Simple evaluation method for radiation streaming through bent ducts having arbitrary rectangular sections and a plane radiation source

The simplified method for radiation streaming through bent ducts with a point source was previously derived by an optics analogy considering only components passing through the duct without enormous preparatory calculations. In the present paper, a simplified method for radiation streaming through arbitrary rectangular section bent ducts including slits with a plane source is derived on the basis of the same consideration as earlier. Dose rates estimated by the derived method are compared with experimental values in 90° one-stage, bent square ducts with 10 cm and 20 cm sides, and 2 cm×40 cm, 2.08 cm×40 cm and 5.25 cm×40 cm slits in iron and polyethylene shields with 14 MeV neutron source. The dose rates estimated by the method agree with experimental values within a factor of 2. Regarding the two-stage, 45° bent duct with a diameter of 15 cm and the two-stage, 90° bent square duct with a 1.8 m side, estimated dose rates at the positions of the bend points and the exit agree with the experimental values within a factor of 2. From these results, it is concluded that the derived method is applicable to midway shielding design.     

Evaluation Formula for Radiation Duct Streaming

Author's works on an evaluation formula for the radiation duct streaming are reviewed.

The formula, which describes the direct and albedo components, is derived in a generic straight duct geometry. It is expressed by the product of spatial distributions which are represented by twice and eight-time reflected components, and power of an albedo matrix. The computational method of the spatial distributions is outlined.

The application of the formula to bent ducts is then described. The inflow of radiations to downstream at a corner of multibent ducts is formulated with the flux in the upstream leg. Using the obtained inflow current as the source term to the downstream, the formula predicts the radiation flux in the downstream leg.

The estimation of the leakage component is made by a simple point kernel method. When the streaming effect of radiations of the leakage component is important, the formula is combined with the point kernel method.

The formula is then extended to more general duct geometries, where a straight duct penetrating two material zones and a bent duct with general crossing angle are taken up. The estimation method of the flux in these ducts is briefly outlined.

The formula has been tested through the application to a variety of streaming problems, and showed to be highly applicable.     

Improvement of a Simple Evaluation Method for Duct Streaming by Taking Account of the Deep Penetration Component

A simplified formula for radiation streaming through bent ducts was previously derived by optics analogy, considering only components passing through the duct and without enormous preparatory calculations. Now, by taking account of deep penetrating component, the streaming formula is improved in order to estimate not only γ-ray flux but also neutron flux with accuracy, factor 2–3, applicable to preliminary shielding designs. Dose rates at the exit estimated by the improved formula are compared with experimental values. As for γ-ray evaluated values become conservative by improvement, while evaluated values were underestimated before improvement. Also for neutron, the discrepancy between the calculated values by the present formula and observed ones becomes within factor 2 after improvement from factor 3 before improvement. From these results, it is concluded that the improved formula is applicable to preliminary shielding designs.     

Albedo Component of Gamma-Ray Streaming through Straight Cylindrical Ducts

An experimental study was performed to investigate the effect of γ-ray streaming through cylindrical steel and lead ducts having various lengths and radii. In the experiment, ⁶⁰Co and ¹³⁷Cs point sources, and small sized detector (pulse dosimeter with plastic scintillator) was adopted, source and detector being located at the centers of opposite duct mouths. The albedo-to-direct component ratio is about 30% for steel ducts and 10% for lead at γ-ray energies of 1.25 and 0.661 MeV, with L/R ₀~10.

The measured results were compared with those calculated by the Monte Carlo method; they showed a good agreement. Consequently, the prediction of doses for straight cylindrical ducts having various length-to-radius ratios can be made by the procedures based on the Monte Carlo calculations.     

Proposal of Method to Estimate Criticality Correction for Anisotropic Diffusion in Plate Lattice Fast Assembly

Measurements on the neutron and γ-ray attenuation in multi-legged air filled ducts have been carried out using a Cockroft-Walton type neutron generator. The measured spectra were obtained with an NE-213 liquid scintillator using pulse shape discrimination to resolve neutron and γ-ray pulse height data and using a spectral unfolding code to convert these data to energy spectra. Results are given for two rectangular duct geometries. A series of integral measurements have also been carried out with a long counter for fast neutrons and a radiation survey meter to measure the γ-ray dose. The biological dose arising due to the streaming of neutrons was measured with a rem counter. The results show the relative decrement of the integral counts along the axis of the duct. The spectral measurements reveal the presence of resonant windows of oxygen present in the concrete medium. Also the 14 MeV component is found to be dominant in the first leg of the duct whereas degraded neutrons dominate in the second leg. The results are compared with MORSE calculations.     

Gamma-Ray Streaming through Multibend Cylindrical Ducts

The ray analysis method is applied to the problem of γ-ray streaming through multibend cylindrical ducts. In this application, the total path length of a γ-ray photon through the shielding material is determined by calculating one by one the positions of the intersections between the straight line representing the direction of photon travel and the surfaces of the shield or of the wall of the bent duct, as the photon passes from duct into wall and out of wall into duct.

Based on this procedure of calculation, two computer programs have been prepared for the estimation of collision-free and total γ-ray doses as well as single and total scattered γ-ray doses behind a shield with multibend cylindrical duct.

The results of calculation with these programs are compared with the experimental values obtained by Shure and Marshall, and fairly good agreement is obtained.     

Multiple Scattering Streaming of Gamma-Rays through Cylindrical Ducts

A semi-analytic Monte Carlo calculation has been applied to the duct streaming problem. The first positions of collision of photons are selected systematically in a certain volume surrounding the duct, and further scattering positions determined by the random walks of the photons. The expectation value that the photons arrive at the duct exit from their scattering point are integrated over the volume.

The multiple scattering doses through ducts surrounded by a lead and water shield are presented. The ratio between scattered and unscattered dose is about 40% for duct in water when E ₀=1 MeV and L/R ₀=10; the ratio decreases with increase of E ₀ and L/R ₀. It also decreases with increasing atomic numbers of the shield materials. A large part of the scattered dose consists of once scattered photons. The statistical deviation of the calculated dose is about 5% for 6,000–8,000 histories, when the relative dose attenuation is of the order of 10^-6.

The buildup factors of γ-rays from a point isotropic source for a slab shield are also obtained using the above method; the results agree well with those by the moments method.     

New Efficient Albedo Monte Carlo Method for Neutron Streaming Analyses

An efficient albedo Monte Carlo method newly developed has been studied by analyzing two types of experiments on neutron streaming. The method is characterized by employing the energy-angle dependent doubly differential albedos for slab, which can be calculated in a short computer time with a one-dimensional transport theory, such as the Sn method and more efficient invariant imbedding method. This paper describes the features of the present albedo Monte Carlo method, including fundamental formulas. In the analyses of the neutron streaming experiments, the calculated results agreed with the measured data within a factor of 2 for a benchmark experiment at the YAYOI reactor and within a factor of 3 for an SNR sodium duct mock-up experiment.

It is concluded that the present albedo Monte Carlo method is practical and applicable to the reactor shielding analysis concerning radiation streaming.     

Dose Enhancement Effects in a Shepherd Model 81-22 Cobalt-60 Irradiator

The magnitude of the low energy component of the photon energy spectrum inside a J. L. Shepherd Model 81-22 Co-60 irradiator has been experimentally determined. Thermoluminescent dosimeters and a specialized ionization chamber were used to verify the usefulness of a lead (0.06") and aluminum (0.03") filter box to harden the Co-60 photon energy spectrum. The area of uniform dose as a function of distance from the Co-60 source inside the irradiator also has been determined. The results suggest two recommendations for use of the irradiator: the Pb/Al box should be used for all irradiated samples where low energy radiation can be a factor, and samples should be placed as close to the source as possible, while maintaining a uniform dose rate over the entire sample area.     

Gamma-Rays and Neutrons Streaming through Air Filled Cylindrical Ducts in Water with Off-Set Alignment

Radiation streaming effects through off-set duct was studied by systematically changing four geometrical factors in a parametric survey by calculation and experiment. The calculation were based on the ray analysis method.

The four geometric parameters of the duct were (a) wall radius, (b) total length, (c) location of the off-set point, and (d) distance of off-set.

The distribution of the leakage dose rate was measured in the direction across the duct outlet, parallel to that of off-set. The calculations took into account such effects as those of neutron capture f-rays, dose buildup for γ-rays transmitted through shield, and photo-neutrons in the water.

Satisfactory agreement was obtained between calculation and experiment for γ-rays. For neutrons the agreement was not good, and the discrepancy is attributed to roughness in the approximations adopted for the calculations and to insufficient precision afforded by the method of measurement. Nevertheless it is believed that the calculations still provide results useful for obtaining a rough outline estimate of neutron streaming and which should serve for reference in shield design.     

超临界压力水冷包层方案第一壁的热与应力分析

针对超临界水冷包层中第一壁的运行工况,利用数值计算软件ANSYS中CFX和Workbench两个模块对第一壁结构中的固体域和流体域进行数值分析研究。对比矩形管道和圆形管道内传热及热应力分布发现,矩形管道四个角域强化了壁面流体和主流流体的动量和热量的交换,使传热性能优于圆形管道,而四个角域的存在也造成了该处的应力集中,使结构材料的最大应力明显高于圆形管道。进一步研究冷却剂流向和冷却管道几何结构参数对第一壁结构温度场和应力场的影响发现,在ITER运行工况下,冷却剂流向影响很小,增大冷却管道直径和减小冷却管道最小壁厚均能改善第一壁结构材料中的最高温度,而这两个几何结构参数对第一壁应力的影响较为复杂。     

Calibration factors for the SNOOPY - NP-100 neutron dosimeter

Within CANDU nuclear power facilities, only a small fraction of workers are exposed to neutron radiation. For these individuals, roughly 4.5% of the total radiation equivalent dose is the result of exposure to neutrons. When this figure is considered across all workers receiving external exposure of any kind, only 0.25% of the total radiation equivalent dose is the result of exposure to neutrons. At many facilities, the NP-100 neutron dosimeter, manufactured by Canberra Industries Incorporated, is employed in both direct and indirect dosimetry methods. Also known as “SNOOPY”, these detectors undergo calibration, which results in a calibration factor relating the neutron count rate to the ambient dose equivalent rate, using a standard Am-Be neutron source. Using measurements presented in a technical note, readings from the dosimeter for six different neutron fields in six source-detector orientations were used, to determine a calibration factor for each of these sources. The calibration factor depends on the neutron energy spectrum and the radiation weighting factor to link neutron fluence to equivalent dose. Although the neutron energy spectra measured in the CANDU workplace are quite different than that of the Am-Be calibration source, the calibration factor remains constant - within acceptable limits - regardless of the neutron source used in the calibration; for the specified calibration orientation and current radiation weighting factors. However, changing the value of the radiation weighting factors would result in changes to the calibration factor. In the event of changes to the radiation weighting factors, it will be necessary to assess whether a change to the calibration process or resulting calibration factor is warranted.     

Dose field in hiroshima and the risk of radiation-induced cancer

The dose distribution of γ-neutron radiation over the earth's surface as a result of the atomic explosion in Hiroshima is of decisive value for establishing the risk coefficients for death from radiation-induced cancer. At the same time, this radiation field is unique, since the ratio of its components varies both with distance from the hypocenter and with dose. For the DS 86 dosimetric system, the relation between mortality from solid cancer and dose has a linear, zero-threshold character. For this reason, the linear zero-threshold dependence is used as a basis for the entire system of radiation protection. However, the DS 86 results depend strongly on the choice of the coefficients of relative biological effectiveness of neutrons, which have not been established adequately. New data, which could change the dosimetric system and lead to a reexamination of the risk coefficients at low dose, are analyzed. 3 figures, 1 table, 7 references. State Science Center of, the Russian Federation-Institute of Biophysics. Translated from Atomnaya énergiya, Vol. 87, No. 4, pp. 294–297, October, 1999.     

Call for papers to be presented at the 4th International Conference on: Structural Mechanics in Reactor Technology: San Francisco,California, USA, 15–19 August 1977

The structural response of liquid metal fast breeder reactor (LMFBR) subassemblies to local accidental events is of interest in assessing the safety of such systems. Problems to be resolved include failure propagation modes from pin to pin and from subassembly to subassembly. Factors which must be considered include: (a) the geometry of the structure, (b) uncertainty of the pressure-energy source, (c) uncertainty of materials properties under reactor operating conditions, and (d) the difficulty in performing in-pile or out-of-pile experiments which would simulate the above conditions. The main effort in evaluating the subassembly response has been centered around the development of appropriate analyses based on the finite element technique. Analysis has been extended to include not only the subassembly duct structure itself, but also the fluid environment, both within subassemblies and between them. These models and codes have been devised to cover a wide range of accident loading conditions, and can treat various materials as their properties become known. The effort described here is centered mainly around an experimental effort aimed at verifying, modifying or extending the models used in treating subassembly damage propagation.To verify the finite element codes under development, a series of out-of-pile room temperature experiments has been performed on LMFBR-type subassembly ducts under various loading conditions. The duct sections were instrumented to measure internal pressure, duct midflat strains and deflection of the mid-flat and corners. Since moderate deflections were expected, and effect influence on the radial deformation would occur over a relatively short length. Preliminary calculations and subsequent static and dynamic tests demonstrated that for the range of deformations expected in single subassembly prior to failure, a shortened duct section of only 30.48 cm in length was sufficient to provide a central section over which axially uniform conditions prevailed. As a result, with axial motion of the end plates constrained, the deformation over the uniform deflection range corresponds to two-dimensional, plane-strain conditions and a two-dimensional, finite element computer code could be applied. Tests were subsequently performed on several ducts made of type 316 stainless steel which were either annealed or 50% cold-worked. Material properties of the ducts used in the experiments were determined by testing samples obtained from each duct. Also, diamond point hardness measurements were obtained across the subassembly duct flats in order to establish that the material properties were uniform. Comparisons were then made between the code calculations and experimental results which demonstrated remarkable agreement, thus lending confidence to the code's ability to predict duct response, at least under quasi-static loading. Further preliminary work was performed on the dynamic response of hexcans to a pressure pulse designed to duplicate a postulated local event.     

Scalar flux spectra for 6.2 MeV source photons penetrating cyclindrical ducts in concrete shields

Experimentally-determined benchmark scalar flux spectra are presented for 6.2 MeV photons from a 0.86 m diameter disc source penetrating cylindrical ducts of varying diameters up to 500 mm, machined through concrete shields of thicknesses up to 4.74 m.f.p. Measurements taken with a 76 mm cylindrical NaI spectrometer were unfolded using the code radak. Spectral and exposure dose-rate characteristics have been analysed for shields with and without the ducts. The introduction of a duct into the concrete shield causes an increase in the emerging dose rate which is more pronounced with increasing shield thickness and duct diameter.     

A New Target Material for the Production of I-131

The spacial distributions of neutrons and photons have been studied in a cylindrical duct through a simulated iron-water multi-layered shield, filled with an iron plug on which six helical grooves had been cut at regular intervals and with pitches of 22, 44, 88 or 176 cm.

The results obtained are discussed in relation to the three aspects of axial distribution, cross-sectional distribution and streaming. It is shown that the axial distributions of the fast neutrons and photons accord with an exponential law, while those of the thermal and epithermal neutrons are the sum of two exponential functions. It has been found that the cross-sectional distributions of neutrons and photons have quite a sharp peak on the beam axis, indicating that the transmission through the duct is more important than the scattering and diffusion in the duct. The streaming effect along the helical grooves is not very marked in any of the cases, because of the large volume fraction of the iron plug to the total duct volume (79–89%).     

Measurement of dose rate dependence of radiation induced damage tothe current gain in bipolar transistors

We report the study of radiation induced change in the current gain of bipolar transistors for three different gamma dose rates. The dose rates differed by a factor of 60 with the lowest dose to that anticipated for the LHC, and the highest at a rate we have been routinely using for radiation damage tests. The maximum dose attained was 200 kRad, which is high enough to compare with other measurements. The importance of annealing high dose rate data is demonstrated     

Gamma-Rays and Fast Neutrons Streaming about an Annular Duct

Leakage radiation from an infinite plane source reaching the rear surface of a shield provided with an annular duct has been evaluated numerically. The plane of the infinite source is divided into eight zones, for the convenience of integration, and the contribution of each zone is calculated separately by ray analysis. The distribution of the radiation flux and dose rate around the duct have been systematically obtained for different shielding materials and source energies, and the effect of changing duct geometry is also analyzed.

Among the contributions from the different zones, it is found that the predominant portion is provided by zones outside the annular area, and that this predominance extends even to the region of shield surface directly above the duct. It is also demonstrated that Rockwell's expression based on Line-of-Sight method is included as a special case of the present formula, which is applicable not only to γ-rays, but also to fast neutrorts.     

Effect of Lithium on Pressure of Hydrogen in Liquid Sodium

Abstract

Neutron spectra in a cylindrical straight duct and in bent ducts with angles of 30°, 60° and 90° have been measured by the multiple foil activation and thermoluminescence dosimetry methods. Two-dimensional discrete ordinates and three-dimensional Monte Carlo calculations are executed, and the results are compared with the measurements. The flow rate at the duct entrance calculated by the DOT3.5 code is underestimated by approximately 30%, due to a conversion of the core and reflector geometry from XY to RZ geometry. The fast neutron flux in the ducts is underestimated by 20% by the MORSE-SGC/S code due to a too coarse angular mesh of the source, which does not properly represent the actual angular distribution of the fast flux, which is highly peaked forwardly into the ducts. The thermal neutron flux was overestimated by the Monte Carlo calculation. A method is proposed to calculate the angular distribution of the flow rate at the duct entrance and to calculate the source strength and the angular distribution of the flow rate at the entrance of the second leg of the duct. The results are compared with those of the transport calculations. Generally, the agreement is quite satisfactory.     

中国成年女性精细乳房模型的建立及其在外照射防护中的应用

乳腺是对辐射致癌最敏感的器官之一,2007年国际放射防护委员会(ICRP)将乳腺的组织权重因子由0.05提升为0.12,使乳腺剂量的准确评估变得更为重要。本文利用数学建模方法建立了精细乳房模型,包括输乳窦、输乳管、小叶和脂肪组织,模型体素化后与中国成年女性参考人体素模型进行了融合。利用该模型计算了光子AP照射时乳腺腺体的剂量转换系数,并与原来模型的计算结果进行对比。结果显示,光子能量较低时两个模型的计算结果相差30%,光子能量在0.03 MeV以上时差别不大。