RADIATION PROTECTION

Discussion on some critical issues in the emergency monitoring program of environmental radiation in nuclear power plant

HUANG Yanjun,HE Yi,SHANGGUAN Zhihong,CHEN Chaofeng,ZHAO Feng

RADIATION PROTECTION. 2019, 39(5): 355-364.

Abstract ( 444 )

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As one of the supporting technical documents for the emergency plan, the emergency monitoring program for environmental radiation of nuclear power plant (NPP) is acted as a realistic guide for the emergency preparedness and response. In this paper, the critical issues in the formulation of the emergency monitoring program, including the monitoring scheme, the monitoring methodology, the capacities of facility and equipment, the monitoring response, the capacity maintainability, and the quality assurance, were analyzed and summarized. And technical suggestions for preliminary standardized specification were presented, which would provide a reference for the establishment of the emergency monitoring program for NPP and push forward the standardization progress.

Radiation protection monitoring and control under the condition of fuel defects for PWR

YANG Junwu, ZHOU Liangfa, GAO Xing, LUN Zhengming, LIU Qiang, ZHOU Zhihui, XU Qiang, LI Wei

RADIATION PROTECTION. 2019, 39(5): 365-371.

Abstract ( 204 )

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Monitoring and control plan for gaseous fission products, as wel as radiation protection measures during power operation and refueling outage when fuel defect occurred in PWR, are introduced in this paper. The overall implementation of the plan and measures during sixth outage of Lingao unit 2 (L206) is presented. The monitoring data of air pollution and α contamination are analyzed and evaluated.

Analysis of sampling representativeness of radioactive airborne effluents

YANG Chuan, HE Zeyin, ZHANG Kun, YIN Shirong, SUN Shizheng

RADIATION PROTECTION. 2019, 39(5): 372-378.

Abstract ( 458 )

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In order to obtain a single point continuous monitoring and sampling location of radioactive airborne effluents of nuclear facilities, a gas-solid multiphase turbulent coupling method based on the Discrete Random Walk model is proposed to solve the representative sampling location. The continuous phase is simulated with the k-epsilon turbulent model. Discrete particle model was introduced to simulate the discrete phase. A gas-solid multiphase turbulent coupling calculation method in the stack based on the DRW model is established. Flow field distribution of airborne effluents in the stack is calculated. The relationships between the cyclone Angle, air velocity, tracer gas concentration, aerosol particle concentration and pipe height are analyzed. Analysis results indicated that cyclone angle, COV of air velocity, COV of tracer gas concentration, deviation between the maximum and average tracer gas concentration gradually declined and stabilized. The COV of aerosol particle concentration met the requirements at cross section 6 and 8. The representative sampling location can be determined quickly based on computational fluid dynamics. It provides a theoretical reference for field test of single point sampling of airborne effluents.

Radiation effects and risk assessment for non-human species arising from nuclear power at island

HAO Rui, ZHAO Feng, SHA Xiangdong, JIANG Jun

RADIATION PROTECTION. 2019, 39(5): 379-385.

Abstract ( 259 )

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Island nuclear power plant is one important trend for site selection currently. “Technical guidelines for environmental impact assessment format and content of environmental impact reports for nuclear power plants” (HJ 808—2016) explicitly requires radiation estimation caused by nuclear power plants on non-human organisms. But there is little research in the area of island nuclear power plant radiation effects on the surrounding island organisms. In this paper, we use designed gaseous emission of a Nuclear power plant and ERICA program, along with the results of island local ecological survey, to estimate the radiation and risk of radioactive effluent on island organisms, and carry out the “Three Key Analysis”. The results show that the radiation dose rate of the nuclear power plants on island organisms is less than 10 μGy/h of ERICA dose screening values, and the overall radiation risk is quite low. In view of the applicability of some parameters in ERICA program (such as dose conversion coefficient and concentration ratio), the future research direction is suggested.

Study on testining methods of removal effect of radioactive dust on concrete surface by decontamination detergents

WANG Jing, LI Jian, LV Linmei, Zhao Li

RADIATION PROTECTION. 2019, 39(5): 386-390.

Abstract ( 156 )

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Based on the analysis of testing principles, technical characteristics and surface suitability of various testing methods for removal effect of radioactive contamination, atesting method suitable for removing dust on concrete surface is proposed. The key control factors which influence the testing results have been identified and studied. The influences of detergent dosage, dust dosage and other experimental factors on removal effect are revealed, and some reasonable control principles and methods of experimental factors are obtained. This study may effectively improve the scientific nature, repeatability and reliability of testing methods for evaluating removal effect of radioactive contamination on concrete surface.

Study on primary coolant radioactivity requirement at the time of pressure vessel head opening in PWR

HUANG Qianqian, LV Weifeng, XIONG Jun

RADIATION PROTECTION. 2019, 39(5): 391-395.

Abstract ( 151 )

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The limits of radioactivity concentration in primary coolant during the time of the pressure vessel head opening for shutdown are the important design parameters of PWR, which lack the design methodology in a long time. In this paper, the evaluation methods of radioactive concentration control value in the primary coolant during the time of RPV opening are established for PWR,based on the analysis of the sources of radiation risks after pressure vessel head opening,. This methodology has been verified by comparing with original ERP design results.

Study on pyrolysis treatment technology of radioactive spent ion exchange resin

XU Wei, ZHANG Yu, CHU Haoran, HOU Bonan

RADIATION PROTECTION. 2019, 39(5): 396-402.

Abstract ( 223 )

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A high temperature pyrolysis treatment technology for spent anion and cation exchange resins was studied through the thermal gravimetric analysis (TGA) and bench experiments. The results show that the high temperature pyrolysis of resin can be achieved through heating some metal balls in a reactor by electromagnetic induction and assisted stirring. Air is more suitable than nitrogen and water vapor as reaction atmosphere. In air atmosphere conditions of 1 kg/h resin treatment capacity, 2 m³/h air flow rate, 600-700 ℃ reaction temperature, and CuSO₄·5H₂O additive, the waste residual rates of anion and cation exchange resins are about 8% and 12% respectively, and the final residual rate by pyrolysis can reach about 3%-5%. Distinct difference between pyrolysis reactions of anion and cation exchange resins was observed. The anion exchange resin is more sensitive to heat, and the temperature and air flow required for its pyrolysis are lower. Its reaction is more intense with larger amount of flue gas produced.

Sensitivity analysis of near-field radionuclide release rate for buffer material parameters

LING Hui, WANG Ju, LIU Yuemiao, GAO Yufeng, CHEN Weiming, TONG Qiang

RADIATION PROTECTION. 2019, 39(5): 403-409.

Abstract ( 114 )

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The buffer material is critical to long-term safety for geological repository of high level radioactive waste disposal. Based on safety analysis of long-term evolution scenario for repository, this paper focuses on sensitivity analysis of buffer material parameters such as the buffer thickness, the density of buffer material, and the distribution coefficient of radionuclide in buffer material through the Monte Carlo stochastic simulation method. The simulated results indicate that the near field radionuclide release rate is sensitive to the buffer thickness within 1000 years after closure of a disposal repository, and less sensitive thereafter. The near field radionuclide release rate is not sensitive to the density of buffer material. When the distribution coefficient of radionuclide in buffer material increases continuously, the corresponding sensitivity increases gradually. The results show that the sensitivity analysis can provide feedback guidance for the design of buffer material.

Study on supercritical water oxidation of waste anion resin used in a nuclear power plant

PAN Yuelong, ZHANG Zhidong, CHAI Tao, GAO Yahua, LAN Shuren, LIU Yucun

RADIATION PROTECTION. 2019, 39(5): 410-415.

Abstract ( 183 )

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A supercritical water oxidation (SCWO)equipment of batch-type was used to study the treatment of anion resin water suspension under supercritical conditions. The effects of reaction temperature, reaction pressure, reaction time, and oxidation coefficient on the removal rate of COD were investigated experimentally. At the same time, the effects of catalyst, reaction pressure and reaction temperature on ammonia removal rate were explored. Through an orthogonal experiment, the most significant factors influencing the reaction system are ranked as follows: reaction temperature>reaction pressure>reaction time>oxidation coefficient. The experiment results show that the removal rate of COD for anion resin is 99.65% under the conditions of reaction temperature of 540 ℃, reaction pressure of 26 MPa, reaction time of 8 min, and oxidation coefficient of 3. The content of NH₃-N in anion resin is high and difficult to be removed, so catalysts CuSO₄, MnO₂, and CeO₂ were added, and the catalytic effect was ranked as CuSO₄>CeO₂>MnO₂. The maximum removal rate of NH₃-N was 96.53%.

Design and numerical simulation of a supercritical water oxidation reactor for spent resins

ZHANG Zhidong, CHI Xiangyu, PAN Yuelong, HE Junshan, WANG Naihua

RADIATION PROTECTION. 2019, 39(5): 416-422.

Abstract ( 123 )

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Supercritical water oxidation offers a viable treatment for ion exchange resins. A novel reactor coupled with heterogeneous chemical reaction and homogeneous chemical reaction is proposed, and a numerical model of this reactor based on porous media model has been established by a computational fluid dynamics method. The flow, heat transfer and chemical reaction processes in the reactor have also been investigated. Results show that organic materials are completely degraded under all working conditions and the production requirements have all been met. The outlet temperature, the maximum temperature of the fluid zone, pressure drop and outlet velocity would gradually increase with the increase of heating power.

Study on the influence of high burn-up on storage and transportation of spent fuel

HONG Zhe, ZHAN Lechang, LIU Zhuo, ZHANG Ou, ZHANG Min, LIU Xinhua

RADIATION PROTECTION. 2019, 39(5): 423-428.

Abstract ( 161 )

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The integrity of the spent fuel cladding structure with high burn-up was analyzed. Several important temperature limits, such as fuel cladding temperature limits, cladding solution temperatures, and DBTT, which will affect the structural integrity of the cladding, were discussed. The method for analyzing the structural integrity of the cladding was given. On this basis, the performance of the container to be stored in the dry storage facility for more than 20 years and the structural integrity of the spent fuel during transportation after storage were analyzed and relevant recommendations were given.

Research on dose to patient receiving main types of X ray examinations by DR in Hunan province

ZHU Guozhen, LI Zhichun, GE Liangquan, CHEN Donghui, XU Zhiyong, TAN Xiong, PENG Junzhe, CAO Zhenwei

RADIATION PROTECTION. 2019, 39(5): 429-433.

Abstract ( 314 )

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To study the dose levels to patient from six main examination types in Hunan province. Using random sampling method, 967 patients were selected as the survey samples in research on the dose to patient receiving the common projection type. The typical dose values to patient were Lumbar(AP) 3.9 mGy,Lumbar spine(LAT) 5.7 mGy,Pelvis (AP) 2.5 mGy,hip joint (AP) 3.8 mGy,Chest (PA) 0.3 mGy,Chest (LAT) 0.7 mGy,thoracic (AP) 1.3 mGy,thoracic (LAT) 3.0 mGy,Skull (PA) 0.7 mGy,Skull (LAT) 0.7 mGy. The doses to Patient from the Chest (PA) and Chest (LAT) are 0.77 mGy and 1.44 mGy when the DR has the function AEC. The doses to Patient from the Chest (PA) and Chest (LAT) are 0.27 mGy and 0.56 mGy when the DR hasn’t the function AEC. The dose to patient receiving six types of X ray examinations by DR is lower than the dose Guidance Level in GB 18871—2002《Basic standards for protection against ionizing radiation and for the safety of radiation sources》in HUNAN province. The function of AEC has positive effect in dose controlling. To some extent,The function of AEC can optimize the dose receiving from X ray examinations by DR.

Research on cognition of nuclear emergency knowledge of college students in Hunan

LI Juan, LI Guili, LIU Yufei, LI Rui, YUE Xiaoqiao, HUANG Bo

RADIATION PROTECTION. 2019, 39(5): 434-438.

Abstract ( 288 )

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To learn about the knowledge of nuclear emergency cognized by college students, we used self-designed questionnaires to investigate 342 college students in university. The scores of nuclear emergency related knowledge provided in schools are low (45.6±11.3 points), as only 19.30%. With no regard to census register, the nuclear emergency knowledge level and concern about the Japanese nuclear leakage accident are different among different ages, genders, and majors. With no consideration of ages, the genders, majors, and the census registers have different influences on students’ knowledge of the distribution of nuclear power plants in China. It shows that different genders and census registers, different levels of attention to nuclear accidents. For levels of interest in school nuclear emergency training, there is also significant difference in ages and census registers. Some effective measures should be taken to improve the knowledge of college students’ nuclear emergency knowledge, so they can adopt correct self-help methods in case of nuclear accidents.

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