YAP multi-crystal gamma camera prototype

The Anger camera principle has shown a practical limit of a few millimeters spatial resolution. To overcome this limit, a new gamma camera prototype has been developed, based on a position-sensitive photomultiplier tube (PSPMT) coupled with a new scintillation crystal. The Hamamatsu R2486 PSPMT is a 76-mm diameter photomultiplier tube in which the electrons produced in the conventional bi-alkali photocathode are multiplied by proximity mesh dynodes and form a charge cloud around the original coordinates of the light photon striking the photocathode. A crossed wire anode array collects the charge and detects the original position. The intrinsic spatial resolution of PSPMT is better than 0.3 mm. The scintillation crystal consists of yttrium aluminum perovskit (YAP:Ce or YAlO₃:Ce). This crystal has a light efficiency of about 38% relative to NaI, no hygroscopicity and a good gamma radiation absorption. To match the characteristics of the PSPMT, a special crystal assembly was produced by the Preciosa Company, consisting of a bundle of YAP:Ce pillars where single crystals have 0.6×0.6 mm² cross section and 3 mm to 18 mm length. Preliminary results from such gamma camera prototypes show spatial resolution values ranging between 0.7 mm and 1 mm with an intrinsic detection efficiency of 37÷65% for 140 keV gamma energy     

A YAP camera 40×40 mm2 with fast readoutelectronics

Previous papers have already demonstrated the suitability for single-photon emission computed tomography (SPECT) tracers biodistribution studies on small animals, of a miniature gamma camera composed by a YAP:Ce scintillating array 4×4×1 cm³ coupled to a 3-in Hamamatsu crossed-wire anode position sensitive photomultiplier (PSPMT). Some modifications have been applied to the imaging system readout in order to adapt its utilization for specific purposes (i.e. SPECT radiopharmaceutical testing) and to reduce construction costs so that a larger scale production may be feasible. In particular, the system introduced here is based on a resistive chain readout and an integrated acquisition electronics. The intrinsic performance of this new version has been analyzed and compared to the previous system, which had a single-wire readout. Results showed that the basic characteristics of the imaging system remain substantially unchanged in this simplified version, while a faster count rate is achieved. A test on a biological sample performed with the YAP camera both in single-wire and in resistive chain readout modality is also presented, which shows the system high performance in comparison to the Auger camera     

Online integrated visual inspection and sorting system for fast reactor fuels

Mixed oxide (MOX) fuel for prototype fast breeder reactor (PFBR) is designed to have initial burn up of 100,000 MWD/T. The major differences from thermal reactor fuel are relatively smaller dimension with central hole and higher plutonium concentration (21% and 28% of PuO₂) MOX pellets which are loaded into 2.5 m long clad tubes with depleted UO₂ blanket pellets at either end of the MOX stack. The relatively smaller dimension of fuel pellets for PFBR results in large volume at fabrication and inspection. To ensure fast and accurate inspection and sorting of as sintered pellets with less radiation exposure to personnel an integrated on line pellet inspection system for remote visual inspection and sorting of pellets based on diameter has been developed. Details of the integrated pellet inspection system developed at Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur along with the results of the performance trials has been described in this paper.     

高功率研究堆低浓化物理特性研究

应用ＦＧ２ＤＢ两维两群扩散燃耗程序和带６９群中子截面库的ＣＥＬＬ栅元少群参数程序，对高功率研究堆低浓化堆芯进行了物理计算。ＬＥＵ燃料元件的铀密度为３．６－７．２ｇ／ｃｍ３，包壳厚度为０．３８－０．５６ｍｍ。结果表明：改变燃料芯体铀密度或厚度在物理上相当；各堆芯方案的控制棒价值等运行安全有关参数都可以接受。部分计算结果被拟合成线性或二次关系式以便于应用。给出了各堆芯的最小临界值、剩余反应性、运行寿期、快热中子通量和积分通量等物理参数。分析这些参数后指出：当Ｕ－２３５含量提高２０％或更多时，ＬＥＵ堆芯与ＨＥＵ堆芯的主要物理性能相近，这时快中子通量几乎不受影响，热中子通量的下降率近似正比于元件Ｕ－２３５含量增加率。但由于ＬＥＵ堆芯运行寿期的延长，对一般同位素生产与燃料元件辐照考验不会有明显影响。     

Detector Identification with Four BGO Crystals on a Dual PMT

The current trend towards higher resolution positron emission tomography (PET) requires the development of small but efficient detectors. This paper presents the initial characterization of a detector composed of four equally spaced, 4.5mm wide bismuth germanate (BGO) crystals coupled to the face of a Hamamatsu R1548 dual photomultiplier tube (PMT). The intrinsic optical cross talk between the two segments of the dual PMT exhibits a spatial dependence that has been exploited for crystal identification. Misidentification of crystals due to the combined effects of compton scattering of gamma rays between crystals, crystal penetration of gamma rays, and the imperfection of the optical cross talk identification scheme has been studied under various conditions. The detector pair spatial resolution on axis was measured to be 2.9mm full width at half maximum (FWHM), and the intrinsic peak efficiency was found to be higher than that of the larger detectors currently in use in the Montreal Neurological Institute's (MNI) Positome III camera.     

Phantom experiments on a PSAPD-based compact gamma camera with submillimeter spatial resolution for small animal SPECT

We demonstrate a position sensitive avalanche photodiode (PSAPD) based compact gamma camera for the application of small animal single photon emission computed tomography (SPECT). The silicon PSAPD with a two-dimensional resistive layer and four readout channels is implemented as a gamma ray detector to record the energy and position of radiation events from a radionuclide source. A 2 mm thick monolithic CsI:Tl scintillator is optically coupled to a PSAPD with a 8mm×8mm active area, providing submillimeter intrinsic spatial resolution, high energy resolution (16% full-width half maximum at 140 keV) and high gain. A mouse heart phantom filled with an aqueous solution of 370 MBq (99m)Tc-pertechnetate (140 keV) was imaged using the PSAPD detector module and a tungsten knife-edge pinhole collimator with a 0.5 mm diameter aperture. The PSAPD detector module was cooled with cold nitrogen gas to suppress dark current shot noise. For each projection image of the mouse heart phantom, a rotated diagonal readout algorithm was used to calculate the position of radiation events and correct for pincushion distortion. The reconstructed image of the mouse heart phantom demonstrated reproducible image quality with submillimeter spatial resolution (0.7 mm), showing the feasibility of using the compact PSAPD-based gamma camera for a small animal SPECT system.     

Development of In-Reactor Fuel Behavior Observation System

A great effort has been made to develop an optical apparatus for direct observation of transient fuel behavior in a water environment in actual in-reactor experiments. There are many difficulties to be overcome, such as high radiation fluence, limited space, shock pressure generation, fission products release.

Through the extensive irradiation tests of various kinds of glasses, optical fibers and film and ex-reactor simulation tests, an in-reactor fuel behavior observation system was designed and fabricated for NSRR (Nuclear Safety Research Reactor) experiments. The system consists of a test capsule and connected upper containments. As an image guide is installed inside the containment a periscope with non-browning lenses whose lower part is covered by a stainless steel pipe with anti-shock window made of quartz in the test capsule, and the top of the periscope is connected with a high speed camera. A high intense lamp is immersed directly into water in the capsule as a light source. The motion pictures taken by the system in the NSRR experiments could record clearly Cěrenkov glow, states of red hot fuel rod, coolant boiling, cladding melting and cracking, bubble formation, and so on which we had never seen before.     

Upgrading of X-ray CT technology for analyses of irradiated FBR MOX fuel

The X-ray CT technology previously developed by JAEA was upgraded. The shape of the X-ray source beam was changed from a circular shape to an elliptical one and the collimator slit width was decreased from 0.3 to 0.1 mm. The X-ray detector was improved by changing a CdWO₄ scintillator to a highly sensitive silicon semiconductor detector. The analysis code of X-ray CT image was revised with respect to the number of points by using two kinds of experimental results and taking into account the effects of crack existence and deviation of the central void position from the radial center of a fuel pellet. As a result, high resolution X-ray CT images could be obtained on the transverse cross section of irradiated fuel assemblies. The error of the dimensional measurement was improved from ±0.1 to ±0.03 mm by upgrading the instrument and revising the analysis code of X-ray CT image. The discriminating accuracy of density difference could be increased, and the low density region (undisturbed region) and high density region (equi-axial and columnar regions) in the X-ray CT image on the cross section of irradiated fuel could be discriminated from each other. The reliability of fuel performance analysis improves because a large number of PIE data can be collected, compared with the conventional destructive PIE.     

Application of X-Ray Computer Tomography for Observing the Deflection and Displacement of Fuel Pins in an Assembly Irradiated in FBR

A nondestructive method making use of X-ray computer tomography (X-ray CT) has been applied to post irradiation examination of fast breeder reactor (FBR) fuel assemblies. In the study, an examination is made of the deflection and displacement of fuel pin in a fuel assembly irradiated to 74.2GWd/t peak burnup in the fast reactor “JOYO.”

In the examination, X-ray CT images of transverse cross sections of fuel pin were obtained at different heights of fuel pin along its axis. Analysis of the resulting images indicated that:

1. The hexagonal wrapper tube had its lateral wall faces slightly bulged outward;

2. The fuel pins loaded in the outermost array were markedly displaced in the direction of wrapper tube, particularly in portions of fuel pin intermediate between positions constrained by wrapping wire.

The latter behavior of fuel pins was substantiated by the contours of fuel pin along its axis, which were derived from cross section images obtained at different levels along axis.

Such fuel pin displacement is surmised to have been caused by thermal stressing of the affected fuel assembly cladding.     

The closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

钍基先进CANDU堆(TACR)钍-铀燃料功率影响研究

在CANDU堆燃料栅元物理的研究中,通常选择堆芯平均的燃料比功率对栅元进行计算模拟,而在TACR中,由于使用了钍燃料,比功率的不同就可能对核反应产生影响,并通过影响棒束栅元的基本截面参数而影响到全堆计算的结果.本文对不同定功率条件下,含全铀燃料和钍-铀燃料棒束的栅元截面参数随辐照值的变化以及钍燃料棒束中233Pa和233U的质量份额进行了计算分析,认为功率会对钍燃料的栅元宏观截面产生影响,在全堆计算中,栅元基本参数应尽量使用基于历史的局部参数法.     

Reconstruction method of homogenized cross sections

The reconstruction method of homogenized cross sections in the direct response matrix method has been developed. In this reconstruction method, homogenized cross sections, which take into consideration the influences of neighboring fuel assemblies, can be reconstructed with the response relationship of incoming neutron partial currents and neutron production rates. Calculations for heterogeneous multi fuel assembly systems were done to verify the developed method. The thermal energy group fuel assembly cell-averaged homogenized cross sections reconstructed by this method agreed with those evaluated by the direct calculation of the whole system using the Monte Carlo method within 0.2%. The effect using the reconstructed fuel assembly cell-averaged homogenized cross sections in a conventional core analysis code using cross sections homogenized in a fuel assembly cell was also investigated. The results obtained showed that the analysis accuracy of k-infinity can be improved by using the cross sections reconstructed by the method. Because almost no influences on the analysis accuracy could be found related to the divided numbers of the surfaces and the angles, and the response relationship with neutron production rates of fuel rods or a fuel assembly cell-averaged neutron production rate, this reconstruction method is applicable to a conventional core analysis code using homogenized cross sections in a fuel assembly cell.     

An effective homogenization method for heterogeneous assemblies

An effective homogenization method has been developed for heterogeneous assemblies such as fuel assemblies with and without control blades in BWR and control-rod channels in FBR. Effective homogenized cross sections are calculated so as to preserve the integrated reaction rates in a heterogeneous assembly in each group by iteratively changing the cross section used in homogeneous super-cell calculations in a model composed of the heterogeneous assembly and a fuel region. The method has been applied to the rod-worth calculation for pin rods in the fast critical assembly ZPPR-10 and to the power-density calculation of a test BWR core.     

核燃料组件无损检测探测系统设计

燃料组件是反应堆的核心部分,在高温、高压及强中子辐射场等复杂环境条件下,燃料棒中芯块会出现肿胀、变形甚至包壳破裂,严重威胁反应堆的安全运行。为了更好地了解燃料组件在反应堆内的变化,研究高燃耗的燃料组件中燃料棒的中心空洞形成和燃料棒的变形情况,高能X射线无损检测是一种有效的技术手段。由于辐照后核燃料组件自身具有强放射性,探测系统设计中必须考虑减弱燃料组件自身辐射对探测采集的影响,因此组件探测系统中探测器阵列及准直器的优化设计十分必要。经过建模及相关模拟计算,得到了探测器单元最佳尺寸,优化了后准直器的结构设计,为提高燃料组件无损检测系统重建图像的质量提供帮助。     

Detection System for the Fuel Assembly Nondestructive Testing

Fuel assemblies are the central components of a reactor. The core fuel pellets in the fuel pins will swell and deform and the fuel cladding may even break under the complex environment of high temperature, high pressure and intense neutron radiation field, which threats the safety of the reactor. To better understand the changes in the behavior of the fuel assembly in the reactor and study the central void formations and deformations of fuel pins in fuel assemblies to high burn-up, high-energy X-ray non-destructive testing is an effective technical means. Irradiated nuclear fuel assembly has a strong radioactivity, it is necessary to optimize the design of the detector system and the collimator to reduce the effect from gamma rays emitted from the irradiated fuel assembly during detection system designing phase. Through modeling, estimating and optimization, the optimal size of the detector unit is obtained and the collimator design is optimized which can lay the foundation to improve the quality of the reconstructed images of the fuel assembly nondestructive system.     

An ISPA-camera for gamma rays with improved energy resolution

An Imaging Silicon Pixel Array (ISPA)-tube with nonfiber optic quartz window has been tested with respect to spatial and energy resolutions. The 122-keV γ-quanta of a Co⁵⁷ source was converted in a planar YAP-disc or in optically separated YAP-pillars of 0.6×0.6-mm² or 0.3×0.3-mm² cross sections. The achieved intrinsic spatial resolution is worsened when compared to the ISPA-version with fiber optic glass window but is still in the submillimeter region. As the YAP emission spectrum is fully covered by the spectral transmission of the quartz window and well adapted to the photocathode quantum efficiency spectrum, the resulting improved energy resolution allows for triggering of ISPA-events within a selected energy window at the 122-keV total absorption peak. This mode of operation, since it strongly reduces the number of unwanted events, is essential for the application of a γ camera in environmental conditions where a wider energy range from other background sources may be present     

Scaling study of in-core boil-off and heating process

A simple analytical method, which describes uncovery and heatup in the core under accident conditions, is derived, tested against experimental data, and used for generating the scaling criteria. Void fraction and core uncovery levels are analytically derived integrating mass and energy equations under the assumption of quasi-steady state. The coolant energy equation in the uncovered region is integrated to convert the partial differential equation for the fuel temperature into an ordinary differential equation through the assumption of the same axial distribution of the amount of energy loss from the fuel to the coolant as that of the decay heat generation rate. The ordinary differential equation for the fuel temperature, combined with the governing equation for cladding oxidation, is analytically solved assuming a linear variation of fuel temperature and oxidation thickness with time over a period.The present analytical model is tested against the Power Bursting Facility Scoping Test (PBF-ST) and SCDAP calculation. The model produces the estimation of inlet flow rates and its results which are in good agreement with the measured levels. There is an overprediction of the fuel temperatures and an underprediction of the rate of increase of the fuel temperatures by the model, presumed to be mainly caused by no consideration of reflux condensation and the higher prediction of radiation energy loss to the shroud through the treatment of one radial region of the bundle.The PBF-ST is examined with the scaling criteria generated by the present model. It is found out that the linear heat generation rate in the PBF should be by four times larger than that in the prototype system and the radiation number is highly distorted in the PBF.     

A new method to predict Grid-To-Rod Fretting in a PWR fuel assembly inlet region

Grid-To-Rod Fretting (GTRF) is one of the main causes of leaking fuel in a Pressurized Water Reactor (PWR). GTRF is caused by grid-to-rod gap, secondary flow, and axial/lateral turbulence caused pressure fluctuations within the fuel assembly, which produces rod vibration and wear. The cross flow and vortex shedding phenomenon produce low frequency vibration forces on fuel rods. In some plants, leaking fuel has been detected at the fuel inlet region of fuel assembly designs that do not have Protective Grid (P-grid) which, in addition to providing debris protection, also provides lateral stability against vibration. In order to understand the root cause of the fuel leaks, a thorough investigation of the flow field at the fuel inlet region is required. Leaking fuel has also been detected in the fuel inlet region in transition cores. In the transitional core arrangement, there are different fuel assembly designs next to each other. Due to the structure difference, there will be cross flow between fuel assemblies, which may be the initiating factor for fuel leaks.A method based on Computational Fluid Dynamics (CFD) has been developed in Westinghouse to predict the GTRF in the fuel inlet region. The fuel inlet region consists of the lower core plate, the bottom nozzle, the fuel rods, the thimble rods, the P-grid, and the bottom grid. This study employed CFD to investigate the unsteady forces on the fuel rods under typical reactor in-core conditions. Two fuel assembly (FA) inlet regions with and without the P-grid were simulated. The time history of the unsteady force components on fuel rods was recorded. Fast Fourier Transform (FFT) analyses were carried out for the force history. Compared to the data from operating plants, the new method predicted synchronized excitation forces on the rods that leaked in real operation. The CFD results also demonstrated the advantage of using the P-grid. GTRF at the fuel inlet region can be significantly reduced when the P-grid is used in Westinghouse fuel assembly designs.     

压水堆乏燃料单棒冷却液膜流动特性实验研究

为避免在极端事故工况下,乏燃料水池会长期失去补水和冷却,第3代核电站如AP1000和CAP1400,引入了喷淋冷却系统。在喷淋条件下,乏燃料棒上液膜流动特性是影响冷却效果的重要因素,国内外学者还未对其做过详细的研究。本文使用光学法研究了在不同雷诺数(Re)条件下,对单根乏燃料棒进行喷淋冷却所形成的液膜厚度随时间和空间的变化。利用CCD相机采集液膜图像,并经过处理得到了清晰的液膜厚度图像和与图像吻合良好的数据。实验结果表明：当Re为608～7 538时,瞬态液膜厚度最大值出现在Re=7 085的条件下,其值为2.36 mm;随着Re的增加,时均液膜厚度会随之增加,并且液膜波动的振幅也会随之增加;在沿棒方向上,随着距棒顶距离的增加,液膜厚度会逐渐减小并趋于平稳,并且随着Re的增加,平稳部分会出现在距棒顶更远的位置。对乏燃料单棒冷却液膜流动特性的研究,为确定具有有效冷却能力的最小喷淋流量奠定了基础。     

数字化胸部X射线摄影中铜滤过对受检者辐射剂量的影响

采用西门子公司生产的数字化X射线透视摄影系统,在不同的附加铜滤过(0、0.1、0.2、0.3 mm)下,对不同厚度的聚甲基丙烯酸甲酯(PMMA)模体进行胸部摄影。利用对比度噪声比评估不同附加铜滤过对图像质量的影响,同时使用辐射剂量仪记录不同铜滤过下的体表入射剂量,以探讨数字化胸部X射线摄影中,附加铜滤过对受检者体表入射...