Download or read book Modeling of Fission Product Release from HTR high Temperature Reactor Fuel for Risk Analyses written by and published by . This book was released on 1989 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The US and FRG have developed methodologies to determine the performance of and fission product release from TRISO-coated fuel particles under postulated accident conditions. The paper presents a qualitative and quantitative comparison of US and FRG models. The models are those used by General Atomics (GA) and by the German Nuclear Research Center at Juelich (KFA/ISF). A benchmark calculation was performed for fuel temperatures predicted for the US Department of Energy sponsored Modular High Temperature Gas Cooled Reactor (MHTGR). Good agreement in the benchmark calculations supports the on-going efforts to verify and validate the independently developed codes of GA and KFA/ISF. This work was performed under the US/FRG Umbrella Agreement for Cooperation on Gas Cooled Reactor Development. 6 refs., 3 figs., 3 tabs.

Download or read book HTR 2014 Paper Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR 1 Safety Tests written by and published by . This book was released on 2001 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Safety tests were conducted on fourteen fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800°C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show an overall over-prediction of the fractional release of these fission products, which is largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. Correction factors to these diffusivities were assessed for silver and cesium in order to enable a better match between the modeling predictions and the safety testing results. In the case of strontium, correction factors could not be assessed because potential release during the safety tests could not be distinguished from matrix content released during irradiation. In the case of krypton, all the coating layers are partly retentive and the available data did not allow to determine their respective retention powers, hence preventing to derive any correction factors.

Download or read book Experiments on the Release of Fission Products from Molten Reactor Fuels written by G. E. Creek and published by . This book was released on 1959 with total page 38 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Data Summary Report for Fission Product Release Test VI 5 written by and published by . This book was released on 1991 with total page 51 pages. Available in PDF, EPUB and Kindle. Book excerpt: Test VI-5, the fifth in a series of high-temperature fission product release tests in a vertical test apparatus, was conducted in a flowing mixture of hydrogen and helium. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium which had been irradiated to a burnup of (approximately)42 MWd/kg. Using a hot cell-mounted test apparatus, the fuel rod was heated in an induction furnace under simulated LWR accident conditions to two test temperatures, 2000 K for 20 min and then 2700 K for an additional 20 min. The released fission products were collected in three sequentially operated collection trains on components designed to measure fission product transport characteristics and facilitate sampling and analysis. The results from this test were compared with those obtained in previous tests in this series and with the CORSOR-M and ORNL diffusion release models for fission product release. 21 refs., 19 figs., 12 tabs.

Download or read book Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR 1 Irradiation Experiment written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive layers of the TRISO-coated particle and potentially through the compact matrix. The release of these fission products was measured in PIE and modeled with PARFUME.

Download or read book Energy Research Abstracts written by and published by . This book was released on 1994-06 with total page 544 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book ERDA Energy Research Abstracts written by and published by . This book was released on 1983 with total page 974 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Correlations for Fission Product Release from N Reactor Fuel Under High temperature Accident Conditions written by and published by . This book was released on 1988 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book HTR 2014 Paper Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR 1 Irradiation Experiment written by and published by . This book was released on 2001 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of fission products silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the SiC layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed silicon carbide (SiC) layers, the over-prediction is by a factor of about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the PARFUME predictions and the modeled diffusivity of strontium in UCO. In the case of silver, the comparisons between PARFUME and PIE are better than for cesium and strontium. They show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons lie in the same order of magnitude.

Download or read book Development of an Improved Model for Predicting Fission Product Release from High Temperature Gas Cooled Reactors written by Daniel A. Smith and published by . This book was released on 1994 with total page 162 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Advances in High Temperature Gas Cooled Reactor Fuel Technology written by International Atomic Energy Agency and published by . This book was released on 2012-06 with total page 639 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Download or read book Modeling of Fission Products Release Using Fuzzy Reasoning written by Adel Abdelrahman and published by LAP Lambert Academic Publishing. This book was released on 2014-03 with total page 112 pages. Available in PDF, EPUB and Kindle. Book excerpt: The fuzzy reasoning is used to model the fractional release of fission products under severe accidents conditions comparing with the data from out of pile experiments conducted in special annealing furnaces, performed on specimens of spent nuclear fuel with high burn up. The physical phenomena of interest are migration of fission products from the severe damaged fuel matrix and its microstructure grains with associated phenomena of sweeping of fission products from inter-granular and intra-granular sites, the increase of grain size on the account of annihilation of smaller grains during high temperature annealing, and a coalescence of pores. These phenomena can be viewed and imaged as global mass transfer of imaginary packages carrying the fission products with time. So incorporating the Package Flow Model PFM to represent the migration of the imaginary fission product packages of grains, bubbles or, pores with adjusted size during the annealing tests. The predicted values have shown a good agreement with the experimental data set from ORNL VI and HI tests, CRL experiments and, CEA experiments.

Download or read book Recent MELCOR and VICTORIA Fission Product Research at the NRC written by and published by . This book was released on 1999 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes in the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.

Download or read book Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR 1 Safety Tests written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book MELCOR 1 8 5 Modeling Aspects of Fission Product Release Transport and Deposition an Assessment with Recommendations written by and published by . This book was released on 2010 with total page 52 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged light water reactor fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and mixed oxide (MOX) fuels. This paper discusses the synthesis of these findings in the MELCOR severe accident code. Based on recent assessments of MELCOR 1.8.5 fission product release modeling against the Phebus FPT-1 test and on observations from the ISP-46 exercise, modifications to the default MELCOR 1.8.5 release models are recommended. The assessments identified an alternative set of Booth diffusion parameters recommended by ORNL (ORNL-Booth), which produced significantly improved release predictions for cesium and other fission product groups. Some adjustments to the scaling factors in the ORNL-Booth model were made for selected fission product groups, including UO2, Mo and Ru in order to obtain better comparisons with the FPT-1 data. The adjusted model, referred to as 'Modified ORNL-Booth, ' was subsequently compared to original ORNL VI fission product release experiments and to more recently performed French VERCORS tests, and the comparisons was as favorable or better than the original CORSOR-M MELCOR default release model. These modified ORNL-Booth parameters, input to MELCOR 1.8.5 as 'sensitivity coefficients' (i.e. user input that over-rides the code defaults) are recommended for the interim period until improved release models can be implemented into MELCOR. For the case of ruthenium release in air-oxidizing conditions, some additional modifications to the Ru class vapor pressure are recommended based on estimates of the RuO2 vapor pressure over mildly hyperstoichiometric UO2. The increased vapor pressure for this class significantly increases the net transport of Ru from the fuel to the gas stream. A formal model is needed. Deposition patterns in the Phebus FPT-1 circuit were also significantly improved by using the modified ORNL-Booth parameters, where retention of lower volatile Cs2MoO4 is now predicted in the heated exit regions of the FPT-1 test, bringing down depositions in the FPT-1 steam generator tube to be in closer alignment with the experimental data. This improvement in 'RCS' deposition behavior preserves the overall correct release of cesium to the containment that was observed even with the default CORSOR-M model. Not correctly treated however is the release and transport of Ag to the FPT-1 containment. A model for Ag release from control rods is presently not available in MELCOR. Lack of this model is thought to be responsible for the underprediction by a factor of two of the total aerosol mass to the FPT-1 containment. It is suggested that this underprediction of airborne mass led to an underprediction of the aerosol agglomeration rate. Underprediction of the agglomeration rate leads to low predictions of the aerosol particle size in comparison to experimentally measured ones. Small particle size leads low predictions of the gravitational settling rate relative to the experimental data. This error, however, is a conservative one in that too-low settling rate would result in a larger source term to the environment. Implementation of an interim Ag release model is currently under study. In the course of this assessment, a review of MELCOR release models was performed and led to the identification of several areas for future improvements to MELCOR. These include upgrading the Booth release model to account for changes in local oxidizing/reducing conditions and including a fuel oxidation model to accommodate effects of fuel stoichiometry. Models such as implemented in the French ELSA code and described by Lewis are considered appropriate for MELCOR. A model for ruthenium release under air oxidizing conditions is also needed and should be included as part of a fuel oxidation model since fuel stoichiometry is a fundamental parameter in determining the vapor pressure of ruthenium oxides over the fuel. There is also a need to expand the MELCOR architecture for tracking fission product classes to allow for more speciation of fission products. An example is the formation of CsI and Cs2MoO4 and potentially CsOH if all Mo is combined with Cs such that excess Cs exists in the fuel. Presently, MELCOR can track only one class combination (CsI) accurately, where excess Cs is assumed to be CsOH. Our recommended interim modifications map the CsOH (MELCOR Radionuclide Class 2) and Mo (Class 7) vapor pressure properties to Cs2MoO4, which approximates the desired formal class combination of Cs and Mo. Other extensions to handle properly iodine speciation from pool/gas chemistry are also needed.

Download or read book Modeling of In vessel Fission Product Release Including Fuel Morphology Effects for Severe Accident Analyses written by Kune Yull Suh and published by . This book was released on 1989 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Assessment of Improved Fission product Transport Models in VICTORIA Against the ORNL HI and VI Tests written by and published by . This book was released on 1991 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: New models for the release of fission-products from fuel have recently been incorporated into a comprehensive code, VICTORIA, for the prediction of radionuclide behavior under severe reactor accident conditions as an alternative to a simple Booth diffusion calculation. It is widely known that the Booth model has severe limitations when used under conditions of changing temperature and power. A new transport model based on a two node diffusive flow formulation has been implemented into VICTORIA. In addition to the diffusive flow model, other mechanisms such as grain growth, grain boundary sweeping and intergranular bubble behavior are taken into account. These physically based models focus on fission-product behavior in intact fuel geometries. While the VICTORIA program is primarily concerned with the behavior of severely degraded fuel geometries, the capability for accurately characterizing fission-product release from intact fuel must be demonstrated before other geometries can be reliably modeled. Results of VICTORIA simulations are compared with the results of in-cell heating tests on irradiated fuel. The fission-product transport models involved in these simulations are assessed and their predictive capability examined. 10 refs., 7 figs., 2 tabs.