Development of calculation methodology for estimation of radionuclide composition in wastes generated at Fukushima Daiichi nuclear power station

ABSTRACT

A calculation methodology for estimating the radionuclide composition in the wastes generated at the Fukushima Daiichi nuclear power station has been developed by constructing a skeleton overview of the distribution of radionuclides considering the material balance in the system and calculation flowcharts of the transportation of radionuclides into the wastes. The wastes have a distinctive feature that their level of contamination includes considerable uncertainties because the process behind the contamination with the radionuclides released from the damaged fuel during and after the accident is not yet fully understood. Here, the developed method can explicitly specify the intrinsic uncertainties as a band to be included in the estimated radionuclide composition in the wastes and can quantitatively describe the uncertainties by calibration using analytically measured data on actual waste samples collected at the site. Further studies to improve the quality of the calculation method, the introduction of a stochastic approach to describe uncertainties, and acquiring a quantitative understanding of the spatial distribution of radionuclides inside the reactor building are suggested as important steps toward reasonable and sustainable waste management as an integral part of the decommissioning of the Fukushima Daiichi nuclear power station.     

大亚湾潮流场拟谱法计算及放射性核素扩散的粒子随机行走模拟

基于拟谱方法计算得到大亚湾海域的潮流场,根据该潮流场的计算结果和大亚湾地区的风速实测值,利用粒子随机行走模型,对大亚湾核电站液态排出物中的放射性核素的扩散进行模拟,给出了放射性核素随风、潮流和湍流扩散的相对浓度分布和放射性核素的浓度分布中心在大亚湾海域的轨迹。这些结果可为大亚湾核电站液态排出物的合理排放提供参考信息,为排出物中的放射性核素对周边辐射环境安全的可能影响提供参考信息。     

江西省修水县石煤矿区放射性环境调查与评价

石煤的开发和综合利用对环境有可能造成一定的污染,其中影响环境放射性水平增高和居民受照剂量增加等辐射方面的污染问题较为敏感以及突出。本次研究对江西省主要石煤资源区——修水县一处典型石煤矿的放射性环境进行调查,通过多介质采样分析以及现场监测γ辐射空气吸收剂量率和地面γ能谱相结合,综合研究该区放射性核素分布与环境γ剂量率之间的关系,利用公式估算石煤矿区居民辐射剂量并评价其环境辐射情况。结果表明,调查范围内的石煤矿区地面γ辐射空气吸收剂量率的平均值达到了662 nGy/h,该处铀矿主要形式为碳硅泥岩型,区内石煤、石煤渣、土壤、水样放射性核素测量值都高于省内相应介质的本底值,石煤矿的开采需接受专业的监管。     

Spectrometry analysis based on approximation coefficients and deep belief networks

A method of spectrometry analysis based on approximation coefficients and deep belief networks was developed. Detection rate and accurate radionuclide identification distance were used to evaluate the performance of the proposed method in identifying radionuclides. Experimental results show that identification performance was not affected by detection time, number of radionuclides, or detection distance when the minimum detectable activity of a single radionuclide was satisfied. Moreover, the proposed method could accurately predict isotopic compositions from the spectra of moving radionuclides. Thus, the designed method can be used for radiation monitoring instruments that identify radionuclides.     

Was the Chazhma Accident a Chernobyl of the Far East?

The 1985 reactivity accident on a submarine in bukhta Chazhma was accompanied by a substantial emission of fission products and activational radionuclides whose total activity reached 5 MCi. Some specialists have compared this emission to the emission resulting from the 1986 accident in Chernobyl, neglecting the large difference in the radionuclide composition: short-lived products of prompt fission of uranium (with an admixture of activational ⁶⁰Co) in Chazhma and long-lived fission products accumulated over the run of the power reactor with an admixture of short-lived nuclides from the spontaneous excursion of the RBMK-1000 reactor in the Chernobyl nuclear power plant. It is shown that the emission of long-lived radioecologically significant radionuclides in Chazhma was approximately 0.79 Ci, while in the Chernobyl accident this emission was 90 MCi.A quantitative comparison is presented of the activity and radionuclide composition between the accidents in Chazhma and Chernobyl taking account of the fraction of long-lived radionuclides and neglecting the radioactive inert gases. These quantitative estimates are used to show that the Chazhma accident is not analogous to the 1986 accident in Chernobyl.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

Analysis of radionuclide wash-off from river basins

The concept of the rate of transport of exchange and non-exchange forms of radionuclides over river catchment is introduced. On this basis, a differential transport equation is proposed for finding the radionuclide content in soil and radionuclide flow into a river with nonuniform contamination of the catchment area. The time dependence of the radionuclide wash-off time constant is determined from the measurements. It is found that transport is slow, so that radionuclide decay plays a large role. A large fraction of the radionuclides, aside from the long-lived transuranium nuclides, decays within the catchment area and does not enter the river. The flow into the river is largely determined by the radionuclide content near river edge. Examples of calculations of radionuclide flow into a river and bukhta are presented.     

PBMR radionuclide source term analysis validation based on AVR operating experience

The determination of radionuclide source terms is vital for any reactor design and licensing safety evaluation. This paper provides an overview of the PBMR analysis tools with specific focus on the modelling of mobile and deposited radionuclide source terms within the pressure boundary of a typical pebble-bed high temperature reactor (HTR). The main focus is on the Dust and Activity Migration and Distribution (DAMD) software code system that models the activation, migration and time-dependent distribution of dust and atomic particles in an HTR such as the AVR and PBMR. Since DAMD provides a time-dependent systems integrated model of HTR designs, most of the obvious physical phenomena relevant to source terms are at play. These include the neutron flux, activation cross-sections, radioactive decay, dust production rates, dust impurity levels, dust filter capabilities, dust particle size distributions, thermal-hydraulic parameters influencing the migration and distribution of particles throughout the main power system and subsystems, and helium coolant leakage and make-up rates.At this stage the DAMD calibration and validation is mainly based on the operational data, experiments and measurements made during 21 years of operating life of the AVR. The comparisons of the DAMD results with various AVR measurements provide confidence in the use of DAMD for the PBMR design and safety evaluations. In addition, sensitivity analyses are performed with DAMD to determine the bounding system parameters that drive the migration and distribution of radionuclides. The use of DAMD to evaluate design configurations, e.g. the effect of the introduction and placement of filters on the radionuclide distribution, is also shown.In conclusion, the importance of a systems modelling approach for radionuclide transport and distribution within the pressure boundary of a typical HTR system, is demonstrated. Since the DAMD code system is calibrated and validated against the AVR measurements it can be concluded that the radionuclide source term phenomena in the AVR, resulting in the measured AVR contamination levels, is taken into account in the design and safety evaluation of the PBMR.     

源于大亚湾核电站气态排出物中的放射性核素浓度计算

本文基于对大亚湾核电站气态排出物放射性核素的分析,并基于2005年11月上旬广东地区的风场,利用高斯模型对源于大亚湾核电站的放射性核素浓度进行了计算,给出了广东省区域内,特别是大亚湾、深圳、珠海和广州地区的放射性核素相对浓度分布的计算结果,为大亚湾核电站正常运行情况下对广东省区域内辐射环境影响及监测提供参考数据,为可能发生的核电站泄漏事件的监测提供参考数据.     

γ射线级联衰变符合相加修正方法研究

γ射线级联衰变引起的符合相加干扰可导致被测量γ射线全能峰计数增加或减少,从而使测量结果偏离真实值。本文从级联符合相加产生的机理出发,以~(134)Cs为例研究了级联符合相加修正因子的数值计算方法。基于蒙特卡罗计算程序MCNP给出了γ射线全能峰效率和总效率的模拟计算方法。通过实验测量和软件计算验证了符合相加修正因子数值计算方法的准确性。针对核电厂周围环境监测中主要关注的人工放射性核素,计算了不同环境介质中γ核素测量的符合相加修正因子。计算结果表明,对于气溶胶、生物灰样品中核素~(110)Ag~m、~(134)Cs符合相加干扰可达20%以上,而对于全能峰效率相对较低的土壤样和水样符合相加干扰可大于10%,其余核素也不同程度地存在符合相加干扰。本文研究结果为γ放射性核素的准确测量和误差来源分析提供了技术参考。     

我国南方某核厂址蟾蜍剂量率评估的比较研究

选择我国南方某核厂址两栖动物——蟾蜍作为参考生物,建立了蟾蜍的生物解剖学模型和外照射环境模型。采用蒙特卡罗模拟技术计算源介质中放射性核素¹³⁷Cs、⁹⁰Sr和²³⁹Pu对靶组织/器官的辐射剂量率,由此计算蟾蜍整体的辐射剂量率。采用ERICA程序和RESRAD-BIOTA程序计算蟾蜍的剂量率,并与解剖学模型进行比较。结果表明:三种方法计算的蟾蜍内照射剂量率基本一致;由于外照射环境模型的不同,外照射剂量率估算结果并不相同,ERICA程序与解剖学模型计算的外照射剂量率结果更接近;解剖学模型关注生物组织/器官的辐射剂量,对于核素分布不均匀的生物个体研究具有重要的意义。     

In Situ Methods for Quantifying Gamma-Radiation Levels and Radionuclide Concentrations

Small perturbations of the environmental radiation field by artificial radionuclides have been successfully quantified using high pressure ionization chambers and in situ semiconductor detector gamma-ray spectra. The calibration and use of these instruments for the detection of ground-deposited and airborne sources of activity is described and general methods for data interpretation are discussed. Specific examples are given in which the exposure rate from fallout radionuclides deposited on the soil surface and from noble gases released by nuclear facilities are determined and unambiguously separated from variations in the underlying background.     

船用反应堆大破口失水事故封闭环境核素扩散研究

基于严重事故分析程序MELCOR耦合计算流体力学软件CFD-FLUENT研究方法,采用MELCOR对船用反应堆失水事故进行分析,结果作为CFD-FLUENT模拟实验的初始条件,对船用反应堆大破口严重失水事故在堆舱内的放射性核素扩散进行研究。研究结果表明,泄漏时间在45 min时,放射性核素扩散至冷却剂进口、出口位置,而在14 min时放射性核素已经开始向安全壳扩散;在51 min时,放射性核素开始从安全壳破口向安全壳外扩散;在87 min时,放射性核素开始向邻舱扩散。本研究计算结果可为核事故的应急决策提供理论支持和数据支撑。      

Radiation Monitoring of Objects in the Environment Using γ Rays from Samples

A method is described for determining the specific activity of a mixture of radionuclides and individual radionuclides in environmental objects on the basis of radiation emitted by samples. This method is suitable for different geometries for measuring the -ray dose rate of the sample and different compositions of the radionuclide mixture in a sample. The conversion factor from the exposure dose rate to the specific activity of radionuclides is determined using a representative sample in which the composition and contribution of radionuclides are determined using spectrometric and radiochemical analyses. 1 table, 2 references.     

Study of the nuclide inventory of operational radioactive waste for the RBMK-1500 reactor

The experimental and modeling results on the radionuclide concentrations in the Ignalina NPP operational waste are presented in the work. The scaling factors between easy-to-measure γ emitters ¹³⁷Cs, ⁶⁰Co and a number of difficult-to-measure radionuclides, the activity measurements of which are related to radiochemical procedures, α and β spectrometry, have been determined. The study shows that the scaling factor method can be applied for RBMK-1500 reactor waste characterization. The scaling factors were used in determination of the nuclide composition of operational radioactive waste and characterization of radioactive waste during the Ignalina NPP decommissioning.     

Analysis of relative release rates of radionuclides from the RBMK-1500 reactor fuel elements

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.     

Radiation monitoring of soil

A method of radiation monitoring of the soil used for growing grains, potatoes and green vegetables is proposed. A derivative specific activity of the radionuclides in soil is calculated to perform this monitoring. For specific activity of a radionuclide in soil equal to the derivative level, the yearly consumption of produce grown in this soil results in radionuclide intake over one year equal to the yearly intake for the general population. The contribution to the yearly intake of this radionuclide can be determined by comparing the upper estimate of the specific activity of a radionuclide in the soil to the derivative. __________ Translated from Atomnaya énergiya, Vol. 102, No. 6, pp. 378–381, June, 2007.     

Measurement of Soil Concentration Profile of Radon-222 and Its Effect on Exposure Rate

The concentration profiles of ²²²Rn in the soil air are measured at several selected sites in different parts of Japan, together with the ²²⁶Ra concentrations in soil, ²²²Rn escape-to-production ratios of soil and the characteristic physical properties of the soil. The resulting data are used to derive the vertical soil concentration profile of ²²²Rn, which is found to differ distinctly from site to site. Measurements are also made of the concentrations in soil of ²³²Th and ⁴⁰K.

Based on these data, the exposure rates due to the nuclides in the soil are calculated for a level 1m above ground at each site. The lowering of exposure rate due to the diffusion loss of ²²²Rn is estimated to be 2~13% of the total exposure rate due to naturally occurring radionuclides in the soil.     

Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well.