RADIATION PROTECTION

The design of radiation protection optimization for HPR1000

MAO Yawei, MI Aijun, WANG Xiaoliang, LIU Xinjian, CHEN Qiaoyan, QIU Lin, GAO Guiling

RADIATION PROTECTION. 2021, 41(1): 1-8.

Abstract ( 203 )

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Optimization of radiation protection is the key issue in the radiation protection design of Nuclear Power Plants. Based on the optimization strategy for radiation protection described in IAEA Guidance, this paper gives a brief overview of the radiation protection optimization design for HPR1000 from the aspects of optimization design strategy, design goal, design content and evaluation, and ensuring continuous improvement. The optimization design contents of HPR1000 are introduced, including source terms, radiation zone, radiation protection after accident, dose assessment. The principle of radiation protection optimization has been effectively implemented in the design of HPR1000.

The development history of uranium mine ventilation in China

LI Xianjie, ZHANG Zhe, HU Penghua, CHEN Gang, REN Jianjun

RADIATION PROTECTION. 2021, 41(1): 9-16.

Abstract ( 186 )

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Ventilation for reducing radon is one of the main parts of radiation protection in uranium mines. The ventilation technology development of uranium mines in China has been compartmentalized into five periods, based on the ventilation technique advancement and radon exhalation study: the 1958-1965 period, the establishment period of ventilation technology for reducing radon, which mainly reflects in learning and introducing the positive pressure ventilation method in the former Soviet Union; the 1966-1977 reposeful period, reflecting in the application verification and experience accumulation of negative pressure ventilation instead of positive pressure ventilation; the 1978-1989 period is the theory and technique developing period, the theory of interaction between seepage-diffusion law of radon exhalation and tunneling method is proposed, which promotes the development of uranium mine ventilation protection technology; the 1990-2002 is the stagnation period, for the purpose of reducing the input of ventilation protection to reduce production costs, the research on uranium mine ventilation protection technology is at a standstill; and the 2003-2016 is the intensified developing period, using seepage-diffusion law of radon exhalation to guide the research and application of ventilation protection technology under different mining methods. The ventilation technology system for reducing radon of uranium mine in China is coming into being.

The calculation of the content per intake of ²⁴¹Am based on ICRPPublication No. 141 and comparison of the new and old models’ results

CHEN Qianlan, LUO Zhiping, LIU Senlin

RADIATION PROTECTION. 2021, 41(1): 17-26.

Abstract ( 171 )

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The content per intake calculating program of the new transuranic elements' biokinetic model in ICRP Publication No.141 is established using Matlab Simulink software. Then program is used to calculate the content per intake of Am-241 inhaled by reference worker (breathing rate 1.2 m³/h, AMAD 5 μm), and the new model’s results are compared with the old models’ results. The new model’s retention indicates: (1) the factors of absorbing type and the AMAD size have big influence on calculation of content per intake in lung, skeleton, liver. It is necessary to make clear the two factors of the radiation source before internal dose evaluation; (2) Type S Am aerosol inhaled mainly remains in lungs, skeleton after 1 000 days; Type F Am aerosol inhaled is mainly absorbed in skeleton and liver; Type M Am aerosol inhaled distributes quite similar amount in lung, skeleton and liver in the early 300 d. Therefore the internal dose direct monitoring methods should be different for Type S, M and F Am inhaled. The comparing of the new and old biokinetic models indicates: (1)for the lung content per intake calculated, the short-term results (＜1 000 days after inhalation) of Type S and M are similar for the two models, the results of the middle and long-term (≥1 000 days after inhalation) from the new model is much higher than that of the old model, and the difference of the two models’ predictions are more obvious for Type S than Type M of Am; for Type F the difference of the two models are very big at any day. (2) For skeleton and liver content per intake calculated, for the Type S Am inhaled, the middle and long-term results of the new model are quite higher than the old model’s, while the two models’ short-term predicts are similar. For Type M and Type F of Am,the skeleton (liver also the same) content per intake predicted by the new model is lower than the old model’s, and the models’ difference of Type F Am is more obvious than that of Type M Am.

Study on the optimization of sample introduction flow rate for aerosoldirect injection device and aerosol loss rate

WU Hao, WANG Chuangao, ZHENG Guowen, PANG Hongchao, LUO Zhiping, CHEN Ran, CHEN Ling, WANG Zhongwen

RADIATION PROTECTION. 2021, 41(1): 27-32.

Abstract ( 138 )

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Considering the impact of oxygen in the air on the measurement results of ICP-MS, the air direct injection measurement could lead to the reduction of ionization degree for nuclides to be measured, which may decrease the sensitivity of ICP-MS or even cause an outage. Therefore, in order to measure air aerosol samples by ICP-MS directly, a set of aerosol direct injection device was designed to achieve the exchange between air and working argon, so as to make sure that aerosol can be carried into ICP-MS for measurement by argon. Conversion efficiency of the device was measured and analyzed. Analyzing the system background levels measured by the on-line system of ICP-MS, and estimating the operating costs the whole system, the optimum programme was chosen while the inlet flowrate of air sample was 0.8 L/min, and the inlet flowrate of argon was 10 L/min. According to study on the aerosol loss rate, it is proved that the aerosol loss does not need to be considered when aerosol passes the direct injection device. With the development and performance test of the aerosol direct injection device, air aerosol can be measured directly and fast by ICP-MS, laying a foundation for rapid and quantitative measurement of various long-life radioactive nuclides in air aerosol by ICP-MS in the future, and providing a new method to measure aerosol continuously and rapidly in sites and effluents.

Tritium distribution characteristics and evaluation of effective publicannual dose around China Nuclear Power Institute

DU Yunwu, DENG Xiaoqin, MAO Wanchong, ZHANG Li

RADIATION PROTECTION. 2021, 41(1): 33-38.

Abstract ( 183 )

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Based on the monitoring data of tritium in the air and water around the nuclear base of nuclear power institute from 2014 to 2017, the effective doses to be accumulated in various pathways of key residential groups were roughly estimated. The results show that there is no significant difference in tritium activity concentration between the upstream, downstream and drinking water, indicating that the sampling time of surface water may be misplaced with the discharge time of liquid effluent, or the tritium of liquid effluent at 1km downstream of the discharge outlet has been diluted to the background level; the tritium activity concentration in air and rain water decreases with the increase of distance from the nuclear base. The average annual tritium intake of adults, teenagers, children, young children and infants in the residential groups near the complex building was 1.52, 1.44, 1.05, 0.681 and 0.562 kBq/a, respectively. And the annual effective doses to be accumulated were 0.027 4, 0.026 1, 0.024 2, 0.021 1 and 0.026 9 μSv/a respectively. The maximum effective dose of tritium in adult group was 0.027 4 μSv/a, but it only accounted for less than 1 ‰ of the target dose (0.25 mSv/a). It can be concluded that tritium has little effect on the surrounding environment of nuclear base under normal operation conditions.

Analysis of thermal neutron shielding performance ofboron-aluminum composite material

LI Changyuan, XIA Xiaobin, YANG Zhongtian, LIU Yuchen, ZHANG Zhihong, WANG Jianhua, CAI Jun, HU Jifeng

RADIATION PROTECTION. 2021, 41(1): 39-43.

Abstract ( 162 )

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Boron-aluminum composite material has been widely studied due to its relatively simple preparation process, good mechanical properties, low price of raw material. Boron-aluminum composite material has been used as thermal neutron absorbing materials in many fields. In this paper, the thermal neutron shielding performance of boron-aluminum composites was analyzed by theoretical formula, MCNP software and experimental measurement. Theoretical calculations show that boron is better as a composite additive than boron carbide when the additive content is the same only in terms of thermal neutron absorption properties of the material. Through simulation calculation of MCNP program and through experimental measurement, it is found that the boron-aluminum composite material has excellent neutron absorption performance with neutron energy less than 10^-7 MeV.

Analysis on radiation protection design of AP1000 unit of Sanmen nuclear power plant

LI Hui, ZANG Yikun, HU Yipeng

RADIATION PROTECTION. 2021, 41(1): 44-49.

Abstract ( 183 )

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AP1000 unit of Sanmen nuclear power plant is the third-generation nuclear power unit. In order to reduce the unit radiation level and occupational exposure dose, the design of radiation protection has adopted many measures, such as adding zinc of primary, operating at higher pH value, shutdown oxidation operation, steam generator primary water chamber electropolishing, optimizing equipment maintenance, optimizing shielding design, wireless dose monitoring, etc. This paper introduces the radiation level and occupational radiation dose data of AP1000 units of Sanmen nuclear power plant during operation and overhaul, and compared with the relevant data of domestic CPR1000 units, to analyse the radiation protection design of AP1000 units. And giving the suggestions on radiation protection operation management and technical improvement of AP1000 units of Sanmen nuclear power plant.

Investigation on performance requirements and test methods oflow and intermediate vitrified radioactive waste

GAN Xueying, XU Chunyan, FANG Lan, WANG Shijun, HE Wei, LIU Xinhua

RADIATION PROTECTION. 2021, 41(1): 50-57.

Abstract ( 188 )

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Vitrification technologies were applied in the treatment of low and intermediate-level radioactive wastes for waste volume reduction and the improvement of waste disposal safety. The relevant standard of performance requirements and test methods for the low and intermediate vitrified radioactive waste are absence in China currently. The performance requirements and test methods were suggested by comparison analysis with several waste form standards and some relevant published test data.

Equipment development and performance experiment ofradioactive waste supercompactor

ZHENG Wei, QIAO Baoquan, LIN Peng, ZHOU Dongsheng, ZOU Liping, WANG Zhaohui, ChenJunjie, LIU Xiajie

RADIATION PROTECTION. 2021, 41(1): 58-63.

Abstract ( 156 )

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In order to achieve the indigenous production of radioactive waste supercompactor, an advanced supercompactor was developed and a serious of compression experiments based on various typical solid wastes of nuclear power plant were conducted. Results showed that, the supercompactor remains stable and there was no liquid leaking into environment during the process of transfer after running for over 200 hours. In addition, the compression efficiency exceeded seven tanks per hour. The supercompactor can be applied in the treatment of compressible radioactive solid wastes in nuclear power plant.

Test on hydrogen explosion accident in high level liquid waste tank

CHEN Lei, WANG Pengyi, SUN Shutang, ZHUANG Dajie, MENG Dongyuan, LIAN Yiren, YAN Feng, ZHANG Jiangang, LI Guoqiang

RADIATION PROTECTION. 2021, 41(1): 64-70.

Abstract ( 135 )

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High level liquid waste generates hydrogen gas during storage. If the hydrogen in the high-level liquid waste tank is not discharged or diluted in time, and when the concentration of hydrogen in the mixed gas reaches the explosion threshold, an explosion will possibly occur. In this paper, the explosion pressure and wall temperature of hydrogen in the high-level liquid waste tank are studied by using the experimental device. The test results show that the maximum explosion over-pressure is about 0.596 5 MPa, and the maximum wall temperature is about 110 ℃ when the hydrogen concentration is 30%(vol), and the ignition position is near the top center of the tank.

Investigation and analysis on frequency of clinical nuclear medicine andradiotherapy in Hunan province in 2016

PENG Junzhe, LI Zhichun, TAN Xiong, CHEN Donghui, XU Zhiyong, ZHU Guozhen

RADIATION PROTECTION. 2021, 41(1): 71-75.

Abstract ( 171 )

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To study the basic status and the frequency of clinical nuclear medicine and radiotherapy in Hunan province, general survey and sampling survey were conducted. We had investigated the basic situation of all medical institutions of clinic nuclear medicine and radiotherapy in Hunan, and had investigated the frequency of gender, age and type of diagnosis and treatment in the sampled medical institutions. The frequency of clinical nuclear medicine diagnostic and treatment were 1.37 and 0.139 per 1 000 population respectively. The frequency of radiotherapy was 0.412 per 1 000 population. The frequency of clinical nuclear medicine and radiotherapy in Hunan province has increased significantly since 1998, which was a sustained growth with the development of society. It is necessary to pay attention to the radiation protection of patients and staff as well as the quality control of diagnosis and treatment.

Analysis of radiation accident cases handled inour hospital from 2003 to 2018

SHANG Wei, WANG Shouzheng, ZHANG Huisheng, LI Li, LING Yan, DU Zhiguo, WEN Wen, LU Chiqiao

RADIATION PROTECTION. 2021, 41(1): 76-80.

Abstract ( 160 )

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In this article multiple radiation accidents were reporte,among them four cases caused overexposure, one case of degree I local acute radiation skin injury, another case of degree IV. Through analysis of the causes of these radiological accidents and medical observations of the patients, we can further understand the radiation hazards caused by different accidents, summarize the medical diagnosis and treatment experience of radiological injuries. It is suggested that radiation protection management for key occupational groups should be strengthened to prevent the occurrence of radiological accidents.

Consideration and enlightenment from the population-related sitingreqirements of NRC for small modular reactors

XIONG Xiaowei, WANG Yichuan, YANG Li, FAN Yuan, CHEN Xiaoqiu, YANG Duanjie, MAO Yuxian, SHANG Zhaorong

RADIATION PROTECTION. 2021, 41(1): 81-87.

Abstract ( 109 )

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This paper introduced the U. S. Nuclear Regulatory Commission (NRC) revision plans for the population-related siting considerations for advanced small modular reactors. Four options including their pros and cons which were developed by NRC for consideration to address siting questions for small advanced reactors had been analyzed in detail. The problems that should be paid attention to during site evaluation process for small modular reactors in China are put forward: (1) Considering the overall social risks and from the perspective of plant site comparison, it is necessary to establish an appropriate distance between the reactor and the boundary of the densely populated center (population center). In addition, taking the defense-in-depth concept into account, the small modular reactor site still needs to keep a proper distance from the densely populated center; (2) based on the radiological consequences and influence scope of the accident of small reactors, it is beneficial to develop an evaluation index with the equivalent social risk between small modular reactors and typical large LWR can be developed such as the collective effective dose around the plant site under accident conditions; (3) the public acceptance and siting policy are the important factors for whether the small modular reactors can be located near the densely populated center.

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