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%.     

10MW高温气冷堆的燃耗测量研究

10MW高温气冷堆的燃耗测量系统是采用非破坏性高纯锗γ谱仪在线监测来确定燃耗值,利用MCNP4A程序对测量系统的衰减因子进行计算,基于核燃料裂变核素的γ射线能谱分析,以137Cs和134Cs核素活度作为测量对象,并对燃耗测量结果进行讨论.     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Validation of 134Cs, 137Cs and 154Eu single ratios as burnup monitors for ultra-high burnup UO2 fuel

A measurement station has been built for the non-destructive investigation of burnt fuel rod segments through high-resolution gamma spectrometry. Four UO₂ 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% (1σ) 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.     

Analysis of cesium isotope compositions in environmental samples by thermal ionization mass spectrometry-3

Cesium was recovered from soil samples obtained in Fukushima prefecture. Isotopic analysis of Cs was performed by γ-spectrometry to determine the activity ratio of ¹³⁴Cs/¹³⁷Cs and thermal ionization mass spectrometry was used to determine the isotopic ratios of ¹³³Cs/¹³⁷Cs and ¹³⁵Cs/¹³⁷Cs. The analytical results showed that both the activity ratio of ¹³⁴Cs/¹³⁷Cs and the isotopic ratio of ¹³⁵Cs/¹³⁷Cs were within the expected values for the Fukushima Daiichi Nuclear Power Plant estimated using the ORIGEN-II code, suggesting that most of the radioactive Cs in the soil sample originated from the Fukushima Daiichi Nuclear Power Plant. The concentration of ¹³⁷Cs and the contribution of radioactive Cs from global fallout were correlated to the distance from the Fukushima Daiichi Nuclear Power Plant, while the contribution of radioactive Cs from each reactor did not show any similar distance dependence.     

Investigation of thermohydraulic parameters during natural convection cooling of TRIGA reactor

Important steady-state thermohydraulic parameters of the TRIGA research reactor operating under natural convection mode of coolant flow were investigated using NCTRIGA computer code. Neutronic parameters used in preparing the input of NCTRIGA were taken from the analysis performed by 3-D Monte Carlo code MCNP4C. Benchmarking of the NCTRIGA calculated results were performed against the experimental data measured by the thermocouples in the instrumented fuel element (IFE) during the steady state operation of the reactor under natural convection mode of coolant flow. Various thermohydraulic parameters like the coolant velocity, flow rate and mass flow rate were generated for the hot channel as well as for the two channels comprising instrumented fuels. Calculated peak fuel temperatures at different power levels were compared with the measured values and also with the calculations performed by PARET code. Axial temperature profile at the fuel centreline, fuel surface and coolant in the hot channel were generated. Fuel surface heat flux, heat transfer coefficient and Reynolds’s number for the hot channel were also calculated. The effect of fuel-cladding gap and the influence of fuel rod spacing were investigated to validate the performance of NCTRIGA code. The investigated results were found to be in good agreement with the experimental values, which indicates that the NCTRIGA code can be used with confidence for TRIGA reactor analysis.     

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C,JENDL-3.3 and ENDF/B-Vl data libraries

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for ^natZr, ^natMo, ^natCr, ^natFe, ^natNi, ^natSi, and ^natMg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ²⁸, δ²⁵, ρ²⁵, and C¹ were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries.     

Distribution of the 134Cs/137Cs ratio around the Fukushima Daiichi nuclear power plant using an unmanned helicopter radiation monitoring system

Large amounts of radioactive substances were released into the environment by the Fukushima Daiichi nuclear power plant (FDNPP) accident. Several research institutes have mapped the distribution of nuclides with long half-lives, such as ¹³⁴Cs and ¹³⁷Cs. Although the ratio of ¹³⁴Cs and ¹³⁷Cs has been believed to be equal without depending on the location of the contaminated area, several researchers report that it is different depending on places quite a little. We measured the energy spectrum of gamma rays in high resolution within an approximately 3-km radius of the FDNPP by using an unmanned helicopter equipped with a LaBr₃(Ce) scintillation detector. Then, we analyzed the ¹³⁴Cs/¹³⁷Cs ratio in the area from these measured data in detail. The results show that the ¹³⁴Cs/¹³⁷Cs ratio is different between the plume trace extending north and the other plume traces. We have obtained valuable data for identification of which radioactive substances were released by individual reactor units.     

Accuracy evaluation of pin exposure calculations in current LWR core design codes

The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at the Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments.     

固定式环境γ辐射剂量率仪现场校准技术

固定式环境γ辐射剂量率仪是承担环境连续监测任务的主要设备,不便于拆卸送往计量实验室进行校准检定,且送检周期较长,影响连续监测点数据的连续性。为按期校准固定式仪表,本文结合蒙特卡罗方法研制了能量补偿型高气压电离室和便携式137 Cs照射装置,利用天然本底辐射(陆地γ射线和宇宙射线)和便携式照射装置产生的137 Csγ射线参考辐射对固定式环境γ辐射剂量率监测仪表开展现场校准实验。结果表明,采用环境比对和现场照射的方法能较好地解决固定式环境γ辐射剂量率仪的校准问题,现场所得校准因子与标准实验室中校准因子的相对偏差小于5%。     

Application of the Monte Carlo method to the analysis of measurement geometries for the calibration of a HP Ge detector in an environmental radioactivity laboratory

A gamma spectrometer including an HP Ge detector is commonly used for environmental radioactivity measurements. The efficiency of the detector should be calibrated for each geometry considered. Simulation of the calibration procedure with a validated computer program is an important auxiliary tool for environmental radioactivity laboratories. The MCNP code based on the Monte Carlo method has been applied to simulate the detection process in order to obtain spectrum peaks and determine the efficiency curve for each modelled geometry. The source used for measurements was a calibration mixed radionuclide gamma reference solution, covering a wide energy range (50-2000 keV). Two measurement geometries - Marinelli beaker and Petri boxes - as well as different materials - water, charcoal, sand - containing the source have been considered. Results obtained from the Monte Carlo model have been compared with experimental measurements in the laboratory in order to validate the model.     

Comparison of the Results of Field Radiometry and Sampling in the Investigation of 137Cs Soil Content in Bryansk Oblast

It is suggeted that field radiometry be used for real-time examination of the radioactive contamination of soil. It is shown that the accuracy and representativeness of the measurements performed with a portable 256-channel gamma spectrometer (RKG-09N) using a collimated detector is higher than that the method of extracting and analyzing soil samples. The results of a comparison of measurements of the ¹³⁷Cs contamination of soil, which were obtained using field radiometry, and soil sampling data for 16 sites at the Scientific and Industrial Association Taifun in Bryansk oblast are presented in detail. The variability of the ¹³⁷Cs content within the sampling site (~2 m²) in anthropogenic and virgin soil in different landscapes is estimated.     

CdZnTe quasi-hemispherical detector for gamma–neutron detection

Quasi-hemispherical CdZnTe detector was manufactured successfully to fully understand the performance in the mixed gamma–neutron detection field. Together with the software of COMSOL, Geant4, and Matlab, the detector structure has been optimized. The CdZnTe detector performs good energy resolutions for ²⁴¹Am, ⁵⁷Co, and ¹³⁷Cs radiation sources, especially for ¹³⁷Cs (10.91 keV full width at half maximum [FWHM] at 662 keV). A linear relationship between the energy positions and spectrum channels indicates that the detector is effective for the precise energy detection from 59.5 to 662 keV. Finally, neutron and gamma events were detected simultaneously at room temperature using ²⁴¹AmBe neutron source. The spectrum shows good energy resolution for neutron capture gamma ray (14.28 keV FWHM at 558 keV). Our work demonstrates that the quasi-hemispherical CdZnTe detector is promising for simultaneous detection of neutrons and gamma radiation.     

Burnup Determination of High Burnup and Dry Processed Fuels Based on Isotope Dilution Mass Spectrometric Measurements

Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors, and of a dry processed fuel fabricated from a spent PWR fuel and irradiated in the Hanaro research reactor. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U, Pu, ¹⁴⁸Nd, ¹⁴⁵Nd+¹⁴⁶Nd, total of the Nd isotopes, ¹³³Cs and ¹³⁷Cs determinations by the isotope dilution mass spectrometric method (IDMS) by using quadrupole spikes (²³³U, ²⁴²Pu, ¹⁵⁰Nd, and ¹³³Cs). The methods involved two sequential anion exchange resin (AG 1X8 and 1X4) separation procedures and a Cs purification with a cation exchange resin (AG 50WX4) separation procedure. The results obtained by the Nd and Cs isotopes from the mass spectrometric measurement were compared with those by the ORIGEN code.     

Development of a low-energy high resolution X-ray camera for high-energy gamma photon background environments

Although a high-energy gamma camera can obtain images of ¹³⁷Cs distribution by detecting the 662-keV gamma photons, its spatial resolution is reduced because high-energy gamma photons penetrate the edge of the pinhole collimator. To solve this problem, we developed a low-energy X-ray camera that detects the characteristic X-ray photons (32–37 keV) that are emitted from ¹³⁷Cs to obtain high resolution images. We used a 45 × 45 × 1-mm-thick NaI(Tl) scintillator that was encapsulated in 0.1-mm-thick aluminum and optically coupled to a 2-inch square, position sensitive photomultiplier tube (Hamamatsu Photonics, PSPMT:H12700 MOD) as an imaging detector. The imaging detector was encased in a 2-cm-thick tungsten alloy container and a pinhole collimator was attached to its camera head. The spatial resolution and sensitivity were ～5 mm full-width at half-maximum and ～0.6 cps/MBq for the 1.5-mm pinhole collimator 10 cm from the collimator surface, respectively. We administered 5 MBq of ¹³⁷Cs to a soybean seedling, imaged the distribution of radionuclides for six hours, and successfully obtained a high resolution image of it with our developed X-ray camera. We believe our camera will be a powerful tool for such ¹³⁷Cs imaging in plants.     

半球形CdZnTe伽马谱仪在放射性"热点"测量中的应用

核设施退役过程中,对退役场所进行"热点"调查(包括"热点"定位,活度测量等)是非常重要的.为了配合伽马相机在热点调查后对"热点"进行更为精确的测量,建立了一套配有铅屏蔽准直器的CdZnTe伽马能谱系统,用Visual c#语言编写了便于现场操作的谱分析软件MiniAnalysis,对现场用CdZnTe谱仪测得的能谱进行...     

通过分析放射性核素活度鉴别燃料初始富集度的可行性研究

高温气冷堆核电站示范工程（HTR-PM）的反应堆在达到平衡状态前要经过一个较长时间的过渡过程。该过程中堆芯将装入两类燃料球,它们在设计上只有燃料初始富集度不同。反应堆运行要求在过渡过程中要鉴别出装有低富集度燃料的燃料球,并按其燃耗水平不同将其卸出。本文针对此问题,讨论了通过分析燃料球中放射性核素活度（或其比值）以鉴别两类燃料球的方法。堆物理分析软件和KORIGEN软件针对过渡过程的计算结果初步肯定了该方法的理论可行性,并可看出最有可能的鉴别指征量是¹³⁴Cs活度、¹²⁵Sb与¹³⁷Cs的活度比值和¹³⁴Cs活度与¹³⁷Cs活度平方的比值。     

Transmutation of Cesium-137 Using Proton Accelerator

Abstract

Two transmutation methods of ¹³⁷Cs using a proton accelerator were evaluated in terms of the effective half life and the transmutation energy. One was the proton method which mainly used high energy proton spallation reaction for transmutation, and the other was the spallation neutron method which mainly used thermal neutron capture reaction. The transmutation energies and the effective half lives for the two transmutation methods were calculated by Monte Carlo simulation codes for particle transport, the NMTC/JAERI code and the MCNP code. The calculated transmutation energies were 510 MeV and 570 MeV for the spallation neutron method and the proton method, respectively, for an effective half life of 2 yr for ¹³⁷Cs.     

Pressurized heavy water reactor fuel behaviour in power ramp conditions

In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.     

In-pile release behavior of metallic fission products in graphite materials of an htgr fuel assembly

Distribution of metallic fission products in the graphite sleeve and block of the fifth OGL-1 fuel assembly was measured by gamma spectrometry with lathe sectioning. Considerably large release fractions of long-lived fission products with smooth axial profiles were observed in the sleeve due to a large failure fraction of coated fuel particles accompanied with failed silicon carbide layers. Nevertheless, a key nuclide^110mAg, whose large release is suspected at increased burnups for low-enriched uranium fuels, was effectively retained within the graphite sleeve. The retention was also observed for¹²⁵Sb, ¹⁵⁴Eu and¹⁵⁵Eu up to a burnup of 3.2% fission per initial metal atom, but was limited for¹³⁴Cs and¹³⁷Cs at high sleeve-temperatures above 900°C. In-pile diffusion coefficients in IG-110 graphite have been evaluated for Cs, Ag and Sb; those for Cs are in reasonable agreement with available in-pile data.