Burnup Indicator Research Articles

Experimental burn-up analysis of ITU Triga Mark II reactor fuel elements

In this study, burn-up values ​​of spent fuels were determined using Non-Destructive Analysis (NDA) with the help of 137Cs burn-up indicator. Four spent fuel elements (F30, D18, B5 and E14) that burned in the core of the Istanbul Technical University Triga Mark II research reactor over the years were investigated with two different methods. In addition, the individual activity of the fuel elements along the axial fuel length was determined by the High Purity Germanium (HPGe) detector. As expected, it was determined that the fuel element B5, located in the center of the reactor core, exhibited the highest burn-up value with both methods. The calculations conducted using method 1 revealed a burn-up rate of 5.40% for the B5 fuel element, while method 2 yielded a burn-up value of 205.59 MWd/tHM. Experimental results were compared with previous simulation burn-up studies.

Accommodation and diffusion of Nd in uranium silicide - U3Si2

Uranium silicide, U3Si2, is considered as an advanced nuclear fuel for commercial light water reactors with improved accident tolerance as well as competitive economics. Nd is employed as a local burnup indicator for conventional oxide fuels due, among other reasons, to its low mobility in the UO2 fuel matrix and its high fission product yield. As part of the studies necessary to determine whether Nd can be considered as a candidate burnup indicator in the U3Si2 concept fuel, we investigate the mobility of Nd in U3Si2. In this work, density functional theory (DFT) calculations are performed to predict the most stable accommodation sites of Nd in U3Si2, found to be within the uranium sublattice. Based on DFT calculations of binding energies and migration activation energies, we investigate Nd diffusion by computing the transport coefficients within the framework of the self-consistent mean-field method. Our calculations predict that the diffusion ratio of Nd to U is smaller in U3Si2 than in UO2. Moreover, at the individual maximum centerline temperature of the fuel, the diffusion of Nd in U3Si2 is much slower than in UO2. From this perspective, Nd represents a good candidate burnup indicator, in similarity to that in UO2.

The irradiated fuel-burnup experiment of Tehran Research Reactor using nondestructive gamma-ray spectroscopy

In this work, a nondestructive gamma scanning technique has been applied to determine the irradiated fuel burnup of Tehran Research Reactor (TRR). Thereby, a system was designed and installed at the rim of the TRR pool, which includes a High-Purity Germanium (HPGe) detector and associated electronics-devices, a proper shield, a collimator, and an elevator to radioactive fuel handling for any longitudinal and transverse fuel movement. In the system, it also was possible to measure fuel burnup for fuels with short cooling times at the minimum distance between the fuel and the detector. Five Standard Fuel Elements (SFEs) have been studied with U3O8Al fuel in burnup range of 14%–60% FIMA and cooling time range of 60 days–550 days. Then, by analyzing the gamma-ray emitted from 137Cs isotope as a fuel burnup indicator, an axial profile of fuel burnup was measured in the active fuel length of 61.5 cm and an active width of 6 cm. Finally, verification of results from reactor physics calculations was conducted based on fuel burnup measurements using a nondestructive gamma scanning technique that represents a well enough agreement between calculations and measurements.

Comparative analysis of VERA depletion benchmark through consistent code-to-code comparison

This paper presents a comparative analysis of the VERA depletion benchmark through consistent code-to-code comparisons between four neutronics analysis codes. An optimization of depletion calculation methods has been performed through an extensive sensitivity study in terms of nuclear transmutation equation solution methods and numerical calculation options such as mesh size, burnup intervals, and recoverable energy per fission (Q-value). An applied comparison was successfully demonstrated to improve the consistency of code-to-code depletion behavior comparisons and mitigate the adverse effects of the different kappa values of different codes by using the number density of 148Nd and a cumulative flux as a burnup indicator.

Re-evaluation of the thermal neutron capture cross section of 147Nd

In this paper we are proposing a re-evaluation of the thermal-neutron induced capture cross section of 147Nd. A unique measurement exists from which this cross section was calculated in 1974. This original calculation is based on an assumed value for a specific gamma-ray fraction (called F2), taken from the neighboring nucleus 145Nd. With the availability of reaction codes such as TALYS, such fraction can nowadays be calculated using specific reaction models and parameters. The new value of F2 indicates a decrease of the thermal cross section by 45%, leading to 243 barns, instead of the 440 barns previously reported. This new cross section impacts the calculation of the number density for the well-known burn-up indicator 148Nd, but as shown, the change is close to the usual experimental uncertainties for the 148Nd number densities, thus having a limited impact on burn-up calculation.

Absolute burnup measurement of LEU silicide fuel plate irradiated in the RSG GAS multipurpose reactor by destructive radiochemical technique

Absolute burnup measurements of a sample taken from the IDA0045 fuel plate of RI-SIE2 low enriched uranium, silicide, experimental fuel element irradiated in the Reaktor Serba Guna G.A. Siwabessy (RSG GAS; previous name MPR-30) multipurpose reactor were conducted using destructive radiochemical technique. The destructive burnup measurement effort is an important part of the continuing efforts related to the post irradiation examination (PIE) project of the Indonesian National Nuclear Energy Agency. The amount of U-235 fissile material left and the Cs-137 fission product in the sample were selected as the burnup indicator for the present work. The measured burnup values of the sample were 72.4∓4.5% loss of U-235 and 78.6∓5.0% loss of U-235 based on the amount of U-235 fissile material and Cs-137 fission product, respectively. The burnup value based on the measured Cs-137 showed an overestimation since it was assumed that all fissions originated from U-235 thermal fissions. Numerical burnup verification was also conducted using the SRAC2006 code system with a PIE dedicated detail burnup chain and a SRAC library based on the JENDL-3.3 nuclear data. The verification results, combined with the previous nondestructive burnup measurement results on the plate wise and axial distributions of the Cs-134/Cs-137 activity ratio, produced burnup value of 75.0∓5.0% loss of U-235 which agreed well with the measured values.

Design, development and utilisation of a tomography station for γ-ray emission and transmission analyses of light water reactor spent fuel rods

A tomographic measurement station has been designed and developed for the investigation of light water reactor spent fuel rod segments. A unique feature of the station is that it allows both γ-ray transmission and emission computerised tomography to be performed on individual fresh or burnt fuel rods. The basics of the tomographic station design, construction and operation are presented together with the results of four burnt UO2 fuel rod segments investigated, with rod average burnups of 54, 71, 91 and 121 MWd kg−1.The combination of transmission and emission tomography techniques performed on an individual object provides a synergetic experimental approach. By performing transmission tomography, not only the average material density but also detailed spatial distributions of the density and of the γ-ray linear attenuation coefficient have been determined for each sample. By performing emission tomography, the determination of within-rod internal radial distributions of 134Cs, 137Cs and 154Eu has also been achieved for each individual pin. The results have indicated large central depressions in the caesium distributions because of isotope migration, with varying magnitude from sample to sample.Finally and in addition to the tomographic reconstructions, the possibility to use single isotope activities, and/or isotopic concentration ratios from 134Cs, 137Cs, and 154Eu, as burnup indicators at very high and ultra-high burnups, has also been investigated. This paper presents in a comprehensive manner the large panel of analysis performed for this interesting set of spent fuel samples which represents a valuable experimental insight in the physics of ultra-high burnups.

Nondestructive burnup verification by gamma-ray spectroscopy of LEU silicide fuel plates irradiated in the RSG GAS multipurpose reactor

Nondestructive burnup verification by gamma-ray spectroscopy of two LEU silicide fuel plates, irradiated in the Indonesian RSG GAS multipurpose reactor, has been successfully conducted. The two fuel plates, one from the central part and another from the outer part, were extracted from the in-house manufactured, experimental silicide fuel assembly which had a declared burnup level of approximately 50% loss of U-235. As a burnup indicator, the Cs-134/Cs-137 activity ratio was selected and measured to obtain the axial distribution of relative burnup, its average values, burnup peak values and peak locations for each fuel plate, as well as the burnup difference between central and outer plates. A detail depletion analysis using SRAC2006 code system and JENDL-3.3 based library was conducted to obtain the Cs-134/Cs-137 activity ratio based on the irradiation history of the fuel element. An excellent agreement between the calculated and measured values of the activity ratio was confirmed.

TRIGA fuel burn-up calculations and its confirmation

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%.

Chemical Isotopic Analysis of Fission Products in PWR-MOX Spent Fuels and Computational Evaluation Using JENDL, ENDF/B, JEF, and JEFF

A chemical isotopic analysis of high-burn-up MOX spent fuels with a burn-up of 45 MWd/kgHM was carried out to accumulate nuclide composition data. Furthermore, computational analysis was performed using the integrated burn-up calculation code system SWAT. The differences between the amounts of fission products obtained by the chemical isotopic analysis and SWAT calculation using JENDL-3.2, JENDL-3.3, ENDF/B-VI.5, ENDF/B-VI.8, JEF-2.2, and JEFF-3.0 were evaluated as the ratios of the calculated values to the experimental ones (C/E ratios). The fission products such as 88Sr, 90Sr, 106Ru, 133Cs, 134Cs, and 135Cs, which are gamma and decay heat sources, neutron absorption nuclides, or burn-up indicators in spent fuels, were further investigated to improve their C/E ratios using the simplified burn-up chains of fission products using JENDL-3.3; consequently, the correction values for the fission yields or capture cross sections of the fission products were estimated using sensitivity coefficients. The C/E ratios for 154Eu, 155Eu, 154Gd, 155Gd, and 156Gd markedly differed among libraries. The reason for this difference was also discussed using the sensitivity coefficient and capture cross section of each fission product, which is in their main sensitive production paths.

Validation of 134Cs, 137Cs and 154Eu single ratios as burnup monitors for ultra-high burnup UO 2 fuel

A measurement station has been built for the non-destructive investigation of burnt fuel rod segments through high-resolution gamma spectrometry. Four UO 2 pressurised water reactor fuel rod segments with different burnup levels between 50 and >100 GWd/t and ⩽10 year cooling time have been experimentally characterised using gamma-ray spectrometry to determine 134Cs, 137Cs and 154Eu and their corresponding concentration ratios. Experimental errors of ∼2% (1 σ) for the 134Cs/ 137Cs ratio were obtained for most of the segments. In parallel, pin cell depletion calculations have been performed for each segment using the deterministic code CASMO-4. Measured and calculated ratios have then been compared with the purpose of deriving and validating pin-averaged single-ratio burnup indicators for very high burnups. It is shown that the 134Cs/ 137Cs ratio, frequently used as a burnup monitor, is considerably less precise for values exceeding 50 GWd/t; discrepancies of ∼16% are found between measured and calculated values, increasing with burnup up to ∼23%. The ratios built with the 154Eu concentration show even much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by just using different basic cross section data.

Computational investigation of on-line interrogation of pebble bed reactor fuel

Pebble bed reactors are characterized by multipass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit (80000-100000 MWD/MTU). For such reactors, the fuel is assayed on-line to ensure that the burnup limit is not breached. We considered assaying the fuel using an HPGe detector to perform passive gamma-ray spectrometry of fission products. Since neither fresh nor irradiated fuel is readily available, computer simulations were utilized to identify the radionuclides that can be used as burnup indicators, and to visualize the gamma-ray spectra at various levels of burnup. Specifically, we used the ORIGEN-MONTEBURNS-MCNP code system. This allowed the establishment of the burnup dependent one-group gas reactor cross-sections for the radionuclides of interest. Subsequently, ORIGEN was used to simulate in-core pebble depletion to establish the irradiated pebble isotopics. Finally, the codes MCNP and SYNTH were used to simulate the response of the HPGe gamma-ray spectrometer. The results show that absolute and relative indicators can be used on-line to determine unambiguously the enrichment and burnup on a pebble-by-pebble basis. The activity of Cs-137 or the activity ratio of Co-60/Cs-134 can be combined with the activity ratio of Np-239/I-132 to yield the enrichment and burnup information. To use the relative indicators, a relative efficiency calibration of the gamma-ray spectrometer can be performed using the La-140 gamma lines that are emitted by the irradiated pebble. I-132, Cs-134, Cs-137, La-140, and Np-239 are produced upon the irradiation of the fuel. Co-60 is produced by doping the fuel with a small amount (/spl sim/100 ppm) of Co-59. Using this approach, the uncertainty in burnup determination due to factors such as power history variation, detector efficiency calibration, and counting statistics is expected to remain in the range of /spl plusmn/5% to /spl plusmn/10%.

{\rtf1\ansi\ansicpg1250\deff0\deflang1038\deflangfe1038\deftab708{\fonttbl{\f0\froman\fprq2\fcharset238{\*\fname Times New Roman;}Times New Roman CE;}} \viewkind4\uc1\pard\f0\fs24 Efficiency self-calibration of the HPGe detector for on-line assay of pebble bed reactor fuel \par }

Summary {\rtf1\ansi\ansicpg1250\deff0\deflang1038\deflangfe1038\deftab708{\fonttbl{\f0\froman\fprq2\fcharset238{\*\fname Times New Roman;}Times New Roman CE;}} \viewkind4\uc1\pard\f0\fs24 Multi-pass pebble bed reactors are loaded with spherical fuel pebbles that are circulated through the core until they reach a proposed burnup limit (~100,000 MWD/MTU). An HPGe detector can be used to assay the fuel on-line to ensure that the burnup limit is not breached. The potential exists for using a relative burnup indicator, which would allow the use of a relative self-calibrating approach for efficiency determination. The utilization of several fission products as built-in on-line relative calibration standards is explored using the computer codes ORIGEN2.1, MCNP4C, and SYNTH to produce simulated gamma-ray spectra of the fuel pebbles at various levels of burnup. \par }

Gamma-ray spectrometry analysis of pebble bed reactor fuel using Monte Carlo simulations

Monte Carlo simulations were used to study the gamma-ray spectra of pebble bed reactor fuel at various levels of burnup. A fuel depletion calculation was performed using the ORIGEN2.1 code, which yielded the gamma-ray source term that was introduced into the input of an MCNP4C simulation. The simulation assumed the use of a 100% efficient high-purity coaxial germanium (HPGe) detector, a pebble placed at a distance of 100cm from the detector, and accounted for Gaussian broadening of the gamma-ray peaks. Previously, it was shown that 137Cs, 60Co (introduced as a dopant), and 134Cs are the relevant burnup indicators. The results show that the 662keV line of 137Cs lies in close proximity to the intense 658keV of 197Nb, which results in spectral interference between the lines. However, the 1333keV line of 60Co, and selected 134Cs lines (e.g., at 605keV) are free from spectral interference, which enhances the possibility of their utilization as relative burnup indicators.

Assessment of on-line burnup monitoring of pebble bed reactor fuel using passive gamma-ray spectrometry

An investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line burnup determination for the Modular Pebble Bed Reactor (MPBR). In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burnup limit (/spl sim/80 000 MWD/MTU). Unlike the situation with conventional light water reactors, depending solely on computational methods to perform in-core fuel management will be highly inaccurate. As a result, an on-line measurement approach becomes the only accurate method to assess whether a particular pebble has reached its end-of-life burnup limit. The results of this investigation indicate that the fission products Cs-137 and Eu-154 have the potential to provide accurate and power-history-resistant signatures that can be directly correlated with burnup. Furthermore, depending on the fuel manufacturing process, artificially introduced dopants (e.g., Co) can provide gamma-ray lines that are usable for burnup monitoring. In fact, it was found that the relative activity of Co-60 to Cs-134 could form a burnup indicator that is resistant to power-history variations. In this case, the use of a relative indicator has several advantages, among them the elimination of the need for absolute knowledge of the detector full-energy peak efficiency curve and the establishment of a system quality-assurance figure of merit based on the peak area ratio of the Co lines.

Kainate and isoproterenol evoke synergistically glutamate release in the mouse striatum

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%.

Determination of neodymium isotopes as burnup indicator of highly irradiated (U, Pu)O2 lmfbr fuel

For the determination of fission product Nd as a burnup indicator of nuclear fuel, a method is proposed. The procedure involves the fuel dissolution in HNO3−HF mixture, the Nd separation on an anion exchange column of Dowex 1-X4 and PbO2, purification on two extraction chromatography columns and determination of the Nd by mass spectrometry. The calculation of the possible contamination with natural Nd and its correction, determination of the average weighted fission yield and the calculation of the atom percent burnup are also described.

Determination by neutron activation analysis of the burn-up indicator neodymium-148 in irradiated uranium dioxide-plutonium dioxide

ADVERTISEMENT RETURN TO ISSUEPREVArticleNEXTDetermination by neutron activation analysis of the burn-up indicator neodymium-148 in irradiated uranium dioxide-plutonium dioxideM. R. Monsecour and Andre C. DemildtCite this: Anal. Chem. 1969, 41, 1, 27–31Publication Date (Print):January 1, 1969Publication History Published online1 May 2002Published inissue 1 January 1969https://doi.org/10.1021/ac60270a009RIGHTS & PERMISSIONSArticle Views65Altmetric-Citations8LEARN ABOUT THESE METRICSArticle Views are the COUNTER-compliant sum of full text article downloads since November 2008 (both PDF and HTML) across all institutions and individuals. These metrics are regularly updated to reflect usage leading up to the last few days.Citations are the number of other articles citing this article, calculated by Crossref and updated daily. Find more information about Crossref citation counts.The Altmetric Attention Score is a quantitative measure of the attention that a research article has received online. Clicking on the donut icon will load a page at altmetric.com with additional details about the score and the social media presence for the given article. Find more information on the Altmetric Attention Score and how the score is calculated. Share Add toView InAdd Full Text with ReferenceAdd Description ExportRISCitationCitation and abstractCitation and referencesMore Options Share onFacebookTwitterWechatLinked InReddit PDF (590 KB) Get e-Alerts Get e-Alerts

Burnup Analysis of Uranium Dioxide Nondestructive Analysis by External Conversion Method

Fundamental experiments on the nondestructive burnup analysis of UO2 were made by external conversion method using a sector type double focusing β-ray spectrometer. Details of the procedure are described. About 100 mg of UO2 enclosed in an Al-tube was irradiated and analyzed by this method. The fission products used as burnup indicators were 95Zr, 95Nb, 140La and 103Ru. A code for fission product calculation developed in our laboratory was used in burnup determination. The burnup thus determined was 53.07 MWD/t.