Table of Content

20 September 2015 Volume 35 Issue 5

  

In-situ gamma-spectrometry measurement of radiological source term for primary system of NPPs based on HPGe detector

Liu Liye, Cao Qinjian, Xiong Wanchun, Xiao Yunshi, Zhao Yuan, Wei Xiaofeng, Xia Sanqiang

RADIATION PROTECTION. 2015, 35(5):  257-261. 

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The technology of in-situ gamma-spectrometry source term measurement for primary system of NPPs was introduced in the paper. The measurement system was developed based on HPGe detector. Numerical efficiency calibration technique was developed based on the Mnote Carlo method. The comparisons between numerical calibration and point source experiments showed that relative deviations were within ±5% and ±10% in the energy of 300—1 408 keV for the HPGe detector without and with collimator, respectively. The measurement system had been applied for the activated corrosion source term measurement campaigns during the outage of PWR-NPPs in China. The on-site measurement results showed that it could perform a good measurement accuracy for typical radionuclides, such as ⁶⁰Co, ⁵⁸Co, ^110mAg, etc., which were the major contributors to radiation field. The relative deviations of the contacted dose rate between calculation and measurement value were mostly within ±40%.

Development of glass formulation for plasma vitrification of combustible solid wastes from nuclear power plants

Chen Mingzhou, Bai Bing, Liu Xiajie, Lv Yonghong, Huang Wenyou

RADIATION PROTECTION. 2015, 35(5):  262-266. 

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To develop glass formulations for plasma vitrification of combustible low and intermediate level radioactive wastes form nuclear power plants, 5 kinds of simulated wastes were selected and then incinerated. The incineration ash was vitrified by high temperature melting. Analysis of the glasses discloses that when the mass ratio of incineration ash, SiO₂, B₂O₃ and Na₂O equals to 0.4∶0.4∶0.1∶0.1(4^#), the glassy waste form shows the best performance with 2.5 g/cm³of density, 90 MPa of compress strength and less than 0.535 g·m^-2·d^-1 of normalized elemental leaching rate for Cs, Sr Co, etc. It can be concluded that the 4^# formulation seems to be the most preferable one for vitrification tests on the 50 kg/h plasma system.

A roadmap and key issues of spent resin wet-oxidation research

Guo Xiliang, Feng Wendong, Gao Chao, Yang Weibin, Liu Jianqin, An Hongxiang

RADIATION PROTECTION. 2015, 35(5):  267-273. 

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After briefing on the advantages of wet oxidation technology, the paper analyzes and compares the advances of wet oxidation technology domestic and aboard. A technical route of wet oxidation and cemented solidification of spent resin for volume reduction is then proposed with discussions on key issues.

Development of integrated nuclear emergency command and decision support system for nuclear power plant

Yang Yapeng, Zhang Jiangang, Tang Rongyao, Feng Zongyang, Xu Xiaoxiao, Jia Linsheng

RADIATION PROTECTION. 2015, 35(5):  274-283. 

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The integrated nuclear emergency command and decision support system is mainly used for reactor operation monitoring, emergency command, reactor condition assessment, environment impact prediction analysis, decision support and emergency data management of nuclear power plant. The system acquires the data of reactor operation, meteorology, environment radiation from relative databases in realtime, monitors the unit operation by picture configuration, then execute the emergency status degree assisted judgment, reactor core damage assessment, source term calculation, environment impact prediction analysis, operation intervention level calculation based on data from database or manual input. The system builds in a protection action analysis model for three phases of accident sequence, which are start emergency, pre-release from containment and release to environment. The system displays the emergency resources and facilities, emergency assessment results (equivalent dose curve, protection action region, and etc.) using WebGIS. The paper introduces the process, functional module design, development plan, and the main models of the system.

Radioprotective effects of nuclear targeting nanomaterial C₆₀(OH)₂₄-SLN-E on mice irradiated by sublethal doses

Xu Ying, Fan Chen, Xu Yujie, Yang Baixia, Zhou Jiebo, Li Hailong

RADIATION PROTECTION. 2015, 35(5):  284-288. 

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To investigate the protective effects of C₆₀(OH)₂₄-SLN-E on mice exposed to sub-lethal doses of ⁶⁰Co γ radiation. SPF ICR mice (male) were given C₆₀(OH)₂₄-SLN-E intraperitoneally (i.p.) continuously for 7 days before whole-body ⁶⁰Co γ radiation. The total doses were 6.0 Gy and the dose rate was 0.42 Gy/min. The survival situation, the change of body weight and peripheral blood cell on the 14^th day after sub-lethal doses of ⁶⁰Co γ irradiation were observed. C₆₀(OH)₂₄-SLN-E pretreated before IR could effectively reduce the decrease of body weight and peripheral blood leukocyte in mice induced by ⁶⁰Co γ radiation, it could also enhance the recovery of hematopoiesis of red bone marrow after irradiation, which had statistical significance compared with the irradiation group (p<0.05). C₆₀(OH)₂₄-SLN-E has efficient capacity of radioprotection in mice exposed to sub-lethal doses of ⁶⁰Co γ radiation.

Radiological monitoring and assessment of liquid effluent release to waterborne environment of inland nuclear power plant at USA

Huang Yanjun, Qin Chunli, Shangguan Zhihong, Zhou Ruming

RADIATION PROTECTION. 2015, 35(5):  289-299. 

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In this paper, a systematic analysis of the radiological monitoring results for the waterborne environment that would possibly affected by liquid effluent release of 38 typical inland NPPs in USA during 2005—2012, including the sediment, surface water, drinking water and aquatic organisms, were presented. It was shown that the release of the liquid effluent would impact little on the radiological waterborne environment, except for the issue of tritium. For the inland nuclear power plant with receiving water of poor dilution and dispersion conditions, the cumulative tritium would be accumulated to a relative higher level, but lower than the guideline concentration in drinking water regulated by the US Environment Protection Agency (EPA). The conservative assessment of the radiological dose on public from the liquid effluent release was presented, and a conclusion can be drawn that the impact is really negligible compared to the background dose for the public in USA. The conclusion was confirmed by analyzing the regulation of US Nuclear Regulatory Commission (NRC) for the requirements of the detection limits and the reporting level for the radiological monitoring data.

Study on criteria for limited impact in european utility requirements

Chen Yan, Guo Ruiping, Chen Haiying, Dang Lei, Jing Jianping, Zhang Chunming, Tong Jiejuan

RADIATION PROTECTION. 2015, 35(5):  300-304. 

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This paper investigates the probabilistic safety targets in European Utility Requirements document and studies the Criteria for Limited Impact, which is applied for the assessment of the acceptability of the design in DECs. Since there are no methods to determinate the criterion and the coefficients of CLI, we try to calculate the criterion and the coefficients of CLI using the accident dose modes, the hypothesis with homogeneous release and data of nuclear power plant in China. It finds that the magnitude of part coefficients is about one order different from that in EUR. Finally, we discuss the relationship between the CLI and large release, and its availability.

Influence of Bayan Obo ores on indoor ²²²Rn, ²²⁰Rnand γ radiation levels

Wang Chunhong, Liu Fudong, Liu Guifang, Chen Ling, Liu Senlin, Shang Bing, Zhu Xinhua, Liu Yanyang, Sheng Mingwei, Wang Ying, Shao Minggang

RADIATION PROTECTION. 2015, 35(5):  305-310. 

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To study the influence of exploitation of Bayan Obo ores on indoor ²²²Rn, ²²⁰Rn and γ radiation levels, measurements of indoor ²²²Rn,²²⁰Rn concentrations were performed using solid state nuclear track detectors for one year, and indoor γ radiation level was performed using TLD for 3 months. The investigation shows that ranges of concentration and averages of indoor ²²²Rn, ²²⁰Rn and γ radiation level are 16.7-125.3 Bq/m³ and 39.3 Bq/m³, 30-153.3 Bq/m³ and 51.2 Bq/m³ and 120-283 nGy/h and 149 nGy/h respectively. Thusly, the paper estimates that the exploitation and utilization of Bayan Obo ores has caused an increase of indoor γ radiation level. The levels of ²²²Rn, ²²⁰Rn and γ in Baiyan Obo area are higher than in Baotou. The concentration of indoor ²²⁰Rn in Baotou Iron and Steel Co. Ltd. (Baogang) is no higher than in other districts of Baotou, but the concentrations of indoor ²²²Rn and the γ radiation levels in Baogang area, Baiyan Obo area and the rooms with slag building materials are higher than in other districts. With the years' increase of the buildings built, the concentrations of ²²²Rn and ²²⁰Rn are increasing notably, and the γ radiation levels have no obvious trend noticed.

Study on the optimization of public dose constraint value ofuranium processing and fabrication facilities

Wang Meng, Chen Hailong, Pan Wei, Gu Zhijie

RADIATION PROTECTION. 2015, 35(5):  311-316. 

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Taking domestic uranium processing and fabrication facilities (hereinafter referred to as factory A) for example, the author analyzes the factors influencing the optimization of public dose constraint value of uranium processing and fabrication facilities. Using methods of engineering judgment and multiple attribute function analysis, factors including individual effective dose, the public collective effective dose, waste management cost and public reactions of the critical population group are considered to calculate the optimal values of public dose constraint of uranium processing and fabrication facilities weighted by different influencing factors, then the optimal value and maximum efficiency function curve of dose constraint value is calculated using least square method. Finally, the average value of optimal target management of uranium processing and fabrication facilities was calculated by adopting of the weighted average method. Strictly speaking, this value depends on various factors. So this is not “optimal solution”, can only be described as a compromised “satisfactory solution”.

Investigations and analysis of radon concentration in Xi'an metro station

Jing Junbo, Lu Xinwei, Li Changzhuo, Fu Luying, Shi Lu, Qiang Long

RADIATION PROTECTION. 2015, 35(5):  317-320. 

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In order to understand the current level of air radon concentration in Xi'an subway stations, radon concentration levels in the Xi'an subway stations of Line 1 and Line 2 were monitored using Model 1027 continuous radon monitor, and radiation exposure to subway staff was assessed according to evaluation methods recommended by the United Nations Scientific Committee on the Effects of Atomic Radiation. The monitoring results show that the average radon level at stations of Line 1 is 60.27 Bq/m³ and 32.54 Bq/m³ for Line 2. The annual effective dose of air radon to staff in Line 1 and Line 2 stations is 0.63 mSv/a and 0.34 mSv/a, respectively, which are lower than the recommended safe value.