Measurement of neutron flux distribution in the irradiation channel in the Ghana Research Reactor-1 using Monte Carlo method

The Monte Carlo method was used to determine the neutron fluxes in the irradiation channels of the Ghana Research Reactor-1. The MCNP5 code was used for this purpose to simulate the radial and axial distribution of the neutron fluxes within all the 10 irradiation channels. After the MCNP simulation, it was observed that axially, the fluxes rise to a peak before falling and then finally leveling out. It was also observed that the fluxes were higher in the center of the irradiation channels; the fluxes got higher as it moved toward the center of the core. The multiplication factor (k_eff) was observed as 1.000397 ± 0.0007. Radially, the thermal, epithermal and fast neutron flux in the inner irradiation channel range from 1.15 × 10¹² n/cm².s ± 0.1018 × 10¹¹ − 1.19 × 10¹² n/cm².s ± 0.1172 × 10¹¹, 1.21 × 10¹² n/cm².s ± 0.1014 × 10¹¹ − 1.36 × 10¹² n/cm².s ± 0.1038 × 10¹¹ and 2.47 × 10¹¹ n/cm².s ± 0.1120 × 10¹⁰ − 2.97 × 10¹¹ n/cm².s ± 0.1255 × 10¹⁰ respectively. For the outer channel, the flux range from 7.14 × 10¹¹ n/cm².s ± 0.1381 × 10¹⁰ − 7.38 × 10¹¹ n/cm².s ± 0.208 × 10¹⁰ for thermal, 1.94 × 10¹¹ n/cm².s ± 0.1014 × 10¹⁰ − 2.51 × 10¹¹ n/cm².s ± 0.1281 × 10¹⁰ for epithermal and 3.69 × 10¹⁰ n/cm².s ± 0.8912 × 10⁸ − 5.14 × 10¹⁰ n/cm².s ± 0.1009 × 10⁹ for fast. The results have shown that there are flux variations within the irradiation channels both axially and radially.     

Design of epicadmium-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate an epicadmium-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one epicadmium covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model which has an epicadmium-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the epicadmium-shielded channel was made. The final k_eff of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new epicadmium designed model was recorded as 1.00332. Also, a final prompt neutron lifetime of 1.5237 × 10⁻⁴ s was recorded for the new epicadmium designed model while a value of 1.5571 × 10⁻⁷ s was recorded for the original MCNP design of the GHARR-1. The neutron energy causing fission for the original MCNP design of the GHARR-1 was 1.3533 × 10⁻² MeV while that of the new epicadmium designed model was 1.3513 × 10⁻² MeV.     

Comparison of calculated and measured reaction rates obtained through foil activation in the subcritical dual spectrum facility YALINA-Booster

Reaction rates were measured by the foil activation technique to obtain neutron spectrum information in a subcritical core driven by an external neutron source. The experimental results are compared with Monte Carlo calculations in order to examine the capability of the Monte Carlo code MCNP together with ENDFB-6.8, JEFF-3.1.1 and CENDL-3.1 neutron cross section libraries to predict the neutron spectrum dependent reaction rates correctly in a subcritical core. The focus lies on fast neutrons. A discrepancy is found in the calculated-to-experimental values of the reaction rates and an inaccurate cross section is identified in CENDL-3.1.     

超热中子活化分析用于燃煤电厂对环境土壤污染的研究

大燃煤电厂每天由烟囱排出的粉尘(飞灰)有几十吨,通过干沉降与湿沉降,它们最终将落在周围土壤表面上。由于地质构成上的不同,一般土壤中的元素浓度与煤飞灰中的元素浓度是不同的,大量飞灰的下落,以及雨水的冲刷、渗透,可能会改变土壤中原有元素的浓度,从而影响人体健康。这是人们关心的问题。     

Comparison of the effects of cadmium-shielded and boron carbide-shielded irradiation channel of the Ghana Research Reactor-1

The MCNP model for the Ghana Research Reactor-1 (GHARR-1) was redesigned to incorporate cadmium-shielded irradiation channel as well as boron carbide-shielded channel in one of the outer irradiation channels. Further investigations were made after initial work in the cadmium-shielded channel to consider the boron carbide-shielded channel and both results were compared to determine the best material for the shielded channel. Before arriving at the final design of only one shielded outer irradiation channel extensive investigations were made into several other possible designs; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model which has a shielded channel is to equip GHARR-1 with the means of performing efficient epithermal neutron activation analysis. The use of epithermal neutron activation analysis can be very useful in many experiments and projects (e.g. it can be used to determine uranium and thorium in sediment samples). After the simulation, a comparison of the results from the boron carbide-shielded channel model for the GHARR-1 and the epicadmium-shielded channel was made. The inner irradiation channels of the two designs recorded peak values of approximately 1.18 × 10¹² ± 0.0036 n/cm² s, 1.32 × 10¹² ± 0.0036 n/cm² s and 2.71 × 10¹¹ ± 0.0071 n/cm² s for the thermal, epithermal and fast neutron flux, respectively. Likewise the outer irradiation channels of the two designs recorded peak values of approximately 7.36 × 10¹¹ ± 0.0042 n/cm² s, 2.53 × 10¹¹ ± 0.0074 n/cm² s and 4.73 × 10¹⁰ ± 0.0162 n/cm² s for the thermal, epithermal and fast neutron flux, respectively. The epicadmium design recorded a peak thermal flux of 7.08 × 10¹¹ ± 0.0033 n/cm² s and an epithermal flux of 2.09 × 10¹¹ ± 0.006 n/cm² s in the irradiation channel where the shield was installed. Also, the boron carbide design recorded no peak thermal flux but an epithermal flux of 1.18 × 10¹¹ ± 0.0079 n/cm² s in the irradiation channel where the shield was installed. The final multiplication factor (k_eff) of the boron carbide-shielded channel model for the GHARR-1 was recorded as 1.00282 ± 0.0007 while that of the epicadmium designed model was recorded as 1.00332 ± 0.0007. Also, a final prompt neutron lifetime of 1.5237 × 10⁻⁴ ± 0.0008 s was recorded for the cadmium designed model while a value of 1.5245 × 10⁻⁴ ± 0.0008 s was recorded for the boron carbide-shielded design of the GHARR-1.     

辐照前预浓集在中子活化分析中的应用

本文综述近期辐照前预浓集在中子活化分析中的应用,包括冷冻干燥法、吸附法、萃取法、共沉淀法、火试金法及其它方法。     

Transient analysis of Ghana Research Reactor-1 using PARET/ANL thermal–hydraulic code

PARET/ANL (Version 7.3 of 2007) thermal–hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1. The reactivities inserted were 2.1 mk, 4 mk and 6.71 mk. The results obtained are similar to experiment and theoretical studies performed to demonstrate that the reactor is safe to operate. The PARET/ANL (Version 7.3 of 2007) could not simulate the reactivities above 5 mk insertions which were successfully performed in earlier theoretical and experimental studies. This may be attributed to different fluid flow and heat transfer regimes within the flow channels of the reactor that were considered by the codes.     

Effect of updated WIMSD libraries on neutron energy spectrum at irradiation site of Pakistan Research Reactor-1 using 3D modeling

International Atomic Energy Agency (IAEA) has recently released new WIMSD libraries based on current cross-section evaluations. Using these libraries the effect of different evaluated data sets on effective multiplication factor and neutron energy spectrum was studied with the help of 3D reactor simulation code CITATION. Simulation methodology adopted in this work was validated by analyzing IAEA 10 MW benchmark reactor.The k_eff values obtained using all newly released libraries are within 0.45% to the experimental value, while the old library released in 1981 resulted in calculated value 1.05% larger than experimental. The flux spectrum obtained for standard fuel element using 3D modeling is smaller in fast energy range and higher in thermal energy range than is calculated using the 1D model for the standard cell. In the flux trap, differences of about −4% to 13% were found in thermal flux using the newly released libraries as compared to that obtained using 1981 WIMSD library. The major differences in the flux spectra between newly available libraries and the 1981 WIMSD library in thermal energy range are due to the differences in cross-sections of hydrogen bound-in-water. The use of only newly available cross-sections of hydrogen bound-in-water with 1981 WIMSD library resulted in significant improvement in value of k_eff as well as in the flux spectrum. Moreover the differences among new libraries in the thermal energy range are also due to these cross-sections. Difference in fission spectra from different libraries is responsible for differences of flux spectra in the fast energy range. These differences in flux are reduced significantly in the fast energy range by only replacement of fission spectra.     

APCP分析程序使中子活化实验条件最佳化

反应堆中子活化分析(NAA)是近年来迅速发展的核分析技术,由于该方法取样量少、快速、准确、灵敏度高、具有多元素同时分析等优点,适合于生物样品中微量元素的分析研究。实验中,如果感兴趣的元素很多,往往要全面考虑到各个待测元素的放射性核素,选用合适的中子通量和辐照时间、衰变时间、计数时间。在这种情况下,正确合理选择最佳化实验条件是十分重要的。传统的方法是通过反复实验尝试和误差方法来取得,这将变得非常费时费     

Chemical characterization of ancient pottery from the greater Accra region of Ghana using neutron activation analysis

Archaeology in Ghana has a long and respectable tradition. Despite this encouraging situation, significant gaps still exist in our understanding of the history of some early societies in Ghana. Accumulated evidence revealed that the Ga (Ayawaso), Dangme-Shai and the Wullf had trade and other cultural contacts with their Akan and Guan neighbours as well as the various European factors that traded and established footholds in the Accra coast. In an attempt to reconstruct the early history of the Ga, Dangme-Shai and Wullf, the archaeological material remains recovered from these communities during excavation have been studied. In all, 15 trace elements were determined in 40 pottery shards using instrumental neutron activation analysis. The elemental concentrations were processed using multivariate statistical methods, such as cluster, factor and discriminant analyses. The results revealed patterns of trade between these communities and also classified the 40 samples into two major groups based on variations in elemental compositions. The groupings suggested a clear separation between the shards from Shai and Ayawaso. The shards from Wullf scattered amongst the two groups, consistent with the archaeological findings that the Wullf community never produced their own pots.     

Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% ²³⁵U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required ⁹⁹Mo/^99mTc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.     

中子活化分析中钐与铀的相互干扰

在钐与铀的不破坏中子活化分析中主要测量46.4小时~(153)Sm的103.2keVγ的光峰和56.4小时~(259)Np的277.6keVγ的光峰。但是,在~(239)Np的衰变中,由于内转换还有较大的几率发射99.5与103.7keV的X射线(~(239)Pu),它们与~(239)Np的106.1keVγ严重干扰钐的测量。用SCORPIO/SPECTRAN软件包中的γ谱单峰分析程序分析由SCORPIO-3000计算机程控Ge(Li)γ谱仪测得的经中子活化的纯铀的γ(X)谱,在所得的结果中,     

Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

Low enriched uranium foil (19.99% ²³⁵U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required ⁹⁹Mo/^99mTc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.     

中子活化分析在变质岩原岩类型判别中的应用

应用中子活化分析结合地球化学方法讨论了四川盐边冷水箐的斜长角闪岩和变粒岩的原岩。结果表明,该区的斜长角闪岩是由大洋拉斑玄武岩变质而成,变粒岩则是由酸性沉积岩变质而成。     

循环中子活化分析茶叶中的氟含量

应用堆循环中子活化分析(Cyclic neutron activation analysis,CNAA)和三种传统氟测量方法分析了含氟标准参考物质茶叶(GBW10016)中氟含量。结果表明,CNAA具有准确性好、灵敏度高、省时、用量少、溯源性强、抗干扰性好等优势。CNAA分析四类16种商品茶中氟含量的结果显示,茶叶中氟含量依次为黑茶、红茶、乌龙茶、绿茶,其原因与茶叶的选料及加工工艺密切相关。个别黑茶氟含量高达276 mg/kg,超出国家限量标准(≤200 mg/kg),存在饮用安全风险。     

Ground deposition assessment of radionuclides following a hypothetical release from Ghana Research Reactor-1 (GHARR-1) using atmospheric dispersion model

With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. There is a challenge today in assessing radiological dose from nuclear reactor using a more reliable computer tool in addressing the released radionuclide to the atmosphere and ground effectively and to compute the dose rates. As such the dealing of atmospheric dispersion of radionuclide release from a nuclear facility has become very imperative. This has enhanced the idea of revisiting the safety features of the existing nuclear plants and particularly research reactors. One of such kind of research reactors whose safety is of concern now is the 30 kW Ghana Research Reactor-1 (GHARR-1) which uses a Highly Enrich Uranium (HEU) fuel. In connection with conversion of GHARR-1 from HEU fuel to the use of Low Enrich Uranium (LEU) fuel; assessment of a postulated radiological dose from possible radionuclides released using computer technology is essential. An effective computer model which is based on a reliable atmospheric transport and dispersion theory can help address such drawbacks. Atmospheric dispersion modeling and radiological safety analysis were performed for a postulated accident scenario of the HEU fuel of the GHARR-1 core. The simulation was performed using a reliable health physics atmospheric dispersion code called HotSpot. The HotSpot code which employs a Gaussian plume technique was used to perform the atmospheric transport modeling which was then applied to determine the ground deposition of radionuclides and to estimate the Total Effective Dose Equivalent (TEDE) of release radionuclides. The source term was generated from an inventory of peak radioisotope activities released by using the Oak Ridge isotope generation code ORIGEN-2. The adopted methodology used was based on the predominant site-specific meteorological data. Some selected radionuclides were evaluated to prove whether their release may have radiological effect on the public. Nonetheless, prudence requires assessing the effect on the public during such events. The results indicate that the maximum ground deposition value of 1.5E-04 kBq/m² occurred at 96 m distance and the maximum TEDE value of 1.9E-02 mSv occurred at 93 m from the reactor. It was observed that the values were far below the NRC acceptable limit of the 0.1 rem (1 mSv) for the public in a year even in the event of worse accident scenario.     

用放射性示踪技术和中子活化法研究一种新型钙剂在兔子体内的分布和吸收

用４７Ｃａ放射性示踪剂和中子活化分析技术研究了一种由等离子体法制备的新型钙剂在兔子体内的药代动力学、分布、吸收及生物利用度。实验结果表明，这种新型营养品中的钙在口服６至７天后主要沉积在骨骼中，其吸收率约为３０％—４０％。     

Design of a prompt-gamma neutron activation analysis system on China Advanced Research Reactor

In this paper,computational methods are used to optimize the design of a prompt-gamma neutron activation analysis(PGNAA) system on China Advanced Research Reactor(CARR).Approaches are adopted for obtaining accurate neutron beam parameter and saving the computing time.For the radiation shielding design,the optimizing factors include the cost,weight,volume,machining convenience and background radiation at the detector position.Low background spectrum and high sensitivity are expected.The simulation results...     

中子活化分析在大气环境及水环境研究中的应用

利用中子活化分析方法测试了成都市的大气环境及水环境样品。研究了有害污染元素分析的特征。