinides and fission products are analyzed. For an extended characterization program this can increase to up to approximately 50 isotopes

  16. Analysis of burnup credit on spent fuel storage

    International Nuclear Information System (INIS)

    Chemical analyses were carried out on high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins. Measured data of the composition of nuclides from 234U to 242Pu were used for evaluation of ORIGEN-2/82 code. Criticality calculations were executed for the casks which were being designed to store 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for (1) axial and horizontal profiles of burnup, and void history (BWR), (2) operational histories such as control rod insertion history, BPR insertion history and others, and (3) calculational accuracy of ORIGEN-2/82 code on the composition of nuclides. Present evaluation shows that introduction of burnup credit has a substantial merit in criticality safety analysis of the cask, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for present reactivity bias evaluation and showed a possibility of simplifying the reactivity bias evaluation in burnup credit. Finally, adapting procedures of burnup credit such as the burnup meter were evaluated. (author)

  17. Effects of axial burnup distributions on the reactivity of spent fuel

    International Nuclear Information System (INIS)

    Criticality safety analyses for spent fuel shipping casks will eventually need to take credit for the decreased reactivity of spent fuel assemblies resulting from burnup. In order to do so, it will be necessary to assess the reactivity effects of the multitude of burnup shapes that can characterize spent fuel. A computer program, CASAX, has been written that allows the analyst to quickly evaluate the reactivity effects of actual and simplified axial burnup distributions on a group of PWR fuel assemblies. CASAX employs one dimensional, two group diffusion calculations to determine the k-effective of a cluster of assemblies. Assembly average, burnup dependent, two group cross sections for CASAX were obtained from CASMO3 using physical properties representative of Westinghouse 17 x 17 assemblies. Reactivity results are presented in terms of (k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup)/(k for a uniform axial burnup distribution at the assembly average burnup). Axial burnup distributions can have both positive and negative effects on the calculated k-effective. Positive reactivity effects generally result at high assembly average burnups and for axial distributions with low burnups in the assembly's tips

  18. Protein Structure Determination Using Chemical Shifts

    DEFF Research Database (Denmark)

    Christensen, Anders Steen

    In this thesis, a protein structure determination using chemical shifts is presented. The method is implemented in the open source PHAISTOS protein simulation framework. The method combines sampling from a generative model with a coarse-grained force field and an energy function that includes che...... residues. For Rhodopsin (225 residues) a structure is found at 2.5 Å CA-RMSD from the experimental X-ray structure, and a structure is determined for the Savinase protein (269 residues) with 2.9 Å CA-RMSD from the experimental X-ray structure....

  19. Effect of Self-Shielding on Burn-Up Calculation of ETRR-2 Reactor

    International Nuclear Information System (INIS)

    There exist two approaches for burn-up calculation. The first on is to use cell parameters generated using cell calculation code at different degrees of burn-up. The other is to use microscopic cross sections with self-shielding in order to compensate for the variation of spectrum at different degree of burn-up. The effect of using different forms of self-shielding factors on burn-up calculation for ETRR-2 reactor has been determined. The results of the two approaches are inter-compared up to 50% burn-up

  20. Application of a burnup verification meter to actinide-only burnup credit for spent PWR fuel

    International Nuclear Information System (INIS)

    A measurement system to verify reactor records for burnup of spent fuel at pressurized water reactors (PWR) has been developed by Sandia National Laboratories and tested at US nuclear utility sites. The system makes use of the Fork detector designed at Los Alamos National Laboratory for the safeguards program of the International Atomic Energy Agency. A single-point measurement of the neutrons and gamma- rays emitted from a PWR assembly is made at the center plane of the assembly while it is partially raised from its rack in the spent fuel pool. The objective of the measurements is to determine the variation in burnup assignments among a group of assemblies, and to identify anomalous assemblies that might adversely affect nuclear criticality safety. The measurements also provide an internal consistency check for reactor records of cooling time and initial enrichment. The burnup verification system has been proposed for qualifying spent fuel assemblies for loading into containers designed using burnup credit techniques. The system is incorporated in the US Department of Energy's.''Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages'' [DOE/RW 19951

  1. Fuel burnup characteristics for the NRU research reactor

    International Nuclear Information System (INIS)

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U3Si, consisting of particles of U3Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  2. Fuel burnup characteristics for the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C., E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U{sub 3}Si, consisting of particles of U{sub 3}Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  3. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    International Nuclear Information System (INIS)

    , initial 235U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed

  4. Long-term safety of radioactive waste disposal: Chemical reaction of fabricated and high burnup spent UO2 fuel with saline brines. Final report

    International Nuclear Information System (INIS)

    This is the final report of a large EU-research project on spent fuel stability in saline repository environments. Static dissolution experiments with high burnup spent fuel samples and unirradiated UO2 were performed for about two years in anaerobic NaCl solutions and deionized water with and without container material (iron) being present. Experiments performed at 25 and 150 C gave similar results. Dissolution rates were similar to those measured in the Swedish, or Canadian program for granite media. Rates are strongly influenced by the specific sample surface area, probably related to the mass balance of consumption and production of radiolytic oxidants. In the competition between the oxidizing effect of radiolysis and the reducing effect of iron, the metal corrosion process dominates. Processes controlling radionuclide release are matrix dissolution, solubility, coprecipitation sorption phenomena and colloid formation. In the absence of iron release rates of Sr90, Tc99, Np237, Sb125 and at low reaction progress Ru106 were controlled by matrix dissolution whereas concentrations of tetra-, hexa-, and trivalent actinides (U, Pu, Am, Cm) were controlled by solubility or coprecipitation. The presence of iron did effectively reduce the rates of fuel dissolution and the concentration of many, though not all radionuclides. Solubilities of U were similar for uniradiated UO2 and for spent fuel both in the case of oxidizing and reducing conditions. In contrast, due to the effect of radiolysis, reaction rates of spent fuel were higher than UO2 dissolution rates. (orig.)

  5. Advanced Burnup Method using Inductively Coupled Plasma Mass Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Hilton, Bruce A. [Idaho Natonal Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Glagolenko, Irina; Giglio, Jeffrey J.; Cummings, Daniel G

    2009-06-15

    Nuclear fuel burnup is a key parameter used to assess irradiated fuel performance, to characterize the dependence of property changes due to irradiation, and to perform nuclear materials accountability. For advanced transmutation fuels and high burnup LWR fuels that have multiple fission sources, the existing Nd-148 ASTM burnup determination practice requires input of calculated fission fractions (identifying the specific fission source isotope and neutron energy that yielded fission, e.g., U-235 from thermal neutron, U-238 from fast neutron) from computational neutronics analysis in addition to the measured concentration of a single fission product isotope. We report a novel methodology of nuclear fuel burnup determination, which is completely independent of model predictions and reactor types. The proposed method leverages the capability of Inductively Coupled Plasma Mass Spectrometry (ICP-MS) to quantify multiple fission products and actinides and uses these data to develop a system of burnup equations whose solution is the fission fractions. The fission fractions are substituted back in the equations to determine burnup. This technique requires high fidelity fission yield data, which is not uniformly available for all fission products. We discuss different means that can potentially assist in indirect determination, verification and improvement (refinement) of the ambiguously known fission yields. A variety of irradiated fuel samples are characterized by ICP-MS and the results used to test the advanced burnup method. The samples include metallic alloy fuel irradiated in fast spectrum reactor (EBRII) and metallic alloy in a tailored spectrum and dispersion fuel in the thermal spectrum of the Advanced Test Reactor (ATR). The derived fission fractions and measured burnups are compared with calculated values predicted by neutronics models. (authors)

  6. Advanced Burnup Method using Inductively Coupled Plasma Mass Spectrometry

    International Nuclear Information System (INIS)

    Nuclear fuel burnup is a key parameter used to assess irradiated fuel performance, to characterize the dependence of property changes due to irradiation, and to perform nuclear materials accountability. For advanced transmutation fuels and high burnup LWR fuels that have multiple fission sources, the existing Nd-148 ASTM burnup determination practice requires input of calculated fission fractions (identifying the specific fission source isotope and neutron energy that yielded fission, e.g., U-235 from thermal neutron, U-238 from fast neutron) from computational neutronics analysis in addition to the measured concentration of a single fission product isotope. We report a novel methodology of nuclear fuel burnup determination, which is completely independent of model predictions and reactor types. The proposed method leverages the capability of Inductively Coupled Plasma Mass Spectrometry (ICP-MS) to quantify multiple fission products and actinides and uses these data to develop a system of burnup equations whose solution is the fission fractions. The fission fractions are substituted back in the equations to determine burnup. This technique requires high fidelity fission yield data, which is not uniformly available for all fission products. We discuss different means that can potentially assist in indirect determination, verification and improvement (refinement) of the ambiguously known fission yields. A variety of irradiated fuel samples are characterized by ICP-MS and the results used to test the advanced burnup method. The samples include metallic alloy fuel irradiated in fast spectrum reactor (EBRII) and metallic alloy in a tailored spectrum and dispersion fuel in the thermal spectrum of the Advanced Test Reactor (ATR). The derived fission fractions and measured burnups are compared with calculated values predicted by neutronics models. (authors)

  7. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.)

  8. Compressive creep of simulated burnup fuel

    International Nuclear Information System (INIS)

    In order to study the nitride fuel mechanical properties, we measured the compressive steady state creep rates of uranium mononitride (UN) and UN containing neodymium which was simulated burnup fuel. The stress exponent n'' and the apparent activation energy ''Q'' of the creep rate were determined in the range of 27.5 ≤ σ ≤ 200.0 MPa and 950 ≤ T ≤ 1500 degC. (author)

  9. Burnup span sensitivity analysis of different burnup coupling schemes

    International Nuclear Information System (INIS)

    Highlights: ► The objective of this work is the burnup span sensitivity analysis of different coupling schemes. ► Three kinds of schemes have been implemented in a new MCNP–ORIGEN linkage program. ► Two kinds of schemes are based predictor–corrector technique and the third is based on Euler explicit method. ► The analysis showed that the predictor–corrector approach better accounts for nonlinear behavior of burnup. ► It is sufficiently good to use the Euler method at small spans but for large spans use of second order scheme is mandatory. - Abstract: The analysis of core composition changes is complicated by the fact that the time and spatial variations in isotopic composition depend on the neutron flux distribution and vice versa. Fortunately, changes in core composition occur relatively slowly and hence the burnup analysis can be performed by dividing the burnup period into some burnup spans and assuming that the averaged flux and cross sections are constant during each burn up span. The burnup span sensitivity analysis attempts to find how much the burnup spans could be increased without any significant change in results. This goal has been achieved by developing a new MCNP–ORIGEN linkage program named MOBC (MCNP–ORIGEN Burnup Calculation). Three kinds of coupling scheme have been implemented in MOBC. Two of these are based on second order predictor–corrector technique and enable us to choose larger time steps, whilst the third one is based on Euler explicit first order method and is faster than the other two. The validity of the developed program has been evaluated by the code vs. code comparison technique. Two different types of codes are employed. The first one is based on deterministic two dimensional transport method, like CASMO-4 and HELIOS codes, and the second one is based on Monte Carlo method, like MCODE code. Only one coupling technique is employed in each of these state of the art codes, while the MOBC excels in its ability to

  10. High burnup experience in PWRs

    International Nuclear Information System (INIS)

    The purpose of this paper is to summarize the high burnup experience of Westinghouse PWR fuel. The emphasis is on two regions of commercial PWR fuel that attained region average burnups greater than 36,000 MWD/MTU. One region operated under load follow conditions. The other region operated at base load conditions with a high average linear heat rating. Coolant activity data and post irradiation data were obtained. The post-irradiation data consisted of visual examinations, crud sampling, rod-to-rod dimensional changes, fuel column length changes, rod and assembly growth, assembly bow, fuel rod profilometry, grid spring relaxation, and fuel assembly sipping tests. The data showed that the fuel operated reliably to this burnup. Plans for irradiation to higher burnups are also discussed

  11. Research on burnup physics

    International Nuclear Information System (INIS)

    One of the major problems in burnup studies is the reasonably fast and accurate calculation of the space-and-energy dependent neutron flux and reaction rates for realistic power reactor fuel geometries and compositions, and its optimal integration in the global reactor calculations. The scope of the present research was to develop improved methods trying to satisfy the above requirements. In the epithermal region, simple and efficient approximation is proposed which allows the analytical solution for the space dependence of the spherical harmonics flux moments, and hence the derivation of the recurrence relations between he flux moments at successive lethargy pivotal points. A new matrix formalism to invert the coefficient matrix of band structure resulted in a reduce computer time and memory demands. The research on epithermal region is finalized in computing programme SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, and the related integral quantities as reaction rates and resonance integrals. For partial verification of the above methods a Monte Carlo procedure was developed. Using point-wise representation of variables, a flexible and fast convergent integral transport method SEPT i developed. Expanding the neutron source and flux in finite series of arbitrary polynomials, the space-and-energy dependent integral transport equation is transformed into a general linear algebraic form, which is solved numerically. A simple and efficient procedure for deriving multipoint equations and constructing matrix is proposed and examined, and no unwanted oscillations were noticed. The energy point method was combined with the spherical harmonics method as well. A multi zone few-group program SPECTAR for global reactor calculations was developed. For testing, the flux distribution, neutron leakage and effective multiplication factor for the PWR reactor of the power station San Onofre were calculated. In order to verify

  12. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work

  13. EPRI R and D perspective on burnup credit

    International Nuclear Information System (INIS)

    'Burnup credit' refers to taking credit for the burnup of nuclear fuel in the performance of criticality safety analyses. Historically, criticality safety analyses for transport of spent nuclear fuel have assumed the fuel to be unirradiated (i.e. 'fresh' fuel). In 1999, the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office issued Interim Staff Guidance - 8 (ISG-8) with recommendations for the use of burnup credit in storage and transportation of pressurized water reactor (PWR) spent fuel. The use of burnup credit offers an opportunity to reduce the number of spent nuclear fuel shipments by ∼30%. A simple analysis shows that the increased risk of a criticality event associated with properly using burnup credit is negligible. Comparing this negligible risk component with the reduction in common transport risks due to the reduced number of spent fuel shipments (higher capacity casks for transporting PWR spent fuel) leads to the conclusion that using 'burnup credit' is preferable to using the 'fresh fuel' assumption. A specific objective of the EPRI program is to support the Goals of the U.S. Industry. These goals are consistent with the original U.S. Department of Energy (DOE) goal defined in 1988: a burnup credit methodology that takes credit for the negative reactivity that is practical (all fissile actinides, most neutron absorbing actinides, and a subset of the fission products that account for the majority of the available credit from all fission products). The determination of the optimum number of fission products to consider in a practical burnup credit methodology validates the approach advocated by researchers from France to first focus on a handful of isotopes that include Sm-149; Rh-103; Nd-143; Gd-155; and Sm-152. (author)

  14. Instrumentation for measuring the burnup of spent nuclear fuel

    International Nuclear Information System (INIS)

    Many different methods or procedures have been developed to measure reactivity of fissil materials. Few of these, however, have been designed specifically for light water reactor fuel or have actually been used to measure the reactivity of spent fuel. The methods that have been used to make measurements of related systems are the 252Cf source-driven noise analysis method, a noise analysis method using natural neutron sources, subcritical assembly measurements, and pulsed neutron techniques. Several different approaches to directly measuring burnup have been developed by various organizations. The experimental work on actual spent nuclear fuel utilizing reactivity measurement techniques is insufficient to provide conclusive evidence of the applicability of these techniques for verifying fuel burnup. The work with burnup meters indicates, however, that good correlations can be obtained with any of the systems. A burnup meter's primary function would be a secondary assurance that the administrative records are not grossly in error. Reactivity measurements provide information relating to the reactivity of the fuel only under the conditions measured. Criticality prevention design requirements will necessitate that casks accommodate a minimum burnup level for a given initial enrichment (i.e., a maximum reactivity). Direct measurement of the burnup will enable an easy determination of whether a particular fuel assembly can be shipped in a specific cask with a minimum number of additional correlations

  15. Burnup credit considerations in dry spent-fuel storage licensing

    International Nuclear Information System (INIS)

    Burnup credit has been allowed in reactor basin spent-fuel storage at pressurized water reactors for a number of years. However, such storage occurs under strict administrative, procedural, and design controls. In recent years, dry spent-fuel storage cask vendors have expressed interest in designing cask fuel baskets with allowance for burnup credit. At last year's American Nuclear Society Winter Meeting, an ad hoc session was organized and authorized on burnup credit for dry storage and transportation casks. It has become clear that some utilities are interested in burnup credit for dry storage designs. Given this, the US Nuclear Regulatory Commission (NRC) staff is examining the technical issues involved in allowing burnup credit. Analytical work focused on the development of branch technical positions for determination of burnup credit for dry spent-fuel storage technology designs has begun. Procedural and administrative issues will be examined, based on licensing experience, and will also be the subject of branch technical positions. At an appropriate time, preparation of regulatory guides will be considered

  16. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  17. PWR AXIAL BURNUP PROFILE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  18. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  19. CANDU lattice uncertainties during burnup

    International Nuclear Information System (INIS)

    Uncertainties associated with fundamental nuclear data accompany evaluated nuclear data libraries in the form of covariance matrices. As nuclear data are important parameters in reactor physics calculations, any associated uncertainty causes a loss of confidence in the calculation results. The quantification of output uncertainties is necessary to adequately establish safety margins of nuclear facilities. In this work, microscopic cross-section has been propagated through lattice burnup calculations applied to a generic CANDU® model. It was found that substantial uncertainty emerges during burnup even when fission yield fraction and decay rate uncertainties are neglected. (author)

  20. A technique of melting temperature measurement and its application for irradiated high-burnup MOX fuels

    International Nuclear Information System (INIS)

    A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. (author)

  1. Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

    International Nuclear Information System (INIS)

    A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO2), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO2/UO2), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin

  2. Chemical interaction between the oxide and the clad in PHENIX fuel at burnup up to 60,000 MWd/t

    International Nuclear Information System (INIS)

    In every fuel element there is a potential problem of chemical interaction between the fissile portion and the clad. As a matter of fact, even if the choice of materials is made after having established a satisfactory chemical compatibility between the fuel- (UO2 (U,Pu)O2, (U,Pu) C, . . .) and the clad (stainless steel, zircaloy, . . . ) out of pile, it is difficult to guarantee this compatibility after operation in the reactor due, on one hand, to the presence of fission products and, on the other hand, to impurities which are always present in the fuel to a greater or lesser degree. The fuel element currently chosen for the sodium-cooled fast reactors ((U,Pu)O2 in stainless steel clad) does not avoid this problem, in particular because of the relatively high temperatures envisioned for this type of reactor - the clad temperature is about 650 deg. C. Since it is considered as a demonstration reactor, Phenix should be able to provide additional information on this phenomenon, and one will see that we have been able to shed light on some points which the experiments or irradiations made to date have been unable to explain. However, before presenting the experimental results obtained with Phenix fuel end drawing conclusions, we shall give a brief resume of the expected behavior of this fuel with respect to the phenomenon of interest. (author)

  3. Fuel burnup monitor for nuclear reactors

    International Nuclear Information System (INIS)

    An in-service detector is designed using the principle of comparing temperatures in the fuel element and in the detector material. The detector consists of 3 metallic heat conductors insulated with ceramic insulators, two of them with uranium fuel spheres at the end. One sphere is coated with zirconium, the other with zirconium and gold. The precision of measurement of the degree of fuel burnup depends on the precision of the measurement of temperature and is determined from the difference in temperature gradients of the two uranium fuel spheres in the detector. (M.D.)

  4. Determination of electroless deposition by chemical nickeling

    Directory of Open Access Journals (Sweden)

    M. Badida

    2013-07-01

    Full Text Available Increasing of technical level and reliability of machine products in compliance with the economical and ecological terms belongs to the main trends of the industrial development. During the utilisation of these products there arise their each other contacts and the interaction with the environment. That is the reason for their surface degradation by wear effect, corrosion and other influences. The chemical nickel-plating allows autocatalytic deposition of nickel from water solutions in the form of coherent, technically very profitable coating without usage of external source of electric current. The research was aimed at evaluating the surface changes after chemical nickel-plating at various changes of technological parameters.

  5. 78 FR 55326 - Determinations Regarding Use of Chemical Weapons in Syria Under the Chemical and Biological...

    Science.gov (United States)

    2013-09-10

    ... Determinations Regarding Use of Chemical Weapons in Syria Under the Chemical and Biological Weapons Control and..., 22 U.S.C. 5604(a), that the Government of Syria has used chemical weapons in violation of... Under Secretary of State for Political Affairs: (1) Determined that the Government of Syria has...

  6. Characterisation of high-burnup LWR fuel rods through gamma tomography

    International Nuclear Information System (INIS)

    Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235U enrichment of the fresh UO2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

  7. Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias

    International Nuclear Information System (INIS)

    Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)

  8. Determination of electroless deposition by chemical nickeling

    OpenAIRE

    Badida, M.; M. Gombár; L. Sobotová; J. Kmec

    2013-01-01

    Increasing of technical level and reliability of machine products in compliance with the economical and ecological terms belongs to the main trends of the industrial development. During the utilisation of these products there arise their each other contacts and the interaction with the environment. That is the reason for their surface degradation by wear effect, corrosion and other influences. The chemical nickel-plating allows autocatalytic deposition of nickel from water solutions in the fo...

  9. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

  10. A burnup credit calculation methodology for PWR spent fuel transportation

    International Nuclear Information System (INIS)

    A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)

  11. Burnup measurements with the Los Alamos fork detector

    International Nuclear Information System (INIS)

    The fork detector system can determine the burnup of spent-fuel assemblies. It is a transportable instrument that can be mounted permanently in a spent-fuel pond near a loading area for shipping casks, or be attached to the storage pond bridge for measurements on partially raised spent-fuel assemblies. The accuracy of the predicted burnup has been demonstrated to be as good as 2% from measurements on assemblies in the United States and other countries. Instruments have also been developed at other facilities throughout the world using the same or different techniques, but with similar accuracies. 14 refs., 2 figs., 2 tabs

  12. Power excursion analysis for BWR`s at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Neymoith, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  13. Burn-up measurements at TRIGA fuel elements containing strong burnable poison

    International Nuclear Information System (INIS)

    The reactivity method of determining the burn-up of research reactor fuel elements is applied to the highly enriched FLIP elements of TRIGA reactors. In contrast to other TRIGA fuel element types, the reactivity of FLIP elements increases with burn-up due to consumption of burnable poison. 33 fuel elements with burn-up values between 3% and 14% were investigated. The experiments showed that variations in the initial fuel composition significantly influence the reactivity and, consequently, increase the inaccuracy of the burn-up measurements. Particularly important are variations in the initial concentration of erbium, which is used as burnable poison in FLIP fuel. A method for reducing the effects of the material composition variations on the measured reactivity is presented. If it is applied, the accuracy of the reactivity method for highly poisoned fuel elements becomes comparable to the accuracy of other methods for burn-up determination. (orig.)

  14. Minor Actinide Transmutation Performance in Fast Reactor Metal Fuel. Isotope Ratio Change in Actinide Elements upon Low-Burnup Irradiation

    International Nuclear Information System (INIS)

    Metal fuel alloys containing 5 wt% or less minor actinide (MA) and rare earth (RE) were irradiated in the fast reactor Phénix. After nondestructive postirradiation tests, a chemical analysis of the alloys irradiated for 120 effective full power days was carried out by the inductively coupled plasma - mass spectrometry (ICP-MS) technique. From the analysis results, it was determined that the discharged burnups of U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, and U-19Pu-10Zr-5MA were 2.17, 2.48, and 2.36 at.%, respectively. Actinide isotope ratio analyses before and after the irradiation experiment revealed that Pu, Am, and Cm nuclides added to U-Pu-Zr alloy and irradiated up to 2.0 - 2.5 at.% burnups in a fast reactor are transmuted properly as predicted by ORIGEN2 calculations. (author)

  15. TRIGA fuel burn-up calculations and its confirmation

    International Nuclear Information System (INIS)

    The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.

  16. Measurement techniques for verifying burnup

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.I. (Sandia National Lab., Albuquerque, NM (US)); Bierman, S.R. (Pacific Northwest Lab., Richland, WA (US))

    1992-05-01

    Measurements of the nuclear radiation from spent reactor fuel are being considered to qualify assemblies for loading into casks that will be used to transport spent fuel from utility sites to a federal storage facility. To ensure nuclear criticality safety, the casks are being designed to accept assemblies that meet restrictions as to burnup, initial enrichment and cooling time. This paper reports that measurements could be used to ensure that only fuel assemblies that meet the restrictions are selected for loading.

  17. Measurement techniques for verifying burnup

    International Nuclear Information System (INIS)

    Measurements of the nuclear radiation from spent reactor fuel are being considered to qualify assemblies for loading into casks that will be used to transport spent fuel from utility sites to a federal storage facility. To ensure nuclear criticality safety, the casks are being designed to accept assemblies that meet restrictions as to burnup, initial enrichment and cooling time. This paper reports that measurements could be used to ensure that only fuel assemblies that meet the restrictions are selected for loading

  18. Establishing a PWR burn-up library

    International Nuclear Information System (INIS)

    Starting out from data file ENDF/B IV /1/, a cross-section library has been established for the calculation of operating conditions in pressurized water reactors of the type used in BIBLIS B. The library includes macroscopic, homogenized 2-group cross-sections for all types of fuel elements used in this reactor, including those equipped with boron glass rods. For their calculation the previous irradiation of the fuel has been taken into consideration by approximation. Information on fuel consumption from cell burn-up calculations has been stored in a separate data file. It was designed as a base for the determination of cross sections to be used in the calculation of the incident ''main-steam pipe fracture''. For this library the description of cross sections as a function of the moderator status chose the water densities at 3000C/155 bar, 1900C/140 bar and 1000C/100 bar as fixed values. The burn-up library has been tested by a three-dimensional calculation for the 1sup(st) cycle of the BIBLIS B-reactor using program QUABOX /2/. This showed variances with the anticipated course concerning critically, which can be explained almost quantitatively by known deficiencies of the ENDF/b-IV library. (orig.)

  19. Nondestructive analysis of RA reactor fuel burnup, Program for burnup calculation base on relative yield of 106Ru, 134Cs and 137Cs in the irradiated fuel

    International Nuclear Information System (INIS)

    Burnup of low enriched metal uranium fuel of the RA reactor is described by two chain reactions. Energy balance and material changes in the fuel are described by systems of differential equations. Numerical integration of these equations is base on the the reactor operation data. Neutron flux and percent of Uranium-235 or more frequently yield of epithermal neutrons in the neutron flux, is determined by iteration from the measured contents of 106Ru, 134Cs and 137Cs in the irradiated fuel. The computer program was written in FORTRAN-IV. Burnup is calculated by using the measured activities of fission products. Burnup results are absolute values

  20. Future disposal burnup credit process and effort

    International Nuclear Information System (INIS)

    The United States Department of Energy's Office of Civilian Radioactive Waste Management has developed a risk-informed, performance based methodology for disposal criticality analyses. The methodology is documented in the Disposal Criticality Analysis Methodology Topical Report, YMP/TR-004Q (YMP 2000). The methodology includes taking credit for the burnup of irradiated commercial light water reactor fuel in criticality analyses, i.e., burnup credit. This paper summarizes the ongoing and planned future burnup credit activities associated with the methodology. (author)

  1. Burnup credit activities in the United States

    International Nuclear Information System (INIS)

    This report covers progress in burnup credit activities that have occurred in the United States of America (USA) since the International Atomic Energy Agency's (IAEA's) Advisory Group Meeting (AGM) on Burnup Credit was convened in October 1997. The Proceeding of the AGM were issued in April 1998 (IAEA-TECDOC-1013, April 1998). The three applications of the use of burnup credit that are discussed in this report are spent fuel storage, spent fuel transportation, and spent fuel disposal. (author)

  2. Phenomena and Parameters Important to Burnup Credit

    International Nuclear Information System (INIS)

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given

  3. Phenomena and parameters important to burnup credit

    International Nuclear Information System (INIS)

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water- reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the United States and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given. (author)

  4. Ultrasonic measurement of high burn-up fuel elastic properties

    International Nuclear Information System (INIS)

    The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment

  5. Issues for effective implementation of burnup credit

    International Nuclear Information System (INIS)

    In the United States, burnup credit has been used in the criticality safety evaluation for storage pools at pressurized water reactors (PWRs) and considerable work has been performed to lay the foundation for use of burnup credit in dry storage and transport cask applications and permanent disposal applications. Many of the technical issues related to the basic physics phenomena and parameters of importance are similar in each of these applications. However, the nuclear fuel cycle in the United States has never been fully integrated and the implementation of burnup credit to each of these applications is dependent somewhat on the specific safety bases developed over the history of each operational area. This paper will briefly review the implementation status of burnup credit for each application area and explore some of the remaining issues associated with effective implementation of burnup credit. (author)

  6. High burnup fuel development program in Japan

    International Nuclear Information System (INIS)

    A step wise burnup extension program has been progressing in Japan to reduce the LWR fuel cycle cost. At present, the maximum assembly burnup limit of BWR 8 Χ 8 type fuel (B. Step II fuel) is 50GWd/t and a limited numbers of 9 Χ 9 type fuel (B. Step III fuel) with 55GWd/t maximum assembly burnup has been licensed by regulatory agencies recently. Though present maximum assembly burnup limit for PWR fuel is 48GWd/t (P. Step I fuel), the licensing work has been progressing for irradiation testing on a limited number of fuel assemblies with extended burnup of up to 55GWd/t (p. Step II fuel) Design of high burnup fuel and fabrication test are carried out by vendors, and subsequent irradiation test of fuel rods is conducted jointly by utilities and vendors to prepare for licensing. It is usual to make an irradiation test for vectarion, using lead use assemblies by government to confirm fuel integrity and reliability and win the public confidence. Nuclear Power Engineering Corporation (NUPE C) is responsible for verification test. The fuel are subjected to post irradiation examination (PIE) and no unfavorable indications of fuel behavior have found both in NUPE C verification test and joint irradiation test by utilities and vendors. Burnup extension is an urgent task for LWR fuel in Japan in order to establish the domestic fuel cycle. It is conducted in joint efforts of industries, government and institutes. However, watching a situation of burnup extension in the world, we are not going ahead of other countries in the achievement of burnup extension. It is due to a conservative policy in the nuclear safety of the country. This is the reason why the burnup extension program in Japan is progressing 'slow and steady' As for the data obtained, no unfavorable indications of fuel behavior have found both in NUPE C verification test and joint irradiation test by utilities and vendors until now

  7. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  8. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development

  9. High burnup in DIONISIO code

    International Nuclear Information System (INIS)

    When the residence time of nuclear fuel rods exceeds a given threshold value, several properties of the pellet material suffer changes and hence the posterior behaviour of the rod is significantly altered. Structural modifications start at the pellet periphery, which is usually referred to as rim zone. It is presently believed that these changes are a consequence of the localized absorption of epithermal neutrons by 238U, which effective cross section presents resonant peaks. Due to the chain of nuclear reactions that take place, several Pu isotopes are born especially at the rim. In particular, the fissile character of 239Pu and 241Pu is the cause of the increased number of fission events that occur in the pellet periphery. For this reason, the power generation rate and the burnup adopt a non uniform distribution in the pellet, reaching at the rim values two or three times higher than the average [1]. The rim zone starts to form for a burnup threshold value of about 50-60 MWd/kgHM and its width increases as the irradiation progresses. The microstructure of this zone is characterized by the presence of small grains, with a typical size of 200 nm, and large pores, of some μm. Even though the rim zone is very thin, it has a significant effect on the mechanical integrity of the pellet, particularly when it makes contact with the cladding, and on the temperature distribution in the whole pellet, because of its low thermal conductivity [1,2]. The numerical codes designed to simulate fuel behaviour under irradiation must include the phenomena associated to high burnup if they aim at extending the prediction range, and this is the purpose with our DIONISIO code. But a detailed analysis of the phenomena that take place in this region demands the use of neutronic codes that solve the Boltzmann transport equations [3] in a number of energy intervals (groups), including adequate considerations in the region of the resonant absorption peaks of 238U. These cell codes predict

  10. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  11. Physical mechanism analysis of burnup actinide composition in light water reactor MOX fuel and its application to uncertainty evaluation

    International Nuclear Information System (INIS)

    Highlights: • We discuss physical mechanisms for burnup actinide compositions in LWR’s MOX fuel. • Mechanisms of 244Cm and 238Pu productions are analyzed in detail with sensitivity. • We can evaluate the indirect effect on actinide productions by nuclear reactions. • Burnup sensitivity is applied to uncertainty evaluation of nuclide production. • Actinides can be categorized into patterns according to a burnup sensitivity trend. - Abstract: In designing radioactive waste management and decommissioning facilities, understanding the physical mechanisms for burnup actinide composition is indispensable to satisfy requirements for its validity and reliability. Therefore, the uncertainty associated with physical quantities, such as nuclear data, needs to be quantitatively analyzed. The present paper illustrates an analysis methodology to investigate the physical mechanisms of burnup actinide composition with nuclear-data sensitivity based on the generalized depletion perturbation theory. The target in this paper is the MOX fuel of the light water reactor. We start with the discussion of the basic physical mechanisms for burnup actinide compositions using the reaction-rate flow chart on the burnup chain. After that, the physical mechanisms of the productions of Cm-244 and Pu-238 are analyzed in detail with burnup sensitivity calculation. Conclusively, we can identify the source of actinide productions and evaluate the indirect influence of the nuclear reactions if the physical mechanisms of burnup actinide composition are analyzed using the reaction-rate flow chart on the burnup chain and burnup sensitivity calculation. Finally, we demonstrate the usefulness of the burnup sensitivity coefficients in an application to determine the priority of accuracy improvement in nuclear data in combination with the covariance of the nuclear data. In addition, the target actinides and reactions are categorized into patterns according to a sensitivity trend

  12. Development and Applications of a Prototypic SCALE Control Module for Automated Burnup Credit Analysis

    International Nuclear Information System (INIS)

    Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale burnup credit analyses that can be performed in a reasonable amount of time. STARBUCS is a new prototypic analysis sequence being developed for the SCALE code system to perform automated criticality calculations of spent fuel systems employing burnup credit. STARBUCS is designed to help analyze the dominant burnup credit phenomena including spatial burnup gradients and isotopic uncertainties. A search capability also allows STARBUCS to iterate to determine the spent fuel parameters (e.g., enrichment and burnup combinations) that result in a desired keff for a storage configuration. Although STARBUCS was developed to address the analysis needs for spent fuel transport and storage systems, it provides sufficient flexibility to allow virtually any configuration of spent fuel to be analyzed, such as storage pools and reprocessing operations. STARBUCS has been used extensively at Oak Ridge National Laboratory (ORNL) to study burnup credit phenomena in support of the NRC Research program

  13. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  14. Taking burnup credit for interim storage and transportation system for BWR fuels

    International Nuclear Information System (INIS)

    In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)

  15. Determination of OH groups by wet chemical methods

    Czech Academy of Sciences Publication Activity Database

    Kuráň, P.; Janoš, P.; Madronová, L.; Novák, František

    New York : Nova Science Publisher, 2011 - (Madronová, L.), s. 47-60 ISBN 978-1-61668-965-0. - (Chemistry Research and Applications) Institutional research plan: CEZ:AV0Z60660521 Keywords : determination * OH groups * wet chemical methods Subject RIV: CB - Analytical Chemistry, Separation

  16. Burnup credit issues in transportation and storage

    International Nuclear Information System (INIS)

    Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the US experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed

  17. Burnup credit issues in transportation and storage

    International Nuclear Information System (INIS)

    Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the U.S. experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed. (author)

  18. Chemical methods for the determination of composition of cryolite

    International Nuclear Information System (INIS)

    Preparation of uranium and plutonium alloys containing aluminium involves the use of cryolite and many times, cryolite which may be contaminated with alpha activity has to be analysed for its purity. In view of this, chemical methods for the determination of composition of commercial cryolite samples have been developed. Methods are standardised for the determination of individual constituents of cryolite viz., aluminium, sodium, fluoride and major impurities, calcium and magnesium. Studies on the dissolution of the sample, effect of one or more components on the determination of the other and their elimination are carried out. Aluminium and sodium are determined gravimetrically as oxinate and triple acetate respectively. Fluoride is determined by a volumetric procedure after cation exchange separtion of soluble fluoride. Calcium and magnesium are determined by a sequential pH-metri titration. This report describes the details of the procedures and the results of these studies for two commercial cryolite samples. (author). 7 tabs

  19. Addressing the Axial Burnup Distribution in PWR Burnup Credit Criticality Safety

    International Nuclear Information System (INIS)

    This paper summarizes efforts related to developing a technically justifiable approach for addressing the axial burnup distribution in PWR burnup-credit criticality safety analyses. The paper reviews available data on the axial variation in burnup and the effect of axial burnup profiles on reactivity in a SNF cask. A publicly available database of profiles is examined to identify profiles that maximize the neutron multiplication factor, keff, assess its adequacy for general PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. For this assessment, a statistical evaluation of the keff values associated with the profiles in the axial burnup profile database was performed that identifies the most reactive profiles as statistical outliers that are not representative of typical discharged SNF assemblies. The impact of these bounding profiles on the neutron multiplication factor for a high-density burnup credit cask is quantified. Finally, analyses are presented to quantify the potential reactivity consequence of assemblies with axial profiles that are not bounded by the existing database. The paper concludes with findings for addressing the axial burnup distribution in burnup credit analyses

  20. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  1. Technical Development on Burn-up Credit for Spent LWR Fuel

    International Nuclear Information System (INIS)

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report

  2. Extended burnup: fuel development and performance

    International Nuclear Information System (INIS)

    Fuel Performance for the B and W 15 x 15 (Mark B) and 17 x 17 (Mark C) fuel assembly designs is examined on a plant by plant basis. An extensive data base of fuel assembly and rod bow measurements and tests which demonstrate that these phenomena should not limit the high burnup capability of B and W fuel is presented. Post-irradiation measurements to date for fuel rod and assembly growth show that these phenomena are behaving as predicted and can be adequately evaluated and designed for in high burnup fuel assemblies. Clad creep and ductility data as a function of burnup for B and W fuel is presented with emphasis on their effects on our high burnup targets. Finally, fission gas release and waterside corrosion measurements results are presented

  3. Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages

    International Nuclear Information System (INIS)

    review process for these SNF storage and transportation cask applications. The DOE will also reference NRC-accepted topical reports in its license application for a geologic repository. DOE is requesting NRC acceptance for two general aspects of the actinide-only burnup credit methodology. First, data is sufficient to validate the burnup credit criticality analysis methodology presented in this topical report. This includes the chemical assay data used to validate the spent fuel isotopic concentration calculations and critical experiments used to validate the burnup credit criticality calculations. Second, the conservative methodology in utilizing this data for burnup credit is acceptable. A detailed breakdown of what the DOE is specifically seeking NRC acceptance of is presented in Section 1.6

  4. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    review process for these SNF storage and transportation cask applications. The DOE will also reference NRC-accepted topical reports in its license application for a geologic repository. DOE is requesting NRC acceptance for two general aspects of the actinide-only burnup credit methodology. First, data is sufficient to validate the burnup credit criticality analysis methodology presented in this topical report. This includes the chemical assay data used to validate the spent fuel isotopic concentration calculations and critical experiments used to validate the burnup credit criticality calculations. Second, the conservative methodology in utilizing this data for burnup credit is acceptable. A detailed breakdown of what the DOE is specifically seeking NRC acceptance of is presented in Section 1.6.

  5. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

  6. VVER-related burnup credit calculations

    International Nuclear Information System (INIS)

    The calculations related to a VVER burnup credit calculational benchmark proposed to the Eastern and Central European research community in collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmark Working Group (working under WPNCS - Working Party on Nuclear Criticality Safety) are described. The results of a three-year effort by analysts from the Czech Republic, Finland, Germany, Hungary, Russia, Slovakia and the United Kingdom are summarized and commented on. (author)

  7. REBUS: A burnup credit experimental programme

    International Nuclear Information System (INIS)

    An international programme called REBUS (REactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK-CEN and Belgonucleaire. At present it is sponsored by USNRC, EdF from France and VGB, representing German nuclear utilities. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark would qualify the codes to perform calculations of the burn-up credit. The benchmark exercise will investigate the following fuel types with associated burn-up. 1. Reference absorber test bundle, 2. Fresh commercial PWR UO2 fuel, 3. Irradiated commercial PWR UO2 fuel (50 GWd/tM), 4. Fresh PWR UO2 fuel, 5. Irradiated PWR UO2 fuel (30 GWd/tM). Reactivity effects will be measured in the critical facility VENUS. The accumulated burn-up of all rods will be measured non-destructively by gamma-spectrometry. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-18 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). The experimental implementation of the programme will start in 2000. (author)

  8. Burnup credit implementation in spent fuel management

    International Nuclear Information System (INIS)

    The criticality safety analysis of spent fuel management systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. The concept of allowing reactivity credit for spent fuel offers economic incentives. Burnup Credit (BUC) could reduce mass limitation during dissolution of highly enriched PWR assemblies at the La Hague reprocessing plant. Furthermore, accounting for burnup credit enables the operator to avoid the use of Gd soluble poison in the dissolver for MOX assemblies. Analyses performed by DOE and its contractors have indicated that using BUC to maximize spent fuel transportation cask capacities is a justifiable concept that would result in public risk benefits and cost savings while fully maintaining criticality safety margins. In order to allow for Fission Products and Actinides in Criticality-Safety analyses, an extensive BUC experimental programme has been developed in France in the framework of the CEA-COGEMA collaboration. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Independent measurement systems, e.g. gamma spectrum detection systems, are needed to perform a true independent measurement of assembly burnup, without reliance on reactor records, using the gamma emission signatures fission products (mainly Cesium isotopes). (author)

  9. Activity ratio measurement and burnup analysis for high burnup PWR fuels

    International Nuclear Information System (INIS)

    Applying burnup credit to spent fuel transportation and storage system is beneficial. To take burnup credit to criticality safety design for a spent fuel transportation cask and storage rack, the burnup of target fuel assembly based on core management data must be confirmed by experimental methods. Activity ratio method, in which measured the ratio of the activity of a nuclide to that of another, is one of the ways to confirm burnup history. However, there is no public data of gamma-ray spectrum from high burnup fuels and validation of depletion calculation codes is not sufficient in the evaluation of the burnup or nuclide inventories. In this study, applicability evaluation of activity ratio method was carried out for high burnup fuel samples taken from PWR lead use assembly. In the gamma-ray measurement experiments, energy spectrum was taken in the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA), and 134Cs/137Cs and 154Eu/137Cs activity ratio were obtained. With the MVP-BURN code, the activity ratios were calculated by depletion calculation tracing the operation history. As a result, 134Cs/137Cs and 154Eu/137Cs activity ratios for UO2 fuel samples show good agreements and the activity ratio method has good applicability to high burnup fuels. 154Eu/134Cs activity ratio for Gd2O3+UO2 fuels also shows good agreements between calculation results and experimental results as well as the activity ratio for UO2 fuels. It also becomes clear that it is necessary to pay attention to not only burnup but also axial burnup distribution history when confirming the burnup of UO2+Gd2O3 fuel with 134Cs/137Cs activity ratios. (author)

  10. Determination of some chemical and microbiological characteristics of Kaymak

    OpenAIRE

    Ökten, Sevtap; Ünal, Gülfem; Gönç, Siddik; Sibel Akalin, Ayse

    2006-01-01

    Kaymak is a kind of concentrated cream, which is traditionally manufactured from buffalo or cow’s milk in Turkey. It is generally consumed with honey at breakfast and some traditional Turkish desserts. The aim of this study was to determine some chemical and microbiological properties of kaymak. The samples were obtained from different dairy plants producing kaymak from cow’s milk and local markets located in Zmir. They were examined for total solids and fat contents, acidity, pH ...

  11. Determination of chemical resistance of glass, enamels and glazes

    International Nuclear Information System (INIS)

    The determination is described of chemical resistance of glass, enamels and glazes. 85Kr is incorporated by diffusion or implantation in the material investigated. The material is then exposed to the effects of a corrosion atmosphere or a liquid medium. The decrease in the activity of the radioactive kryptonates in the materials or the activity of the released radioactive krypton is measured. (B.S.)

  12. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

  13. Development of Inverse Estimation Program of Burnup Histories for Nuclear Spent Fuel Based on ORIGEN-S

    International Nuclear Information System (INIS)

    The purpose of this work is to develop a computer program which can accurately estimate burnup histories of spent fuels based on the environmental sample measurements. The burnup histories of spent fuels include initial uranium enrichment, discharge burnup, cooling time after discharge, and nuclear reactor type in which the spent fuel was burnt. The methodologies employed in our program are based on the formulations developed by M. R. Scott1 but we developed a stable bi-section method to correct initial uranium enrichment and used a simplified algorithm without burnup correction. Also, ORIGEN-S2 rather than ORIGEN-23 was used in our program to improve the accuracies by using the new capabilities of burnup dependent cross section libraries of ORIGEN-S. Our program is applied to several benchmark problems including realistic Mihama-3 problems to test the accuracies. We developed a computer program to determine the burnup history such as initial uranium enrichment, burnup, cooling time, and reactor type by using the results of sample measurements as input. Our methodologies are based on the methodologies given in Ref. 1 but we devised a new stable bisection method for the correction of initial uranium enrichment and we used ORIGEN-S rather than ORIGEN-2 to utilize the new capabilities of ORIGEN-S such as burnup dependent cross sections which can be prepared by using SCALE6

  14. Review of high burn-up RIA and LOCA database and criteria

    International Nuclear Information System (INIS)

    This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

  15. The applications of burnup credit and the measurement techniques of burnup verification

    International Nuclear Information System (INIS)

    The factors of influencing criticality safety, implementing criticality control conditions, the calculation methods for predicting criticality, casks design and cask loading graph are described. The problems in the application of burnup credit and the dominant error in burnup credit operation are analysed. In order to avoid the operation error, requirements of measurement techniques and the most suitable measurement method are introduced

  16. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  17. The Fork+ burnup measurement system: Design and first measurement campaign

    International Nuclear Information System (INIS)

    Previous work with the original Fork detector showed that burnup as determined by reactor records could be accurately allocated to spent nuclear fuel assemblies. The original Fork detector, designed by Los Alamos National Laboratory, used an ion chamber to measure gross gamma count and a fission chamber to measure neutrons from an activation source, 244Cm. In its review of the draft Topical Report on Burnup Credit, the US Nuclear Regulatory Commission indicated it felt uncomfortable with a measurement system that depended on reactor records for calibration. The Fork+ system was developed at Sandia National Laboratories under the sponsorship of the Electric Power Research Institute with the aim of providing this independent measurement capability. The initial Fork+ prototype was used in a measurement campaign at the Maine Yankee reactor. The campaign confirmed the applicability of the sensor approach in the Fork+ system and the efficiency of the hand-portable Fork+ prototype in making fuel assembly measurements. It also indicated potential design modifications that will be necessary before the Fork+ can be used effectively on high-burnup spent fuel

  18. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  19. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    International Nuclear Information System (INIS)

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. Fifty-seven UO2, UO2/Gd2O3, and UO2/PuO2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on keff (which can be a function of the trending parameters) such that the biased keff, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading

  20. Determining the chemical composition of cloud condensation nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Williams, A.L.; Rothert, J.E.; McClure, K.E. (Illinois State Water Survey, Champaign, IL (United States)); Alofs, D.J.; Hagen, D.E.; White, D.R.; Hopkins, A.R.; Trueblood, M.B. (Missouri Univ., Rolla, MO (USA). Cloud and Aerosol Science Lab.)

    1992-02-01

    This second progress report describes the status of the project one and one-half years after the start. The goal of the project is to develop the instrumentation to collect cloud condensation nuclei (CCN) in sufficient amounts to determine their chemical composition, and to survey the CCN composition in different climates through a series of field measurements. Our approach to CCN collection is to first form droplets on the nuclei under simulated cloud humidity conditions, which is the only known method of identifying CCN from the background aerosol. Under cloud chamber conditions, the droplets formed become larger than the surrounding aerosol, and can then be removed by inertial impaction. The residue of the evaporated droplets represents the sample to be chemically analyzed. Two size functions of CCN particles are collected by first forming droplets on the large particles are collected by first forming droplets on the large CCN in a haze chamber at 100% relative humidity, and then activating the remaining CCN at 1% supersaturation in a cloud chamber. The experimental apparatus is a serious flow arrangement consisting of an impactor to remove the large aerosol particles, a haze chamber to form droplets on the remaining larger CCN, another impactor to remove the haze droplets containing the larger CCN particles for chemical analysis, a continuous flow diffusion (CFD) cloud chamber to form droplets on the remaining smaller CCN, and a third impactor to remove the droplets for the small CCN sample. Progress is documented here on the development of each of the major components of the flow system. Chemical results are reported on tests to determine suitable wicking material for the different plates. Results of computer modeling of various impactor flows are discussed.

  1. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  2. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

  3. Radionuclide Release from High Burnup Fuel

    International Nuclear Information System (INIS)

    In this paper we investigate the production, evolution and release of radioactive fission products in a light water reactor. The production of the nuclides is determined by the neutronics, their evolution in the fuel by local temperature and by the fuel microstructure and the rate of release is governed by the scenario and the properties of the microstructure where the nuclides reside. The problem combines fields of reactor physics, fuel behaviour analysis and accident analysis. Radionuclide evolution during fuel reactor life is also important for determination of instant release fraction of final repository analysis. The source term problem is investigated by literature study and simulations with reactor physics code Serpent as well as fuel performance code ENIGMA. The capabilities of severe accident management codes MELCOR and ASTEC for describing high burnup structure effects are reviewed. As the problem is multidisciplinary in nature the transfer of information between the codes is studied. While the combining of the different fields as they currently are is challenging, there are some possibilities to synergy. Using reactor physics tools capable of spatial discretization is necessary for determining the HBS inventory. Fuel performance studies can provide insight how the HBS should be modelled in severe accident codes, however the end effect is probably very small considering the energetic nature of the postulated accidents in these scenarios. Nuclide release in severe accidents is affected by fuel oxidation, which is not taken into account by ANSI/ANS-5.4 but could be important in some cases, and as such, following the example of severe accident models would benefit the development of fuel performance code models. (author)

  4. COGEMA/TRANSNUCLEAIRE's experience with burnup credit

    International Nuclear Information System (INIS)

    Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)

  5. Fission gas release and pellet microstructure change of high burnup BWR fuel

    International Nuclear Information System (INIS)

    UO2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

  6. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    International Nuclear Information System (INIS)

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent 235U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU)

  7. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    Energy Technology Data Exchange (ETDEWEB)

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

  8. Burnup analysis of the power reactor, 3

    International Nuclear Information System (INIS)

    The atomic number densities of uranium and transuranium were measured for JPDR-1. For the purpose of the study, the program has been prepared. It solves the burnup equation by the exponential matrix method. The void fraction and exposure distribution of the required data were calculated by three-dimensional nuclear-thermal-hydro-dynamic program FLORA under the operating conditions. The distribution of each atomic number density was obtained. The results agree with the measured values. The programs calculating nuclear constants in the cell were evaluated by obtaining the effective cross sections from the atomic number densities and the burnup. (auth.)

  9. Fission gas release modelling at high burnup

    International Nuclear Information System (INIS)

    A large quantity of experimental data on fission gas release is now available in the public domain. It covers a wide variety of fuel types and burnups of up to more than 70 GWd/tU. This data, together with gas release measurements from British Energy's AGRs, has been used to build a comprehensive validation database for the fuel performance code ENIGMA. Validation of ENIGMA version 5.11 against this database has identified a requirement for model development to improve predictions at high burnup. A modified gas release model has been produced and tested. (author)

  10. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

  11. Siemens PWR burnup credit criticality analysis methodology: Depletion code and verification methods

    International Nuclear Information System (INIS)

    Application of burnup credit requires knowledge of the reactivity state of the irradiated fuel for which burnup credit is taken. The isotopic inventory of the irradiated fuel has to be calculated, therefore, by means of depletion codes. Siemens performs depletion calculations for PWR fuel burnup credit applications with the aid of the code package SAV. This code package is based on the first principles approach, i.e., avoids cycle or reactor specific fitting or adjustment parameters. This approach requires a general and comprehensive qualification of SAV by comparing experimental with calculational results. In the paper on hand the attention is focused mainly on the evaluation of chemical assay data received from different experimental programmes. (author)

  12. Detailed Burnup Calculations for Testing Nuclear Data

    Science.gov (United States)

    Leszczynski, F.

    2005-05-01

    A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross

  13. Studies on future application of burnup credit in Hungary

    International Nuclear Information System (INIS)

    This paper describes the present status of the fuel storage and the possible future applications of burnup credit in wet and dry storage systems in Hungary. It gives a survey of the activities planned in AERI concerning the burnup credit. Some part of these investigations dealing with the influence of the axial changing of the assembly burnup are given in more details. (author)

  14. Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One

    International Nuclear Information System (INIS)

    Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design

  15. Modeling of WWER-440 fuel pin behavior at extended burn-up

    Energy Technology Data Exchange (ETDEWEB)

    El-Koliel, Moustafa S. E-mail: moustafa_elkoliel@yahoo.com; Abou-Zaid, Attya A.; El-Kafas, A.A

    2004-04-01

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the 'rim effect'. High burn-up phenomena in WWER-440 UO{sub 2} fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO{sub 2} fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.

  16. Modeling of WWER-440 fuel pin behavior at extended burn-up

    International Nuclear Information System (INIS)

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the 'rim effect'. High burn-up phenomena in WWER-440 UO2 fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO2 fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

  17. Non destructive assay of nuclear LEU spent fuels for burnup credit application

    International Nuclear Information System (INIS)

    Criticality safety analysis devoted to spent fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as Burnup credit. To be used, Burnup credit involves obtaining evidence of the reactivity loss with a Burnup measurement. Many non destructive assays (NDA) based on neutron as well as on gamma ray emissions are devoted to spent fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously and the link to Burnup is a power link. As a result, burn-up determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels but the convenience of the gamma spectrometric methods is very dependent on characteristics of spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several Nuclear Power Plants in western Europe, gives the average Burnup within a 5% uncertainty and also the extremity Burnup, 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (Active Neutron Interrogation, Passive Neutron

  18. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    International Nuclear Information System (INIS)

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that keff and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  19. Transient behaviour of high burnup fuel

    International Nuclear Information System (INIS)

    The main subjects of the meeting were the discussion of regulatory background, integral tests and analysis, plant calculations, separate-effect test and analysis, concerning high burnup phenomena during RIA accidents in reactors, especially LWR, BWR and PWR type reactors. 32 papers were abstracted and indexed individually for the INIS database. (R.P.)

  20. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

    Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor

  1. Determination of Reference Chemical Potential Using Molecular Dynamics Simulations

    Directory of Open Access Journals (Sweden)

    Krishnadeo Jatkar

    2010-01-01

    Full Text Available A new method implementing molecular dynamics (MD simulations for calculating the reference properties of simple gas hydrates has been proposed. The guest molecules affect interaction between adjacent water molecules distorting the hydrate lattice, which requires diverse values of reference properties for different gas hydrates. We performed simulations to validate the experimental data for determining Δ0, the chemical potential difference between water and theoretical empty cavity at the reference state, for structure II type gas hydrates. Simulations have also been used to observe the variation of the hydrate unit cell volume with temperature. All simulations were performed using TIP4P water molecules at the reference temperature and pressure conditions. The values were close to the experimental values obtained by the Lee-Holder model, considering lattice distortion.

  2. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  3. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  4. Development of advanced cladding material for burnup extension

    International Nuclear Information System (INIS)

    The development of new cladding materials is one of the critical issues on burnup extension. The practical life of Zircaloy would be limited by the growth of oxide films and by the ductility loss due to hydride precipitation, oxygen absorption and radiation damage. In the case of high burnup using MOX fuels, the low neutron adsorption cross section of Zircaloy is not a dominant factor for selecting the cladding material, because MOX fuels can be enriched up to 20%Pu. Austenitic stainless steel, titanium alloy, niobium alloy, ferritic steel and nickel base superalloy are considered as candidate materials. The corrosion resistance, mechanical properties and the irradiation resistance of these materials were examined for evaluating the practical possibility as a cladding material. The austenitic stainless steel with high g phase stability was selected as the primary candidate material. However, it is required to improve the resistance to irradiation associated stress corrosion cracking through the experience in LWR plants. In the JAERI, the austenitic stainless steel with intergranular corrosion resistance has been developed by the adjustment of the chemical composition, the modification of the metallographic structure by thermo-mechanical treatment and the purification by electron beam melting. (author)

  5. Kinetic parameter calculation as function of burn-up of candu reactor

    International Nuclear Information System (INIS)

    Kinetic parameter calculation as function of burn-up of candu reactor. Kinetic marameter calculation as function of burp-up of CANDU reactor with Canflex fuel type-CANDU has been done. This type of fuel is currently being develop, so kinetic parameter such as effective delay neutron fraction (.......), delay neutron decay constant ( .... ) and prompt neutron generation time ( ...... ) are very important for analysis of reactor operation safety. WIMS-CRNL code was used to generate macroscopic cross section and reaction rate based on transport theory. Fast and thermal neutron velocity and macroscopic cross section fission product of the unit cell were determined by KINETIC Code. The result of calculation showed that the value of effective delay neutron fraction was 7,785616 x 10-3 at the beginning of operation at burn-up of 0 MWD/T and after the reactor operated at burn-up of 7,2231 x 10-3 MWD/T was 4,962766 x 10-3, or reduced by 36%. The value of prompt generation time was 9,982703 x 10-4 s at the beginning of operation at burn-up of 0 MWD/T and 8,965416 x 10-4 s after the reactor operated at burn-up of 7,2231 x 103 MWD/T, or reduced by 10%. The result of calculation showed that the values of effective delay neutron fraction and prompt neutron generation time are still great enough

  6. Effect of spent fuel burnup and composition on alteration of the U(Pu)O2 matrix

    International Nuclear Information System (INIS)

    For a potential performance assessment of direct disposal of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified. Leaching experiments were performed with five spent fuel samples from French power reactors (four UO2 fuel samples with burnup ratings of 22, 37, 47 and 60 GWd.THM-1 and a MOX fuel sample irradiated to 47 GWd.THM-1) to determine the release kinetics of the matrix containing most (over 95%) of the radionuclides. The experiments were carried out with granitic groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25 deg C) in a hot cell. After 1000 or 2000 days of leaching, the Sr/U congruence ratios for all the UO2 fuel samples ranged from 1 to 2, allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO2 fuel matrix. The daily strontium release factor was approximately 1 x 10-7 d-1 for UO2 fuel, and five to six times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings. Copyright (2001) Material Research Society

  7. Determination of some chemical and microbiological characteristics of Kaymak

    Directory of Open Access Journals (Sweden)

    Ökten, Sevtap

    2006-12-01

    Full Text Available Kaymak is a kind of concentrated cream, which is traditionally manufactured from buffalo or cow’s milk in Turkey. It is generally consumed with honey at breakfast and some traditional Turkish desserts. The aim of this study was to determine some chemical and microbiological properties of kaymak. The samples were obtained from different dairy plants producing kaymak from cow’s milk and local markets located in Zmir. They were examined for total solids and fat contents, acidity, pH and peroxide values, as well as counts of coliform bacteria, E. coli, yeast and moulds, and Staphylococci. Chemical characteristics of the samples were generally favorable for Turkish Food Codex. However, microbiological properties of some samples were very poor. Careful considerations should be given by the kaymak industry during manufacturing and storage of the product.Kaymak es una clase de crema concentrada, que se fabrica tradicionalmente de la leche del búfalo o de la vaca en Turquía. Se consume generalmente con la miel en el desayuno y en algunos postres turcos tradicionales. El objetivo de este estudio fue determinar algunas características químicas y microbiológicas del kaymak. Las muestras fueron obtenidas de diversas instalaciones lecheras productoras de kaymak de leche de vaca y de mercados locales situados en Zmir. Se analizó el contenido en sólidos totales y grasas, acidez, pH y valores de peróxido, además del conteo de tan bien como cuentas de las bacterias coliformes, E. coli, levadura y mohos, y estafilococos. Las características químicas de las muestras fueron generalmente aceptables para el Turkish Food Codex. Sin embargo, las características microbiológicas de algunas muestras fueron muy malas. La industria del kaymak debe ser extremadamente cuidadosa durante la fabricación y el almacenaje del producto.

  8. BISON, 1-D Burnup and Transport in Slab, Cylindrical, Spherical Geometry

    International Nuclear Information System (INIS)

    1 - Description of problem or function: BISON-1.5 solves the one- dimensional Boltzmann transport equation for neutron and gamma-rays and transmutation equations for fuel nuclides. 2 - Method of solution: In the transport calculation stage the one- dimensional Boltzmann transport equation is solved by the discrete ordinates method. In the burnup calculation stage, transmutation equations for fuel nuclides are solved by Bateman's method. The neutron flux obtained in the transport calculation stage is used to determine the transmutation rates in the burnup calculation stage. Both stages are repeated in tandem till the end of the burnup cycle. 3 - Restrictions on the complexity of the problem: A 42-group neutron and 21-group gamma-ray cross section library is prepared in the code package. Core storage for array variables is dynamically allocated by the code, so there are no restrictions on the size of each array

  9. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  10. Criticality safety analysis of WWER-440 spent fuel cask with radial and axial burnup profile implementation

    International Nuclear Information System (INIS)

    The impact of radial and axial burnup profile on the criticality of WWER-440 spent fuel cask is presented in the paper. The calculations are performed based on two AER Benchmark problems for WWER-440 irradiated fuel assembly. The radial zonewise dependent spent fuel inventory has been calculated by the NESSEL - NUKO code system. The axial dependent isotope concentrations have been determined by the modular code system SCALE4.4. For criticality calculations the SCALE4.4 has been applied. Calculations have been carried out for cask with 30 WWER-440 fuel assemblies with initial enrichment 3.6% of 235U and burnup up to 40 MWd/kgU. The influence of radial and axial burnup credit on the cask criticality has been evaluated

  11. On the theories, techniques, and computer codes used in numerical reactor criticality and burnup calculations

    International Nuclear Information System (INIS)

    The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented

  12. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease. (author)

  13. In-core fuel management amd attainable fuel burn-up in TRIGA

    International Nuclear Information System (INIS)

    The principles of in-core fuel management in research reactors, and especially in TRIGA, are discussed. Calculations made to determine the attainable fuel burn-up values of various fuel element types in the Otaniemi TRIGA Mark II reactor are described and the results obtained are given. Recommendations are given of how to perform the in-core fuel management to achieve good fuel utilization. The results obtained indicate that burn-up values of up to 5 and 2.5 MWd/element can be achieved for the 8 wt-% U Al clad and the 8.5 wt-% U SS clad elements, respectively. (author)

  14. Development of three-dimensional burnup code system based on discrete ordinates (SN) transport method

    International Nuclear Information System (INIS)

    The burnup analysis program based on three dimensional discrete ordinates (SN) neutron/photon transport method has been developed by the FDS team, China, to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The program uses output parameters generated by three-dimensional SN trans- port code to determine the isotopic inventory and anisotropic flux distribution as a function of time. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. The IAEA benchmark test problem has been correctly calculated and analyzed to validate the system. (authors)

  15. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    OpenAIRE

    Muhammad Atta; Iqbal Masood; Mahmood Tayyab

    2011-01-01

    The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determin...

  16. CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback

    International Nuclear Information System (INIS)

    1 - Description of problem or function: CARMEN is a system of programs developed for the neutronic calculation of PWR cycles. It includes the whole chain of analysis from cell calculations to core calculations with burnup. The core calculations are based on diffusion theory with cross sections depending on the relevant space-dependent feedback effects which are present at each moment along the cycles. The diffusion calculations are in one, two or three dimensions and in two energy groups. The feedback effects which are treated locally are: burnup, water density, power density and fission products. In order to study in detail these parameters the core should be divided into as many zones as different cross section sets are expected to be required in order to reproduce reality correctly. A relevant difference in any feedback parameter between zones produces different cross section sets for the corresponding zones. CARMEN is also capable to perform the following calculations: - Multiplication factor by burnup step with fixed boron concentration - Buckling and control rod insertion - Buckling search by burnup step - Boron search by burnup step - Control rod insertion search by burnup step. 2 - Method of solution: The cell code (LEOPARD-TRACA) generates the fuel assembly cross sections versus burnup. This is the basic library to be used in the CARMEN code proper. With a planar distribution guess for power density, water density and fluxes, the macroscopic cross sections by zone are calculated by CARMEN, and then a diffusion calculation is done in the whole geometry. With the distribution of power density, heat accumulated in the coolant and the thermal and fast fluxes determined in the diffusion calculation, CARMEN calculates the values of the most relevant parameters that influence the macroscopic cross sections by zone: burnup, water density, effective fuel temperature and fission product concentrations. If these parameters by zone are different from the reference

  17. Water effect on peroxy radical measurement by chemical amplification: Experimental determination and chemical mechanism

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    The water effect on peroxy radical measurement by chemical amplification was determined experimentally for HO2 and HO2+OH, respectively at room temperature (298±2) K and atmospheric pressure (1×105 Pa). No significant difference in water effect was observed with the type of radicals. A theoretical study of the reaction of HO2·H2O adduct with NO was performed using density functional theory at CCSD(T)/6-311 G(2d, 2p)//B3LYP/6-311 G(2d, 2p) level of theory. It was found that the primary reaction channel for the reaction is HO2·H2O+NO→HNO3+H2O (R4a). On the basis of the theoretical study, the rate constant for (R4a) was calculated using Polyrate Version 8.02 program. The fitted Arrenhnius equation for (R4a) is k = 5.49×107 T 1.03exp(?14798/T) between 200 and 2000 K. A chemical model incorporated with (R4a) was used to simulate the water effect. The water effect curve obtained by the model is in accordance with that of the experiment, suggesting that the water effect is probably caused mainly by (R4a).

  18. A non-chemical spectroscopic determination of atmospheric beryllium

    International Nuclear Information System (INIS)

    Beryllium in the atmosphere is determined by emission spectroscopy using a non-chemical method of analysis. Long term effects of beryllium poisoning result in respiratory and skin disease, and this is partly reflected by the low threshold limits (0.002 mg/m3). In comparison the threshhold values for lead and cadmium are 0.2 and 0.16 mg/m3 respectively. Air samples are collected at 2 litres/ minute using cellulose filters, and sampling time is dependent on the individual process being monitored, but can be as short as five minutes, eg. dental laboratories. The filters are initially divided in two parts, and one portion is carefully pelletised using a steel press. The pellet is placed in an electrode cup and 'wetted' using isopropanol and ethylene glycol. Wetting is necessary because the pellets tended to explode out of the arcing zone. Calibration graphs were produced using an internal cobalt standard, and the 234.8 nm, 313.0 nm emission lines were used. No spectral and inter-element effects were observed, and the minimum detection limit was one nanogram. Under normal working conditions a 25% precision was obtained. (author)

  19. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  20. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    International Nuclear Information System (INIS)

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports

  1. Burnup dependent core neutronic analysis for PBMR

    International Nuclear Information System (INIS)

    The strategy for core neutronics modeling is based on SCALE4.4 code KENOV.a module that uses Monte Carlo calculational methods. The calculations are based on detailed unit cell and detailed core modeling. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and the fuel kernels in the pebble. The core is then modeled by placing these pebbles randomly throughout the core, yet not loosing track of any one of them. For the burnup model, a cyclic manner is adopted by coupling the KENOV.a and ORIGEN-S modules. Shifting down one slice at each discrete time step, and inserting fresh fuel from the top, this cyclic calculation model continues until equilibrium burnup cycle is achieved. (author)

  2. Program package for 2D burnup calculation

    International Nuclear Information System (INIS)

    The program package for 2 dimension burnup calculation was developed for TRIGA Mark III reactor. The package consists of 3 modules: PRESIX, SIXTUS-2, and BURN; 1 library, and 2 input files. PRESIX module prepared cross sections for diffusion calculation. SIXTUS-2 module, a two dimensional diffusion code in hexagonal geometry, calculates keff, neutron fluxes and power distributions. BURN module performs the burnup of fuel elements and stored the result in the ELEM.DAT file. PRESIX.LIB is two group cross section library for major reactor core components prepared using WIMS-D4 code. PRES.INP, the first input file, reads information on reactor power and core loading pattern. ELEM.DAT, the second input file, is prepared for specific TRIGA reactor and dependent on operation history. To verify the reactor model and computational methods, the calculated excess reactivities were compared to the measurement. The results are in good agreement. (author)

  3. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    The authors approach theoretical treatment of isotopic composition changement for nuclear fuel in nuclear reactors. They show the difficulty of exhaustive treatment of burn-up problems and introduce the principal simplifying principles. Due to these principles they write and solve analytically the evolution equations of the concentration for the principal nuclides both in the case of fast and thermal reactors. Finally, they expose and comment the results obtained in the case of a power fast reactor. (author)

  4. Burnup and plutonium distribution of WWER-440 fuel pin at extended burnup

    International Nuclear Information System (INIS)

    The formation of rim region in LWR UO2 based nuclear fuel at high burnup is a common observation. This region has very high porosity due to excessive gas release. Such a region is also characterized by a significantly high plutonium concentration and high local burnup compared to the internal fuel region. Spatial distribution of these parameters has been incorporated with fuel behavior and performance analysis codes by using mostly empirical relations. Variation of these parameters depends on the neutron flux as well as neutron energy spectrum. Detailed neutronics analysis is necessary for the accurate prediction of these parameters. This study is performed by MCNP4B Monte Carlo code for the calculation of local neutron flux, ORIGEN2 for burnup and depletion calculations, and MONTEBURNS for coupling these codes. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell. Fuel pin is divided into a number of radial segments. A relatively small mesh size is used at the region near the surface to reveal the rim effect. The variation of plutonium and local burnup are obtained for high burnup. Results are compared with existing experimental observations for WWER-440 fuel and other theoretical predictions

  5. OECD/NEA Burnup Credit Criticality Benchmark

    International Nuclear Information System (INIS)

    The report describes the final result of the phase-1A of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. The phase-1A benchmark problem is an infinite array of a simple PWR spent fuel rod. The analysis has been performed for the PWR spent fuels of 30 and 40 GWd/t after 1 and 5 years of cooling time. In total, 25 results from 19 institutes of 11 countries have been submitted. For the nuclides in spent fuel, 7 major actinides and 15 major fission products (FP) are selected for the benchmark calculation. In the case of 30 GWd/t burnup, it is found that the major actinides and the major FPs contribute more than 50% and 30% of the total reactivity loss due to burnup, respectively. Therefore, more than 80% of the reactivity loss can be covered by 22 nuclides. However, the larger deviation among the reactivity losses by participants has been found for cases including EPs than the cases with only actinides, indicating the existence of relatively large uncertainties in FP cross sections. The large deviation seen also in the case of the fresh fuel has been found to reduce sufficiently by replacing the cross section library from ENDF-B/IV with that from ENDF-B/V and taking the known bias of MONK6 into account. (author)

  6. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  7. Determination of gold by chemical hydride generation atomic absorption spectrometry

    International Nuclear Information System (INIS)

    Complete text of publication follows. The chemical vapour generation (CVG) of transition and noble metals opens a novel route for introduction of these elements into atomic spectrometric sources. It can be accomplished by merging an acidic sample with tetrahydroborate reductant solution (Y. L. Feng et al., J. of Anal. At. Spectrom., 20 (2005) 255-265). There have been some studies for determination of Au; however, only mg L-1 levels of gold have been determined by CVG - Atomic Absorption Spectrometry (AAS) (G. Ertas et al., Applied Spectroscopy, 60 (2006) 423-429). Volatile Au species were generated in flow injection arrangement from acid environment in presence of surfactants. The core of the system is a mixing manifold based on 3 concentric capillaries (T. Matousek et al., J. of Anal. At. Spectrom., 18 (2003) 487-494) protruding into the glass gas-liquid separator (glass, volume 3 ml). Optimum flow rate of Ar as a carrier gas was found at 240 mL/min. The study of generation parameters as well as the use of reaction modifiers-surfactants and dithiocarbamate- will be presented. Quartz tube multiatomizer for AAS was employed for atomization. Atomization conditions including composition of carrier gases and their flow rates and atomization temperature were optimized. 900 deg C was found as the optimum atomization temperature; over 900 deg C, peak area of Au signal decreased; in addition, peak shape was altered. A sharp maximum of 6 mL/min oxygen as the outer gas was observed. Another important point was that hydrogen-rich atmosphere caused signal depression. Analytical performance of this approach to generation and atomization will be discussed and perspectives of its future will be outlined. This work was supported by the GA ASCR (grant No. A400310507 and IAA400310704) and Institute of Analytical Chemistry of the ASCR, v.v.i. (project no. AV0Z40310501). This work also was supported from OYP (Faculty Development Program) from the Middle East Technical University

  8. Key issues in nuclear fuel cycle concerning high burn-up strategy

    International Nuclear Information System (INIS)

    In the present high burn-up strategy in Japan, the economic efficiency and reduction of the spent nuclear fuel have been in progress. On the other hand, in the further progress of the strategy, several issues may appear. The amount and activity of nuclides, heat generation, and radiation for a fuel pin in the typical 17x17 PWR assembly were calculated as functions of burn-up and cooling time, using the SWAT code system. Waste loading in glass waste forms from spent UO2 fuel and MOX fuel were discussed, assuming the number of glass canisters of 150 liter per THM is 1.25 at 45 GWd/THM. The number of glass canisters per GWd is almost constant in the range of burn-up up to 70 GWd/THM. The amount of molybdate from Pu-239 fissions linearly increases as a function of burn-up similarly like increase from U-235 fissions. The current vitrification technology may not face serious situation to be required substantial reduction in waste loading relating to molybdate up to 70 GWd/THM. The initial cooling period prior to vitrification, the waste loading and the interim storage period prior to final disposal are major factors which determine the way of storage and final disposal. The higher burn-up above 45 GWd/THM may require pretreatment of HLLW or substantial reduction in waste loading to retain the integrity of the ceramic melter for e.g. five years. Further promotion of high burn-up strategy should be consistent with nuclear fuel cycle including waste management. A potential approach, a conceptual new reprocessing system for thermal reactors is described. (author)

  9. Burn-up and cycle length optimization project of the robust fuel programme

    International Nuclear Information System (INIS)

    The Spanish electric sector (UNESA) takes part in the Robust Fuel programme in the different work groups set up by EPRI. Iberinco, with the collaboration of Iberdrola Generacion (TECNO and Cofrentes NPP) and Soluziona Ingenieria, has created a stable multidisciplinary group to assimilate and follow up this program, analyzing in detail the technology generated and evaluating the conclusions to provide the most suitable recommendations for application. Along these lines, one of the most promising projects within technical group 3 (High burn properties) has been the one called Burn-up and cycle Length Optimization. In January 2000 Duke Power published a study on the plants it owns (PWR type) and 18-month cycles, to establish the optimum unloading burn-up of fuel. The conclusion it reached is that the fuel cost drops t a minimum for average unload burn-ups of between 60 and 70 GWd/MTU. As an extension to this study and covering a wider base of considerations, Exelon, with the support of Westinghouse and the University of Pennsylvania, released a study in December 2001 on different reference cores with different cycle lengths. In this study, the optimum burn-up without exceeding current maximum enrichment limits (5%) is determined. Publication of the results of the second phase, considering higher enrichments, was due in the summer of 2002. The design of the core to be refueled and economic analyzes show that both pressurized water reactors (PWR) and boiling water reactors (BWR) can obtain significant benefits by increasing the fuel unloading burn-up above currently licensed limits. However, the optimum unload burn-up level is not reached without exceeding the current enrichment limit of 5% . (Author)

  10. Taking burnup credit for interim storage and transportation system for BWR fuels

    International Nuclear Information System (INIS)

    In the back-end issues of nuclear fuel cycle, selection of reprocessing or one-through is a big issue. For both of the cases, a reasonable interim storage and transportation system is required. This study proposes an advanced practical monitoring and evaluation system. The system features the followings: (l) Storage racks and transportation casks taking credit for burnup. (2) A burnup estimation system using a compact monitor with Cd- Te detectors and fission chambers. (3) A neutron emission-rate evaluation methodology, especially important for high burnup MOX fuels. (4) A nuclear materials management system for safeguards. Current storage system and transport casks are designed on the basis of a fresh fuel assumption. The assumption is too conservative. Taking burnup credit gives a reasonable design while keeping conservatism. In order to establish a reasonable burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of some modules such as TGBLA, ORIGEN, CITATION, MCNP and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. The code takes operational history such as, power density, void fraction into account. This code is applied to the back-end issues for a more accurate design of a storage and a transportation system. The ORIGEN code is well-known one-point isotope depletion code. In the calculation system, the code calculates isotope compositions using libraries generated from the TGBLA code. The CITATION code, the MCNP code, and the KENO code are three dimensional diffusion code, continuous energy Monte Carlo code, discrete energy Monte Carlo code, respectively. Those codes calculate k- effective of the storage and transportation systems using isotope compositions generated from the ORIGEN code. The CITATION code and the KENO code are usually used for practical designs. The MCNP code is used for reference

  11. Burnup effects of MOX fuel pincells in PWR - OECD/NEA burnup credit benchmark analysis -

    International Nuclear Information System (INIS)

    The burnup effects were analyzed for various cases of MOX fuel pincells of fresh and irradiated fuels by using the HELIOS, MCNP-4/B, CRX and CDP computer codes. The investigated parameters were burnup, cooling time and combinations of nuclides in the fuel region. The fuel compositions for each case were provided by BNFL (British Nuclear Fuel Limited) as a part of the problem specification so that the results could be focused on the calculation of the neutron multiplication factor. The results of the analysis show that the largest saving effect of the neutron multiplication factor due to burnup credit is 30 %. This is mainly due to the consideration of actinides and fission products in the criticality analysis

  12. Program Helps To Determine Chemical-Reaction Mechanisms

    Science.gov (United States)

    Bittker, D. A.; Radhakrishnan, K.

    1995-01-01

    General Chemical Kinetics and Sensitivity Analysis (LSENS) computer code developed for use in solving complex, homogeneous, gas-phase, chemical-kinetics problems. Provides for efficient and accurate chemical-kinetics computations and provides for sensitivity analysis for variety of problems, including problems involving honisothermal conditions. Incorporates mathematical models for static system, steady one-dimensional inviscid flow, reaction behind incident shock wave (with boundary-layer correction), and perfectly stirred reactor. Computations of equilibrium properties performed for following assigned states: enthalpy and pressure, temperature and pressure, internal energy and volume, and temperature and volume. Written in FORTRAN 77 with exception of NAMELIST extensions used for input.

  13. Simulation of the behaviour of nuclear fuel under high burnup conditions

    International Nuclear Information System (INIS)

    Highlights: • Increasing the time of nuclear fuel into reactor generates high burnup structure. • We analyze model to simulate high burnup scenarios for UO2 nuclear fuel. • We include these models in the DIONISIO 2.0 code. • Tests of our models are in very good agreement with experimental data. • We extend the range of predictability of our code up to 60 MWd/KgU average. - Abstract: In this paper we summarize all the models included in the latest version of the DIONISIO code related to the high burnup scenario. Due to the extension of nuclear fuels permanence under irradiation, physical and chemical modifications are developed in the fuel material, especially in the external corona of the pellet. The codes devoted to simulation of the rod behaviour under irradiation need to introduce modifications and new models in order to describe those phenomena and be capable to predict the behaviour in all the range of a general pressurized water reactor. A complex group of subroutines has been included in the code in order to predict the radial distribution of power density, burnup, concentration of diverse nuclides and porosity within the pellet. The behaviour of gadolinium as burnable poison also is modelled into the code. The results of some of the simulations performed with DIONISIO are presented to show the good agreement with the data selected for the FUMEX I/II/III exercises, compiled in the NEA data bank

  14. Implementation of burnup credit in spent fuel management systems

    International Nuclear Information System (INIS)

    Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate safety margin. The motivation for using burnup credit in criticality safety applications is based on economic considerations and additional benefits contributing to public health and safety and resource conservation. Interest in the implementation of burnup credit has been shown by many countries. In 1997, the International Atomic Energy Agency (IAEA) started a task to monitor the implementation of burnup credit in spent fuel management systems, to provide a forum to exchange information, to discuss the matter and to gather and disseminate information on the status of national practices of burnup credit implementation in the Member States. The task addresses current and future aspects of burnup credit. This task was continued during the following years. (author)

  15. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations

    International Nuclear Information System (INIS)

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor keff (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  16. Theory analysis and simple calculation of travelling wave burnup scheme

    International Nuclear Information System (INIS)

    Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)

  17. The US department of energy's transportation burnup credit program

    International Nuclear Information System (INIS)

    Aspects of the U. S. Department of Energy's (DOE's) transportation burnup credit program, the Department's motivation for conducting the program, and the status of burnup credit activities are presented. The benefits, technical, and regulatory considerations associated with using burnup credit for transport of irradiated nuclear fuel are discussed. The methods used in the DOE's actinide-only topical report are described in terms of the technical and regulatory issues. (authors)

  18. Core burnup characteristics of high conversion light water reactor, (1)

    International Nuclear Information System (INIS)

    In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in the reactor (HCLWR-J1) under study. The burnup calculations were carried out under the assumption of three batch and out-in fuel loading from the first cycle to the equilibrium cycle. A detailed evaluation was made for discharge burnup, conversion ratio, power distribution, and reactivity coefficients and so on. (author)

  19. Using chemical benchmarking to determine the persistence of chemicals in a Swedish lake.

    Science.gov (United States)

    Zou, Hongyan; Radke, Michael; Kierkegaard, Amelie; MacLeod, Matthew; McLachlan, Michael S

    2015-02-01

    It is challenging to measure the persistence of chemicals under field conditions. In this work, two approaches for measuring persistence in the field were compared: the chemical mass balance approach, and a novel chemical benchmarking approach. Ten pharmaceuticals, an X-ray contrast agent, and an artificial sweetener were studied in a Swedish lake. Acesulfame K was selected as a benchmark to quantify persistence using the chemical benchmarking approach. The 95% confidence intervals of the half-life for transformation in the lake system ranged from 780-5700 days for carbamazepine to <1-2 days for ketoprofen. The persistence estimates obtained using the benchmarking approach agreed well with those from the mass balance approach (1-21% difference), indicating that chemical benchmarking can be a valid and useful method to measure the persistence of chemicals under field conditions. Compared to the mass balance approach, the benchmarking approach partially or completely eliminates the need to quantify mass flow of chemicals, so it is particularly advantageous when the quantification of mass flow of chemicals is difficult. Furthermore, the benchmarking approach allows for ready comparison and ranking of the persistence of different chemicals. PMID:25565241

  20. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  1. High Burnup Fuel Performance and Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

    2007-03-15

    The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

  2. The burnup dependence of light water reactor spent fuel oxidation

    International Nuclear Information System (INIS)

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO2 to higher oxides. The oxidation of UO2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO2 to UO2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

  3. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  4. Use of burnup credit in criticality safety design analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    Full text: It is well known that the use of Burnup Credit (BUC) in criticality safety design analysis of spent fuel storage systems significantly impacts the design of the system. BUC is defined as the consideration of the change in the fuel's isotopic composition and hence in its reactivity due to the irradiation of the fuel. Using BUC means to identify that isotopic composition and hence that burnup which just results in the maximum neutron multiplication factor allowable for the system, including all mechanical and calculational uncertainties. This burnup is the minimum burnup necessary for fuel to be loaded in the system. Since the isotopic composition at given burnup depends on the initial enrichment of the fuel, the minimum burnup is usually given as a function of the initial enrichment. The graph of this function is commonly named as 'loading curve'. Thus, application of BUC to a spent fuel storage system consists in implementation of three key steps: Determination of the isotopic composition as a function of burnup and initial enrichment; Criticality calculation and evaluation of the loading curve; Quantification and verification of the actual burnup of the fuel to be loaded into the system. The main considerations of the first and the second step will be discussed. The isotopic composition is predicted by means of depletion calculations. To perform such calculations the parameters describing the fuel design characteristics and the fuel depletion conditions have to be defined. In addition the cooling time that may be credited (e.g., in BUC applications to spent fuel storage/transport cask systems) has to be specified. These parameters will be discussed with particular attention being given to the sensitivity of the neutron multiplication factor of the storage system to variations in the parameters and conditions characterizing the depletion conditions. These parameters and conditions are: Specific power and operating history, fuel temperature, moderator

  5. Study on the criticality safety evaluation method for burnup credit in JAERI

    International Nuclear Information System (INIS)

    In relation to burnup credit, three tasks have been carried out at the Japan Atomic Energy Research Institute (JAERI) for establishing the evaluation method of criticality safety for a spent-fuel system, such as storage age ponds and transport casks. The first task is to prepare a benchmark database of criticality experiments and nuclide compositions of spent fuels. The database of nuclide composition is formed by data treasured at JAERI and data collected from the literature. For the database of criticality experiments, the effective multiplication factor of a spent-fuel assembly has been measured at JAERI and data collected from the literature. For the database of criticality experiments, the effective multiplication factor of a spent-fuel assembly has been measured at JAERI. The next task is to develop computer codes. The burnup and criticality codes have been developed and validated by analyzing a large number of benchmarks stored in the aforementioned database. The last task needed to establish the methodology in order to confirm the subcriticality of a spent-fuel system applying burnup credit is described. A reference fuel assembly is introduced so that the criticality of a system can be evaluated by using it, instead of modeling all fuel assemblies explicitly. To determine the nuclide composition of a spent fuel, a simple method is studied utilizing a large number of nuclide composition data stored in the database. Further, the effects of the axial burnup profile and calculation errors are discussed, and the remaining tasks are identified

  6. Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

    International Nuclear Information System (INIS)

    The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.

  7. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    International Nuclear Information System (INIS)

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  8. Power excursion analysis for high burnup cores

    International Nuclear Information System (INIS)

    A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling water reactor, the events identified as having significant energy deposition in the fuel were a rod drop accident, a recirculation flow control failure, and the overpressure events; in a pressurized water reactor, they were a rod ejection accident and boron dilution events. The RDA analysis was done with RAMONA-4B, a computer code that models the space- dependent neutron kinetics throughout the core along with the thermal hydraulics in the core, vessel, and steamline. The results showed that the calculated maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important uncertainties in each of these categories are discussed in the report

  9. BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies

    International Nuclear Information System (INIS)

    BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inventory as a function of time and power density. The code allows for multiple fueled regions to be analyzed. The companion code, RELOAD, can be used to shuffle fueled regions or reload regions with fresh fuel. BURNCAL can be used to study the reactivity effects and isotopic inventory as a function of time for a nuclear reactor system. Neutron transmutation, fission, and radioactive decay are included in the modeling of the production and removal terms for each isotope of interest. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. Fueled and un-fueled regions, such as cladding and moderator, can be analyzed simultaneously. The nuclides analyzed are limited only by the neutron cross section availability in the MCNP cross-section library. BURNCAL is unique in comparison to other burnup codes in that it does not use the calculated neutron flux as input to other computer codes to generate the nuclide mixture for the next time step. Instead, BURNCAL directly uses the neutron absorption tally/reaction information generated by MCNP for each nuclide of interest to determine the nuclide inventory for that region. This allows for the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed

  10. BASIS FOR DETERMINATION OF CHEMICAL STABILITY and COMPATIBILITY OF SOLID WASTE CHEMICAL COMPATIBILITY TECHNICAL BASIS

    International Nuclear Information System (INIS)

    Solid wastes must be managed to prevent inadvertent reactions, explosion and degradation of waste containers per the ''Washington State Department of Ecology Dangerous Waste Regulations'' (WAC 173-303). An understanding of chemical compatibility principles and a consistent approach for implementing compatibility requirements is essential for complying with the regulations. This document explains the technical basis for ensuring chemical compatibility for solid wastes that are stored on site at on-site TSD facilities and for solid waste that will go to off-site TSD facilities. The document applies directly to the following aspects of chemical compatibility: (1) Ensuring that hazardous waste is not chemically reactive or unstable such that it cannot be safely transported or stored; (2) Ensuring that lab packs (i.e., drums containing multiple inner containers of differing types of hazardous waste) are packaged such that incompatible chemicals are not placed into the same drum; (3) Selecting containers and liners that are compatible with the waste they contain. This document does not cover individual TSD requirements, or specific offsite TSD requirements. This document does not cover chemical compatibility and segregation requirements for shipping wastes on-site or off-site. This document does not cover radiological hazards associated with radioactive waste or mixed wastes. Evaluation of compatibility for comingling and treating solid waste is beyond the scope of this document. In addition, heat generation and gas generation as they apply to the Hanford waste acceptance criteria are not covered in this document

  11. Radium-226 determination in different substances without preliminary chemical isolation

    International Nuclear Information System (INIS)

    A method of radium-226 determination, based on the measurement of radon-222 quantity, extracted from the substance analyzed, is developed. Determination limit of radium, equal to 6.6x10-15 mass.%, is achieved at the expense of application of low-background proportional counter and special vacuum installation to fill the detector with working gas mixture

  12. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  13. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  14. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

    Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1/2 truck cask, which has been relicensed to carry pressurized water reactor (PWR) fuel with 56,000 MWd/ton U burnup at 450 days of cooling time. New cask designs should consider the effects of high burnup for future use, even though it is not commercially desirable to include currently unneeded capability. In conclusion, the increased heat and gamma radiation of high-burnup fuels can be accommodated by additional cooling time, but the increased neutron radiation source cannot be accommodated unless the balance of neutron and gamma contributions to the overall dose rate is properly chosen in the initial cask design. Criticality control of high-burnup fuels is possible with heavily poisoned baskets, but burnup credit in licensing is a much more direct means of demonstrating criticality safety

  15. Implementation of burnup credit in PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    Implementation of burnup credit in spent fuel storage of LWR fuel at nuclear power plants is approved in Germany since the beginning of 2000. The burnup credit methods applied have to comply with the newly developed German criticality safety standard DIN 25471 passed in November 1999 and published in September 2000, cp. (orig.)

  16. Method of compensating distribution of reactor burnup degree

    International Nuclear Information System (INIS)

    An object of the present invention is to attain an appropriate power distribution and a burnup degree distribution during an operation cycle, thereby improving the succeeding operation cycle in a BWR type reactor. That is, a deviation between a distribution of an actual axial burnup degree and that of an aimed axial burnup degree in a reactor core is measured upon completion of the operation cycle by using a burnup degree distribution measuring device. Then, the content of burnable poisons in fresh fuels to be charged to the reactor core is controlled in accordance with the deviation, to compensate the distribution of the axial burnup degree in the reactor core in the next operation cycle. Accordingly, the distribution of the axial burnup degree in the reactor core can be made closer to the aimed distribution of the burnup degree in the next operation cycle. Further, appropriate power distribution and a burnup degree distribution can be obtained by improving the axial power distribution in the reactor core with the characteristics of the fresh fuels themselves to be loaded, without depending only on changes of a control rod pattern. Accordingly, fuel economy and operation performance can be improved. (I.S.)

  17. Determination of chemical states of sulphur 35 obtained from the 35Cl (n, p)35S

    International Nuclear Information System (INIS)

    The chemical states of sulphur-35 obtained from the 35Cl(n,p)35S reaction by the irradiation of potassium chloride without any previous treatment and with previous heating under vacuum, were determined. The influence of irradiation time and temperature after irradiation was examined. Paper electrophoresis technique was employed for the determination of the chemical states. (Author)

  18. Analysis of determination modalities concerning the exposure and emission limits values of chemical and radioactive substances

    International Nuclear Information System (INIS)

    This document presents the generic approach adopted by various organizations for the determination of the public exposure limits values to chemical and radioactive substances and for the determination of limits values of chemical products emissions by some installations. (A.L.B.)

  19. Finnish contribution to the CB4 burnup credit benchmark

    International Nuclear Information System (INIS)

    The CB4 phase of the WWER burnup credit benchmark series studies the effect of flat and realistic axial burnup profiles on the multiplication factor of a conceptual WWER cask loaded with spent fuel. The benchmark was calculated at VTT Energy with MCNP4C, using mainly ENDF/B-V1 cross sections. According to the calculation results the effect of the axial homogenization on the keff estimate is complex. At low burnups the use of a axial profile overestimates keff but at high burnups the reverse is the case. Ignoring fission products leads to conservative keff and the effect of axial homogenization on the multiplication factor is similar to a reduction of the burnup (Authors)

  20. Probabilistic assessment of dry transport with burnup credit

    International Nuclear Information System (INIS)

    The general concept of probabilistic analysis and its application to the use of burnup credit in spent fuel transport is explored. Discussion of the probabilistic analysis method is presented. The concepts of risk and its perception are introduced, and models are suggested for performing probability and risk estimates. The general probabilistic models are used for evaluating the application of burnup credit for dry spent nuclear fuel transport. Two basic cases are considered. The first addresses the question of the relative likelihood of exceeding an established criticality safety limit with and without burnup credit. The second examines the effect of using burnup credit on the overall risk for dry spent fuel transport. Using reasoned arguments and related failure probability and consequence data analysis is performed to estimate the risks of using burnup credit for dry transport of spent nuclear fuel. (author)

  1. Non-destructive burn-up degree evaluation method for nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Makoto; Kumanomido, Hironori

    1998-01-06

    The present invention concerns a non-destructive burn-up degree evaluation method for spent fuels by a spontaneous neutron releasing rate method. Namely, an equation (1) is provided as: S = ({phi}/P)x(1-k) where {phi} is spontaneous neutron flux, P is the proportional coefficient, S is neutron releasing rate and k is neutron effective multiplication factor. S is further given by an equation (2): S = S4{sub 0}x(1+S2/S4{sub 0})xVxT where S2 is releasing rate from Cm242, S4{sub 0} is releasing rate from other nuclides, v is a void ratio of coolants and T is a time decaying effect, and the equations (1) and (2) are joined. P is determined by theoretical calculation, and S2/S4{sub 0} is determined based on a half decay characteristics of Cm242 to determine a correction amount. S4{sub 0} and V are determined as a correlational function of the burn-up degree: x based on burning calculation while using the Pu enrichment degree {epsilon}, Pu compositional ratio f, and concrete void ratio v. k is determined as a correlational function of v. A first appropriate value of x is obtained while having the burnup degree x{sup (0)} as an initial value. x is determined successively by repeating calculation based on modified k in this case. (I.S.)

  2. Burnup-dependent cross section data for research reactors

    International Nuclear Information System (INIS)

    Studies currently in progress consider research and test reactors which commonly have burnups of 50 atom percent in 235-U and may reach as high a 70 atom percent. At these levels of burnup changes in cross-section data with burnup become significant. Some preliminary studies of these effects lead to the development of a modified version of REBUS-2 which supports changes in cross-section data with burnup. This version of REBUS-2 allows for changes in the cross-section data only at each time sub-interval in the problem, and these cross-section changes for capture and fission are based on a least squares polynomial fit as a function of burnup. In this paper an attempt is made to evaluate the importance of burnup dependent data for the various isotopes and/or groups, and to assess the accuracy of this method by comparing the REBUS-2 results with results obtained from PDQ-7. The 10 MW IAEA benchmark problem has been selected for this study. A description of the reactor and the XY model can be found in the IAEA Guidebook. The EPRI-CELL4 code was used to generate burnup dependent cross section data for use with both REBUS-2 and PDQ-7. Cross-section data were generated at 10 time steps to a burnup of approximately 50 atom percent in 235-U. The agreement between the PDQ-7 results and the REBUS-2 results with fitted burnup dependent cross-section data are quite good. Burnup dependent cross sections are essential for accurate estimates of cycle lengths and reactivities, and low order polynomial fits of capture and fission data for selected isotopes and energy groups can provide this capability

  3. A validated methodology for evaluating burnup credit in spent fuel casks

    International Nuclear Information System (INIS)

    The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the U.S. Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in keff. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. (Author)

  4. A validated methodology for evaluating burnup credit in spent fuel casks

    International Nuclear Information System (INIS)

    The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in keff. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs

  5. Analytical and numerical study of radiation effect up to high burnup in power reactor fuels

    International Nuclear Information System (INIS)

    In the present work the behavior of fuel pellets for power reactors in the high burnup range (average burnup higher than 50 MWd/kgHM) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup, as long as a new microstructure develops, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behaviour. The evolution of porosity in the high burnup structure (HBS) is assumed to be determinant of the retention capacity of the fission gases released by the matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Starting from several works published in the open literature, a model was developed to describe the behaviour and evolution of porosity at local burnup values ranging from 60 to 300 MWd/KgHM. The model is mathematically expressed by a system of non-linear differential equations that take into account the open and closed porosity, the interactions between pores and the free surface and phenomena like pore's coalescence and migration and gas venting. Interactions of different orders between open and closed pores, growth of pores radius by vacancies trapping, the evolution of the pores number density, the internal pressure and over pressure within the pores, the fission gas retained in the matrix and released to the free volume are analyzed. The results of the simulations performed in the present work are in excellent agreement with experimental data available in the open literature and with results calculated by other authors (author)

  6. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    International Nuclear Information System (INIS)

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are 149Sm, 151Sm, and 155Gd

  7. Benefits of the delta K of depletion benchmarks for burnup credit validation

    International Nuclear Information System (INIS)

    Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, keff. The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

  8. Parametric Study of Control Rod Exposure for PWR Burnup Credit Criticality Safety Analyses

    International Nuclear Information System (INIS)

    The Interim Staff Guidance on burnup credit (ISG-8) for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office, recommends the use of analyses that provide an ''adequate representation of the physics'' and notes particular concern with the ''need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted.'' In the absence of readily available information on the extent of control rod (CR) usage in U.S. PWRs and the subsequent reactivity effect of CR exposure on discharged SNF, NRC staff have indicated a need for greater understanding in these areas. In response, this paper presents results of a parametric study of the effect of CR exposure on the reactivity of discharged SNF for various CR designs (including Axial Power Shaping Rods), fuel enrichments, and exposure conditions (i.e., burnup and axial insertion). The study is performed in two parts. In the first part, two-dimensional calculations are performed, effectively assuming full axial CR insertion. These calculations are intended to bound the effect of CR exposure and facilitate comparisons of the various CR designs. In the second part, three-dimensional calculations are performed to determine the effect of various axial insertion conditions and gain a better understanding of reality. The results from the study demonstrate that the reactivity effect increases with increasing CR exposure (e.g., burnup) and decreasing initial fuel enrichment (for a fixed burnup). Additionally, the results show that even for significant burnup exposures, minor axial CR insertions (e.g., eff of a spent fuel cask

  9. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    International Nuclear Information System (INIS)

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155

  10. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  11. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  12. Moessbauer spectroscopic determination of chemical state of iron in bauxite

    International Nuclear Information System (INIS)

    The chemical state of iron contained in several kinds of bauxite, which are utilized as a raw material in the aluminum industry in Japan, were investigated by Moessbauer spectroscopy. The main compounds of iron were identified from the results, which showed variations of the Moessbauer absorption spectra with calcination and measuring temperature. Although the absorption intensities of the spectra differed significantly, major species identified were paramagnetic or superparamagnetic α-Fe2O3 in all of these bauxite samples. The superparamagnetic α-Fe2O3 was found mainly in the gibbsite-type bauxite, but not in the boehmite/gibbsite-type or the boehmite-type bauxite. The Moessbauer absorption spectra of red mud and its calcined products were also given. (author)

  13. Determination of solute organic concentration in contaminated soils using a chemical-equilibrium soil column system

    DEFF Research Database (Denmark)

    Gamst, Jesper; Kjeldsen, Peter; Christensen, Thomas Højlund

    determination of solute concentration in a contaminated soil were developed; (1) a chemical Equilibrium and Recirculation column test for Volatile organic chemicals (ER-V) and (2) a chemical Equilibrium and Recirculation column test for Hydrophobic organic chemicals (ER-H). The two test systems were evaluated...... 80) an unacceptable recovery was found (9%). The contact time needed for obtaining chemical equilibrium was tested in the ER-H system by performing five test with different duration (1, 2, 4, 7 and 19 days) using the low organic carbon soil. Seven days of contact time appeared sufficient for......Groundwater risk assessment of contaminated soils implies determination of the solute concentration leaching out of the soil. Determination based on estimation techniques or simple experimental batch approach has proven inadequate. Two chemical equilibrium soil column leaching tests for...

  14. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  15. Establishing the fuel burn-up measuring system for 106 irradiated assemblies of Dalat reactor by using gamma spectrometer method

    International Nuclear Information System (INIS)

    The fuel burn-up is an important parameter needed to be monitored and determined during a reactor operation and fuel management. The fuel burn-up can be calculated using computer codes and experimentally measured. This work presents the theory and experimental method applied to determine the burn-up of the irradiated and 36% enriched VVR-M2 fuel type assemblies of Dalat reactor. The method is based on measurement of Cs-137 absolute specific activity using gamma spectrometer. Designed measuring system consists of a collimator tube, high purity Germanium detector (HPGe) and associated electronics modules and online computer data acquisition system. The obtained results of measurement are comparable with theoretically calculated results. (author)

  16. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    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

  17. The commercial impact of burnup increase

    International Nuclear Information System (INIS)

    Deregulation has a dramatic effect on competition in the electricity markets. This will lead to a continued pressure on the prices in virtually all areas of the nuclear fuel cycle and will encourage further optimization, technical and technological progress and innovations with respect to further cost reductions of power production. The permission of direct disposal, in Germany legally granted in 1994 as an alternative to the reprocessing path, made possible cost savings and has consequently resulted in a decline of reprocessing prices. In addition, suppliers as well as operators are making considerable efforts to reduce the disposal costs fraction by optimizing disposal technologies and concepts. The increase of discharge has essentially contributed to the reduction the disposal cost fraction. Compared to former scenarios, the economic potential of burn-up increase is decreasing

  18. Chemical and microbiological farm milk quality determination in three Croatian regions

    OpenAIRE

    Neven Antunac; Jasna Đermadi; Miljenko Konjačić; Rajka Božanić; Zoran Bašić; Vera Volarić

    2012-01-01

    The purpose of this study was to determine chemical and microbiological quality of raw milk from 30 farms of different sizes from eastern, central and southern Croatian regions. Samples of fresh raw milk (n=360) are determined by the content of fat, protein, total solids, and the number of microorganisms and somatic cells. Analysed milk derived from Holstein, Simmental and Brown Swiss cows, and their crossbred. Chemical composition of milk was determined by infrared spectrophotometry, microbi...

  19. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

  20. Experimental programmes related to high burnup fuel

    International Nuclear Information System (INIS)

    The experimental programmes undertaken at IGCAR with regard to high burn-up fuels fall under the following categories: a) studies on fuel behaviour, b) development of extractants for aqueous reprocessing and c) development of non-aqueous reprocessing techniques. An experimental programme to measure the carbon potential in U/Pu-FP-C systems by methane-hydrogen gas equilibration technique has been initiated at IGCAR in order to understand the evolution of fuel and fission product phases in carbide fuel at high burn-up. The carbon potentials in U-Mo-C system have been measured by this technique. The free energies and enthalpies of formation of LaC2, NdC2 and SmC2 have been measured by measuring the vapor pressures of CO over the region Ln2O3-LnC2-C during the carbothermic reduction of Ln2O3 by C. The decontamination from fission products achieved in fuel reprocessing depends strongly on the actinide loading of the extractant phase. Tri-n-butyl phosphate (TBP), presently used as the extractant, does not allow high loadings due to its propensity for third phase formation in the extraction of Pu(IV). A detailed study of the allowable Pu loadings in TBP and other extractants has been undertaken in IGCAR, the results of which are presented in this paper. The paper also describes the status of our programme to develop a non-aqueous route for the reprocessing of fast reactor fuels. (author)

  1. Light a CANDLE. An innovative burnup strategy of nuclear reactors

    International Nuclear Information System (INIS)

    CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)

  2. Current Status of Burnup Evaluation for Test Fuel at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Park, Seung Jae; Shin, Yoon Taeg; Choo, Kee Nam; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For the research reactor, 8 mini plate fuels were irradiation-tested during 4 irradiation cycles. 2 more irradiation capsules were fabricated for additional test of plate type fuel. Also fission Mo target for the performance verification and the demonstration of Mo-99 extraction process will be irradiated at HANARO. It is important to evaluate the burnup history of test fuel. The burnup of test fuel has been calculated using HANARO Fuel Management System (HANAFMS). Although it is proper to evaluate the burnup of HANARO fuel, it is difficult to accurately calculate the burnup of test fuel due to the limitation of HANAFMS model. Therefore, the improvement of burnup evaluation for the recent irradiated test fuel is conducted and reported in this paper. To evaluate the burnup of test fuel, HANAFMS has been used; however, HANAFMS model is not proper to apply plate type fuel. Therefore, MCNP burned core model was developed for HAMP-1 burnup calculation. Throughout the comparison of fuel assembly power, MCNP burned core model showed the good agreement with HANAFMS.

  3. Calculation study of TNPS spent fuel pool using burnup credit

    International Nuclear Information System (INIS)

    Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks, the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied. The MONK9A code was used to analyze keff, of different enrichment fuels at different burnups. A reference loading curve was proposed in accordance with the system keff's changing with the burnup of different initially enriched nuclear fuels. The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC. (authors)

  4. Burnup calculation methodology in the serpent 2 Monte Carlo code

    International Nuclear Information System (INIS)

    This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

  5. BNFL assessment of methods of attaining high burnup MOX fuel

    International Nuclear Information System (INIS)

    It is clear that in order to maintain competitiveness with UO2 fuel, the burnups achievable in MOX fuel must be enhanced beyond the levels attainable today. There are two aspects which require attention when studying methods of increased burnups - cladding integrity and fuel performance. Current irradiation experience indicates that one of the main performance issues for MOX fuel is fission gas retention. MOX, with its lower thermal conductivity, runs at higher temperatures than UO2 fuel; this can result in enhanced fission gas release. This paper explores methods of effectively reducing gas release and thereby improving MOX burnup potential. (author)

  6. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    In fact, till now, burnup credit has not be applied in Ukrainian nuclear power for spent fuel management systems (storage and transport). However, application of advanced fuel at VVER reactors, arising spent fuel amounts, represent burnup credit as an important resource to decrease spent fuel management costs. The paper describes spent fuel management status in Ukraine from viewpoint of subcriticality assurance under spent fuel storage and transport. It also considers: 1. Regulation basis concerning subcriticality assurance, 2. Basic spent fuel and transport casks characteristics, 3. Possibilities and demands for burnup credit application at spent fuel management systems in Ukraine. (author)

  7. Estimate of preliminary experiments to study the burn-up of gadolinium as a poison

    International Nuclear Information System (INIS)

    Full text: Proposed preliminary experiments to determine the burn-up of Gd2O3 as a poison in different reactors are discussed. Estimates are given of parameters such as the weight of the sample to be irradiated, irradiation and decay times, expected activity and photon spectrum. 1 g samples of natural UO2 with 8 % of Gd2O3, 3 days irradiation time and 30 days decay time are recommended

  8. A Genesis breakup and burnup analysis in off-nominal Earth return and atmospheric entry

    Science.gov (United States)

    Salama, Ahmed; Ling, Lisa; McRonald, Angus

    2005-01-01

    The Genesis project conducted a detailed breakup/burnup analysis before the Earth return to determine if any spacecraft component could survive and reach the ground intact in case of an off-nominal entry. In addition, an independent JPL team was chartered with the responsibility of analyzing several definitive breakup scenarios to verify the official project analysis. This paper presents the analysis and results of this independent team.

  9. Characterization of chemical-waste-site contamination and determination of its extent using bioassays

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, J.M.; Skalski, J.R.; Cline, J.F.; McShane, M.C.; Miller, W.E.

    1986-01-01

    The purpose of using bioassays to evaluate soils, soil elutriates, and surface and subsurface water from hazardous chemical waste sites is to provide a more-direct, integrated estimate of environmental toxicity. Based on bioassay data, chemical waste sites can be ranked according to their toxic potential or mapped for cleanup operations. The objectives of the study were to (a) assess the comparative sensitivity of test organisms to known chemicals, (b) determine if the chemical components in field soil and water samples of unknown composition could be inferred from laboratory studies using pure chemicals and (c) investigate kriging (a relatively new statistical mapping technique) of bioassay results as a method to define the areal extent of contamination. In support of these objectives, data are presented on the response of the organisms listed in the Hazardous Materials Assessment Team (HMAT) test protocol (3) to pure chemicals from three chemical subgroups (heavy metals, insecticides, and herbicides).

  10. Experimental Determination of Chemical Diffusion within Secondary Organic Aerosol Particles

    Energy Technology Data Exchange (ETDEWEB)

    Abramson, Evan H.; Imre, D.; Beranek, Josef; Wilson, Jacqueline; Zelenyuk, Alla

    2013-02-28

    Formation, properties, transformations, and temporal evolution of secondary organic aerosols (SOA) particles strongly depend on particle phase. Recent experimental evidence from a number of groups indicates that SOA is in a semi-solid phase, the viscosity of which remained unknown. We find that when SOA is made in the presence of vapors of volatile hydrophobic molecules the SOA particles absorb and trap them. Here, we illustrate that it is possible to measure the evaporation rate of these molecules that is determined by their diffusion in SOA, which is then used to calculate a reasonably accurate value for the SOA viscosity. We use pyrene as a tracer molecule and a-pinene SOA as an illustrative case. It takes ~24 hours for half the pyrene to evaporate to yield a viscosity of 10^8 Pa s for a-pinene. This viscosity is consistent with measurements of particle bounce and evaporation rates. We show that viscosity of 10^8 Pa s implies coalescence times of minutes, consistent with the findings that SOA particles are spherical. Similar measurements on aged SOA particles doped with pyrene yield a viscosity of 10^9 Pa s, indicating that hardening occurs with time, which is consistent with observed decrease in water uptake and evaporation rate with aging.

  11. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  12. Point reactivity burnup code DELIGHT-4 for high temperature, gas-cooled reactor cells

    International Nuclear Information System (INIS)

    The code DELIGHT-4 has been developed for analizing burnup characteristics of the graphite moderated reactor cells and producing the few-group constants. Calculation models for the code are as follows: (1) The number of neutron energy groups is 61 for fast neutrons (10 MeV -- 2.38 eV) and 50 for thermal neutrons (2.38 eV -- 0 eV). (2) The doubly space-heterogeneous effect of fuel (dispersion of coated fuel particles in fuel compacts and regular array of fuel rods in graphite blocks) is considered in the calculation of resonance absorption. (3) The double heterogenity of burnable poison (dispersion of absorber grains in rods) can be considered. (4) The chemical binding effect of graphite is introduced in the scattering of thermal neutrons. (5) The calculations of criticality and burnup are by a few-energy-group models (up to 10 groups for both fast and thermal neutrons), and nuclide chains of thorium-uranium and uranium-plutonium are used for burnup calculation. (6) Neutron streaming effect through holes and gaps in cells can be considered in criticality calculation. (7) The flux distribution in cells can be calculated. The cell-averaged few group constants can be produced in card form for 1-D transport approximation code SLALOM, 2-D S sub( n) code TWOTRAN, 1-D diffusion code BRIQUET, 2-D diffusion code ZADOC-3 and 3-D diffusion code CITATION-DEGA. (author)

  13. Evaluation and Selection of Boundary Isotopic Composition for Burnup Credit Criticality Safety Analysis of RBMK Spent Fuel Management

    International Nuclear Information System (INIS)

    The on-site wet-type spent fuel storage facility ISF-1 is currently used for interim storage of spent nuclear fuel removed from Chernobyl NPP power units. The results of ISF-1 preliminary criticality analyses demonstrated the need for using the burnup credit principle in nuclear safety analysis. This paper provides results from the selection and testing of computer codes for determining the isotopic composition of RBMK spent fuel. Assessment is carried out and conclusions are made on conservative approaches to fuel burnup credit in subsequent ISF-1 safety assessment. (author)

  14. Computational Chemical Imaging for Cardiovascular Pathology: Chemical Microscopic Imaging Accurately Determines Cardiac Transplant Rejection

    Science.gov (United States)

    Tiwari, Saumya; Reddy, Vijaya B.; Bhargava, Rohit; Raman, Jaishankar

    2015-01-01

    Rejection is a common problem after cardiac transplants leading to significant number of adverse events and deaths, particularly in the first year of transplantation. The gold standard to identify rejection is endomyocardial biopsy. This technique is complex, cumbersome and requires a lot of expertise in the correct interpretation of stained biopsy sections. Traditional histopathology cannot be used actively or quickly during cardiac interventions or surgery. Our objective was to develop a stain-less approach using an emerging technology, Fourier transform infrared (FT-IR) spectroscopic imaging to identify different components of cardiac tissue by their chemical and molecular basis aided by computer recognition, rather than by visual examination using optical microscopy. We studied this technique in assessment of cardiac transplant rejection to evaluate efficacy in an example of complex cardiovascular pathology. We recorded data from human cardiac transplant patients’ biopsies, used a Bayesian classification protocol and developed a visualization scheme to observe chemical differences without the need of stains or human supervision. Using receiver operating characteristic curves, we observed probabilities of detection greater than 95% for four out of five histological classes at 10% probability of false alarm at the cellular level while correctly identifying samples with the hallmarks of the immune response in all cases. The efficacy of manual examination can be significantly increased by observing the inherent biochemical changes in tissues, which enables us to achieve greater diagnostic confidence in an automated, label-free manner. We developed a computational pathology system that gives high contrast images and seems superior to traditional staining procedures. This study is a prelude to the development of real time in situ imaging systems, which can assist interventionists and surgeons actively during procedures. PMID:25932912

  15. Computational chemical imaging for cardiovascular pathology: chemical microscopic imaging accurately determines cardiac transplant rejection.

    Directory of Open Access Journals (Sweden)

    Saumya Tiwari

    Full Text Available Rejection is a common problem after cardiac transplants leading to significant number of adverse events and deaths, particularly in the first year of transplantation. The gold standard to identify rejection is endomyocardial biopsy. This technique is complex, cumbersome and requires a lot of expertise in the correct interpretation of stained biopsy sections. Traditional histopathology cannot be used actively or quickly during cardiac interventions or surgery. Our objective was to develop a stain-less approach using an emerging technology, Fourier transform infrared (FT-IR spectroscopic imaging to identify different components of cardiac tissue by their chemical and molecular basis aided by computer recognition, rather than by visual examination using optical microscopy. We studied this technique in assessment of cardiac transplant rejection to evaluate efficacy in an example of complex cardiovascular pathology. We recorded data from human cardiac transplant patients' biopsies, used a Bayesian classification protocol and developed a visualization scheme to observe chemical differences without the need of stains or human supervision. Using receiver operating characteristic curves, we observed probabilities of detection greater than 95% for four out of five histological classes at 10% probability of false alarm at the cellular level while correctly identifying samples with the hallmarks of the immune response in all cases. The efficacy of manual examination can be significantly increased by observing the inherent biochemical changes in tissues, which enables us to achieve greater diagnostic confidence in an automated, label-free manner. We developed a computational pathology system that gives high contrast images and seems superior to traditional staining procedures. This study is a prelude to the development of real time in situ imaging systems, which can assist interventionists and surgeons actively during procedures.

  16. Evaluation of Fission Product Critical Experiments and Associated Biases for Burnup Credit Validation

    International Nuclear Information System (INIS)

    One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13% k/k.

  17. Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1993-06-01

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author).

  18. Investigation of research and development subjects for the Very High Burnup Fuel

    International Nuclear Information System (INIS)

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)

  19. Reactivity loss validation of high-burnup PWR fuels with pile-oscillation experiments in MINERVE

    International Nuclear Information System (INIS)

    The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a pressurized water reactor (PWR) between five and seven cycles, and also on the experimental validation of the spent fuel reactivity loss with burnup, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and the nuclear data responsible for the reactivity loss. This program also offers unique experimental data for fuels with a burnup reaching 85 GWd/tonne, as spent fuels in French PWRs have never exceeded 70 GWd/tonne up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first step, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists of the self-shielding of cross sections on the 281-energy-group SHEM mesh, followed by flux calculation by the method of characteristics in a two-dimensional exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between experiment and calculation shows satisfactory results with the JEFF3.1.1 library, which predicts the reactivity loss within 2% for burnup of ∼75 GWd/tonne and within 4% for burnup of ∼85 GWd/tonne. (authors)

  20. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of keff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

  1. Final evaluation of the CB3+burnup credit benchmark addition

    International Nuclear Information System (INIS)

    In 1966 a series of benchmarks focused on the application of burnup credit in WWER spent fuel management system was launched by L.Markova (1). The four phases of the proposed benchmark series corresponded to the phases of the Burnup Credit Criticality Benchmark organised by the OECD/NEA.These phases referred as CB1, CB2, CB3 and CB4 benchmarks were designed to investigate the main features of burnup credit in WWER spent fuel management systems. In the CB1 step, the multiplication factor of an infinite array of spent fuel rods was calculated taking the burnup, cooling time and different group of nuclides as parameters. The fuel compositions was given in the benchmark specification (Authors)

  2. Prediction of fission gas release at high burn-up

    International Nuclear Information System (INIS)

    Reliable design of LWR fuel rods requires the fission gas release to be predicted as accurately as possible. Indeed that physical phenomenon governs both the fuel temperatures and the inner gas pressure. Fission gas release data have been reviewed by the NRC and it has been concluded that a fission gas release enhancement occurs at burn-up above 20 GWd/tM. To correct deficient fission gas release models which do not include burn-up dependence, the NRC developed an empirical correction method to describe burn-up enhancement effect. BELGONUCLEAIRE has developed its own fission gas release model which is utilized in licensing calculation through the COMETHE code. Fission gas release predictions at high burn-up are confronted to the experimental data as well as to the predictions of the NRC correlation. The physics of the fission gas release phenomenon is discussed

  3. Impact of extended burnup on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The Advisory Group Meeting was held in Vienna from 2 to 5 December 1991, to review, analyse, and discuss the effects of burnup extension in both light and heavy water reactors on all aspects of the fuel cycle. Twenty experts from thirteen countries participated in this meeting. There was consensus that both economic and environmental benefits are driving forces toward the achievement of higher burnups and that the present trend of burnup extension may be expected to continue. The extended burnup has been considered for the three main stages of the fuel cycle: the front end, in-reactor issues and the back end. Thirteen papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  4. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

  5. TRIGA criticality experiment for testing burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz [Jozef Stefan Institute, Reactor Physics Division, Ljubljana (Slovenia)

    1999-07-01

    A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)

  6. Comparisons of the predicted and measured isotopic composition for high burnup PWR spent fuels

    International Nuclear Information System (INIS)

    Comparisons between the calculated and measured isotopic composition for high burnup Korean PWR spent fuel samples were carried out. Spent fuel samples used in this study were obtained from commercial Korean PWRs, Ulchin unit 2 and Yonggwang unit 1. A radiochemical analysis of the spent fuel samples was performed to determine the isotopic compositions of U, Pu, and Nd. The depletion calculations which were carried out using the SAS2H control module in Version 5.1 of the SCALE code system were compared with the results of the radiochemical analyses. The results derived from the measured and calculated concentrations for each isotope of the corresponding samples were generally consistent with the earlier studies and the results were different within a few percent. The validity of the SAS2H control module in Version 5.1 of the SCALE code system could be confirmed in a high burnup spent fuel above 45 GWd/MTU

  7. Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

    International Nuclear Information System (INIS)

    Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanium (HPGe) gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the 137Cs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% (1σ). The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors. (authors)

  8. A guide introducing burnup credit, preliminary version. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)

  9. A burn-up module coupling to an AMPX system

    International Nuclear Information System (INIS)

    The Reactors and Neutrons Division of the Bariloche Atomic Center uses the AMPX system for the study of high conversion reactors (HCR). Such system allows to make neutronic calculations from the nuclear data library (ENDF/B-IV). The Nuclear Engineering career of the Balseiro Institute developed and implemented a burn-up module at a μ-cell level (BUM: Burn-up Module) which agrees with the requirement to be coupled to the AMPX system. (Author)

  10. A Burnup Analysis of PBMR-400MWth Reactor Core

    International Nuclear Information System (INIS)

    The purpose of this study is to analyze the burnup characteristics of 400MWth PBMR using Monte Carlo method. In the world, the deterministic method is widely used to model such that system but it still has a disadvantage which is not flexible in simulating the burnup cycle. Although this method applies some techniques to increase the accuracy of calculation results but it is necessary to model this system by a suitable computer code that can verify and validate the results of the deterministic method. A method which uses a Monte Carlo technique for simulating the burnup cycle was performed. A reactor physics computer code uses in this method is MONTEBURN 2.0 which accurately and efficiently computes the neutronic and material properties of the fuel cycle. MONTEBURN is a fully automated tool that links the MCNP Monte Carlo transport code with a radioactive decay and burnup code ORIGEN. In this model, the calculations are based on a detailed core modeling using MCNP. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and fuel kernels in the pebble. For the burnup model, a start-up core was studied with considering the movement of pebbles. By shifting down one layer at each discrete time step and inserting fresh fuel from the top, this cyclic calculation is continued until equilibrium burnup cycle is achieved. In this study, the time dependence of multiplication factor keff, the spatial dependence of flux profile, power distribution, burnup, and inventory of isotopes in the start up process are analyzed. The results will provide the basis data of the burnup process and be also utilized as the verified data to validate a compute code for PBMR core analysis which will be developed in near future

  11. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  12. A guide introducing burnup credit, preliminary version. Contract research

    International Nuclear Information System (INIS)

    It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)

  13. Application of Candle burnup to small fast reactor

    International Nuclear Information System (INIS)

    A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. An equilibrium state was obtained for a large fast reactor (core radius is 2 m and reflector thickness is 0.5 m) successfully by using a newly developed direct analysis code. However, it is difficult to apply this burnup strategy to small reactors, since its neutron leakage becomes large and neutron economy becomes worse. Fuel enrichment should be increased in order to sustain the criticality. However, higher enrichment of fresh fuel makes the CANDLE burnup difficult. We try to find some small reactor designs, which can realize the CANDLE burnup. We have successfully find a design, which is not the CANDLE burnup in the strict meaning, but satisfies qualitatively its characteristics mentioned at the top of this abstract. In the final paper, the general description of CANDLE burnup and some results on the obtained small fast reactor design are presented.(author)

  14. Research on irradiation behavior of superhigh burnup fuel

    International Nuclear Information System (INIS)

    In Japan Atomic Energy Research Institute, the special team for LWR future technology development project was organized in Tokai Research Establishment from October, 1991 to the end of fiscal year 1993. Due to the delay of the introduction of fast reactors, LWRs are expected to be used for considerably long period also in 21st century, therefore, it aimed at the further advancement of LWRs, and as one of its embodiments, the concept of superhigh burnup fuel was investigated. The superhigh burnup fuel aims at the attainment of 100 GWd/t burnup, and it succeeded the achievement of the conceptual design study on 'superlong life LWRs'. It is generally recognized that the development of the new material that substitutes for zircaloy is indispensable for superhigh burnup fuel. The concept of superhigh burnup core and the specification of fuel, the research and development of superhigh burnup fuel, the research on the irradiation behavior and irradiation damage of fuel and the damage by ion irradiation, and the method and the results of the irradiation experiment using a tandem accelerator are reported. (K.I.)

  15. Research on irradiation behavior of superhigh burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-03-01

    In Japan Atomic Energy Research Institute, the special team for LWR future technology development project was organized in Tokai Research Establishment from October, 1991 to the end of fiscal year 1993. Due to the delay of the introduction of fast reactors, LWRs are expected to be used for considerably long period also in 21st century, therefore, it aimed at the further advancement of LWRs, and as one of its embodiments, the concept of superhigh burnup fuel was investigated. The superhigh burnup fuel aims at the attainment of 100 GWd/t burnup, and it succeeded the achievement of the conceptual design study on `superlong life LWRs`. It is generally recognized that the development of the new material that substitutes for zircaloy is indispensable for superhigh burnup fuel. The concept of superhigh burnup core and the specification of fuel, the research and development of superhigh burnup fuel, the research on the irradiation behavior and irradiation damage of fuel and the damage by ion irradiation, and the method and the results of the irradiation experiment using a tandem accelerator are reported. (K.I.).

  16. Triton burnup study in JT-60U

    International Nuclear Information System (INIS)

    The behavior of 1 MeV tritons produced in the d(d,p)t reaction is important to predict the properties of D-T produced 3.5 MeV alphas because 1 MeV tritons and 3.5 MeV alphas have similar kinematic properties, such as Larmor radius and precession frequency. The confinement and slowing down of the fast tritons were investigated by measuring the 14 MeV and the 2.5 MeV neutron production rates. Here the time resolved triton burnup measurements have been performed using a new type 14 MeV neutron detector based on scintillating fibers, as part of a US-Japan tokamak collaboration. Loss of alpha particles due to toroidal ripple is one of the most important issues to be solved for a fusion reactor such as ITER. The authors investigated the toroidal ripple effect on the fast triton by analyzing the time history of the 14 MeV emission after NB turn-off

  17. Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

    Science.gov (United States)

    Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

    2013-02-01

    Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with 50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for pre-hydrided cladding.

  18. Systemization of burnup sensitivity analysis code (2) (Contract research)

    International Nuclear Information System (INIS)

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

  19. Fuel Modelling at Extended Burnup (Fumex-II). Report of a Coordinated Research Project 2002-2007

    International Nuclear Information System (INIS)

    to fuel licensing. This report describes the results of the coordinated research project on fuel modelling at extended burnup (FUMEX-II). This programme was initiated in 2000 and completed in 2006. It followed previous programmes on fuel modelling, D-COM which was conducted between 1982 and 1984, and the FUMEX programme which was conducted between 1993 and 1996. The participants used a mixture of data, derived from actual irradiation histories, in particular those with PIE measurements from high burnup commercial and experimental fuels, combined with idealized power histories intended to represent possible future extended dwell, commercial irradiations, to test code capabilities at high burnup. All participants have carried out calculations on the six priority cases selected from the 27 cases identified to them at the first research coordination meeting (RCM). At the second RCM, three further priority cases were identified and have been modelled. These priority cases have been chosen as the best available to help determine which of the many high burnup models used in the codes best reflect reality. The participants are using the remaining cases for verification and validation purposes as well as inter-code comparisons. The codes participating in the exercise have been developed for a wide variety of purposes, including predictions for fuel operation in PWR, BWR, WWER, the pressurized HWR type, CANDU and other reactor types. They are used as development tools as well as for routine licensing calculations, where code configuration is strictly controlled.

  20. Post-irradiation examination of uranium-molybdenum dispersion fuel irradiated to high burn-up in NRU

    International Nuclear Information System (INIS)

    UMo dispersion fuels are promising candidates for research and test reactors. Mini-elements containing U7Mo and U10Mo (7 and 10 wt% Mo in U alloy) fuel particles dispersed in aluminium have been fabricated with a nominal loading of 4.5 gU/cm3. In order to compare the performance of the different UMo alloys, the mini-elements were irradiated adjacent to each other under nominally identical conditions in the National Research Universal (NRU) reactor. Maximum element linear ratings up to 100 kW/m and discharge burnups up to 80 atom% 235U were achieved. The experiment was conducted in phases such that adjacent pairs of mini-elements could be removed for post-irradiation examinations (PIE) after 20, 40, 60 and 80 atom% 235U burnup. PIE included underwater inspections, visual examinations and photography in the hot cells, gamma spectroscopy, dimensional measurements, immersion density measurements, metallography, and chemical burnup analysis. The results from the high burnup fuels are presented in this paper. The assessments compare the microstructural changes, porosity formation and fuel swelling in the two UMo dispersion fuels. The results indicate that U7Mo fuel is less stable that U10 Mo fuel under the conditions tested in NRU. (author)

  1. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  2. Indirect Determination of Chemical Composition and Fuel Characteristics of Solid Waste

    DEFF Research Database (Denmark)

    Riber, Christian; Christensen, Thomas Højlund

    Determination of chemical composition of solid waste can be performed directly or indirectly by analysis of combustion products. The indirect methodology instrumented by a full scale incinerator is the only method that can conclude on elements in trace concentrations. These elements are of great...... interest in evaluating waste management options by for example LCA modeling. A methodology description of indirect determination of chemical composition and fuel properties of waste is provided and validated by examples. Indirect analysis of different waste types shows that the chemical composition is...... significantly dependent on waste type. And the analysis concludes that the transfer of substances in the incinerator is a function of waste chemical content, incinerator technology and waste physical properties. The importance of correct representation of rare items in the waste with high concentrations of...

  3. Detailed balance method for chemical potential determination in Monte Carlo and molecular dynamics simulations

    International Nuclear Information System (INIS)

    We present a new, nondestructive, method for determining chemical potentials in Monte Carlo and molecular dynamics simulations. The method estimates a value for the chemical potential such that one has a balance between fictitious successful creation and destruction trials in which the Monte Carlo method is used to determine success or failure of the creation/destruction attempts; we thus call the method a detailed balance method. The method allows one to obtain estimates of the chemical potential for a given species in any closed ensemble simulation; the closed ensemble is paired with a ''natural'' open ensemble for the purpose of obtaining creation and destruction probabilities. We present results for the Lennard-Jones system and also for an embedded atom model of liquid palladium, and compare to previous results in the literature for these two systems. We are able to obtain an accurate estimate of the chemical potential for the Lennard-Jones system at higher densities than reported in the literature

  4. Burnup instabilities in the full-core HTR model simulation

    International Nuclear Information System (INIS)

    Highlights: • We performed full-core burnup calculation coupled with Monte Carlo code. • Depletion instabilities have been detected for HTR system at high burnup. • We assess the stability of time step models in application to core calculation. • Discussion of the modeling factors related to burnup core simulation is presented. - Abstract: The phenomenon of numerical instabilities present in the Monte Carlo burnup calculations has been shown and explained by many authors using models of LWR, often simplified. Some theoretical considerations about origins of oscillations are very general, however it may be difficult to apply it easily to other models as a prediction of stability. Physics of HTR core differs significantly from the properties of light water system and the reliable extrapolation of the current numerical results is not possible. Moreover, most of the works concerning HTR burnup calculations put no emphasis on the spatial stability of the simulation and apply very long time steps. The awareness in this field of research seems to be not sufficient. In this paper, we focus on the demonstration of depletion instabilities in the simulations of HTR core dedicated for deep burnup of plutonium and minor actinides. We apply various methodology of time step implemented in advanced Continuous Energy Monte Carlo burnup code MCB version 5. Stability analysis is very rare for the full core calculations and the awareness of the oscillation’s problem is obligatory for the reliable modeling of a fuel cycle. In the summary of this work we systematize and discuss factors related to the stability of depletion and review available solutions

  5. Transnucleaire's experience with burnup credit in transport operations

    International Nuclear Information System (INIS)

    Facing a continued increase in fuel enrichment values, Transnucleaire has progressively implemented a burnup credit programme in order to maintain or, where possible, to improve the capacity of its transport packagings without physical modification. Many package design approvals, based on a notion of burnup credit, have been granted by the French competent authority for transport since the early eighties, and many of these approvals have been validated by foreign competent authorities. Up to now, these approvals are restricted to fuel assemblies made of enriched uranium and irradiated in pressurized water reactors (PWR). The characterization of the irradiated fuel and the reactivity of the package are evaluated by calculation, performed using qualified French codes developed by the CEA (Commisariat a l'Energie Atomique/French Atomic Energy Commission): CESAR as a depletion code and APOLO-MORET as a criticality code. The approvals are based on the hypothesis that the burnup considered is that applied on the least irradiated region of the fuel assemblies, the conservative approach being not to take credit for any axial profile of burnup along the fuel assembly. The most reactive configuration is calculated and the burnup credit is also restricted to major actinides only. On the operational side and in compliance with regulatory requirements, verification is made before transport, in order to meet safety objectives as required by the transport regulations. Besides a review of documentation related to the irradiation history of each fuel assembly, it consists of either a qualitative (go/no-go) verification or of a quantitative measurement, depending on the level of burnup credit. Thus the use of burnup credit is now a common practice with Transnucleaire's packages, particularly in France and Germany. New improvements are still in progress and qualifications of the calculation code are now well advanced, which will allow in the near future the use of six selected

  6. Dissolution of low burnup Fast Flux Test reactor fuel

    International Nuclear Information System (INIS)

    The first Fast-Flux Test Facility reactor fuel [mixed (U,Pu)O2 composition] has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 19970C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 950C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 290C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 950C dissolution contained the equivalent of 198 mg of 239Pu per 100 g of hulls, while the cladding from the 290c experiments contained only 0.21 mg of 239Pu per 100 g of hulls. 9 references, 5 figures

  7. Determination of Chemical Characteristics of Saffron in Different Area of Iran

    OpenAIRE

    Ahmad Kalbasi; Farahnaz MotamediSedeh; Hamid Tavakolipour; Saeed Rajabifar; Naimeh Khazaei; Mohammad Jouki

    2012-01-01

    In this research, saffron samples collected from 11 regions of Khorasan-Iran and chemical characteristics of them such as color, flavor and aroma were studied. Chemical characteristics of saffron (Crocus sativus L.) were determined by spectrophotometric device Using 255, 325 and 440 nm wavelength for three components, picrocrocin, safranal and croicn which are responsible for flavor, aroma and color parameters respectively. Spectrophotometric analysis showed that maximum absorption were 1/928...

  8. The SMOPY system: Quantitative burn-up measurement monitor combining gamma spectrometry and neutron measurement for safeguards applications

    International Nuclear Information System (INIS)

    Full text: IAEA uses today FORK Detector in attended and unattended mode for the verification of spent fuels. This system uses a neutron fission chamber and an ionisation chamber to combine total neutron counting and total gamma counting. CANBERRA now proposes the new SMOPY system, which enhances performance as it combines a fission chamber and a CZT gamma spectrometer. This new measurement capability associated with a depletion code embedded in the interpretation software of the system allows a complete identification of the burn-up of any type of fuel (also MOX for example). A first prototype of the SMOPY system was developed in collaboration between AREVA NC CEA and CANBERRA for safeguards but also burn-up credit applications. This prototype has been already used by the IAEA. CANBERRA has now completed the industrialization of this system adding new functionalities. The system allows also axial scanning of the fuel assembly instead of a single point measurement. Two types of interpretation of the measurement have been developed. The first one requires the irradiation history to determine very precisely the burn-up of the fuel assembly, thus allow to verify the operator's declaration. The second method is less precise but doesn't require any data of the fuel to determine the cooling time, burn-up, and the fuel type (MOX or LEU). The new CANBERRA industrialized SMOPY system will allow new possibilities of IAEA verifications and will also permit to address new scenarios of IAEA safeguards activities. (author)

  9. The impact of burn-up credit in criticality studies

    International Nuclear Information System (INIS)

    Nowadays optimization goes with everything. So French engineering firms try to demonstrate that fuel transport casks and storage pools are able to receive assemblies with higher 235U initial enrichments. Fuel Burnup distribution contributes to demonstrate it. This instruction has to elaborate a way to take credit of burnup effects on criticality safety designs. The calculation codes used are CESAR 4.21-APOLLO 1-MORET III. The assembly studied (UO2) is irradiated in a French Pressurized Water Reactor like EDF nuclear power reactor: PWR 1300 MWe, 17 x 17 array. Its initial enrichment in 235U equals 4.5%. The studies exposed in this report have evaluated the effects of: i) the 15 fission products considered in Burnup Credit (95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 133Cs, 143Nd, 145Nd, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 153Eu, 155Gd), ii) the calculated abundances corrected or not by fixed factors, iii) the choice of one cross sections library used by CESAR 4.21, iu) the zone number elected in the axial burnup distribution zoning, u) the kind of cut applied on (regular/optimized). Two axial distribution profiles are studied: one with 44 GWd/t average burnup, the other with 20 GWd/t average burnup. The second one considers a shallow control rods insertion in the upper limit of the assembly. The results show a margin in reactivity about 0.045 with consideration of the 6 most absorbent fission products (103Rh, 133Cs, 143Nd, 149Sm, 152Sm, 155Gd), and about 0.06 for all Burnup Credit fission products whole. Those results have been calculated with an average burnup of 44 GWj/t. In a conservative approach, corrective factors must be apply on the abundance of some fission products. The cross sections library used by CESAR 4.21 (BBL 4) is sufficient and gives satisfactory results. The zoning of the assembly axial distribution burnup in 9 regular zones grants a satisfying calculation time/result precision compromise. (author)

  10. Applications of ''candle'' burn-up strategy to several reactors

    International Nuclear Information System (INIS)

    The new burn-up strategy CANDLE is proposed, and the calculation procedure for its equilibrium state is presented. Using this strategy, the power shape does not change as time passes, and the excess reactivity and reactivity coefficient are constant during burn-up. No control mechanism for the burn-up reactivity is required, and power control is very easy. The reactor lifetime can be prolonged by elongating the core height. This burn-up strategy can be applied to several kinds of reactors whose maximum neutron multiplication factor changes from less than unity to more than unity, and then to less than unity. In the present paper it is applied to some fast reactors, thus requiring some fissile material such as plutonium for the nuclear ignition region of the core, but only natural uranium is required for the other region of the initial reactor and for succeeding reactors. The drift speed of the burning region for this reactor is about 4 cm/year, which is a preferable value for designing a long-life reactor. The average burn-up of the spent fuel is about 40%; that is, equivalent to 40% utilisation of the natural uranium without the reprocessing and enrichment. (author)

  11. Triton burnup measurements by neutron activation at JT-60U

    International Nuclear Information System (INIS)

    This paper describes measurements on triton burnup in a deuterium plasma by the detection of the 2.5 MeV neutrons (from DD fusion) and the 14 MeV neutrons (from DT fusion). The 2.5 MeV neutrons have been measured by fission chambers and activation of indium foils while the 14 MeV neutrons have been detected by activation of silicon, aluminum, and copper foils. The measured yields of the 2.5 MeV neutrons utilizing In foils are similar 20-40% higher than the yields obtained from fission chambers depending on what calibration factors are used. The deviation decreases with the plasma major radius (or increasing plasma volume). When the triton burnup is measured by utilizing neutron threshold reactions (En>2.5 MeV) and In foils, then systematic errors in the calibration factors cancel and the maximum deviation between the measured triton burnup for different calibration factors is reduced to similar 5%. The measurements indicate that triton burnup increases with the 14 MeV neutron yield, indicating that the relative yield of 14 MeV neutrons increases depending on the time duration of the deuterium neutral beam injection (NBI). Furthermore, the triton burnup decreases with an increased plasma major radius, indicating increased triton ripple losses, and increases with plasma current, indicating reduced banana orbit losses. (orig.)

  12. Disposal criticality analysis methodology's principal isotope burnup credit

    International Nuclear Information System (INIS)

    This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)

  13. The commercial and technological impact of high burnup

    International Nuclear Information System (INIS)

    Deregulation of electricity markets is driving prices downward. Consequently utilities continue to demand the minimization of electrical production costs. Fuel cycle cost savings are valued as a strong contributor, although directly representing only about one third of electricity generating costs. Burnups consistent with the current enrichment limit of 5 w/0 will be required. Significant progress has already been achieved by Siemens in meeting the demands of utilities for increased fuel burnup. The technological challenges imposed are mainly related to corrosion and hydrogen pickup of the clad, the properties of the fuel and the dimensional changes of the structure. Clad materials with increased corrosion resistance have been developed. The high burnup behaviour of the fuel has been extensively investigated and the decrease of thermal conductivity, the rim effect and the increase of fission gas release can be described, with good accuracy, in fuel rod computer codes. Advanced statistical design methods have been developed and introduced. In summary, most of the questions about the fuel operational behaviour and reliability in the high burnup range have been solved or the solutions are visible. The main licensing challenges for high burnup fuel are currently seen for accident condition analyses, especially for RIA and LOCA. (author)

  14. Burnup credit in nuclear waste transport: An economic analysis

    International Nuclear Information System (INIS)

    The US DOE is responsible for transporting nuclear spent fuel from commercial reactors to monitored retrievable storage (MRS) facilities and/or to repositories. Current plans call for approximately 110,000 metric tons uranium (MTU) to be transported over approximately 40 years beginning in 1998. Because of the large volume of spent fuel to be transported, new generations of spent fuel transportation casks are being planned. These casks will embody the latest technology and will be designated to accommodate the spent fuel in a way that maximizes the overall efficiency of the cask. In planning for the new generation of transport casks, the DOE is investigating the possibility of tailoring the cask design for the extent to which spent fuel has been used in the reactors, or, for spent fuel burnup. Granting design credit for burnup would allow one to fabricate casks with relatively larger capacities than would be possible otherwise. The remainder of the paper discusses the economic implications of using burnup credit in cask design, discusses the approach used in analyzing the economics of burnup credit, describes the results of the analysis, and offers some conclusions about the economic value of the burnup credit option

  15. Burnup monitoring of VVER-440 spent fuel assemblies

    International Nuclear Information System (INIS)

    This paper reports on the results of the experiments performed on spent VVER-440 fuel assemblies at the Paks Nuclear Power Plant (NPP), Hungary. The fuel assemblies submerged in the service pit were examined by high-resolution gamma spectrometry (HRGS). The assemblies were moved to the front of a collimator tube built in the concrete wall of the pit in the reactor block at the NPP, and lifted down and up under water for scanning by the refueling machine. The HPGe detector was placed behind the collimator in an outside staircase. The measurements involved scanning of the assemblies along their length of all the 6 sides, at 5-12 measurement positions side by side. Axial and azimuthal burnup profiles were taken in this way. Assembly groups for measurements were selected according to their burnup (10–50 GWd/tU) and special positions (e. g. control assembly, neighbour of control assembly). Burnup differences were well observable between assembly sides looking towards the center of the core and opposite directions. Also, burnup profiles were different for control assemblies and normal (working) fuel assemblies. The ratio of the measured activities of Cs-134 and Cs-137 was evaluated by relative efficiency (intrinsic) calibration. Measurement uncertainty is around 3 %. Taking into account irradiation history and cooling time (i. e.the time elapsed since the discharge of the assembly out of the core), the activity ratio Cs-134/Cs-137 shows good correlation with the declared burnup.

  16. Fuel cycle economical improvement by reaching high fuel burnup

    International Nuclear Information System (INIS)

    Improvements of fuel utilization in the light water reactors, burnup increase have led to a necessity to revise strategic approaches of the fuel cycle development. Different trends of the fuel cycle development are necessary to consider in accordance with the type of reactors used, the uranium market and other features that correspond to the nuclear and economic aspects of the fuel cycle. The fuel burnup step-by-step extension Program that successfully are being realized by the leading, firms - fuel manufacturers and the research centres allow to say that there are no serious technical obstacles for licensing in the near future of water cooling reactors fuel rod burnup (average) limit to 65-70 MWd/kgU and fuel assembly (average) limit to (60-65) MWd/kgU. The operating experience of Ukrainian NPPs with WWER-1000 is 130 reactor * years. At the beginning of 1999, a total quantity of the fuel FA discharged during all time of operation of 11 reactors was 5819 (110 fuel cycles). Economical improvement is reached by increase of fuel burn-up by using of some FA of 3 fuel cycles design in 4th fuel loading cycle. Fuel reliability is satisfactory. The further improvement of FA is necessary, that will allow to reduce the front-end fuel cycle cost (specific natural uranium expenditure), to reduce spent fuel amount and, respectively, the fuel cycle back end costs, and to increase burn-up of the fuel. (author)

  17. Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

    International Nuclear Information System (INIS)

    For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B4C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability

  18. Fuel cycle cost considerations of increased discharge burnups

    International Nuclear Information System (INIS)

    Evaluations are presented that indicate the attainment of increased discharge burnups in light water reactors will depend on economic factors particular to individual operators. In addition to pure resource conserving effects and assuming continued reliable fuel performance, a substantial economic incentive must exist to justify the longer operating times necessary to achieve higher burnups. Whether such incentive will exist or not will depend on relative price levels of all fuel cycle cost components, utility operating practices, and resolution of uncertainties associated with the back-end of the fuel cycle. It is concluded that implementation of increased burnups will continue at a graduated pace similar to past experience, rather than finding universal acceptance of particular increased levels at any particular time

  19. Mechanical Property Evaluation of High Burnup PHWR Fuel Clads

    International Nuclear Information System (INIS)

    Assurance of clad integrity is of vital importance for the safe and reliable extension of fuel burnup. In order to study the effect of extended burnup of 15,000 MW∙d/tU on the performance of Pressurised Heavy Water Reactor (PHWR) fuel bundles of 19-element design, a couple of bundles were irradiated in Indian PHWR. The tensile property of irradiated cladding from one such bundle was evaluated using the ring tension test method. Using a similar method, claddings of mixed oxide (MOX) fuel elements irradiated in the pressurized water loop (PWL) of CIRUS to a burnup of 10,000 MW∙d/THM were tested. The tests were carried out both at ambient temperature and at 300°C. The paper will describe the test procedure, results generated and discuss the findings. (author)

  20. Study on the conservative factors for burnup credit criticality calculation

    International Nuclear Information System (INIS)

    When applies the burnup credit technology to perform criticality safety analysis for spent fuel storage or transportation problems, it is important for one to confirm that all the conditions adopted are adequate to cover the severest conditions that may encounter in the engineering applications. Taking the OECD/NEA burnup credit criticality benchmarks as sample problems, we study the effect of some important factors that may affect the conservatism of' the results for spent fuel system criticality safety analysis. Effects caused by different nuclides credit strategy, different cooling time and axial burnup profile are studied by use of the STARBUCS module of SCALE5. 1 software package, and related conclusions about the conservatism of these factors are drawn. (authors)

  1. High-burnup fuel and the impact on fuel management

    International Nuclear Information System (INIS)

    Competition in the electric utility industry has forced utilities to reduce cost. For a nuclear utility, this means a reduction of both the nuclear fuel cost and the operating and maintenance cost. To this extent, utilities are pursuing longer cycles. To reduce the nuclear fuel cost, utilities are trying to reduce batch size while increasing cycle length. Yankee Atomic Electric Company has performed a number of fuel cycle studies to optimize both batch size and cycle length; however, certain burnup-related constraints are encountered. As a result of these circumstances, longer fuel cycles make it increasingly difficult to simultaneously meet the burnup-related fuel design constraints and the technical specification limits. Longer cycles require fuel assemblies to operate for longer times at relatively high power. If utilities continue to pursue longer cycles to help reduce nuclear fuel cost, changes may need to be made to existing fuel burnup limits

  2. Validation issues for depletion and criticality analysis in burnup credit

    International Nuclear Information System (INIS)

    This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage, and disposal. The issues discussed are ones that have been identified by one or more constituents of the United States technical community (national laboratories, licensees, and regulators) that have been exploring the use of burnup credit. There is not necessarily agreement on the importance of the various issues, which sometimes is what creates the issue. The broad issues relate to the paucity of available experimental data (radiochemical assays and critical experiments) covering the full range and characteristics of spent nuclear fuel in away-from-reactor systems. The paper will also introduce recent efforts initiated at Oak Ridge National Laboratory (ORNL) to provide technical information that can help better assess the value of different experiments. The focus of the paper is on experience with validation issues related to use of burnup credit for transport and dry storage applications. (author)

  3. Burnup credit methodology validation against WWER experimental data

    International Nuclear Information System (INIS)

    A methodology for criticality safety analyses with burnup credit application has been developed for WWER spent fuel management facilities. This methodology is based on two worldwide used code systems: SCALE 4.4 for depletion and criticality calculations and NESSEL-NUKO - for depletion calculations. The methodology is in process of extensive validation for WWER applications. The depletion code systems NESSEL-NUKO and SCALE4.4 (control module SAS2H) have been validated on the basis of comparison with the calculated results obtained by other depletion codes for the CB2 Calculational Burnup Credit Benchmark. The validation of these code systems for WWER-440 and WWER-1000 spent fuel assembly depletion analysis based on comparisons with appropriate experimental data commenced last year. In this paper some results from burnup methodology validation against measured nuclide concentration given in the ISTC project 2670 for WWER-440 and from ORNL publication for WWER-1000 are presented. (authors)

  4. Metal Oxide Nanoparticles: The Importance of Size, Shape, Chemical Composition, and Valence State in Determining Toxicity

    Science.gov (United States)

    Dunnick, Katherine

    Nanoparticles, which are defined as a structure with at least one dimension between 1 and 100 nm, have the potential to be used in a variety of consumer products due to their improved functionality compared to similar particles of larger size. Their small size is associated with increased strength, improved catalytic properties, and increased reactivity; however, their size is also associated with increased toxicity in vitro and in vivo. Numerous toxicological studies have been conducted to determine the properties of nanomaterials that increase their toxicity in order to manufacture new nanomaterials with decreased toxicity. Data indicates that size, shape, chemical composition, and valence state of nanomaterials can dramatically alter their toxicity profile. Therefore, the purpose of this dissertation was to determine how altering the shape, size, and chemical composition of various metal oxide nanoparticles would affect their toxicity. Metal oxides are used in variety of consumer products, from spray-sun screens, to food coloring agents; thus, understanding the toxicity of metal oxides and determining which aspects affect their toxicity may provide safe alternatives nanomaterials for continued use in manufacturing. Tungstate nanoparticles toxicity was assessed in an in vitro model using RAW 264.7 cells. The size, shape, and chemical composition of these nanomaterials were altered and the effect on reactive oxygen species and general cytotoxicity was determined using a variety of techniques. Results demonstrate that shape was important in reactive oxygen species production as wires were able to induce significant reactive oxygen species compared to spheres. Shape, size, and chemical composition did not have much effect on the overall toxicity of these nanoparticles in RAW 264.7 cells over a 72 hour time course, implicating that the base material of the nanoparticles was not toxic in these cells. To further assess how chemical composition can affect toxicity

  5. Criticality safety evaluation for the direct disposal of used nuclear fuel. Preparation of data for burnup credit evaluation (Contract research)

    International Nuclear Information System (INIS)

    In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. In the conventional criticality safety evaluation of UNF where the fresh fuel composition is conservatively assumed, neutron multiplication factor is becoming overestimated as the fuel enrichment increases. The recent development of higher-enrichment fuel has therefore enhanced the benefit of the application of burnup credit. When applying the burnup credit to the criticality safety analysis of the disposed fuel system, the safe-side estimation of the reactivity is required taking into account the factors which affect the neutron multiplication factor of the burnt fuel system such as the nuclide composition uncertainties. In this report, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study. (author)

  6. An empirical formulation to describe the evolution of the high burnup structure

    International Nuclear Information System (INIS)

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  7. An empirical formulation to describe the evolution of the high burnup structure

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-15

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  8. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  9. Economics of VVER Fuel Cycles Leading to High Discharge Burnup

    International Nuclear Information System (INIS)

    Economic characteristics of equilibrium VVER fuel cycles leading to high discharge burnup are investigated by supposing two scenarios named optimistic and pessimistic. The optimistic and pessimistic terms are used in the sense whether the high burnup fuel cycles are economically advantageous or the increasing enrichment cost can increase the specific fuel cycle cost above a certain discharge burnup value. Therefore in case of the optimistic scenario, maximum fabrication and back end costs and minimum enrichment and raw uranium costs were applied, while in case of the pessimistic scenario vice-versa. The applied costs are detailed in Table 1. Table1 Cost data of the two different scenarios. Concerning the transport and storage during the front end fuel cycle, it was assumed that application of burnable poison solves the criticality problems caused by the increased enrichment. By using the advantage of the burnup credit, the subcriticality of the spent fuel storage and transport devices can also be proved. Large reserve in the biological shielding is supposed. According to the above argumentation, fixed cost of the front and back end fuel cycle was used in the calculations, except the enrichment, but a 700 $/pin extra fabrication cost of the burnable poison was taken into account. Instead of fixed batch fraction, fixed cycle length was assumed which is advantageous for maximizing the discharge burnup and for minimizing the burnable poison extra cost but disadvantageous concerning the availability factor, which is constant in the given calculations. Beside the economic characteristics, the feasibility of the cycles are investigated from the point of view of the most important safety related parameters like reactivity coefficients and shut down margin. The figure below shows the burnup dependent fuel cycle cost for the above two scenarios. (author)

  10. Burnup credit implementation plan and preparation work at JAERI

    International Nuclear Information System (INIS)

    Application of the burnup credit concept is considered to be very effective to the design of spent fuel transport and storage facilities. This technology is all the more important when considering construction of the intermediate spent fuel storage facility, which is to be commissioned by 2010 due to increasing amount of accumulated spent fuel in Japan. Until reprocessing and recycling all the spent fuel arising, they will be stored as an energy stockpile until such time as they can be reprocessed. On the other hand, the burnup credit has been partly taken into account for the spent fuel management at Rokkasho Reprocessing Plant, which is to be commissioned in 2005. They have just finished the calibration tests for their burnup monitor with initially accepted several spent fuel assemblies. Because this monitoring system is employed with highly conservative safety margin, it is considered necessary to develop the more rational and simplified method to confirm burnup of spent fuel. A research program has been instituted to improve the present method employed at the spent fuel management system for the Spent Fuel Receiving and Storage Pool of Rokkasho Reprocessing Plant. This program is jointly performed by Japan Nuclear Fuel Limited (JNFL) and JAERI.This presentation describes the current status of spent fuel accumulation discharged from PWR and BWR in Japan and the recent incentive to introduce burnup credit into design of spent fuel storage and transport facilities. This also includes the content of the joint research program initiated by JNFL and JAERI. The relevant study has been continued at JAERI. The results by these research programs will be included in the Burnup Credit Guide Original Version compiled by JAERI. (author)

  11. Determination of the Amino Acid and Chemical Composition of Canned Smoked Mussels (Mytilus galloprovincialis, L.)

    OpenAIRE

    Şengör, Gülgün F.; Gün, Hüseyin; Kalafatoğlu, Hanife

    2008-01-01

    In this research smoking and canning techniques were applied to cultured mussels (Mytilus galloprovincialis, L.) from the Çanakkkale Strait in Eceabat, Turkey. Mussels that were smoked by liquid and traditional methods were canned in different sauces. The chemical composition and amino acid composition of the canned smoked mussels were determined by the results of laboratory analyses. As a result of smoking and canning mussels, a food with high nutritional value was obtained. It was determine...

  12. WO3/W Nanopores Sensor for Chemical Oxygen Demand (COD) Determination under Visible Light

    OpenAIRE

    Xuejin Li; Jing Bai; Qiang Liu; Jianyong Li; Baoxue Zhou

    2014-01-01

    A sensor of a WO3 nanopores electrode combined with a thin layer reactor was proposed to develop a Chemical Oxygen Demand (COD) determination method and solve the problem that the COD values are inaccurately determined by the standard method. The visible spectrum, e.g., 420 nm, could be used as light source in the sensor we developed, which represents a breakthrough by limiting of UV light source in the photoelectrocatalysis process. The operation conditions were optimized in this work, and t...

  13. Isothermal Chemical Denaturation to Determine Binding Affinity of Small Molecules to G-Protein Coupled Receptors

    OpenAIRE

    Ross, Patrick; Weihofen, Wilhelm; Siu, Fai; Xie, Amy; Katakia, Hetal; Wright, S. Kirk; Hunt, Ian; Brown, Richard K; Freire, Ernesto

    2014-01-01

    The determination of accurate binding affinities is critical in drug discovery and development. Several techniques are available for characterizing the binding of small molecules to soluble proteins. The situation is different for integral membrane proteins. Isothermal chemical denaturation (ICD) has been shown to be a valuable biophysical method to determine in a direct and label-free fashion the binding of ligands to soluble proteins. In this communication, the application of isothermal che...

  14. The determination and interpretation of chemical abundances from HII region spectra in galaxies

    International Nuclear Information System (INIS)

    An overview is given of the determination of element abundances from HII region emission lines in external galaxies. The variation of abundances - particularly O, S and N - with type of, and position in a galaxy is discussed. Some aspects of chemical evolution which may have led to these variations are investigated, introducing a ''throughflow'' model to show some effects of gas flow. (author)

  15. Chemical modifiers in electrothermal atomic absorption determination of Platinum and Palladium containing preparations in blood serum

    Directory of Open Access Journals (Sweden)

    Аntonina Alemasova

    2012-11-01

    Full Text Available The biological liquids matrixes influence on the characteristic masses and repeatability of Pt and Pd electrothermal atomic absorption spectroscopy (ETAAS determination was studied. The chemical modifiers dimethylglyoxime and ascorbic acid for matrix interferences elimination and ETAAS results repeatability improvement were proposed while bioliquids ETAAS analysis, and their action mechanism was discussed.

  16. DETERMINATION OF REGIMES FOR DIPHTHERIA EXOTOXIN MODIFICATION BY CHEMICAL AND PHYSICOCHEMICAL METHODS

    OpenAIRE

    Antusheva T.I.; Pluhator T.M.; Ryabovol O.V.; Sklyar N.I.,; Ryzhkova T.A.,; Kalinichenko S.V; Babych E.M.; Panova C.V.

    2011-01-01

    The possibility of diphtheria toxoid obtaining using chemical (amino sugars, organic acids) and physicochemical (amino sugars, organic acids, ultrasound, temperature) factors was studied. It was established that modifiers (including formaldehyde) volume content decreasing didn’t have significant influence on diphtheria toxin derived modifications specific activity. It was experimentally determined that diphtheria toxin modifications obtained by the instrumentality of modifier number 1 with or...

  17. Development of high burnup fuel data-base

    International Nuclear Information System (INIS)

    Development of high burnup fuel data base (HBDB) was studied, which stores various performance data of high burnup fuels using a personal computer. Data items of the data base and storing and display methods of time-depending data such as power history were studied. It was shown that compound systems of a personal computer and an engineering work station have capacity for constructing the data base with much efficiency and small cost. And comparison of data items between the data base and the EPRI fuel base FPDB was discussed. (author)

  18. WWER fuel behaviour and characteristics at high burnup

    International Nuclear Information System (INIS)

    The increase of fuel burnup in fuel rods is a task that provides a considerable cost reduction of WWER fuel cycle in case of its solution. Investigations on fuel and cladding behaviour and change in fuel characteristics under irradiation are carried out in the Russian Federation for standard and as well as for experimental fuel rods to validate the reliable and safe operation of the fuel rods at high burnups. The paper presents the results of examinations on cracking, dimensional, structural and density changes of fuel pellets as well as the results of examination on corrosion and mechanical properties of WWER-440 and WWER-1000 fuel rod claddings. (author)

  19. CB2 result evaluation (VVER-440 burnup credit benchmark)

    International Nuclear Information System (INIS)

    The second portion of the four-piece international calculational benchmark on the VVER burnup credit (CB2) prepared in the collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmarks Working Group and proposed to the AER research community has been evaluated. The evaluated results of calculations performed by analysts from Cuba, the Czech Republic, Finland, Germany, Russia, Slovakia and the United Kingdom are presented. The goal of this study is to compare isotopic concentrations calculated by the participants using various codes and libraries for depletion of the VVER-440 fuel pin cell. No measured values were available for the comparison. (author)

  20. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

    The examinations carried out on the FRAGEMA fuel of EDF reactors show its good behavior in service. The results of research and development programs developed by EDF, FGA and the CEA show that this fuel can be irradiated up to a high burnup, and allow to point out the axies of research to improve still the performance of the product in a more and more soliciting environment (increase of power and burnup coupled with load following). Among the solutions considered, there are the design and fabrication adjustments (geometry, initial pressurization), more fundamental changes concerning fuel cans and fuel pellets, which need still research and development programs

  1. Perturbation and sensitivity theory for reactor burnup analysis

    International Nuclear Information System (INIS)

    Perturbation theory is developed for the nonlinear burnup equations describing the time-dependent behavior of the neutron and nuclide fields in a reactor core. General aspects of adjoint equations for nonlinear systems are first discussed and then various approximations to the burnup equations are rigorously derived and their areas for application presented. In particular, the concept of coupled neutron/nuclide fields (in which perturbations in either the neutron or nuclide field are allowed to influence the behavior of the other field) is contrasted to the uncoupled approximation

  2. Performance of fast reactor irradiated fueled emitters at goal burnup

    International Nuclear Information System (INIS)

    UO2-fueled W emitters were examined that had been irradiated to goal burnups of approximately 4 at.% at emitter surface temperatures to 1820 K in a fast reactor to establish their performance for use in thermionic reactors with power levels from tens of kilowatts to multimegawatts. The examinations provided first-time data on structural integrity, dimensional stability, component compatibility, and fuel and fission product behavior. The data are consistent with similar measurements at approximately 2 at.% burnup with the exception of one emitter which breached the W during irradiation

  3. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  4. Analysis of high burnup spent nuclear fuel by ICP-MS

    International Nuclear Information System (INIS)

    Inductively coupled plasma mass spectrometry (ICP-MS) as the primary tool for determining concentrations of a suite of nuclides in samples excised from high-burnup spent nuclear fuel rods taken from light water nuclear reactors. The complete analysis included the determination of 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 137Cs, 143Nd, 145Nd, 148Nd, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 151Eu, 153Eu, 155Eu, 155Gd, 237Np, 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 242mAm, and 243Am. The isotopic composition of fissiogenic lanthanide elements was determined using high-performance liquid chromatography (HPLC) with ICP-MS detection. These analytical results allow the determination of fuel burn-up based on 148Nd, Pu, and U content, as well as provide input for storage and disposal criticality calculations. Results show that ICP-MS along with HPLC-ICP-MS are suitable of performing routine determinations of most of these nuclides, with an uncertainty of ±10% at the 95% confidence level. (author)

  5. Chiral Random Matrix Model at Finite Chemical Potential: Characteristic Determinant and Edge Universality

    CERN Document Server

    Liu, Yizhuang; Zahed, Ismail

    2016-01-01

    We derive an exact formula for the stochastic evolution of the characteristic determinant of a class of deformed Wishart matrices following from a chiral random matrix model of QCD at finite chemical potential. In the WKB approximation, the characteristic determinant describes a sharp droplet of eigenvalues that deforms and expands at large stochastic times. Beyond the WKB limit, the edges of the droplet are fuzzy and described by universal edge functions. At the chiral point, the characteristic determinant in the microscopic limit is universal. Remarkably, the physical chiral condensate at finite chemical potential may be extracted from current and quenched lattice Dirac spectra using the universal edge scaling laws, without having to solve the QCD sign problem.

  6. Chiral random matrix model at finite chemical potential: Characteristic determinant and edge universality

    Science.gov (United States)

    Liu, Yizhuang; Nowak, Maciej A.; Zahed, Ismail

    2016-08-01

    We derive an exact formula for the stochastic evolution of the characteristic determinant of a class of deformed Wishart matrices following from a chiral random matrix model of QCD at finite chemical potential. In the WKB approximation, the characteristic determinant describes a sharp droplet of eigenvalues that deforms and expands at large stochastic times. Beyond the WKB limit, the edges of the droplet are fuzzy and described by universal edge functions. At the chiral point, the characteristic determinant in the microscopic limit is universal. Remarkably, the physical chiral condensate at finite chemical potential may be extracted from current and quenched lattice Dirac spectra using the universal edge scaling laws, without having to solve the QCD sign problem.

  7. EDXRF for determination of chemical elements in the beetle Alphitobius diaperinus

    International Nuclear Information System (INIS)

    Energy Dispersion X-Ray Fluorescence (EDXRF) spectrometry has been widely employed for chemical element determination of biological matrices, including insects. The beetle Alphitobius diaperinus is a major problem in poultry production, thereby infesting poultry litter and stored grains. Up to now, little is known about the behavior, physiology and environmental interactions of this insect. In this paper, EDXRF was applied to quantify the main chemical elements in A. diaperinus. For the quality of the analytical protocol, certified reference materials produced by National Institute of Standards and Technology - NIST were analyzed together with the samples. The technique was able to quantify Cl, P, S and Zn in this insect, presenting no significant variation at the 95% confidence level among the repetitions (n = 4). A different pattern of chemical element accumulation in this beetle was noticed compared to other Coleoptera species, in which the concentration of the chemical elements were markedly lower in A. diaperinus, probably associated to the restricted availability of chemical elements in food. Since no result has been found in the literature before, A. diaperinus was firstly chemically characterized in this paper. (author)

  8. EDXRF for determination of chemical elements in the beetle Alphitobius diaperinus

    Energy Technology Data Exchange (ETDEWEB)

    Cantinha, Rebeca S.; Farias, Emerson E.G. de; Magalhaes, Marcelo L.R. de; Franca, Elvis J. de, E-mail: rebecanuclear@gmail.com, E-mail: emersonemiliano@yahoo.com.br, E-mail: marcelo_rlm@hotmail.com, E-mail: ejfranca@cnen.gov.br [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Cunha, Franklin M. da; Zacarias, Vyvyane L., E-mail: ukento@yahoo.com.br, E-mail: vyvyanebiologicas@gmail.com [Universidade Federal Rural de Pernambuco (UFRPE), Recife, PE (Brazil)

    2015-07-01

    Energy Dispersion X-Ray Fluorescence (EDXRF) spectrometry has been widely employed for chemical element determination of biological matrices, including insects. The beetle Alphitobius diaperinus is a major problem in poultry production, thereby infesting poultry litter and stored grains. Up to now, little is known about the behavior, physiology and environmental interactions of this insect. In this paper, EDXRF was applied to quantify the main chemical elements in A. diaperinus. For the quality of the analytical protocol, certified reference materials produced by National Institute of Standards and Technology - NIST were analyzed together with the samples. The technique was able to quantify Cl, P, S and Zn in this insect, presenting no significant variation at the 95% confidence level among the repetitions (n = 4). A different pattern of chemical element accumulation in this beetle was noticed compared to other Coleoptera species, in which the concentration of the chemical elements were markedly lower in A. diaperinus, probably associated to the restricted availability of chemical elements in food. Since no result has been found in the literature before, A. diaperinus was firstly chemically characterized in this paper. (author)

  9. Determination of Chemical Characteristics of Saffron in Different Area of Iran

    Directory of Open Access Journals (Sweden)

    Ahmad Kalbasi

    2012-01-01

    Full Text Available In this research, saffron samples collected from 11 regions of Khorasan-Iran and chemical characteristics of them such as color, flavor and aroma were studied. Chemical characteristics of saffron (Crocus sativus L. were determined by spectrophotometric device Using 255, 325 and 440 nm wavelength for three components, picrocrocin, safranal and croicn which are responsible for flavor, aroma and color parameters respectively. Spectrophotometric analysis showed that maximum absorption were 1/928 and 2/760 for pricrocrocin and crocin respectively for samples which are collected in TorbateHeydariyeh county and maximum absorption for safranal was 1/008 for samples which are collected in sheshtamad.

  10. Using ORIGEN and MCNP to calculate reactor criticals and burnup effects

    International Nuclear Information System (INIS)

    The purpose of this modeling effort was to verify the applicability of using ORIGEN-S and MCNP to the analysis of spent fuel of various enrichments and burnups. By comparing the results of criticality studies using MCNP and ORIGEN-S with the measured keff of 1.0, the suitability of the coupled ORIGEN-S/ MCNP package was determined. This study presents the results of the benchmark modeling of five pressurized water reactor (PWR) critical configurations. For these analyses, a combination of ORIGEN-S and MCNP was used to analyze the fuel depletion and criticality of five power reactor core configuration

  11. Validation of IRBURN calculation code system through burnup benchmark analysis

    International Nuclear Information System (INIS)

    Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.

  12. PWR fuel performance and burnup extension programme in Japan

    International Nuclear Information System (INIS)

    Since the first PWR nuclear power plant Mihama Unit 1 initiated commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts on improving the technology of PWRs. The results can already be seen by the significantly improved performance of the PWR plants now in operation. Mitsubishi Heavy Industries, Ltd supplied the nuclear fuel assemblies, which now amount to almost 5000. Although some trouble with fuel was experienced in the beginning, the progressive efforts made to improve the fuel design and manufacturing technology have resulted in the superior performance of Mitsubishi fuels. Since fuel of current design should comply with the limitation set in Japan for a maximum discharged fuel assembly average burnup of less than 39,000 MW·d/t, the maximum burnup is now around 37,000 MW·d/t. However, an increase in this burnup limitation has been strongly requested by Japanese utilities in order to make nuclear power more economic and thus more competitive with other power generation methods. A summary is given of the design improvements made on Mitsubishi fuel, as well as demonstration programmes of current design fuel to prove its superior reliability and to prepare the database for a future extension of burnup. (author)

  13. Fast reactor 3D core and burnup analysis using VESTA

    Energy Technology Data Exchange (ETDEWEB)

    Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

  14. Extension of the TRANSURANUS burn-up model

    International Nuclear Information System (INIS)

    The validation range of the model in the TRANSURANUS fuel performance code for calculating the radial power density and burn-up in UO2 fuel has been extended from 64 MWd/kgHM up to 102 MWd/kgHM, thereby improving also its precision. In addition, the first verification of calculations with post-irradiation examination data is reported for LWR-MOX fuel with a rod average burn-up up to 45 MWd/kgHM. The extension covers the inclusion of new isotopes in order to account for the production of 238Pu. The corresponding one-group cross-sections used in the equations rely on results obtained with ALEPH, a new Monte Carlo burn-up code. The experimental verification is based on electron probe microanalysis (EPMA) and on secondary ion mass spectrometry (SIMS) as well as radiochemical data of fuel irradiated in commercial power plants. The deviations are quantified in terms of frequency distributions of the relative errors. The relative errors on the burn-up distributions in both fuel types remain below 12%, corresponding to the experimental scatter

  15. Prediction of fission gas pressure from high burnup oxide fuel

    International Nuclear Information System (INIS)

    The ELESIM fuel performance code incorporates a fundamentally based treatment of the relevant physical processes affecting fission gas release. The fission gas release model treats fission gas diffusion, formation and subsequent interlinkage of intergranular bubbles, grain boundary storage of gas, grain growth and fuel swelling. The latter case considers the contributions of thermal expansion, densification, solid fission products, and gas bubbles. The effect of porosity on fuel thermal conductivity is taken into account. Previously we showed predictions of the gas release model agreed well with measured values for oxide fuel with burnups to about 300 MW.h/kg U. The applicability of the model to high burnup fuel is examined using examples from the literature. The fission gas release range considered is about 1-100% for burnups to 1000 MW.h/kg U in thermal reactor fuel and 2400 MW.h/kg U in fast reactor fuel. Predicted and measured releases are shown to be in good agreement, suggesting that the fundamental model is correct. In some models, empirical correction factors are required at high burnup to achieve agreement between predicted and measured release values; no such factor is required in ELESIM. (auth)

  16. Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

    International Nuclear Information System (INIS)

    The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit

  17. Chemical and microbiological farm milk quality determination in three Croatian regions

    Directory of Open Access Journals (Sweden)

    Neven Antunac

    2012-12-01

    Full Text Available The purpose of this study was to determine chemical and microbiological quality of raw milk from 30 farms of different sizes from eastern, central and southern Croatian regions. Samples of fresh raw milk (n=360 are determined by the content of fat, protein, total solids, and the number of microorganisms and somatic cells. Analysed milk derived from Holstein, Simmental and Brown Swiss cows, and their crossbred. Chemical composition of milk was determined by infrared spectrophotometry, microbiological quality by milk epifluorescence flow cytometry, and the number of somatic cells in milk was determined by fluoro-opto-electronic method. The results of chemical quality of milk - of milk fat and protein entirely, for all groups of large and small farms in eastern, central and southern regions meet the requirements of the “Regulations on the Quality of Fresh Raw Milk”. Looking at the value of the total number of microorganisms, only a group of small agricultural holdings of the southern region do not meet the requirements prescribed by the Regulations on the Quality of Fresh Raw Milk in 2000. Small agricultural holdings of the southern region had significantly higher total number of microorganisms (P<0.0001 in relation to the other two groups. There was a statistically significant difference (P<0.01 between somatic cell count (SCC in milk of all three region’s large agricultural holdings, while the SCC in milk of small agricultural holdings of different regions did not show statistically significant difference (P<0.05.

  18. Need for higher fuel burnup at the Hatch Plant

    International Nuclear Information System (INIS)

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch's operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about

  19. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

    Beckhman, J.T. [Georgia Power Co., Birmingham, AL (United States)

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.

  20. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  1. Dimensionally reduced expression for the QCD fermion determinant at finite temperature and chemical potential

    International Nuclear Information System (INIS)

    A dimensionally reduced expression for the QCD fermion determinant at finite temperature and chemical potential is derived which sheds light on the determinant's dependence on these quantities. This is done via a partial zeta regularization, formally applying a general formula for the zeta determinant of a differential operator in one variable with operator-valued coefficients. The resulting expression generalizes the known one for the free fermion determinant, obtained via Matsubara frequency summation, to the case of a general background gauge field; moreover there is no undetermined overall factor. Rigorous versions of the result are obtained in a continuous time-lattice space setting. The determinant expression reduces to a remarkably simple form in the low temperature limit. A program for using this to obtain insight into the QCD phase transition at zero temperature and nonzero density is outlined

  2. The measurement of burn-up level in HTR-10

    International Nuclear Information System (INIS)

    Without shutting down the HTR-10, each fuel ball unloaded from the core must be measured. Judgment is made whether the desired burn-up level is reached. A fuel ball should be reloaded into the core when its burn-up level is less than 72,000 Mwd/tu. Since the measurement of burn-up level for a ball containing 0.9g 235U at most must be nondestructive, a γ spectroscopy method with high-resolution for the fission product 137Cs is typically chosen. However, because the HTR-10 is not provided with any kind of external calibrating source, it is impossible to achieve the goal that the accuracy of measurement is up to 2% by the above method. The method measuring burn-up levels without external source uses the ratio of 134Cs to 137Cs and gets bogged down in something unusual that both successful and failed examples have alternated in publications. The inner calibrating method is proposed in the paper. It is necessary to solve the following problems: a) The simple relationship between 134Cs/137Cs and burn-up level is held only in a specified range, but not for a spent ball. b) The migration must be ignored. c) How to deal with the neutron spectrum within HTR-10? The paper also introduces such useful method as neutron spectrum correction, picking out the graphite balls and extraction of 137Cs from γ spectra of reactor. The appropriate instrumentation that discriminates out the spent fuel balls and picks out the graphite balls is described. (author)

  3. Considerations on burn-up dependent RIA and LOCA criteria

    International Nuclear Information System (INIS)

    For RIA transients, a fuel failure threshold has been derived and compared with recent experimental data relevant for BWR and PWR fuel. The threshold can be applied to HZP and CZP transients, account taken for the different initial enthalpy and for the lower ductility at cold conditions. It can also be used for non-zero power transients, provided that a term accounting for the initial power is incorporated. The proposed threshold predicts reasonably well the results obtained in the CABRI and NSRR tests when the different state of the cladding, i.e. ductile or brittle, is taken into account. Apart from some exceptions discussed in the paper, such as the effect of oxide spalling, one should consider ductile state for HZP conditions and brittle state for CZP conditions. The threshold applies equally well to UO2 and MOX fuel, but the database on MOX is limited. For LOCA transients, the cladding limit may decrease with burn-up due to cladding corrosion and hydrogen pick-up. A provisional criterion shows that the predicted burn-up effect is moderate or negligible if one uses the results obtained with actual high burn-up cladding. On the other hand, a large effect is predicted based on the results obtained with non-irradiated, pre-hydrided cladding specimens. There is a question however on as to whether these specimens can be representative for high burn-up material. The experimental evidence is still scarce and more data on high burn-up cladding is needed in order to arrive to firm conclusions. Most of the data currently available relates to Zr-4 cladding. The experiments made on ZIRLO and M5 cladding show that these alloys have a RIA and LOCA behaviour similar to or better than Zr-4. However, the data is limited, especially for LOCA conditions, where only un-irradiated specimens have been tested so far. (author)

  4. Review of the Literature on Determinants of Chemical Hazard Information Recall among Workers and Consumers

    Directory of Open Access Journals (Sweden)

    Farzana Sathar

    2016-05-01

    Full Text Available In many low and middle income countries (LMIC, workers’ and consumers’ only access to risk and hazard information in relation to the chemicals they use or work with is on the chemical label and safety data sheet. Recall of chemical hazard information is vital in order for label warnings and precautionary information to promote effective safety behaviors. A literature review, therefore, was conducted on determinants of chemical hazard information recall among workers and consumers globally. Since comprehension and recall are closely linked, the determinants of both were reviewed. Literature was reviewed from both online and print peer reviewed journals for all study designs and countries. This review indicated that the level of education, previous training and the inclusion of pictograms on the hazard communication material are all factors that contribute to the recall of hazard information. The influence of gender and age on recall is incongruent and remains to be explored. More research is required on the demographic predictors of the recall of hazard information, the effect of design and non-design factors on recall, the effect of training on the recall among low literate populations and the examining of different regions or contexts.

  5. Review of the Literature on Determinants of Chemical Hazard Information Recall among Workers and Consumers

    Science.gov (United States)

    Sathar, Farzana; Dalvie, Mohamed Aqiel; Rother, Hanna-Andrea

    2016-01-01

    In many low and middle income countries (LMIC), workers’ and consumers’ only access to risk and hazard information in relation to the chemicals they use or work with is on the chemical label and safety data sheet. Recall of chemical hazard information is vital in order for label warnings and precautionary information to promote effective safety behaviors. A literature review, therefore, was conducted on determinants of chemical hazard information recall among workers and consumers globally. Since comprehension and recall are closely linked, the determinants of both were reviewed. Literature was reviewed from both online and print peer reviewed journals for all study designs and countries. This review indicated that the level of education, previous training and the inclusion of pictograms on the hazard communication material are all factors that contribute to the recall of hazard information. The influence of gender and age on recall is incongruent and remains to be explored. More research is required on the demographic predictors of the recall of hazard information, the effect of design and non-design factors on recall, the effect of training on the recall among low literate populations and the examining of different regions or contexts. PMID:27258291

  6. Behaviour of fuel rods of the second generation at high burnup WWER-440 fuel cycles. Aspects for attainment of burnup 70 MWd/kgU

    International Nuclear Information System (INIS)

    In this report an analysis of WWER-440 fuel of the second generation supplied by Russian JSC TVEL for high burnup fuel cycle is presented. The certificated code START-3 is applied to modeling of fuel rod operation parameters. Reliability of high-burnup fuel on the base of 5-6 year operation is demonstrated. Special attention is paid to aspects for attainment of burnup 70 MWd/kgU, including experimental and fuel modeling support and fuel operation experience

  7. Proton chemical shift tensors determined by 3D ultrafast MAS double-quantum NMR spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Rongchun; Mroue, Kamal H.; Ramamoorthy, Ayyalusamy, E-mail: ramamoor@umich.edu [Biophysics and Department of Chemistry, The University of Michigan, Ann Arbor, Michigan 48109-1055 (United States)

    2015-10-14

    Proton NMR spectroscopy in the solid state has recently attracted much attention owing to the significant enhancement in spectral resolution afforded by the remarkable advances in ultrafast magic angle spinning (MAS) capabilities. In particular, proton chemical shift anisotropy (CSA) has become an important tool for obtaining specific insights into inter/intra-molecular hydrogen bonding. However, even at the highest currently feasible spinning frequencies (110–120 kHz), {sup 1}H MAS NMR spectra of rigid solids still suffer from poor resolution and severe peak overlap caused by the strong {sup 1}H–{sup 1}H homonuclear dipolar couplings and narrow {sup 1}H chemical shift (CS) ranges, which render it difficult to determine the CSA of specific proton sites in the standard CSA/single-quantum (SQ) chemical shift correlation experiment. Herein, we propose a three-dimensional (3D) {sup 1}H double-quantum (DQ) chemical shift/CSA/SQ chemical shift correlation experiment to extract the CS tensors of proton sites whose signals are not well resolved along the single-quantum chemical shift dimension. As extracted from the 3D spectrum, the F1/F3 (DQ/SQ) projection provides valuable information about {sup 1}H–{sup 1}H proximities, which might also reveal the hydrogen-bonding connectivities. In addition, the F2/F3 (CSA/SQ) correlation spectrum, which is similar to the regular 2D CSA/SQ correlation experiment, yields chemical shift anisotropic line shapes at different isotropic chemical shifts. More importantly, since the F2/F1 (CSA/DQ) spectrum correlates the CSA with the DQ signal induced by two neighboring proton sites, the CSA spectrum sliced at a specific DQ chemical shift position contains the CSA information of two neighboring spins indicated by the DQ chemical shift. If these two spins have different CS tensors, both tensors can be extracted by numerical fitting. We believe that this robust and elegant single-channel proton-based 3D experiment provides useful atomistic

  8. Determination of Chemical Compositions on Adult Kidney Stones—A Spectroscopic Study

    Science.gov (United States)

    Raju, K.; Rakkappan, C.

    2008-11-01

    The chemical compositions of the kidney stones of both the sexes of patients, aged from 40 to 70, living in and around Chidambaram town are determined by using FT-IR and X-RD technique. The kidney stone samples used in the present study were procured from the Rajah Muthiah Medical College and Hospital, Annamalai University. The FT-IR spectra of different kidney stone samples were recorded in the range of 4000-400 cm-1. By identifying the characteristic frequency, the chemical compositions of the samples are determined. The results analyzed by FTIR technique were confirmed by X-RD method, in which the recorded X-ray diffractogram are compared with JCPDS files using search match method. Further analysis of XRD pattern also reveals the same.

  9. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this

  10. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    International Nuclear Information System (INIS)

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from ∼ ±40% at beginning of life to ∼ ±10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this case, a self

  11. Physical, chemical, and mineralogical characterization of vertisols to determine their parent material

    OpenAIRE

    Erasto Domingo Sotelo Ruiz; María del Carmen Gutiérrez Castorena; Carlos Alberto Ortiz Solorio

    2013-01-01

    Haplusterts, Typic Haplusterts, and Mollic Ustifluvents. Sedimentary origin soils were classified as Chromic Calciusterts The response of soils to weathering processes depends upon their parent material. Proper identification of the primary and secondary minerals in Vertisols provides information about the parent material that gives origin to these soils. Thus, the objec-tives of this study were 1) to determine the physical and chemi-cal properties of Vertisols in order to characterize and cl...

  12. Evaluation of Three Flow Injection Analysis Methods for the Determination of Chemical Oxygen Demand

    OpenAIRE

    Korenaga, Takashi; Moriwake, Tosio; Takahashi, Teruo

    1984-01-01

    Three methods for determining chemical oxygen demand (COD) by means of flow injection analysis (FIA) with potassium permanganate, potassium dichromate, or cerium(IV) sulfate as oxidant, developed in this laboratory, are described from the point of view of their operating properties. The permanganate method is the most sensitive and common, but forms manganese(IV) oxide precipitate which blocks the FIA lines and connectors. Addition of phosphoric acid in the reagent system is, however, effecti...

  13. The Importance of Determination of some Physical – Chemical Properties of Wheat and Flour

    Directory of Open Access Journals (Sweden)

    Husejin Keran

    2009-12-01

    In this work, some physical – chemical properties are determined and some comparations of characteristics were performed in both wheat and flour. Characteristics that were observed in this work are moisture content, ash content, protein content, Zeleny sedimentation value, gluten content and water adsorption values. On the base of results obtained in this work, some conclusions are made that could be useful for milling industry.

  14. A laboratory manual for the determination of inorganic chemical contaminants and nutrients in sewage sludges

    International Nuclear Information System (INIS)

    In addition to a brief discussion on sewage sludge disposal, sludge contaminants, and the potential beneficial and adverse effects of the various inorganic chemical contaminants and nutrients commonly present in sewage sludge, this technical guide presents a scheme of analysis for the determination of the major inorganic contaminants and nutrients. Safety and simplicity were the main criteria considered in the selection of the various sample pretreatment procedures and analytical techniques

  15. Revaluation on measured burnup values of fuel assemblies by post-irradiation experiments at BWR plants

    International Nuclear Information System (INIS)

    Fuel composition data for 8x8 UO2, Tsuruga MOX and 9x9-A type UO2 fuel assemblies irradiated in BWR plants were measured. Burnup values for measured fuels based on Nd-148 method were revaluated. In this report, Nd-148 fission yield and energy per fission obtained by burnup analyses for measured fuels were applied and fuel composition data for the measured fuel assemblies were revised. Furthermore, the adequacies of revaluated burnup values were verified through the comparison with burnup values calculated by the burnup analyses for the measured fuel assemblies. (author)

  16. Investigation of several methods to set burnup for criticality safety assessment of spent fuel transport casks

    International Nuclear Information System (INIS)

    Several currently available methods to set burnup for depletion calculation are reviewed and discussed about its adequacy for criticality safety assessment of spent fuel (SF) transport casks by taking burnup credit (BC) into accounts. Various errors associated with BC criticality analyses are evaluated and converted to equivalent burnup to compare each other. Methods are proposed to use some reduced burnups equivalent to compensation of these associated errors. Effects of assumption of axial burnup distribution on criticality calculation and irradiation history parameter variation on depletion calculation are evaluated with OECD/NEA BC international benchmark data. (author)

  17. Determination of temperature and transverse flow velocity at chemical freeze-out in relativistic nuclear interactions

    International Nuclear Information System (INIS)

    We propose a parameter-free method to determine the temperature of a thermalized state in relativistic nuclear interactions, using the experimental μq/T and μs/T values, obtained from strange particle ratios. The hadron gas formalism and strangeness neutrality are employed to relate the quark-chemical potential μq and μs to the temperature and thus determine its value at chemical freeze-out. This temperature, together with the inverse slope parameter from mT distributions, enable the determination of the transverse flow velocity of the fireball matter, thus disentangling the thermal and flow effects. We study several nucleus-nucleus interactions from AGS and SPS and obtain the temperature, transverse flow velocity, and quark-chemical potentials. Extrapolating the systematics we predict the values of these quantities for ongoing and future experiments at AGS, SPS, and RHIC. We discuss the possibility of reaching the conditions for quark deconfinement and QGP formation and give distinct and identifiable signature. copyright 1996 The American Physical Society

  18. Chemical dispersants and pre-treatments to determine clay in soils with different mineralogy

    Directory of Open Access Journals (Sweden)

    Cristiane Rodrigues

    2011-10-01

    Full Text Available Knowledge of the soil physical properties, including the clay content, is of utmost importance for agriculture. The behavior of apparently similar soils can differ in intrinsic characteristics determined by different formation processes and nature of the parent material. The purpose of this study was to assess the efficacy of separate or combined pre-treatments, dispersion methods and chemical dispersant agents to determine clay in some soil classes, selected according to their mineralogy. Two Brazilian Oxisols, two Alfisols and one Mollisol with contrasting mineralogy were selected. Different treatments were applied: chemical substances as dispersants (lithium hydroxide, sodium hydroxide, and hexametaphosphate; pre-treatment with dithionite, ammonium oxalate, and hydrogen peroxide to eliminate organic matter; and coarse sand as abrasive and ultrasound, to test their mechanical action. The conclusion was drawn that different treatments must be applied to determine clay, in view of the soil mineralogy. Lithium hydroxide was not efficient to disperse low-CEC electropositive soils and very efficient in dispersing high-CEC electronegative soils. The use of coarse sand as an abrasive increased the clay content of all soils and in all treatments in which dispersion occurred, with or without the use of chemical dispersants. The efficiency of coarse sand is not the same for all soil classes.

  19. Determination of sulfonamides in meat by liquid chromatography coupled with atmospheric pressure chemical ionization mass spectrometry

    International Nuclear Information System (INIS)

    Liquid chromatography/atmospheric pressure chemical ionization-mass spectrometry (LC-APCI-MS) has been used for the determination of sulfonamides in meat. Five typical sulfonamides were selected as target compounds, and beef meat was selected as a matrix sample. As internal standards, sulfapyridine and isotope labeled sulfamethazine (13C6-SMZ) were used. Compared to the results of recent reports, our results have shown improved precision to a RSD of 1.8% for the determination of sulfamethazine spiked with 75 ng/g level in meat

  20. Rapid Determination of the Chemical Oxygen Demand of Water Using a Thermal Biosensor

    OpenAIRE

    Na Yao; Jinqi Wang; Yikai Zhou

    2014-01-01

    In this paper we describe a thermal biosensor with a flow injection analysis system for the determination of the chemical oxygen demand (COD) of water samples. Glucose solutions of different concentrations and actual water samples were tested, and their COD values were determined by measuring the heat generated when the samples passed through a column containing periodic acid. The biosensor exhibited a large linear range (5 to 3000 mg/L) and a low detection limit (1.84 mg/L). It could tolerat...

  1. Evaluation technology for burnup and generated amount of plutonium by measurement of Xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

    International Nuclear Information System (INIS)

    The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas (DOG) at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant (TRP) during BWR fuel (approx. 30GWD/MTU) reprocessing campaign. Xenon isotopic ratio was determined with Gas Chromatography/Mass Spectrometry. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Laboratory. Inferred burnup evaluated by Xe isotopic measurements and NOVA were in good agreement with those of the declared burnup in the range from -3.8% to 7.1%. Also, the inferred amount of Pu in spent fuel was in good agreed with those of the declared amount of Pu calculated by ORIGEN code in the range from -0.9% to 4.7%. The evaluation technique is applicable for both burnup credit to achieve efficient criticality safety control and a new measurement method for safeguards inspection. (author)

  2. Parameter-free determination of actual temperature at chemical freeze-out in nuclear interactions

    International Nuclear Information System (INIS)

    We propose a method to determine the actual temperature at chemical freeze-out in relativistic nucleus-nucleus collisions, using the experimental μq/T and μs/T values, obtained from strange particle ratios. We employ the Hadron Gas formalism, assuming only local thermal equilibration, to relate the quarkchemical potential and temperature. This relation constrains the allowed values of μq/T, μs/T and T, enabling the determination of the actual temperature. Comparison of the inverse slope parameter of the mT-distributions with the actual temperature determines the transverse flow velocity of the fireball matter. Knowledge of these quantities is essential in determining the EoS of nuclear matter and in evaluating interactions with regard to a possible phase transition to QGP. copyright 1995 American Institute of Physics

  3. Burnup measurement and its relation to the amounts of 149Sm and 150Sm and the ratios of 154Eu/155Eu and 154Eu/152Eu in spent fuel of a power reactor

    International Nuclear Information System (INIS)

    The amounts of 148Nd, 149Sm and 150Sm by the isotope dilution mass spectrography and the ratios of 154Eu/155Eu and 154Eu/152Eu by the γ-spectrography in the spent fuel of a power reactor have been measured. The amouns of 150Sm is directly proportional to that of 148Nd and burnup can be determined by 150Sm. Relation of the ratios of 154Eu/155Eu with respect to the burnup is plotted

  4. An economic evaluation of a storage system for casks with burnup credit

    International Nuclear Information System (INIS)

    It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

  5. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    International Nuclear Information System (INIS)

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO2 fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  6. Transient fission gas release from UO2 fuel for high temperature and high burnup

    International Nuclear Information System (INIS)

    In the present paper it is assumed that the fission gas release kinetics from an irradiated UO2 fuel for high temperature is determined by the kinetics of grain growth. A well founded assumption that Vitanza curve describes the change of uranium dioxide re-crystallization temperature and the experimental results referring to the limiting grain size presented in the literature are used to modify the grain growth model. Algorithms of fission gas release due to re-crystallization of uranium dioxide grains are worked out. The defect trap model of fission gas behaviour described in the earlier papers is supplemented with the algorithms. Calculations of fission gas release in function of time, temperature, burn-up and initial grain sizes are obtained. Computation of transient fission gas release in the paper is limited to the case where steady state of irradiation to accumulate a desired burn-up is performed below the temperature of re-crystallization then the subsequent step temperature increase follows. There are considered two kinds of step temperature increase for different burn-up: the final temperature of the step increase is below and above the re-crystallization temperature. Calculations show that bursts of fission gas are predicted in both kinds. The release rate of gas liberated for the final temperature above the re-crystallization temperature is much higher than for final temperature below the re-crystallization temperature. The time required for the burst to subside is longer due to grain growth than due to diffusion of bubbles and knock-out release. The theoretical results explain qualitatively the experimental data but some of them need to be verified since this sort of experimental data are not found in the available literature. (author)

  7. EDXRF applied to the chemical element determination of small invertebrate samples

    Energy Technology Data Exchange (ETDEWEB)

    Magalhaes, Marcelo L.R.; Santos, Mariana L.O.; Cantinha, Rebeca S.; Souza, Thomas Marques de; Franca, Elvis J. de, E-mail: marcelo_rlm@hotmail.com, E-mail: marianasantos_ufpe@hotmail.com, E-mail: rebecanuclear@gmail.com, E-mail: thomasmarques@live.com.pt, E-mail: ejfranca@cnen.gov.br [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2015-07-01

    Energy Dispersion X-Ray Fluorescence - EDXRF is a fast analytical technique of easy operation, however demanding reliable analytical curves due to the intrinsic matrix dependence and interference during the analysis. By using biological materials of diverse matrices, multielemental analytical protocols can be implemented and a group of chemical elements could be determined in diverse biological matrices depending on the chemical element concentration. Particularly for invertebrates, EDXRF presents some advantages associated to the possibility of the analysis of small size samples, in which a collimator can be used that directing the incidence of X-rays to a small surface of the analyzed samples. In this work, EDXRF was applied to determine Cl, Fe, P, S and Zn in invertebrate samples using the collimator of 3 mm and 10 mm. For the assessment of the analytical protocol, the SRM 2976 Trace Elements in Mollusk produced and SRM 8415 Whole Egg Powder by the National Institute of Standards and Technology - NIST were also analyzed. After sampling by using pitfall traps, invertebrate were lyophilized, milled and transferred to polyethylene vials covered by XRF polyethylene. Analyses were performed at atmosphere lower than 30 Pa, varying voltage and electric current according to the chemical element to be analyzed. For comparison, Zn in the invertebrate material was also quantified by graphite furnace atomic absorption spectrometry after acid treatment (mixture of nitric acid and hydrogen peroxide) of samples have. Compared to the collimator of 10 mm, the SRM 2976 and SRM 8415 results obtained by the 3 mm collimator agreed well at the 95% confidence level since the E{sub n} Number were in the range of -1 and 1. Results from GFAAS were in accordance to the EDXRF values for composite samples. Therefore, determination of some chemical elements by EDXRF can be recommended for very small invertebrate samples (lower than 100 mg) with advantage of preserving the samples. (author)

  8. EDXRF applied to the chemical element determination of small invertebrate samples

    International Nuclear Information System (INIS)

    Energy Dispersion X-Ray Fluorescence - EDXRF is a fast analytical technique of easy operation, however demanding reliable analytical curves due to the intrinsic matrix dependence and interference during the analysis. By using biological materials of diverse matrices, multielemental analytical protocols can be implemented and a group of chemical elements could be determined in diverse biological matrices depending on the chemical element concentration. Particularly for invertebrates, EDXRF presents some advantages associated to the possibility of the analysis of small size samples, in which a collimator can be used that directing the incidence of X-rays to a small surface of the analyzed samples. In this work, EDXRF was applied to determine Cl, Fe, P, S and Zn in invertebrate samples using the collimator of 3 mm and 10 mm. For the assessment of the analytical protocol, the SRM 2976 Trace Elements in Mollusk produced and SRM 8415 Whole Egg Powder by the National Institute of Standards and Technology - NIST were also analyzed. After sampling by using pitfall traps, invertebrate were lyophilized, milled and transferred to polyethylene vials covered by XRF polyethylene. Analyses were performed at atmosphere lower than 30 Pa, varying voltage and electric current according to the chemical element to be analyzed. For comparison, Zn in the invertebrate material was also quantified by graphite furnace atomic absorption spectrometry after acid treatment (mixture of nitric acid and hydrogen peroxide) of samples have. Compared to the collimator of 10 mm, the SRM 2976 and SRM 8415 results obtained by the 3 mm collimator agreed well at the 95% confidence level since the En Number were in the range of -1 and 1. Results from GFAAS were in accordance to the EDXRF values for composite samples. Therefore, determination of some chemical elements by EDXRF can be recommended for very small invertebrate samples (lower than 100 mg) with advantage of preserving the samples. (author)

  9. Accuracy and precision of protein–ligand interaction kinetics determined from chemical shift titrations

    International Nuclear Information System (INIS)

    NMR-monitored chemical shift titrations for the study of weak protein–ligand interactions represent a rich source of information regarding thermodynamic parameters such as dissociation constants (KD) in the micro- to millimolar range, populations for the free and ligand-bound states, and the kinetics of interconversion between states, which are typically within the fast exchange regime on the NMR timescale. We recently developed two chemical shift titration methods wherein co-variation of the total protein and ligand concentrations gives increased precision for the KD value of a 1:1 protein–ligand interaction (Markin and Spyracopoulos in J Biomol NMR 53: 125–138, 2012). In this study, we demonstrate that classical line shape analysis applied to a single set of 1H–15N 2D HSQC NMR spectra acquired using precise protein–ligand chemical shift titration methods we developed, produces accurate and precise kinetic parameters such as the off-rate (koff). For experimentally determined kinetics in the fast exchange regime on the NMR timescale, koff ∼ 3,000 s−1 in this work, the accuracy of classical line shape analysis was determined to be better than 5 % by conducting quantum mechanical NMR simulations of the chemical shift titration methods with the magnetic resonance toolkit GAMMA. Using Monte Carlo simulations, the experimental precision for koff from line shape analysis of NMR spectra was determined to be 13 %, in agreement with the theoretical precision of 12 % from line shape analysis of the GAMMA simulations in the presence of noise and protein concentration errors. In addition, GAMMA simulations were employed to demonstrate that line shape analysis has the potential to provide reasonably accurate and precise koff values over a wide range, from 100 to 15,000 s−1. The validity of line shape analysis for koff values approaching intermediate exchange (∼100 s−1), may be facilitated by more accurate KD measurements from NMR-monitored chemical shift

  10. Chemical aspects of the precise and accurate determination of uranium and plutonium from nuclear fuel solutions

    International Nuclear Information System (INIS)

    A method for the simultaneous or separate determination of uranium and plutonium has been developed. The method is based on the sorption of uranium and plutonium as their chloro complexes on Dowex 1x10 column. When separate uranium and plutonium fractions are desired, plutonium ions are reduced to Pu (III) and eluted, after which the uranium ions are eluted with dilute HCl. Simultaneous stripping of a mass ratio U/Pu approximately 1 fraction for mass spectrometric measurements is achieved by proper choice of eluant HC1 concentration. Special attention was paid to the obtaining of americium free plutonium fractions. The distribution coefficient measurements showed that at 12.5-M HCl at least 30 % of americium ions formed anionic chloro complexes. The chemical aspects of isotopic fractionation in a multiple filament thermal ionization source were also investigated. Samples of uranium were loaded as nitrates, chlorides, and sulphates and the dependence of the measured uranium isotopic ratios on the chemical form of the loading solution as well as on the filament material was studied. Likewise the dependence of the formation of uranium and its oxide ions on various chemical and instrumental conditions was investigated using tungsten and rhenium filaments. Systematic errors arising from the chemical conditions are compared with errors arising from the automatic evaluation of of spectra. (author)

  11. Probabilistic Approach to Determining Unbiased Random-coil Carbon-13 Chemical Shift Values from the Protein Chemical Shift Database

    International Nuclear Information System (INIS)

    We describe a probabilistic model for deriving, from the database of assigned chemical shifts, a set of random coil chemical shift values that are 'unbiased' insofar as contributions from detectable secondary structure have been minimized (RCCSu). We have used this approach to derive a set of RCCSu values for 13Cα and 13Cβ for 17 of the 20 standard amino acid residue types by taking advantage of the known opposite conformational dependence of these parameters. We present a second probabilistic approach that utilizes the maximum entropy principle to analyze the database of 13Cα and 13Cβ chemical shifts considered separately; this approach yielded a second set of random coil chemical shifts (RCCSmax-ent). Both new approaches analyze the chemical shift database without reference to known structure. Prior approaches have used either the chemical shifts of small peptides assumed to model the random coil state (RCCSpeptide) or statistical analysis of chemical shifts associated with structure not in helical or strand conformation (RCCSstruct-stat). We show that the RCCSmax-ent values are strikingly similar to published RCCSpeptide and RCCSstruct-stat values. By contrast, the RCCSu values differ significantly from both published types of random coil chemical shift values. The differences (RCCSpeptide-RCCSu) for individual residue types show a correlation with known intrinsic conformational propensities. These results suggest that random coil chemical shift values from both prior approaches are biased by conformational preferences. RCCSu values appear to be consistent with the current concept of the 'random coil' as the state in which the geometry of the polypeptide ensemble samples the allowed region of (φ,ψ)-space in the absence of any dominant stabilizing interactions and thus represent an improved basis for the detection of secondary structure. Coupled with the growing database of chemical shifts, this probabilistic approach makes it possible to refine

  12. Verification of the OREST (HAMMER-ORIGEN) burn-up program system in the post-irradiation analyses of fuel elements BE-168, 170, 171 and 176 of the Obrigheim reactor

    International Nuclear Information System (INIS)

    The burn-up code OREST has a spectrum code assigned to it, which determines the neutron spectrum in the actual fuel element mixture at the start and during burn-up and carries out the resonance treatment for the most important uranium and transuranic element isotopes. The reliability of the OREST system is shown for UO2 burn-up in PWR's. Post-irradiation analyses of five UO2 fuel elements of KWO with an initial enrichment of 3.13% by weight of U235 and a mean burn-up of 28.4 GWd/tV are used for comparison. The reliability of OREST information for UO2 fuel in PWR's is proved by the good agreement between experiment and calculation, also compared with KfK's results. (orig./HP)

  13. Chemical effects of lanthanides and actinides in glasses determined with electron energy loss spectroscopy

    International Nuclear Information System (INIS)

    Chemical and structural environments of f-electron elements in glasses are the origin of many of the important properties of materials with these elements; thus oxidation state and chemical coordination of lanthanides and actinides in host materials is an important design consideration in optically active glasses, magnetic materials, perovskite superconductors, and nuclear waste materials. We have made use of the line shapes of Ce to determine its oxidation state in alkali borosilicate glasses being developed for immobilization of Pu. Examination of several prototype waste glass compositions with EELS shows that the redox state of Ce doped to 7 wt% could be varied by suitable choice of alkali elements. EELS for a Pu-doped glass illustrate the small actinide N4/N5 intensity ratio and show that the Pu-N4,5 white line cross section is comparable to that of Gd M4,5

  14. Collaborative study of the Food Chemicals Codex method for the determination of the neutralizing value of sodium aluminum phosphate.

    Science.gov (United States)

    Park, D L

    1976-01-01

    Fifteen laboratories participated in a collaborative study to evaluate the Food Chemicals Codex method for the determination of the neutralizing value of sodium aluminum phosphate. The AOAC method for determining the neutralizing value of sodium acid pyrophosphate, sec. 8.010, was also included in the study. The precisions of the Food chemicals Codex method, based on the between-replicate standard deviation and on one collaborator making one determination, are 1.16 and 3.66, respectively. The Food Chemicals Codex method for the determination of the neutralizing value of sodium aluminum phosphate has been adopted as official first action. PMID:2581

  15. CHEMICALS

    CERN Multimedia

    Medical Service

    2002-01-01

    It is reminded that all persons who use chemicals must inform CERN's Chemistry Service (TIS-GS-GC) and the CERN Medical Service (TIS-ME). Information concerning their toxicity or other hazards as well as the necessary individual and collective protection measures will be provided by these two services. Users must be in possession of a material safety data sheet (MSDS) for each chemical used. These can be obtained by one of several means : the manufacturer of the chemical (legally obliged to supply an MSDS for each chemical delivered) ; CERN's Chemistry Service of the General Safety Group of TIS ; for chemicals and gases available in the CERN Stores the MSDS has been made available via EDH either in pdf format or else via a link to the supplier's web site. Training courses in chemical safety are available for registration via HR-TD. CERN Medical Service : TIS-ME :73186 or service.medical@cern.ch Chemistry Service : TIS-GS-GC : 78546

  16. Determining treatment frequency for controlling weeds on traffic islands using chemical and non-chemical weed control

    DEFF Research Database (Denmark)

    Rask, Anne Merete; Larsen, S.U.; Andreasen, Christian;

    2013-01-01

    Many public authorities rely on the use of non-chemical weed control methods, due to stringent restrictions on herbicide use in urban areas. However, these methods usually require more repeated treatments than chemical weed management, resulting in increased costs of weed management. In order to...

  17. Determining the chemical composition of cloud condensation nuclei. Second progress report

    Energy Technology Data Exchange (ETDEWEB)

    Williams, A.L.; Rothert, J.E.; McClure, K.E. [Illinois State Water Survey, Champaign, IL (United States); Alofs, D.J.; Hagen, D.E.; White, D.R.; Hopkins, A.R.; Trueblood, M.B. [Missouri Univ., Rolla, MO (USA). Cloud and Aerosol Science Lab.

    1992-02-01

    This second progress report describes the status of the project one and one-half years after the start. The goal of the project is to develop the instrumentation to collect cloud condensation nuclei (CCN) in sufficient amounts to determine their chemical composition, and to survey the CCN composition in different climates through a series of field measurements. Our approach to CCN collection is to first form droplets on the nuclei under simulated cloud humidity conditions, which is the only known method of identifying CCN from the background aerosol. Under cloud chamber conditions, the droplets formed become larger than the surrounding aerosol, and can then be removed by inertial impaction. The residue of the evaporated droplets represents the sample to be chemically analyzed. Two size functions of CCN particles are collected by first forming droplets on the large particles are collected by first forming droplets on the large CCN in a haze chamber at 100% relative humidity, and then activating the remaining CCN at 1% supersaturation in a cloud chamber. The experimental apparatus is a serious flow arrangement consisting of an impactor to remove the large aerosol particles, a haze chamber to form droplets on the remaining larger CCN, another impactor to remove the haze droplets containing the larger CCN particles for chemical analysis, a continuous flow diffusion (CFD) cloud chamber to form droplets on the remaining smaller CCN, and a third impactor to remove the droplets for the small CCN sample. Progress is documented here on the development of each of the major components of the flow system. Chemical results are reported on tests to determine suitable wicking material for the different plates. Results of computer modeling of various impactor flows are discussed.

  18. Actinide-only burnup credit for spent fuel transport

    International Nuclear Information System (INIS)

    A conservative methodology is described that would allow taking credit for burn up in the criticality safety analysis of spent nuclear fuel packages. Requirements for its implementation include isotopic and criticality validation, generation of package loading criteria using limiting parameters, and assembly burn up verification by measurement. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentrations with burnup. No credit for fission product neutron absorbers is taken. Analyses are included regarding the methodology's financial benefits and conservative margin. It is estimated that the proposed actinide-only burnup credit methodology would save 20% of the transport costs. Nevertheless, the methodology includes a substantial margin. Conservatism due to the isotopic correction factors, limiting modelling parameters, limiting axial profiles and exclusion of the fission products ranges from 10 to 25% k. (author)

  19. Advances In Burnup Credit Criticality Safety Analysis Methods And Applications

    International Nuclear Information System (INIS)

    An International Workshop on “Advances in Applications of Burnup Credit for Spent Fuel Storage, Transport, Reprocessing, and Disposition” organized by the Nuclear Safety Council of Spain (CSN) in cooperation with the International Atomic Energy Agency (IAEA) was held at Córdoba, Spain, on October 27– 30, 2009. The objectives of this workshop were to identify the benefits that accrue from recent improvements of the burnup credit (BUC) analysis methodologies, to analyze the implications of applying improved BUC methodologies, focusing on both the safety-related and operational aspects, and to foster the exchange of international experience in licensing and implementation of BUC applications. In the paper on hand the attention is focused on the improvements of BUC analysis methodologies. (author)

  20. Burnup calculations using serpent code in accelerator driven thorium reactors

    International Nuclear Information System (INIS)

    In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)

  1. Burnup calculations using serpent code in accelerator driven thorium reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.

    2013-07-15

    In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)

  2. Testing of a burnup measuring prototype. Final report

    International Nuclear Information System (INIS)

    The analyses of gamma spectroscopy measurements of spherical fuel elements are reported, which have been performed in order to prove the feasibility of burnup measurement by way of Cs-137 spectroscopy. The detailed analysis and evaluation of the FRJ2-KA2) irradiation experiment carried out in the DIDO reactor supplied clear evidence that it is possible to measure an HTR-Modul fuel pebble with a target burnup ob 80,000 MWd/t after a decay time of 55 h, within a period of 10 seconds and with an accuracy of <5%, with 1 σ confidence, and that there is no need for developing measuring instruments with a higher counting rate. There are commercial peak unfolding programs available. (orig./HP)

  3. Transient behaviour of high burnup fuel. Status report

    International Nuclear Information System (INIS)

    This Status Report is a follow-on to the CSNI Specialist Meeting on Transient Behaviour of High Burnup Fuel which was held in Cadarache, France, from September 12. to 14., 1995. The Status Report identifies the needs and rationale for any further work to better understand the transient behaviour of high burnup fuel. The different options to perform that work, from analytical to experimental activities, and discussion on the potential benefits of performing new integral tests are also addressed. A brief description of the major on-going and short-term planned activities in this field is included as additional information. The main conclusions from this effort are highlighted. (K.A.)

  4. OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN

    International Nuclear Information System (INIS)

    1 - Description of program or function: In OREST, the 1-dimensional lattice code HAMMER and the isotope generation and depletion code ORIGEN are directly coupled for burnup simulation in light-water reactor fuels (GRS recommended). Additionally heavy water and graphite moderated systems can be calculated. New version differs from the previous version in the following features: An 84-group-library LIB84 for up to 200 isotopes is used to update the 3-group -POISON-XS. LIB84 uses the same energy boundaries as THERMOS and HAMLET in . In this way, high flexibility is achieved in very different reactor models. The coupling factor between THERMOS and HAMLET is now directly transferred from HAMMER to THERES and omits the equation 4 (see page 6 of the manual). Sandwich-reactor fuel reactivity and burnup calculations can be started with NGEOM = 1. Thorium graphite reactivity and burnup calculations can be started with NLIBE = 1. High enriched U-235 heavy water moderated reactivity and burnup calculations can be started. HAMLET libraries in for U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-242, Am-241, Am-243 and Zirconium are updated using resonance parameters. NEA-1324/04: A new version of the module hamme97.f has replaced the old one. 2 - Method of solution: For the user-defined irradiation history, an input data processor generates program loops over small burnup steps for the main codes HAMMER and ORIGEN. The user defined assembly description is transformed to an equivalent HAMMER fuel cell. HAMMER solves the integral neutron transport equation in a four-region cylindrical or sandwiched model with reflecting boundaries and runs with fuel power calculated rod temperatures. ORIGEN runs with HAMMER-calculated cross sections and neutron spectra and calculates isotope concentrations during burnup by solving the buildup-, depletion- and decay-chain equations. An output data processor samples the outputs of the program modules and generates tabular works for the

  5. OREST - The hammer-origen burnup program system

    International Nuclear Information System (INIS)

    Reliable prediction of the characteristics of irradiated light water reactor fuels (e.g., afterheat power, neutron and gamma radiation sources, final uranium and plutonium contents) is needed for many aspects of the nuclear fuel cycle. Two main problems must be solved: the simulation of all isotopic nuclear reactions and the simulation of neutron fluxes setting the reactions in motion. In state-of-the-art computer techniques, a combination of specialized codes for lattice cell and burnup calculations is preferred to solve these cross-linked problems in time or burnup step approximation. In the program system OREST, developed for official and commercial tasks in the Federal Republic of Germany nuclear fuel cycle, the well-known codes HAMMER and ORIGEN and directly coupled with a fuel rod temperature module

  6. Methods For The Calculation Of Pebble Bed High Temperature Reactors With High Burnup Plutonium And Minor Actinide Based Fuel

    International Nuclear Information System (INIS)

    The graphite moderated Modular High Temperature Pebble Bed Reactor enables very flexible loading strategies and is one candidate of the Generation IV reactors. For this reactor fuel cycles with high burnup (about 600 MWd/kg HM) based on plutonium (Pu) and minor actinides (MA) fuel will be investigated. The composition of this fuel is defined in the EU-PuMA-project which aims the reduction of high level waste. There exist nearly no neutronic full core calculations for this fuel composition with high burnup. Two methods (deterministic and Monte Carlo) will be used to determine the neutronics in a full core. The detailed results will be compared with respect to the influence on criticality and safety related parameters. (authors)

  7. Application of depletion perturbation theory to fuel cycle burnup analysis

    International Nuclear Information System (INIS)

    Over the past several years static perturbation theory methods have been increasingly used for reactor analysis in lieu of more detailed and costly direct computations. Recently, perturbation methods incorporating time dependence have also received attention, and several authors have demonstrated their applicability to fuel burnup analysis. The objective of the work described here is to demonstrate that a time-dependent perturbation method can be easily and accurately applied to realistic depletion problems

  8. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

    International Nuclear Information System (INIS)

    The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for

  9. The implementation of burnup credit in VVER-440 spent fuel

    International Nuclear Information System (INIS)

    The countries using Russian reactors VVER-440 cooperate in reactor physics in Atomic Energy Research (AER). One of topic areas is 'Physical Problems of Spent Fuel, Radwaste and Decommissioning' (Working Group E). In this article, in the first part is an overview about our activity for numerical and experimental verification of codes which participants use for calculation of criticality, isotopic concentration, activity, neutron and gamma sources and shielding is shown. The set of numerical benchmarks (CB1, CB2, CB3 and CB4) is very similar (the same idea, the VVER-440) to the OECD/NEA/NSC Burnup Credit Criticality Benchmarks, Phases 1 and 2. In the second part, verification of the SCALE 4.4 system (only criticality and nuclide concentrations) for VVER-440 fuel is shown. In the third part, dependence of criticality on burnup (only actinides and actinides + fission products) for transport cask C30 with VVER-440 fuel by optimal moderation is shown. In the last part, current status in implementation burnup credit in Slovakia is shown. (author)

  10. Value of 236U to actinide-only burnup credit

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) submitted a topical report to the US Nuclear Regulatory Commission (NRC) in May 1995 in order to gain approval of a method for criticality analysis of transport packages that takes account for the change in actinide isotopes with burnup [pressurized water reactors (PWRs) only]. Historically, the NRC has conservatively assumed that the fuel was in its initial conditions (without any burnable absorbers). In order to permit credit for the changes in actinide content, the NRC has required validation of the depletion and criticality codes for spent nuclear fuel, justification of conservative depletion modeling, and finally confirmation measurements before loading. The NRC requested additional information on March 22, 1996. The DOE responded by a revision of the topical report in May 1997. The NRC again responded with another set of requests of additional information in April 1998. In that set of questions, the NRC challenged the use of 236U in burnup credit. Uranium-236 is not found in any significant amount in any available critical experiments. The authors explore the value of 236U to actinide-only burnup credit

  11. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

    Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel

  12. Evolution of the ELESTRES code for application to extended burnups

    International Nuclear Information System (INIS)

    The computer code ELESTRES is frequently used at Atomic Energy of Canada Limited to assess the integrity of CANDU fuel under normal operating conditions. The code also provides initial conditions for evaluating fuel behaviour during high-temperature transients. This paper describes recent improvements in the code in the areas of pellet expansion and of fission gas release. Both of these are very important considerations in ensuring fuel integrity at extended burnups. Firstly, in calculations of pellet expansion, the code now accounts for the effect of thermal stresses on the volume of gas bubbles at the boundaries of UO2 grains. This has a major influence on the expansion of the pellet during power-ramps. Secondly, comparisons with data showed that the previous fission gas package significantly underpredicted the fission gas release at high burnups. This package has now been improved via modifications to the following modules: distance between neighbouring bubbles on grain boundaries; diffusivity; and thermal conductivity. The predictions of the revised version of the code show reasonable agreement with measurements of ridge strains and of fission gas release. An illustrative example demonstrates that the code can be used to identify a fuel design that would: reduce the sheath stresses at circumferential ridges by a factor of 2-10; and keep the gas pressure at very high burnups to below the coolant pressure

  13. The REBUS experimental programme for burn-up credit

    International Nuclear Information System (INIS)

    An international programme called REBUS for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Centre SCK·CEN and Belgonucleaire with the support of EdF and IRSN from France and VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. Recently also ORNL from the U.S. jointed the programme. The programme aims to establish a neutronic benchmark for reactor physics codes in order to qualify the codes for calculations of the burn-up credit. The benchmark exercise investigate the following fuel types with associated burn-up: reference fresh 3.3% enriched UO2 fuel, fresh commercial PWR UO2 fuel and irradiated commercial PWR UO2 fuel (54 GWd/tM), fresh PWR MOX fuel and irradiated PWR MOX fuel (20 GWd/tM). The experiments on the three configurations with fresh fuel have been completed. The experiments show a good agreement between calculation and experiments for the different measured parameters: critical water level, reactivity effect of the water level and fission-rate and flux distributions. In 2003 the irradiated BR3 MOX fuel bundle was loaded into the VENUS reactor and the associated experimental programme was carried out. The reactivity measurements in this configuration with irradiated fuel show a good agreement between experimental and preliminary calculated values. (author)

  14. Observations on the CANDLE burn-up in various geometries

    International Nuclear Information System (INIS)

    We have looked at all geometrical conditions under which an auto catalytically propagating burnup wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning 'spaghetti' reactor and the azimuthally burning 'pancake' reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometry, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed. A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology

  15. CEA contribution to power plant operation with high burnup level

    International Nuclear Information System (INIS)

    High level burnup in PWR leads to investigate again the choices carried out in the field of fuel management. French CEA has studied the economic importance of reshuffling technique, cycle length, discharge burnup, and non-operation period between two cycles. Power plants operators wish to work with increased length cycles of 18 months instead of 12. That leads to control problems because the core reactivity cannot be controlled with the only soluble boron: moderator temperature coefficient must be negative. With such cycles, it is necessary to use burnable poisons and for economic reasons with a low penalty in end of cycle. CEA has studied the use of Gd2O3 mixed with fuel or with inert element like Al2O3. Parametric studies of specific weights, efficacities relatively to the fuel burnup and the fuel enrichment have been carried out. Particular studies of 1 month cycles with Gd2O3 have shown the possibility to control power distribution with a very low reactivity penalty in EOC. In the same time, in the 100 MW PWR-CAP, control reactivity has been made with large use of gadolinia in parallel with soluble boron for the two first cycles

  16. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  17. Pore pressure calculation of the UO2 high burnup structure

    International Nuclear Information System (INIS)

    Highlights: • Pore pressure is calculated based on local burnup, density and porosity. • Ronchi's equations of state are used instead of van der Waals’ equation. • Pore pressure increases as HBS transformation begins and then stays constant. • A best approximated parameter used for pore pressure calculation is recommended. -- Abstract: UO2 high burnup structure has an important impact on fuel behavior, especially in case of reactivity initiated accident (RIA). Pore relaxation enhances local fuel swelling and puts additional load to the fuel cladding, which makes fuel more susceptible to pellet–cladding mechanical interaction induced failure. Therefore, pore pressure calculation becomes vital when evaluating the fuel failure. In this paper pore pressure is calculated as a function of pellet radial local burnup based on the basic characteristics of HBS using Ronchi's correlation. The results indicate that pore pressure will approach a stable value as HBS is developing. A best approximated C value of 55 N/m is recommended for pore pressure calculation

  18. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  19. Determination of densities from chemical composition and X-Ray diffraction

    International Nuclear Information System (INIS)

    X-ray diffraction method applied to retained austenite measurements gives volume per cent results, whereas the same kind of measurement made by Moessbauer Effect gives iron percentages. To compare both results one needs to convert the volume % to weight % or vice-versa. This necessitates, among other things, in determining the densities of the α and #betta# phases in the steel being studied. A method for calculating the densities, based on the application of the definition of density to just one unit cell, using X-ray diffraction and chemical results, are described. (Author)

  20. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  1. Practical issues with implementation of burnup credit in the USA for storage and transportation

    International Nuclear Information System (INIS)

    The US NRC issued an interim staff guidance (ISG8 rev1) allowing for burnup credit applications for storage and transport casks in July of 1999. In over two and a half years there has still not been a license submittal using burnup credit. ISG8 rev1 does not provide sufficient burnup credit to allow loading of 5 wt% enriched fuel in a 32 PWR assembly cask without the addition of absorber rod inserts. Pressure to allow all assemblies to contain inserts from the utility, force continued investigation into alternative levels of burnup credit. Utilities do not wish to measure to confirm burnup. This measurement costs, which range form $10 000 to $50 000 per cask and must be done prior to loading. Since burnup credit is actually only needed for transport, and transport is not expected for many years, many utilities are considering keeping the money in the bank until the time of transport. In order to address the need perceived for additional burnup credit beyond actinide-only burnup credit (ISG8), investigations have moved beyond into assuming moderator exclusion during transport and the use of burnup credit to cover a beyond design basis accident assumption of flooding. Burnup credit analysis requirements for a beyond design basis accident should be less than that for criticality control for normal operation. It is proposed that burnup credit analysis to cover the beyond design basis accident of flooding should be consistent with the beyond design basis dilution event in PWR spent fuel pools. The US NRC precedence for this type of burnup credit allows for all isotopes, a 5% reduction in the delta k of burnup, and an allowable keff of less than 1.0 after biases and uncertainties. (author)

  2. Biological and chemical tests of contaminated soils to determine bioavailability and environmentally acceptable endpoints (EAE)

    International Nuclear Information System (INIS)

    The understanding of the concept of bioavailability of soil contaminants to receptors and its use in supporting the development of EAE is growing but still incomplete. Nonetheless, there is increased awareness of the importance of such data to determine acceptable cleanup levels and achieve timely site closures. This presentation discusses a framework for biological and chemical testing of contaminated soils developed as part of a Gas Research Institute (GRI) project entitled ''Environmentally Acceptable Endpoints in Soil Using a Risk Based Approach to Contaminated Site Management Based on Bioavailability of Chemicals in Soil.'' The presentation reviews the GRI program, and summarizes the findings of the biological and chemical testing section published in the GRI report. The three primary components of the presentation are: (1) defining the concept of bioavailability within the existing risk assessment paradigm, (2) assessing the usefulness of the existing tests to measure bioavailability and test frameworks used to interpret these measurements, and (3) suggesting how a small selection of relevant tests could be incorporated into a flexible testing scheme for soils to address this issue

  3. An evaluation of chemical shift index-based secondary structure determination in proteins: Influence of random coil chemical shifts

    Energy Technology Data Exchange (ETDEWEB)

    Mielke, S.P.; Krishnan, V.V. [Biophysics Graduate Group, University of California, Davis (United States)], E-mail: krish@llnl.gov

    2004-10-15

    Random coil chemical shifts are commonly used to detect protein secondary structural elements in chemical shift index (CSI) calculations. Though this technique is widely used and seems reliable for folded proteins, the choice of reference random coil chemical shift values can significantly alter the outcome of secondary structure estimation. In order to evaluate these effects, we present a comparison of secondary structure content calculated using CSI, based on five different reference random coil chemical shift value sets, to that derived from three-dimensional structures. Our results show that none of the reference random coil data sets chosen for evaluation fully reproduces the actual secondary structures. Among the reference values generally available to date, most tend to be good estimators only of helices. Based on our evaluation, we recommend the experimental values measured by Schwarzinger et al. (2000), and statistical values obtained by Lukin et al. (1997), as good estimators of both helical and sheet content.

  4. Study on Determination of Chemical Oxygen Demand in Water with Ion Chromatography

    Institute of Scientific and Technical Information of China (English)

    ZHANG Zhong-Hai; DING Hong-Chun; FANG Yan-Ju; XIAN Yue-Zhong; JIN Li-Tong

    2007-01-01

    A new method for determining chemical oxygen demand (COD) value in water using ion chromatography coupled with nano TiO2-K2S2O8 co-existing system was described. The photocatalytic oxidation system and nano TiO2-K2S2O8 co-existing system could degrade the organic compounds in water. All sulfur-containing species in the reactive solution were eventually transformed to sulfate which could be determined by conductivity detector in ion chromatography. The change of conductivity of sulfate was proportional to COD value. The optimal experimental conditions and the mechanism of the detection were discussed. The application range was 10.0-300.0 mg·L -1 and the lowest limit of detection was 3.5 mg·L -1. It was considered that the value obtained could be reliably correlated with the COD value obtained using the conventional methods.

  5. Chemical separation and AES determination of rare earths in thorium oxide

    International Nuclear Information System (INIS)

    A chemical separation method has been developed for the separation of rare earths like, Ce, Sm, Eu, Gd, Tb, Dy, Ho, Er, Yb, Lu and La, Y from ThO2 matrix and their determination by emission spectrographic method. Cyanex-272 ie. [bis(2,4,4-trimethyl pentyl) phosphinic acid] /xylene/ HNO3 extraction system has been used for separation of thorium. The recovery of rare earths as determined by emission spectrographic method was found to be quantitative within experimental error. The estimation range for the analytes lie between 0.02μg-4μg based on 100 mg ThO2. (author). 3 refs., 1 tab

  6. Rapid determination of chemical oxygen demand (cod) using microwave digestion followed by titrimetery

    International Nuclear Information System (INIS)

    Chemical oxygen demand (COD) is an important parameter in water-pollution control analysis. It is closely related to the organic contamination level of wastewater. The open-reflux Cr-COD titration method has long been the International standard method for COD determination /sup(1)/ however, it still requires a long time (2h) for the digestion step to be completed in this procedure. The objective of this research was to investigate the availability of a rapid procedure to determine the COD in wastewater by digesting samples in a microwave system followed by ferrous ammonium sulphate titration. The results obtained using this microwave aided digestion system were also compared with those obtained by conventional open reflux method. (author)

  7. OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

    International Nuclear Information System (INIS)

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  8. OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  9. Determination of chemical composition and shelf life of shad (Alosa tanaica Grimm, 1901 in refrigeration conditions

    Directory of Open Access Journals (Sweden)

    Hünkar Avni Duyar

    2012-01-01

    Full Text Available This study, was carried out to determine the shelf life and chemical composition of stored shad (Alosa tanaica Grimm, 1901 in refrigerator conditions (4 ±0.5° C. Crude fat, crude ash, crude protein and moisture were 13 ±0.5%, 1.3 ±0.4%, 17 ±0.2%, 68 ±0.6% at the begining of the fresh shad, respectively. The quality of shad fish during storage were evaluated by pH, TVB-N, TBA, sensory and microbiological analysis. According to the results of sensory analysis, be-ginning to lose the consumability property after the 4th day and inconsumable at 7th day of sto¬red shad fish were determined by panelists. As a result of chemical analysis the amount of TVB-N at 0. day 7.4 ± 0.1 mg/100g and 7th day of stored was determined 37.2 ±0.4 mg/100 g, TBA value at 0. days 2.15 ± 0.3 mg /kg MDA and 7. day 15.22 ± 0.9 mg /kg determined as malondialdehyde, the pH value at 6. day was 7.3 respectively. Made as a result of microbiolo-gical analysis of bacterial load at 0. day 0.47 ± 0.09 log CFU /g and 6. day 6.3 ± 0.1 log CFU/g, respectively, and consumption was found to exceed the limit value. In a survey of con-ditions as a result of the refrigerator (4 ± 0.5°C to maintain the shad (Alosa tanaica Grimm, 1901, the shelf life of 6 days.

  10. The structure determination of Al20Cu2Mn3 by near atomic resolution chemical mapping

    International Nuclear Information System (INIS)

    Highlights: • The structure of Al20Cu2Mn3 with a space group of Bbmm is completely determined. • The actual formula of Al20Cu2Mn3 is Al31Cu3Mn5. • Al20Cu2Mn3 is formed by a parallel tessellation of hexagon subunits. • Al20Cu2Mn3 is prone to twinning by an alternate tessellation of hexagon subunits. • The Al20Cu2Mn3 is coherent with α-Al along its longitudinal axis. - Abstract: Al20Cu2Mn3 phase is one kind of common dispersoids in aluminum alloys; however, the atomic arrangement of Al20Cu2Mn3 has not yet been clearly identified. Combining the atomic resolution high angle annular dark field and chemical composition quantitative results, three structure models of Al20Cu2Mn3 were derived basing on the isostructural Mn11Ni4Al60. The formation enthalpies and total energy were calculated using the first-principles approach. The structure of the Al20Cu2Mn3 phase with the minimal energy was identified, giving a fully relaxed structure with lattice parameters of a = 23.98 Å, b = 12.54 Å, c = 7.66 Å, which belongs to a space group of Bbmm. The determined structure is in excellent agreement with the near atomic resolution chemical mapping results

  11. Challenges and trends in the determination of selected chemical contaminants and allergens in food.

    Science.gov (United States)

    Krska, Rudolf; Becalski, Adam; Braekevelt, Eric; Koerner, Terry; Cao, Xu-Liang; Dabeka, Robert; Godefroy, Samuel; Lau, Ben; Moisey, John; Rawn, Dorothea F K; Scott, Peter M; Wang, Zhongwen; Forsyth, Don

    2012-01-01

    This article covers challenges and trends in the determination of some major food chemical contaminants and allergens, which-among others-are being monitored by Health Canada's Food Directorate and for which background levels in food and human exposure are being analyzed and calculated. Eleven different contaminants/contaminant groups and allergens have been selected for detailed discussion in this paper. They occur in foods as a result of: use as a food additive or ingredient; processing-induced reactions; food packaging migration; deliberate adulteration; and/or presence as a chemical contaminant or natural toxin in the environment. Examples include acrylamide as a food-processing-induced contaminant, bisphenol A as a food packaging-derived chemical, melamine and related compounds as food adulterants and persistent organic pollutants, and perchlorate as an environmental contaminant. Ochratoxin A, fumonisins, and paralytic shellfish poisoning toxins are examples of naturally occurring toxins whereas sulfites, peanuts, and milk exemplify common allergenic food additives/ingredients. To deal with the increasing number of sample matrices and analytes of interest, two analytical approaches have become increasingly prevalent. The first has been the development of rapid screening methods for a variety of analytes based on immunochemical techniques, utilizing ELISA or surface plasmon resonance technology. The second is the development of highly sophisticated multi-analyte methods based on liquid chromatography coupled with multiple-stage mass spectrometry for identification and simultaneous quantification of a wide range of contaminants, often with much less requirement for tedious cleanup procedures. Whereas rapid screening methods enable testing of large numbers of samples, the multi analyte mass spectrometric methods enable full quantification with confirmation of the analytes of interest. Both approaches are useful when gathering surveillance data to determine

  12. Burn-up effect on instant release from an initial corrosion of UO2 and MOX fuel under anoxic conditions

    International Nuclear Information System (INIS)

    The objective of the work is to obtain instant release experimental values for different radionuclides as a function of spent fuel type (UO2 and MOX) and burn-up (from 30 to 63 MWd/kgU) that will be useful for the performance assessment studies related to the behaviour of spent fuel under repository conditions or, in any case, spent fuel conditions in which labile radionuclides can be released. To determine the instant release source terms, sets of leaching experiments were conducted with spent UO2 and MOX fuel with burnups ranging from 30 to 63 MWd/kg U in presence of cladding as the container material. The fuels were leached in carbonated groundwater (CW) having a buffered pH of 7.5 at room temperature. Some observations are also made of the differences in matrix dissolution behaviour of the different fuels based on observed U, Pu and Np concentrations The ultimate issue is to evaluate the differences in the ''instant'' inventory measurements for spent fuels in order to provide experimental data that allow to evaluate the source terms used in the safety-assessment calculations, and to improve the accuracy of such data for the future. It is important to remark that the quality of the experimental results obtained describes the influence of the spent fuel (SF) burn-up on fast release of inventory fraction (release under 200 days). (authors)

  13. Taking burnup credit into account in criticality studies: the situation as it is now and the prospect for the future

    International Nuclear Information System (INIS)

    As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For that purpose, a method was developed considering: (i) partial Uranium-and-Plutonium burnup credit in the criticality studies, and (ii) a conservative assumption concerning the axial profile; this actinides-only method was supported by an experimental program called HTC. The method was accepted by the French Safety Authority. Moreover, in order to reduce again the calculated values of the reactivity for irradiated fuels, a French working group was set up in 1997 to define a conservative method which enables industrial companies to take burnup credit into account with some of the fission products and using a more precise profile. The work of this group has been divided into four tasks related to: the determination of (i) the composition of the fuel, (ii) a conservative profile, (iii) a conservative irradiation history, and (iv) the calculation scheme. This work is also supported by experimental programs related to the validation of the fission products effects, in terms of reactivity

  14. Results of post-irradiation examination to validate WWER-440 and WWER-1000 fuel efficiency at high burnups

    International Nuclear Information System (INIS)

    During the last 10 years on the basis of commercial operation of WWER reactors, a conversion from three to four year fuel cycle operation has been succeeded for WWER-440 fuel. This paper presents the examinations of fuel rods and fuel assemblies operated at different NPPs of Russia and Eastern Europe. Three WWER-440 fuel assemblies with different burnups and different irradiation in the core (3, 4 and 5 years fuel cycle) have passed full-scale examinations including both: destructive and non-destructive methods. The results of examinations have revealed that the irregularity of the field of the energy release may result in increased cladding oxidation. A validation of WWER-440 and WWER-1000 fuel efficiency during 4 and 5 years fuel cycles is also made on the basis of assessment of fuel rod and assembly mechanical state and changes in their geometry. The status of the fuel column including grain size, fuel swelling, rim layer and fission gas release depending on fuel burnup are investigated. During examinations mechanical properties and oxidation of the cladding, mechanical and corrosion state of the spacer grid, ultimate stress, hardness and plasticity of central tube and relaxation of spring unit are studied at the maximal fuel burnup 64 MWd/kgU for WWER-440 and 58 MWd/kgU for WWER-1000. Based on the examination results for the principal parameters determined fuel resource (variation in form, material structure and properties, corrosion resistance of the claddings, FGR form the fuel, fuel cladding interaction degree) reliable fuel operation at the burnups corresponding to four and five fuel cycles may be predicted. None of fuel efficiency factors in up-to date FA design are limited for operation during five fuel cycles

  15. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

  16. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Russian Research Centre, Moscow (Russian Federation)

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  17. Research on Integrity of High Burnup Spent Fuel Under the Long Term Dry Storage

    International Nuclear Information System (INIS)

    Objectives were to acquire the following behaviour data by dynamic load impact tests on high burnup spent fuel rods of BWR and PWR and to improve the guidance of regulation of spent fuel storage and transportation. (1) The limit of load and strain for high burnup fuel in the cask drop accident. (2) The amount of deformation of high burnup fuel rods under dynamic load impact. (3) The amount of fuel pellet material released from fuel rods under dynamic load impact

  18. Nuclear fuel burn-up credit for criticality safety justification of spent nuclear fuel storage systems

    International Nuclear Information System (INIS)

    Burn-up credit analysis of RBMK-1000 an WWER-1000 spent nuclear fuel accounting only for actinides is carried out and a method is proposed for actinide burn-up credit. Two burn-up credit approaches are analyzed, which consider a system without and with the distribution of isotopes along the height of the fuel assembly. Calculations are performed using SCALE and MCNP computer codes

  19. Using Quantitative Reverse Transcriptase PCR and Cell Culture Plaque Assays to Determine Resistance of Toxoplasma gondii Oocysts to Chemical Sanitizers

    Science.gov (United States)

    Toxoplasma gondii oocysts are highly resistant to many chemical sanitizers. Current methods used to determine oocyst infectivity have relied exclusively on mouse, chicken, and feline bioassays. Although considered gold standards, they only provide a qualitative assessment of oocyst infectivity. I...

  20. Study on burn-up credit and minor actinide in post-irradiation analysis

    International Nuclear Information System (INIS)

    Accuracy of burnup calculation for actinide is very important as to the study of burn-up credit. For minor-actinides such as Am243 and Cm244, however, typical burnup calculation codes are not accurate enough. The accuracy for both nuclides was studied by using the SWAT code. The study showed that the C/E values of both nuclides could be improved at the same time by changing the cross section of Pu242. A study of burnup calculation related to the cross section of Pu242 should be performed to improve the accuracy for both nuclides. (author)

  1. Results of the isotopic concentrations of VVER calculational burnup credit benchmark No. 2(CB2)

    International Nuclear Information System (INIS)

    Results of the nuclide concentrations are presented of VVER Burnup Credit Benchmark No. 2(CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is summarized. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2' code and other codes were used for the calculation of the spent fuel concentration. (author)

  2. Effect of fuel burnup history on neutronic characteristics of WWER-1000 core

    International Nuclear Information System (INIS)

    The paper analyzes fuel burnup history effect on neutronic characteristics of WWER-1000 core with use of the DYN3D codes. The DYN3D code employs the local Pu-239 concentration as an indicator of burnup spectral history. The calculations have been performed for the first four fuel loadings of Khmelnitsky NPP unit 2 and stationary fuel loading with TVSA. The effect of fuel burnup history is shown both on macro-characteristics on the reactor core and on local values of burnup and power

  3. LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System

    International Nuclear Information System (INIS)

    1 - Description of program or function: The LOLA-SYSTEM is a part of the JEN-UPM code package for PWR fuel management, scope or design calculations. It is a code package for core burnup calculations using nodal theory based on a FLARE type code. The LOLA-SYSTEM includes four modules: the first one (MELON-3) generates the constants of the K-inf and M2 correlations to be input into SIMULA-3. It needs the K-inf and M2 fuel assembly values at different conditions of moderator temperature, Boron concentration, burnup, etc., which are provided by MARIA fuel assembly calculations. The main module (SIMULA-3) is the core burnup calculation code in three dimensions and one group of energy. It normally uses a geometrical representation of one node per fuel assembly or per quarter of fuel assembly. It has included a thermal hydraulic feedback on flow and voids and criticality searches on boron concentration and control rods insertion. The CONCON code makes the calculation of the albedo, transport factors, K-inf and M2 correction factors to be input into SIMULA-3. The calculation is made in the XY transversal plane. The CONAXI code is similar to CONCON, but in the axial direction. 2 - Method of solution: MELON-3 makes a mean squares fit of K-inf and M2 values at different conditions in order to determine the constants of the feedback correlations. SIMULA-3 uses a modified one-group nodal theory, with a new transport kernel that provides the same node interface leakages as a fine mesh diffusion calculation. CONCON and CONAXI determine the transport and correction factors, as well as the albedo, to be input into SIMULA-3. They are determined by a method of leakages equivalent to the detailed diffusion calculation of CARMEN or VENTURE; these factors also include the heterogeneity effects inside the node. 3 - Restrictions on the complexity of the problem: Number of axial nodes less than or equal 34. Number of material types less than or equal 30. Number of fuel assembly types less

  4. Reevaluation of fuel enthalpy in NSRR test for high burnup fuels

    International Nuclear Information System (INIS)

    This paper describes the recent procedure of evaluation of the fuel enthalpy in the reactivity initiated accident (RIA) simulating tests performed at the nuclear safety research reactor (NSRR), and reports some important updates of the fuel enthalpies in the tests with high burnup PWR fuels. Previously, the fuel enthalpy had been evaluated by the procedure based on the short-life fission product measurement, i.e. a pellet slice was sampled from the test fuel rod after the NSRR test, a chemical separation process was applied to the solution of the pellet slice to separate barium, and the amount of Ba-140 was measured by gamma spectrometry on the separated barium. But a part of the results showed significant scattering even within the similar tests with similar fuels, which should have showed similar fuel enthalpies. The scattering appears to indicate the difficulty in treatment of the short-life nuclides after the completion of the NSRR test and unsuccessful measurement of the amount of fuel dissolved in the specimen preparation. Another difficulty of the procedure is that it is not repeatable for a specimen and so double check of an evaluation is not possible. Hence, an alternative procedure, which is based on the total amount of fissile materials evaluated by mass analysis, was developed and has been applied for the tests after 2003; the amount of fissile materials is input to a well-verified neutron transport calculation model for the NSRR reactor core to calculate a coupling factor of power densities between the test fuel rod and the NSRR driver fuel rods. This procedure does not require quickness and is repeatable, so it is applicable even many years later if the fuel sample is available. The recent procedure was thus applied to the tests before 2003, whose burnups are below 60 GWd/tU. It was shown that the fuel enthalpy had been significantly underestimated in the tests with high burnup PWR fuels: the test series HBO and TK. In this paper, the procedure

  5. First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

    International Nuclear Information System (INIS)

    The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C-E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA-AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies. (author)

  6. Determination of chemical composition of commercial honey by near-infrared spectroscopy.

    Science.gov (United States)

    Qiu, P Y; Ding, H B; Tang, Y K; Xu, R J

    1999-07-01

    The feasibility of using near-infrared spectroscopy to determine chemical composition of commercial honey was examined. The influences of various sample presentation methods and regression models on the performance of calibration equations were also studied. Transmittance spectra with 1 mm optical path length produced the best calibration for all constituents examined. The regression model of modified partial least squares (mPLS) was selected for the calibration of all honey constituents except moisture, for which the optimal calibration was developed with PLS. Validation of the established calibration equations with independent samples showed that the spectroscopic technique could accurately determine the contents of moisture, fructose, glucose, sucrose, and maltose with squared correlation coefficients (R(2)) of 1.0, 0.97, 0.91, 0.86, and 0.93 between the predicted values and the reference values. The prediction accuracy for free acid, lactone, and hydroxymethylfurfural (HMF) contents in honey was poor and unreliable. The study indicates that near-infrared spectroscopy can be used for rapid determination of major components in commercial honey. PMID:10552561

  7. A combined photocatalytic determination system for chemical oxygen demand with a highly oxidative reagent

    International Nuclear Information System (INIS)

    This study focuses on the proposal and validation of a combined photocatalytic (PC) system and a three-parameterized procedure for the determination of chemical oxygen demand (COD; PcCODcombined), with a highly oxidative reagent utilized as a photoelectron scavenger and signal indicator. The PcCODcombined was the functional combination of photon-efficient thin-layer photocatalytic oxidation, conventional bulk-phase photocatalytic oxidation and photocarrier-efficient high-activity photocatalytic reduction in one single photodigestion system, and consequently, this system possessed high photon-utilization efficiency, automatic stirring function and satisfactory determination characteristics. In comparison with the conventional one-parameterized procedure, the three-parameterized procedure introduces the blank and total photocatalytic reduction responses as two of the three significant analytical parameters. Under the optimized pH value of 3.0-4.5 and a rotating rate of 40 rpm, the representative KMnO4 species was used for the PcCODcombined system as the combined high-activity oxidant, and a narrow and reliable analytical linear range of 0-260 mg L-1 was achieved during the 10 min duration of the determinations. No observable interference of Cl- was found at concentration of the ion up to 2000 mg L-1. A real sample analysis indicated that the measured values for the PcCODcombined were all within a relative deviation below 5% of CODCr of the standard method, which further validates the practical feasibility of the proposed PcCODcombined system.

  8. Determination of chemical elements in Eucalyptus grandis, manured with Ballad's, by neutrons activation analysis

    International Nuclear Information System (INIS)

    The biosolid is a mud resulting from the biological treatment of wasted liquids. It is considered as a profitable alternative and important to minimize the environmental impact generated by the sewage thrown in to sanitary lands, in forest cultures like the Eucalyptus grandis. The objective of this work was to detect which chemical elements are present in Eucalyptus grandis samples, fertilized with different quantities of biosolid. The eucalyptuses of Estacao Experimental de Ciencias Florestais of Itatinga were planted in March of 1998 and collected with five years old. The used biosolid was produced by Station of Treatment of Sewer of Barueri - SP, classified as kind B. For the determination of the presence and quantity of chemical elements in the eucalyptus samples, an analysis technique by neutronic activation (NAA) was used followed by gamma rays spectroscopy. The samples were irradiated in the Nuclear Reactor IEA-R1 of IPEN-SP, followed by the measure of induced gamma rays activity, using a Detector HPGe. The presence, mainly of Br, Mn, Na and K, was detected in all analyzed samples. (author)

  9. Determinants of exposure to chemical pollutants in wet X-ray film processing in Iran.

    Science.gov (United States)

    Kakooei, Hossein; Ardakani, Mehdi B; Sadighi, Alireza

    2007-07-15

    The aim of the current study was to measure glutaraldehyde, acetic acid and sulfur dioxide and levels inside wet x-ray processing areas in a developing country and comparing data with those in developed countries. Forty-five radiographers from 10 educational hospitals affiliated to the Tehran University of Medical Sciences (TUMS) in Tehran, Iran participated in this descriptive-analytical study. Exposure to glutaraldehyde (a constituent of developer chemistry), acetic acid (a constituent of fixer chemistry) and sulfur dioxide (a byproduct of sulfites present in both developer and fixer solutions) was measured in all participants as well as area exposure. Average full-shift exposure to glutaraldehyde, acetic acid and sulfur dioxide were 0.0018, 2.65 and 1.64 mg m(-1), respectively. The results showed that the TUMS radiographers full-shift exposures are generally lower than the American Conference of Governmental Industrial Hygienists (ACGIH) recommended levels. The concentration of glutaraldehyde collected by area sampling (darkroom) was almost five times (0.0104 mg m(-3)) greater than taken by personal sampling. Exposure to the chemical pollutants in the currents study were generally higher than in developed countries. Identification of these key exposure determinants is useful in targeting exposure evaluation and controls to reduce developer and fixer chemicals exposures in the radiology departments. Employing of a digital imaging system that do not involve wet x-ray processing of photographic film would be a useful device for radiographers protection. PMID:19070154

  10. Rapid determination of chemical oxygen demand using a focused microwave heating system featuring temperature control

    International Nuclear Information System (INIS)

    This paper demonstrates the use of a microwave heating system, employed in the chemical digestion step, for the determination of chemical oxygen demand in wastewater. The results are first compared with those provided by standard methods using reference substances. The problems arising from abrupt heating of the sample and the potential thermal decomposition of potassium dichromate are examined. Two different approaches to sample digestion involving a gradually increasing irradiation time were tested. First, a constant power strategy is applied, and the second proposes a constant temperature approach by using a temperature control system. By optimising the operating conditions, the digestion time was reduced to 8-60 times with respect to the standard method. The reference digestion time is 5 min. In especially difficult digestions, the proposed approach provides a substantially improved degradation with respect to conventional procedures. The procedure was applied to wastewater from various industries and found to ensure thorough digestion of all samples and to provide favourable results in all cases tested

  11. Application of X-ray fluorescence (WDXRF): thickness and chemical composition determination of thin films

    International Nuclear Information System (INIS)

    In this work a procedure is described for thickness and quantitative chemical composition of thin films by wavelength dispersion X-ray fluorescence (WDXRF) using Fundamental Parameters method. This method was validated according to quality assurance standard and applied sample Al, Cr, TiO2, Ni, ZrO2 (single thickness) and Ni/Cr (double thickness) on glass; Ni on steel and metallic zinc and TiO2 on metallic iron (single thickness), all the sample were prepared for physical deposition of vapor (PVD). The thickness had been compared with Absorption (FRX-A) and Rutherford Backscattering Spectrometry (RBS) methods; the result showed good efficiency of the fundamental parameters method. Sample structural characteristics analyzed by X ray diffraction (XRD) showed any influence in the thickness determinations. (author)

  12. Determining stellar atmospheric parameters and chemical abundances of FGK stars with iSpec

    CERN Document Server

    Blanco-Cuaresma, S; Heiter, U; Jofré, P

    2014-01-01

    Context. An increasing number of high-resolution stellar spectra is available today thanks to many past and ongoing extensive spectroscopic surveys. Consequently, the scientific community needs automatic procedures to derive atmospheric parameters and individual element abundances. Aims. Based on the widely known SPECTRUM code by R. O. Gray, we developed an integrated spectroscopic software framework suitable for the determination of atmospheric parameters (i.e., effective temperature, surface gravity, metallicity) and individual chemical abundances. The code, named iSpec and freely distributed, is written mainly in Python and can be used on different platforms. Methods. iSpec can derive atmospheric parameters by using the synthetic spectral fitting technique and the equivalent width method. We validated the performance of both approaches by developing two different pipelines and analyzing the Gaia FGK benchmark stars spectral library. The analysis was complemented with several tests designed to assess other ...

  13. Phase of the Fermion Determinant for QCD at Finite Chemical Potential

    CERN Document Server

    Splittorff, K

    2008-01-01

    In this lecture we discuss various properties of the phase factor of the fermion determinant for QCD at nonzero chemical potential. Its effect on physical observables is elucidated by comparing the phase diagram of QCD and phase quenched QCD and by illustrating the failure of the Banks-Casher formula with the example of one-dimensional QCD. The average phase factor and the distribution of the phase are calculated to one-loop order in chiral perturbation theory. In quantitative agreement with lattice QCD results, we find that the distribution is Gaussian with a width $\\sim \\mu T \\sqrt V$ (for $m_\\pi \\ll T \\ll \\Lambda_{\\rm QCD}$). Finally, we introduce, so-called teflon plated observables which can be calculated accurately by Monte Carlo even though the sign problem is severe.

  14. Determination of rare earth elements in Taiwan monazite by chemical neutron activation analysis

    International Nuclear Information System (INIS)

    Taiwan monazite is a unique mineral obtained from the heavy sand found in the river floor of Tzuo-suei river and En-suei river. Both rivers are flowing parallel with separated narrow area into the sea at southwestern coast of Taiwan. The characteristic of monazite is that it contains considerable rare earth elements (REEs). REEs are considered very useful elements in the local industries and scientific researches such as ceramic, semiconductors, and glass optics. In this study, chemical neutron activation analysis (CNAA) was used to determine the contents of REEs in Taiwan monazite. A few milligram of monazite was digested in the microwave oven for 25 minutes with mixed acid (conc. HNO3 and HClO4). REEs were preconcentrated by hydrated magnesium oxide and CNAA was performed. (author)

  15. Quantitative determinations of chemical compounds with nutritional value from Inca crops: Chenopodium quinoa ('quinoa').

    Science.gov (United States)

    González, J A; Roldán, A; Gallardo, M; Escudero, T; Prado, F E

    1989-12-01

    Quantitative determinations of total and soluble proteins, total and free sugars, starch, total lipids, tanins, ash (Ca, Na, K, Fe, and P), and caloric value were carried out on quinoa flour. Results show that the amount of soluble proteins was higher than the standard value for wheat and maize and was very close to that of barley's. The yield of free sugars like glucose (4.55%), fructose (2.41%) and sucrose (2.39%) were also of importance. Iron and calcium levels were higher than the reported values for maize and barley. The same occurred for the caloric value (435.5 Kcal/100 g). The content of saponins was also examined since its effect on red blood cells of group A and O has been related as a potential problem of the Andes population. From the chemical analysis a more complete view about quinoa as human food was presented. PMID:2631089

  16. Methods of RECORD, an LWR fuel assembly burnup code

    International Nuclear Information System (INIS)

    The RECORD computer code is a detailed rector physics code for performing efficient LWR fuel assembly calculations, taking into account most of the features found in BWR and PWR fuel designs. The code calculates neutron spectrum, reaction rates and reactivity as a function of fuel burnup, and it generates the few-group data required for use in full scale core simulation and fuel management calculations. The report describes the methods of the RECORD computer code and the basis for fundamental models selected, and gives a review of code qualifications against measured data. (Auth. /RF)

  17. Modelling of fission gas behaviour in high burnup nuclear fuel

    International Nuclear Information System (INIS)

    The safe and economic operation of nuclear power plants (NPPs) requires that the behaviour and performance of the fuel can be calculated reliably over its expected lifetime. This requires highly developed codes that treat the nuclear fuel in a general manner and which take into account the large number of influences on fuel behaviour, in particular the trend of NPP operators to increase the fuel burnup. With higher burnup, more fission events impact the material characteristics of the fuel and significant restructuring can be observed. At local burnups in excess of 60-75 MWd/kgU, the microstructure of nuclear fuel pellets differs markedly from the as-fabricated structure. This high burnup structure (HBS) is characterised by three principal features: 1) low matrix xenon concentration, 2) sub-micron grains and 3) a high volume fraction of micrometer-sized pores. The peculiar features of the HBS affect the fuel performance and safety; the large retention of fission gas within the HBS could lead to significant gas release at high burnups, either through the degradation of thermal conductivity or through direct release. The present work has focussed on the development and evaluation of HBS fission gas transport models, especially on two features: the equilibrium xenon concentration in the matrix of the HBS in UO2 fuel pellets, and the growth of the HBS porosity and its effect on fission gas release. A steady-state fission gas model has been developed to examine the importance of grain boundary diffusion for the gas dynamics in the HBS. It was possible to simulate the ∼0.2 wt% experimentally observed xenon concentration. The value of the grain boundary diffusion coefficient is not important for diffusion coefficient ratios in excess of ∼10”4. The model exhibits a high sensitivity to principally three parameters: the grain diffusion coefficient, the bubble number density and the re-solution rate coefficient. The model can reproduce the observed HBS xenon depletion

  18. The Design Method for the ATR High Burnup MOX Fuel

    International Nuclear Information System (INIS)

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has developed the advanced thermal reactor (ATR). PNC is demonstrating MOX fuel utilization in a prototype of ATR, Fugen (165 MWe), in which 638 MOX fuel assemblies have been loaded without a failure since 1979. PNC is developing the high burn-up MOX fuel for the ATR to contribute to MOX fuels for thermal reactors. The statistical design evaluation method that included the MOX fuel rod performance evaluation code 'FEMAXI-ATR' was developed for the ATR high bum-up MOX fuel rod; it was verified that the integrity of the fuel could be maintained over the whole irradiation period

  19. Development and verification of Monte Carlo burnup calculation system

    International Nuclear Information System (INIS)

    Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)

  20. New results from the NSRR experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide [Japan Atomic Research Institute, Toaki, Ibaraki (Japan)] [and others

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  1. OTTER 3 - A single channel, axial burnup code

    International Nuclear Information System (INIS)

    OTTER 3 is a single channel, axial burnup code, written in Fortran for the KDF 9 computer, and suitable for studying fuel management schemes of the continuous charge/discharge type. A general fuel shuffling scheme is allowed, and both unidirectional and bidirectional fuel feed can be studied. A 2-group neutron diffusion code is incorporated, the flux equations being solved by the forward elimination - backward substitution technique for the inner problem and a source iteration technique accelerated by Chebyshev extrapolation for the outer problem. (author)

  2. Implementation of burnup credit in spent fuel management systems. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report

  3. Simulation of fuel cycles with minor actinide management using a fast burnup calculation tool

    International Nuclear Information System (INIS)

    The paper presents a fast and flexible burnup model for fuel cycle simulations which is based on the description of the one-group cross-sections as analytic functions of the isotopic composition. This was accomplished by multi-dimensional regression based on the results of numerous core calculations. The developed model is able to determine the spent fuel composition in reasonable CPU time, and was integrated into a simplified fuel cycle model containing Gas Cooled Fast Reactors (GFR) and conventional light water reactors (LWRs). The fuel cycle simulations revealed an advantageous effect of increased minor actinide content in the GFR core on the fuel utilization parameters. In order to explore the processes that lay behind this effect the neutronics balance of the GFR was investigated in equilibrium cycle conditions. (author)

  4. Preliminary assessment of the benefits of derating a cask for increasing age/burnup capability

    International Nuclear Information System (INIS)

    This study was performed to determine the extent to which the age/burnup capability of the Babcock and Wilcox BR-100 rail cask could be extended by reducing the number of fuel assemblies. Since cask shielding was seen as the limiting design feature, the criterion used to assess the derating was the calculated dose 2 m from the rail car. The reference calculations were based on the 70% design of the BR-100 cask with 21 PWR fuel assemblies. Seven different basket/assembly loading configurations were investigated. The results indicate that both an alternate 18-assembly basket configuration and a 17-assembly/4-empty-hole configuration for the 21-element basket offer substantial gains over the fully loaded reference 21-element basket configuration

  5. Determination of cadmium in water samples by fast pyrolysis-chemical vapor generation atomic fluorescence spectrometry

    Science.gov (United States)

    Zhang, Jingya; Fang, Jinliang; Duan, Xuchuan

    2016-08-01

    A pyrolysis-vapor generation procedure to determine cadmium by atomic fluorescence spectrometry has been established. Under fast pyrolysis, cadmium ion can be reduced to volatile cadmium species by sodium formate. The presence of thiourea enhanced the efficiency of cadmium vapor generation and eliminated the interference of copper. The possible mechanism of vapor generation of cadmium was discussed. The optimization of the parameters for pyrolysis-chemical vapor generation, including pyrolysis temperature, amount of sodium formate, concentration of hydrochloric acid, and carrier argon flow rate were carried out. Under the optimized conditions, the absolute and concentration detection limits were 0.38 ng and 2.2 ng ml- 1, respectively, assuming that 0.17 ml of sample was injected. The generation efficiency of was 28-37%. The method was successfully applied to determine trace amounts of cadmium in two certified reference materials of Environmental Water (GSB07-1185-2000 and GSBZ 50009-88). The results were in good agreement with the certified reference values.

  6. Rapid Determination of the Chemical Oxygen Demand of Water Using a Thermal Biosensor

    Directory of Open Access Journals (Sweden)

    Na Yao

    2014-06-01

    Full Text Available In this paper we describe a thermal biosensor with a flow injection analysis system for the determination of the chemical oxygen demand (COD of water samples. Glucose solutions of different concentrations and actual water samples were tested, and their COD values were determined by measuring the heat generated when the samples passed through a column containing periodic acid. The biosensor exhibited a large linear range (5 to 3000 mg/L and a low detection limit (1.84 mg/L. It could tolerate the presence of chloride ions in concentrations of 0.015 M without requiring a masking agent. The sensor was successfully used for detecting the COD values of actual samples. The COD values of water samples from various sources were correlated with those obtained by the standard dichromate method; the linear regression coefficient was found to be 0.996. The sensor is environmentally friendly, economical, and highly stable, and exhibits good reproducibility and accuracy. In addition, its response time is short, and there is no danger of hazardous emissions or external contamination. Finally, the samples to be tested do not have to be pretreated. These results suggest that the biosensor is suitable for the continuous monitoring of the COD values of actual wastewater samples.

  7. A portable photoelectrochemical probe for rapid determination of chemical oxygen demand in wastewaters.

    Science.gov (United States)

    Zhang, Shanqing; Li, Lihong; Zhao, Huijun

    2009-10-15

    A photoelectrochemical probe for rapid determination of chemical oxygen demand (COD) is developed using a nanostructured mixed-phase TiO2 photoanode, namely PeCOD probe. A UV-LED light source and a USB mircroelectrochemical station are powered and controlled by a laptop computer, which makes the probe portable for onsite COD analyses. The photoelectrochemical measurement of COD was optimized in terms of light intensity, applied bias, and pH. Under the optimized conditions, the net steady state currents originated from the oxidation of organic compounds were found to be directly proportional to COD concentrations. A practical detection limit of 0.2 ppm COD and a linear range of 0-120 ppm COD were achieved. The analytical method using the portable PeCOD probe has the advantages of being rapid, low cost, robust, user-friendly, and environmental friendly. It has been successfully applied to determine the COD values of the synthetic samples consisting of potassium hydrogen phthalate, D-glucose, glutamic acid, glutaric acid, succinic acid, and malonic acid, and real samples from various industries, such as bakery, oil and grease manufacturer, poultry, hotel, fine food factory, and fresh food producer, commercial bread manufacturer. Excellent agreement between the proposed method and the conventional COD method (dichromate) was achieved. PMID:19921898

  8. A determination of the thick disk chemical abundance distribution: Implications for galaxy evolution

    Science.gov (United States)

    Gilmore, Gerard; Wyse, Rosemary F. G.; Jones, Bryn J.

    1995-01-01

    We present a determination of the thick disk iron abundance distribution obtained from an in situ sample of F/G stars. These stars are faint, 15 less than or approximately = V less than or approximately = 18, selected on the basis of color, being a subset of the larger survey of Gilmore and Wyse designed to determine the properties of the stellar populations several kiloparsecs from the Sun. The fields studied in the present paper probe the iron abundance distribution of the stellar populations of the galaxy at 500-3000 pc above the plane, at the solar Galactocentric distance. The derived chemical abundance distributions are consistent with no metallicity gradients in the thick disk over this range of vertical distance, and with an iron abundance distribution for the thick disk that has a peak at -0.7 dex. The lack of a vertical gradient argues against slow, dissipational settling as a mechanism for the formation of the thick disk. The photometric and metallicity data support a turn-off of the thick disk that is comparable in age to the metal-rich globular clusters, or greater than or approximately = 12 Gyr, and are consistent with a spread to older ages.

  9. Determinants of Price-Earnings Ratio: The Case of Chemical Sector of Pakistan

    Directory of Open Access Journals (Sweden)

    Samya Tahir

    2012-08-01

    Full Text Available Price-to-Earnings (P/E ratio, a relative valuation technique has always remained at the centre of attention of market analysts and investors ever since the origin of discounted dividend growth model of Gordon and Shapiro (1956. The present study attempts to identify the factors explaining variations in P/E ratio for chemical sector of Pakistan by using Ordinary Least Square (OLS regression on pooled data of 25 firms listed at Karachi stock exchange for the period 2005 to 2009. Furthermore, taking into account the volatility in Pakistani stock market during the study period, a time-series analysis has also made by using OLS regression model to examine whether determinants of P/E ratio differ across years or not. Results demonstrate that Dividend payout ratio and Tobin’s Q remain the most important determinants of P/E ratios for pooled as well as time-series analysis. The study is expected to facilitate decision makers to evaluate factors that explain variations in firm’s P/E ratio in order to attract investor’s attention and raise their confidence to select these firms in their portfolios.

  10. WO3/W Nanopores Sensor for Chemical Oxygen Demand (COD Determination under Visible Light

    Directory of Open Access Journals (Sweden)

    Xuejin Li

    2014-06-01

    Full Text Available A sensor of a WO3 nanopores electrode combined with a thin layer reactor was proposed to develop a Chemical Oxygen Demand (COD determination method and solve the problem that the COD values are inaccurately determined by the standard method. The visible spectrum, e.g., 420 nm, could be used as light source in the sensor we developed, which represents a breakthrough by limiting of UV light source in the photoelectrocatalysis process. The operation conditions were optimized in this work, and the results showed that taking NaNO3 solution at the concentration of 2.5 mol·L−1 as electrolyte under the light intensity of 214 μW·cm−2 and applied bias of 2.5 V, the proposed method is accurate and well reproducible, even in a wide range of pH values. Furthermore, the COD values obtained by the WO3 sensor were fitted well with the theoretical COD value in the range of 3–60 mg·L−1 with a limit value of 1 mg·L−1, which reveals that the proposed sensor may be a practical device for monitoring and controlling surface water quality as well as slightly polluted water.

  11. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    International Nuclear Information System (INIS)

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  12. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    International Nuclear Information System (INIS)

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup

  13. NFCSim: A Dynamic Fuel Burnup and Fuel Cycle Simulation Tool

    International Nuclear Information System (INIS)

    NFCSim is an event-driven, time-dependent simulation code modeling the flow of materials through the nuclear fuel cycle. NFCSim tracks mass flow at the level of discrete reactor fuel charges/discharges and logs the history of nuclear material as it progresses through a detailed series of processes and facilities, generating life-cycle material balances for any number of reactors. NFCSim is an ideal tool for analysis - of the economics, sustainability, or proliferation resistance - of nonequilibrium, interacting, or evolving reactor fleets. The software couples with a criticality and burnup engine, LACE (Los Alamos Criticality Engine). LACE implements a piecewise-linear, reactor-specific reactivity model for its criticality calculations. This model constructs fluence-dependent reactivity traces for any facility; it is designed to address nuclear economies in which either a steady state is never obtained or is a poor approximation. LACE operates in transient and equilibrium fuel management regimes at the refueling batch level, derives reactor- and cycle-dependent initial fuel compositions, and invokes ORIGEN2.x to carry out burnup calculations

  14. Fission-product burn-up in fast reactors

    International Nuclear Information System (INIS)

    In fast reactors where breeding is emphasized the burn-up of fission products can be of considerable importance. Statistical estimates of fission-product cross-sections are combined with recent yield data for the various fissionable species to estimate the gross fission-product cross-section as a function of irradiation time in a number of fast reactor spectra with various fuels. Because of gaps in yield data for some of the fuel species, it is necessary to interpolate on the yield curves in some cases. The chain yield for a given mass is then apportioned among the chain members through use of the equal charge displacement recipe. The cross-sections estimated for U235 fission products by previous authors are supplemented by estimates for fission products important for other fuels. A range of such spectra is considered. These spectra are characterized by the index (average (Ε-1/2)) in the spectra. The sensitivity of the gross poisoning and its burn-up with respect to spectrum variations are considered. The results are also expressed in terms of a few pseudo-fission products, so that changes in effective cross-section of fission products with irradiation can be taken into account in a simple computational fashion. (author)

  15. MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis

    International Nuclear Information System (INIS)

    The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2, and the Weapons-Grade Mixed Oxide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data

  16. MTR core loading pattern optimization using burnup dependent group constants

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

    Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

  17. High Burnup UO2 Fuel Pellets with Dopants for WWER

    International Nuclear Information System (INIS)

    The currently achieved level of design and technology developments provided for the implementation of the fuel cycle (4x1) in WWER at the maximal design burnup of 56 MW.day/kgU per FA. Presently in Russia the program is under way to improve the technical and economic parameters of WWER fuel cycles characterized by an increased fuel usability. To meet the requirements placed on the new fuel that ensures the reliable operation under conditions of higher burnups complex activities are under way to optimize the composition and microstructure of fuel pellets as applied to WWER. This paper describes a general approach to providing the stimulated composition and microstructure of fuel via introducing various dopants. Aside from this, the paper presents the experimentally results of studies into the main technologic and operational characteristics of dopant containing fuel pellets including higher grain sizes, pores distribution and oxygen to metal ratio. The results of the experiments made it possible to work out the pilot commercial process of the modified fuel fabrication, to manufacture pellet batches to be semi-commercially operated at NPP with WWER. (author)

  18. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  19. Determination of contact maps in proteins: A combination of structural and chemical approaches

    International Nuclear Information System (INIS)

    Contact map selection is a crucial step in structure-based molecular dynamics modelling of proteins. The map can be determined in many different ways. We focus on the methods in which residues are represented as clusters of effective spheres. One contact map, denoted as overlap (OV), is based on the overlap of such spheres. Another contact map, named Contacts of Structural Units (CSU), involves the geometry in a different way and, in addition, brings chemical considerations into account. We develop a variant of the CSU approach in which we also incorporate Coulombic effects such as formation of the ionic bridges and destabilization of possible links through repulsion. In this way, the most essential and well defined contacts are identified. The resulting residue-residue contact map, dubbed repulsive CSU (rCSU), is more sound in its physico-chemical justification than CSU. It also provides a clear prescription for validity of an inter-residual contact: the number of attractive atomic contacts should be larger than the number of repulsive ones — a feature that is not present in CSU. However, both of these maps do not correlate well with the experimental data on protein stretching. Thus, we propose to use rCSU together with the OV map. We find that the combined map, denoted as OV+rCSU, performs better than OV. In most situations, OV and OV+rCSU yield comparable folding properties but for some proteins rCSU provides contacts which improve folding in a substantial way. We discuss the likely residue-specificity of the rCSU contacts. Finally, we make comparisons to the recently proposed shadow contact map, which is derived from different principles

  20. Physical-chemical determinant properties of biological communities in continental semi-arid waters.

    Science.gov (United States)

    da Rocha, Francisco Cleiton; de Andrade, Eunice Maia; Lopes, Fernando Bezerra; de Paula Filho, Francisco José; Filho, José Hamilton Costa; da Silva, Merivalda Doroteu

    2016-08-01

    Throughout human history, water has undergone changes in quality. This problem is more serious in dry areas, where there is a natural water deficit due to climatic factors. The aims of this study, therefore, were (i) to verify correlations between physical attributes, chemical attributes and biological metrics and (ii) from the biological attributes, to verify the similarity between different points of a body of water in a tropical semi-arid region. Samples were collected every 2 months, from July 2009 to July 2011, at seven points. Four physical attributes, five chemical attributes and four biological metrics were investigated. To identify the correlations between the physicochemical properties and the biological metrics, hierarchical cluster analysis (HCA) and canonical correlation analysis (CCA) were applied. Nine classes of phytoplankton were identified, with the predominance of species of cyanobacteria, and ten families of macroinvertebrates. The use of HCA resulted in the formation of three similar groups, showing that it was possible to reduce the number of sampling points when monitoring water quality with a consequent reduction in cost. Group I was formed from the waters at the high end of the reservoir (points P1, P2 and P3), group II by the waters from the middle third (points P4 and P5), and group III by the waters from the lower part of the reservoir (points P6 and P7). Richness of the phytoplanktons Cyanophyceae, Chorophyceae and Bacillariophyceae was the attribute which determined dissimilarity in water quality. Using CCA, it was possible to identify the spatial variability of the physicochemical attributes (TSS, TKN, nitrate and total phosphorus) that most influence the metrics of the macroinvertebrates and phytoplankton present in the water. Low macroinvertebrate diversity, with a predominance of indicator families for deterioration in water quality, and the composition of phytoplankton showing a predominance of cyanobacteria, suggests greater

  1. Determination of antibacterial, antifungal activity and chemical composition of essential oil portion of unani formulation kulzam

    Directory of Open Access Journals (Sweden)

    K Ashok Kumar

    2011-01-01

    Full Text Available Kulzam is a popular unani, liquid formulation; indicated for several minor ailments like cough, cold, running nose, sore throat, insect bites, earache, tooth ache, etc. by the manufacturer. However, this over the counter formulation has not been scientifically evaluated for its claimed uses. Hence in the present study an attempt has been to check the chemical composition, antibacterial and antifungal activity as most of the above-mentioned conditions are underpinned by microbial activity. The antibacterial and antifungal activity of the formulation was carried out on human pathogenic bacteria Pseudomonas aerogenousa, Escherichia coli, Staphylococcus aureus, Corynebacterium and fungi Candida albicans, Aspergillus fumigates and was compared with standards ciprofloxacin and clotrimazole. Kulzam exhibited strong in vitro inhibition of growth against all the test micro-organisms at both 100 and 150 μl levels of undiluted formulation (test sample and more than that of standard at 150 μl level. The chemical composition of essential oil of the formulation was determined by gas chromatography−mass spectroscopy (GC-MS analysis. Thirteen compounds constituting about 93.56% of the essential oil were identified. The main components were Camphor, menthol, thymol, 2-propenal 3-phenyl-, eugenol, trans-caryophyllene, p-allylanisole, linalool, eucalyptol, l-limonene, 1-methyl-2-isopropylbenzene, and 1S-alpha-pinene. The outcome of this study shows that kulzam contain terpenes and their oxygenated derivatives, which are believed to be highly effective antibacterial, antifungal, analgesic, anti-inflammatory, antioxidant, spasmolytic and immunomodulatory agents. The formulation has been found to possess strong antibacterial and antifungal properties, and it becomes very difficult to pin point the specific compound responsible for studied activities. However, the study positively motivates the use of kulzam for common ailments.

  2. Determination of contact maps in proteins: A combination of structural and chemical approaches

    Energy Technology Data Exchange (ETDEWEB)

    Wołek, Karol; Cieplak, Marek, E-mail: mc@ifpan.edu.pl [Institute of Physics, Polish Academy of Science, Al. Lotników 32/46, 02-668 Warsaw (Poland); Gómez-Sicilia, Àngel [Instituto Cajal, Consejo Superior de Investigaciones Cientificas (CSIC), Av. Doctor Arce, 37, 28002 Madrid (Spain); Instituto Madrileño de Estudios Avanzados en Nanociencia (IMDEA-Nanociencia), C/Faraday 9, 28049 Cantoblanco (Madrid) (Spain)

    2015-12-28

    Contact map selection is a crucial step in structure-based molecular dynamics modelling of proteins. The map can be determined in many different ways. We focus on the methods in which residues are represented as clusters of effective spheres. One contact map, denoted as overlap (OV), is based on the overlap of such spheres. Another contact map, named Contacts of Structural Units (CSU), involves the geometry in a different way and, in addition, brings chemical considerations into account. We develop a variant of the CSU approach in which we also incorporate Coulombic effects such as formation of the ionic bridges and destabilization of possible links through repulsion. In this way, the most essential and well defined contacts are identified. The resulting residue-residue contact map, dubbed repulsive CSU (rCSU), is more sound in its physico-chemical justification than CSU. It also provides a clear prescription for validity of an inter-residual contact: the number of attractive atomic contacts should be larger than the number of repulsive ones — a feature that is not present in CSU. However, both of these maps do not correlate well with the experimental data on protein stretching. Thus, we propose to use rCSU together with the OV map. We find that the combined map, denoted as OV+rCSU, performs better than OV. In most situations, OV and OV+rCSU yield comparable folding properties but for some proteins rCSU provides contacts which improve folding in a substantial way. We discuss the likely residue-specificity of the rCSU contacts. Finally, we make comparisons to the recently proposed shadow contact map, which is derived from different principles.

  3. Determination of contact maps in proteins: A combination of structural and chemical approaches

    Science.gov (United States)

    Wołek, Karol; Gómez-Sicilia, Àngel; Cieplak, Marek

    2015-12-01

    Contact map selection is a crucial step in structure-based molecular dynamics modelling of proteins. The map can be determined in many different ways. We focus on the methods in which residues are represented as clusters of effective spheres. One contact map, denoted as overlap (OV), is based on the overlap of such spheres. Another contact map, named Contacts of Structural Units (CSU), involves the geometry in a different way and, in addition, brings chemical considerations into account. We develop a variant of the CSU approach in which we also incorporate Coulombic effects such as formation of the ionic bridges and destabilization of possible links through repulsion. In this way, the most essential and well defined contacts are identified. The resulting residue-residue contact map, dubbed repulsive CSU (rCSU), is more sound in its physico-chemical justification than CSU. It also provides a clear prescription for validity of an inter-residual contact: the number of attractive atomic contacts should be larger than the number of repulsive ones — a feature that is not present in CSU. However, both of these maps do not correlate well with the experimental data on protein stretching. Thus, we propose to use rCSU together with the OV map. We find that the combined map, denoted as OV+rCSU, performs better than OV. In most situations, OV and OV+rCSU yield comparable folding properties but for some proteins rCSU provides contacts which improve folding in a substantial way. We discuss the likely residue-specificity of the rCSU contacts. Finally, we make comparisons to the recently proposed shadow contact map, which is derived from different principles.

  4. An overview of burnup credit application in spent nuclear fuel management

    International Nuclear Information System (INIS)

    The current status of burnup credit application has been overviewed for spent nuclear fuel management. It was revealed that the use of burnup credit is practically limited to spent nuclear fuel storage, for which selected actinides-only are taken into account

  5. Specific behaviour aspects at extended burnup operation of PHWR nuclear fuels

    International Nuclear Information System (INIS)

    In order to evaluate the influence of burnup extension on PHWR nuclear fuel performance, the paper presents and discusses the specific potentially life-limiting factors at extended burnup for this type of fuel using recent experimental evidence and making a direct comparison with LWR fuel performance. (Author)

  6. Burnup calculations using the ORIGEN code in the CONKEMO computing system

    International Nuclear Information System (INIS)

    This article describes the CONKEMO computing system for kinetic multigroup calculations of nuclear reactors and their physical characteristics during burnup. The ORIGEN burnup calculation code has been added to the system. The results of an international benchmark calculation are also presented. (author)

  7. Microstructural characterization of high burn-up mixed oxide fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Teague, Melissa, E-mail: melissa.teague@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Gorman, Brian; King, Jeffrey [Colorado School of Mines, 1500 Illinois St, Golden, CO 80401 (United States); Porter, Douglas; Hayes, Steven [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2013-10-15

    High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

  8. Determination of the Calibration Curve for the Neutron-Moisture Meter by Chemical Analysis of Soils

    International Nuclear Information System (INIS)

    The main difficulty at present in using the neutron moisture meter in agronomy lies in the establishment of a calibration curve. The normal gravimetric method, whether carried out in-the laboratory or in the field, involves a long and costly operation; moreover, it provides little information on the cause of the variations observed in terms of either the dry density or the type of soil. Theoretical investigations conducted in parallel with experiments seem to offer a good approach to determining the influence of the measuring parameters on the response of the moisture meter. The author used the three-group theory of neutron diffusion to construct a mathematical model representing the moisture meter and the medium studied (defined by its over-all chemical composition). This model is applied using a Fortran IV computer programme, by which it can be adapted in the light of experimental studies on the influence of geometry and the nature of the measuring system. Allowance has already been made for thermal flux depression and spread due to the presence of tubing, the emission spectrum of the source, and the yield and the energy level of detection of the counters. Particular reference must be made to the importance of epicadmium neutrons (energy above 0.4 eV). To compare the theoretical and experimental data, the author defines a method of representation by which it is possible to assess the ability of the model to account for the results obtained with seven different media (siliceous sand, alumina, limestone, dolomite, kaolin, chalky clay and silt) defined by their over-all chemical composition. In the range of utilization in agronomy, the response of the moisture meter coincides with the calculated figures to roughly ± 5%. Study of the effect of dry density suggests a general equation for the calibration curves of the form N = f(Hv, ps, α, β, γ, δ), where Hv = moisture content per volume, ps = dry density, and α, β, γ and δ are constants obtained from

  9. Important fission product nuclides identification method for simplified burnup chain construction

    International Nuclear Information System (INIS)

    A method of identifying important fission product (FP) nuclides which are included in a simplified burnup chain is proposed. This method utilizes adjoint nuclide number densities and contribution functions which quantify the importance of nuclide number densities to the target nuclear characteristics: number densities of specific nuclides after burnup. Numerical tests with light water reactor (LWR) fuel pin-cell problems reveal that this method successfully identifies important FP nuclides included in a simplified burnup chain, with which number densities of target nuclides after burnup are well reproduced. A simplified burnup chain consisting of 138 FP nuclides is constructed using this method, and its good performance for predictions of number densities of target nuclides and reactivity is demonstrated against LWR pin-cell problems and multi-cell problem including gadolinium-bearing fuel rod. (author)

  10. A survey of previous and current industry-wide efforts regarding burnup credit

    International Nuclear Information System (INIS)

    Sandia has examined the matter of burnup credit from the perspective of physics, logistics, risk, and economics. A limited survey of the nuclear industry has been conducted to get a feeling for the actual application of burnup credit. Based on this survey, it can be concluded that the suppliers of spent fuel storage and transport casks are in general agreement that burnup credit offers the potential for improvements in cask efficiency without increasing the risk of accidental criticality. The actual improvement is design-specific but limited applications have demonstrated that capacity increases in the neighborhood of 20 percent are not unrealistic. A number of these vendors acknowledge that burnup credit has not been reduced to practice in cask applications and suggest that operational considerations may be more important to regulatory acceptance than to the physics. Nevertheless, the importance of burnup credit to the nuclear industry as a cask design and analysis tool has been confirmed by this survey

  11. Fuel Element Designs for Achieving High Burnups in 220 MW(e) Indian PHWRs

    International Nuclear Information System (INIS)

    Presently 19-element natural uranium fuel bundles are used in 220 MW(e) Indian PHWRs. The core average design discharge burnup for these bundles is 7000 MW·d/Te U and maximum burnup for assembly goes upto of 15 000 MWD/Te U. Use of fuel materials like MOX, Thorium, slightly enriched uranium etc in place of natural uranium in 19-element fuel bundles, in 220 MW(e) PHWRs is being investigated to achieve higher burnups. The maximum burnup investigated with these bundles is 30 000 MW·d/Te U. In PHWR fuel elements no plenum space is available and the cladding is of collapsible type. Studies have been carried out for different fuel element target burnups with different alternative concepts. Modification in pellet shape and pellet parameters are considered. These studies for the PHWR fuel elements/assemblies have been elaborated in this paper. (author)

  12. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package MTRPC system, using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTRPC Package, Empirical Formula For Fuel Burn-Up.

  13. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    pellet surface than the bulk of the pellet in leaching experiments. Thus, formation of oxidising species and radicals by radiolysis is expected to be disproportionately high as well. Therefore, when discussing high burnup fuel dissolution, the effect of the increased radiation field with burnup, as well as of the influence of the smaller grain size and increased porosity at the rim are mentioned as factors which contribute to increased dissolution rates. A third factor, increased fission product and actinide doping with burnup, has been discussed extensively in connection with increased resistance to air oxidation of the fuel. Samples from four different fuel rods, all operated in Pressurised Water Reactors (PWR), are used in the new series of corrosion experiments. They cover a burnup range from 58 to 75 MWd/kgU. The nuclide inventory of all four samples was determined by means of a combination of experimental nuclide analysis and sample specific modelling calculations. More than 40 different nuclides were analysed by isotope dilution analysis using Inductively Coupled Plasma Mass Spectrometry (ICP-MS), as well as other ICP-MS and gamma spectrometric methods. The content of roughly all fission products and actinides was also calculated separately for each sample. The experiments are performed under oxidising conditions in synthetic groundwater at ambient temperature. In order to make results as comparable as possible to those of the Series 11 experiments, the same procedure and the same leachant is used. At least nine consecutive contact periods of one and three weeks and two, three, six and twelve months are planned. The present report covers the first five contact periods up to a cumulative contact time of one year for all four samples and in addition the sixth period up to a cumulative contact time of two years for two of the samples. The samples, kept in position by a platinum wire spiral, are exposed to synthetic groundwater in a Pyrex flask. After the contact

  14. The Design of a Chemical Virtual Instrument Based on LabVIEW for Determining Temperatures and Pressures

    OpenAIRE

    Wen-Bin Wang; Jang-Yuan Li; Qi-Jun Wu

    2007-01-01

    A LabVIEW-based self-constructed chemical virtual instrument (VI) has been developed for determining temperatures and pressures. It can be put together easily and quickly by selecting hardware modules, such as the PCI-DAQ card or serial port method, different kinds of sensors, signal-conditioning circuits or finished chemical instruments, and software modules such as data acquisition, saving, proceeding. The VI system provides individual and extremely flexible solutions for automatic measurem...

  15. Preparation and Determination of the Physical and Chemical Properties of Margarine

    Directory of Open Access Journals (Sweden)

    Habazin, S.

    2012-02-01

    Full Text Available Nutrition is one of the most basic needs of the human body. It ensures the introduction of substances needed to sustain life of the organism, its growth and proper development. In the food pyramid, fats together with carbohydrates are at the very top. One source of fat in human nutrition is margarine. Margarine comprises at least 82 % vegetable fats and 16 % water. The remainder consists of lecithin, sugar, salt, colours, and vitamins.The margarine production process involves hydrogenation of vegetable fats, assembling the margarine mixture, emulsifying, crystallization and packing.The objective of this study was to show that margarine could be prepared in a school laboratory under conditions that are applicable for such laboratory. Meaning:a In a school laboratory at normal pressure and at elevated temperature with nickel as catalyst, i.e. without the use of an autoclave, carry out the reaction of hydrogenation soybean and palm oil in order to obtain a vegetable fat that is the basic ingredient of margarine. During the preparation of margarine, the hydrogenation reaction was carefully monitored by determining the iodine value.b Preparation of margarine obtained from vegetable fats.c Determination and comparison of selected physical and chemical properties of the product with the same properties of several types of margarines available on the market. The following properties were determined:– Melting point, in order to obtain composition of fat phase and determine suitability for humanuse.– Acid value, as an indicator of the amount of free fatty acids that influence the taste.– Peroxide value, for insight into the oxidative stability of fats.This work has shown that it is possible to make vegetable fat in a school lab by hydrogenation of vegetable oils. Unlike the industrial process of hydrogenation carried out under a pressure of 0.36 to 2 atm, which takes about two hours, our reaction was carried out at atmospheric pressure but with a

  16. Determination of iodine with chemical forms in rain water by fractional sampling/NAA

    International Nuclear Information System (INIS)

    A simple and rapid method has been developed for the fractional determination of particulate, iodide, iodate and non-ionic dissolved iodine in rain waters by using some filter technique and neutron activation analysis. The following procedure was chosen as a result of the tracer experiments. Particulate iodine in rain water (0.1-0.2l) is obtained as the residue on Millipore HAWP filter paper by filtration and determined by INAA. Iodide and iodate ion in half of the filtrate are adsorbed on Expapier F3 anion exchange filter papers and passed through the filter as non-ionic dissolved iodine which is then sealed into a plastic vial for irradiation. The iodate ion fraction is eluted with 15 ml of 0.5 M sodium hydroxide, and iodide ion and total ionic iodine (iodide + iodate) in another fraction are determined by the following method. The irradiated sample is decomposed together with an iodide carrier solution containing I-131 by heating in a sodium hypochlorite solution. After decomposition, the solution is acidified with hydrochloric acid, and the insoluble residue is filtered off. To the filtrate sodium sulfite solution and palladium chloride solution are added, and the precipitate of palladium iodide is separated with a glass fiber filter paper. Iodine contents of samples are calculated from the peak areas under the 443 keV γ-ray of I-128 in the precipitate and comparative standard. Corrections for chemical recovery are applied to them by means of the areas under 365 keV γ-ray of I-131. This method was applied to the rain water in Yokohama. The concentration of particulate, iodide, iodate and non-ionic dissolved iodine were 0.1-0.3, 1.0-3.7, 0.2-1.5 and 0-0.6 μg/l. The recovery of iodine in this procedure was about 70%. About 30 min was required for the radiochemical procedure and the limit of determination was 1 ng/l of iodine in a volume of 0.2l. (author)

  17. Impact of Integral Burnable Absorbers on PWR Burnup Credit Criticality Safety Analyses

    International Nuclear Information System (INIS)

    The concept of taking credit for the reduction in reactivity of burned or spent nuclear fuel (SNF) due to fuel burnup is commonly referred to as burnup credit. The reduction in reactivity that occurs with fuel burnup is due to the net reduction of fissile nuclide concentrations and the production of actinide and fission-product neutron absorbers. The change in the inventory of these nuclides with fuel burnup, and the consequent reduction in reactivity, is dependent upon the depletion environment. Therefore, the use of burnup credit necessitates consideration of all possible fuel operating conditions, including the use of integral burnable absorbers (IBAs). The Interim Staff Guidance on burnup credit [1] issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office recommends licensees restrict the use of burnup credit to assemblies that have not used burnable absorbers (e.g., IBAs or burnable poison rods, BPRs). This restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. The reason for this restriction is that the presence of burnable absorbers during depletion hardens the neutron spectrum, resulting in lower 235U depletion and higher production of fissile plutonium isotopes. Enhanced plutonium production has the effect of increasing the reactivity of the fuel at discharge and beyond. Consequently, an assembly exposed to burnable absorbers may have a slightly higher reactivity for a given burnup than an assembly that has not been exposed to burnable absorbers. This paper examines the effect of IBAs on reactivity for various designs and enrichment/poison loading combinations as a function of burnup. The effect of BPRs, which are typically removed during operation, is addressed elsewhere [2

  18. RAPID program to predict radial power and burnup distribution of UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Song, Jae Sung; Bang, Je Gun; Kim, Dae Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    Due to the radial variation of the neutron flux and its energy spectrum inside UO{sub 2} fuel, the fission density and fissile isotope production rates are varied radially in the pellet, and it becomes necessary to know the accurate radial power and burnup variation to predict the high burnup fuel behavior such as rim effects. Therefore, to predict the radial distribution of power, burnup and fissionable nuclide densities in the pellet with the burnup and U-235 enrichment, RAPID(RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet) program was developed. It considers the specific radial variation of the neutron reaction of the nuclides while the constant radial variation of neutron reaction except neutron absorption of U-238 regardless of the nuclides, the burnup and U-235 enrichment is assumed in TUBRNP model which is recognized as the one of the most reliable models. Therefore, it is expected that RAPID may be more accurate than TUBRNP, specially at high burnup region. RAPID is based upon and validated by the detailed reactor physics code, HELIOS which is one of few codes that can calculates the radial variations of the nuclides inside the pellet. Comparison of RAPID prediction with the measured data of the irradiated fuels showed very good agreement. RAPID can be used to calculate the local variations of the fissionable nuclide concentrations as well as the local power and burnup inside that pellet as a function of the burnup up to 10 w/o U-235 enrichment and 150 MWD/kgU burnup under the LWR environment. (author). 8 refs., 50 figs., 1 tab.

  19. Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

    International Nuclear Information System (INIS)

    Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

  20. Determining airborne concentrations of spatial repellent chemicals in mosquito behavior assay systems.

    Directory of Open Access Journals (Sweden)

    Nicholas J Martin

    Full Text Available BACKGROUND: Mosquito behavior assays have been used to evaluate the efficacy of vector control interventions to include spatial repellents (SR. Current analytical methods are not optimized to determine short duration concentrations of SR active ingredients (AI in air spaces during entomological evaluations. The aim of this study was to expand on our previous research to further validate a novel air sampling method to detect and quantitate airborne concentrations of a SR under laboratory and field conditions. METHODOLOGY/PRINCIPAL FINDINGS: A thermal desorption (TD gas chromatography-mass spectrometry (GC-MS method was used to determine the amount of dichlorodiphenyltrichloroethane (DDT in samples of air. During laboratory experiments, 1 L volumes of air were collected over 10 min intervals from a three-chamber mosquito behavior assay system. Significantly higher levels of airborne DDT were measured in the chamber containing textiles treated with DDT compared to chambers free of AI. In the field, 57 samples of air were collected from experimental huts with and without DDT for onsite analysis. Airborne DDT was detected in samples collected from treated huts. The mean DDT air concentrations in these two huts over a period of four days with variable ambient temperature were 0.74 µg/m(3 (n = 17; SD = 0.45 and 1.42 µg/m(3 (n = 30; SD = 0.96. CONCLUSIONS/SIGNIFICANCE: The results from laboratory experiments confirmed that significantly different DDT exposure conditions existed in the three-chamber system establishing a chemical gradient to evaluate mosquito deterrency. The TD GC-MS method addresses a need to measure short-term (<1 h SR concentrations in small volume (<100 L samples of air and should be considered for standard evaluation of airborne AI levels in mosquito behavior assay systems. Future studies include the use of TD GC-MS to measure other semi-volatile vector control compounds.