RADIATION PROTECTION

CONTENTS

RADIATION PROTECTION. 2020, 40(6): 484-484.

Abstract ( 75 )

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An evaluation of the influence of impurity concentration and composition in concrete on activation and radiation dosein the reactor facility

Jae Hyun KIM, Myeong Hyeon WOO, Chang Ho SHIN, Jong Kyung KIM

RADIATION PROTECTION. 2020, 40(6): 486-490.

Abstract ( 216 )

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The content analysis of radioactive waste and radiation dose evaluation is considered as one of the important factors in the reactor facility design.This kind of buildings consists of the concrete for the most part and uses it as the structure and shield of the building.Generally,the concrete has impurities such as cobalt,europium,nickel,and cesium with specific content depending on the production method or manufacturing company.Dominant radioactive nuclides generated from the fundamental components of concrete are considered that it is less contributed to the radiation dose because they are beta decay nuclides in general.Thus,impurities of irradiated concrete in the reactor facilities,are considered occasionally an important evaluation factor for induced activity.In this study,the influence on the activation of impurities in concrete was evaluated from the radiation dose and induced activity calculations.The calculation was evaluated at the bio-shield which is one of the areas with the highest neutron irradiation among the concrete structure in the reactor facility.The results show that radioactive nuclides with gamma decay were produced in these impurities.Moreover,the radiation dose of concrete with impurities was higher than concrete without impurities.The increased radiation dose was quantified through the content of impurities.

Measurement of niobium reaction rate formaterial surveillance tests in fast reactors

Chikara Ito, Shigetaka Maeda, Toshihiko Inoue, Hideki Tomita, Tetsuo Iguchi

RADIATION PROTECTION. 2020, 40(6): 491-495.

Abstract ( 211 )

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A highly accurate and precise technique for measurement of the ⁹³Nb(n,n’)^93mNb reaction rate was established for the material surveillance tests,etc.in fast reactors.The self-absorption effect on the measurement of the characteristic X-rays emitted by ^93mNb was decreased by the dissolution and evaporation to dryness of niobium dosimeter.A highly precise count of the number of ⁹³Nb atoms was obtained by measuring the niobium solution concentration using inductively coupled plasma mass spectrometry.X-rays of ^93mNb were measured accurately by means of comparing the X-ray intensity of irradiated niobium solution with that of the solution in which stable ⁹³Nb was added.The difference between both intensities indicates the effect of ¹⁸²Ta,which is generated from an impurity tantalum,and the intensity of X-rays from ^93mNb was evaluated.Measurement error of the ⁹³Nb(n,n’)^93mNb reaction rate was reduced to be less than 4%,which was equivalent to the other reaction rate errors of dosimeters used for Joyo dosimetry.In addition,an advanced technique using Resonance Ionization Mass Spectrometry was proposed for the precise measurement of ^93mNb yield,and ^93mNb will be resonance-ionized selectively by discriminating the hyperfine splitting of the atomic energy levels between ⁹³Nb and ^93mNb at high resolution.

An optimization of shielding structure in controllable neutron sourceneutron-gamma well logging instrument

SONG Lei, LI Fusheng, WANG Sheng

RADIATION PROTECTION. 2020, 40(6): 496-503.

Abstract ( 203 )

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Neutron-gamma well logging instrument based on controllable neutron sources is widely used in the exploration of petroleum,shale gas and minerals.During the well logging process,neutron is emitted from neutron source and interacts with the surrounding stratum’s elements.Thus,the specific gamma rays are generated and detectors within the well logging instrument can detect them.After analyzing the spectrum of gamma rays,the component of the stratum can be obtained.However,part of neutrons and gamma rays may go through the shielding structure and reach into the detector.These particles can introduce noise signals into the exploration result and reduce the accuracy of the instrument.Thus,it is important to develop an efficient shielding structure to reduce the flux of neutron and gamma rays which may induce the noise.For this purpose,a method employed genetic algorithms and MCNP code was established to develop the shielding structure.As a typical case,the D-D fusion neutron source and BGO detector were adopted in this study.Three shielding structures with different thickness were obtained.The results showed that,these shielding structures have the better performance than the common shielding structures.

Development of fast estimation code for PWR source term

FENG Zongyang, ZHANG Jiangang, YANG Yapeng, JIA Linsheng, WANG Renze, WANG Ning

RADIATION PROTECTION. 2020, 40(6): 504-509.

Abstract ( 183 )

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Source term released into environment is the basis for accident power plant to determine appropriate emergency response level and protective action decision.Source term estimation based on power plant operating conditions is the most important emergency evaluation content during the emergency response period of nuclear power plant (NPP).Based on the relevant technical documents of IAEA and NRC,the procedures and basic data of accident release source term estimation based on pressurized water reactor (PWR) conditions were introduced,and seven practical accident release source term estimation methods were summarized.Based on these methods,a rapid release source term estimation program for PWR accident is developed.The program provides four release pathways for different estimation methods:containment leakage,containment bypass,steam generator tube rupture (SGTR) and direct environmental release.Reduction effect by decay,retention,spray and filtration etc.during release are considered in different release pathways except direct environmental release.The calculation results are close to that of RASCAL software released by Nuclear Regulatory Committee of United States.

Design of beam dump for the commissioning at SHINE facility

XU Yuhai, WANG Guanghong, LI Zhefu, XU Wenzhen, ZHANG Bintuan, LV Jiongjun, YANG Fubin, XIA Xiaobin

RADIATION PROTECTION. 2020, 40(6): 510-515.

Abstract ( 172 )

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SHINE facility,Shanghai hard Xray free electron laser (XFEL) facility,and jointly built by Shanghai Institute of advanced research and Shanghai Institute of Applied physics,Chinese Academy of Sciences (SINAP,CAS).It is currently under construction in Zhangjiang Science Park,Shanghai.The SHINE facility consists of an 8 GeV continuous-wave (CW) superconducting linear accelerator,a 100 PW laser system,three undulator lines,and ten experimental end-stations at phase-I.A beam dump set at BC3 accelerator section is designed for beam commissioning with beam energy of 7.7 GeV and beam current 0.3 μA based on the preliminary conceptual design.This paper presents the heat analysis of the dump using ANSYS 17.0 and the activation analysis of the beam dump for radiation safety evaluation by using Monte Carlo simulation code FLUKA.It is assumed conservatively that the beam is continuously injected into the dump for 2 000 h during normal operation.The activation analysis showed that the surrounding concrete shielding with the thickness of 55-85 cm for the dump is required,in order to keep the residual dose rate outside the dump below 25 μSv/h at 1 hour after stopping the beam injection.

Study on BP neural network algorithm for predicting neutron shieldingeffect of multi-composition materials

LIN Haipeng, LI Guodong, CHEN Faguo, HAN Yi, LIANG Runcheng

RADIATION PROTECTION. 2020, 40(6): 516-521.

Abstract ( 169 )

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The availability of applying BP neural network algorithm to boost the optimization process of multi-composition neutron shielding materials,in which conventional Monte Carlo simulation would cause severe time consumption,has been discussed.A typical 3-layer BP neural network model has been established with 300 random mass components of composite materials and their corresponding dose values calculated by Monte Carlo as training samples.The absolute deviations between the predicted dose value and the sample value are within ±2.For validation samples outside the training samples,the absolute deviations expend to -6.4 to 5.2.According to the deviation distribution statistics,for over 70% of the total samples,the relative deviation absolute values are within 2%.The calculation accuracy and generalization ability of the neural network model are qualitatively determined to meet the requirements of the optimization algorithm.The cross-validation method can improve the calculation accuracy of the training samples,whereas increasing the calculation deviation of the verification samples,indicating that the balance between the fitting degree and generalization ability of the samples should be considered during the establishment of the neural network.

Rat organ dose calculation for external photon irradiation basedon voxel model based on voxel model

ZHANG Xiaomin, LI Dawei, NING Jing, XIE Xiangdong

RADIATION PROTECTION. 2020, 40(6): 522-526.

Abstract ( 170 )

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To obtain the organ dose of externally exposed rat for studying dose-effect relationship in radiation medicine,a voxel model based on Micro-CT images of a SD rat was developed to calculate rat organ dose for external photon irradiation.The rat voxel model had a size of 0.16 mm×0.16 mm×2 mm,weighed about 323 g and contained most of the key organs or tissues.Monte Carlo simulation was carried out to obtain organ dose conversion coefficients for 21 external monoenergetic parallel photon beams between 10 keV and 10 MeV under four different irradiation geometry conditions.The organ dose with varying photon energy was also discussed and analyzed.

Verification of modified sum-peak method in Monte Carlo simulationof full energy peak and sum-peak efficiencies of a HPGe detector

Tsukasa Aso, Yoshimune Ogata, Hidesuke Itadzu

RADIATION PROTECTION. 2020, 40(6): 527-532.

Abstract ( 146 )

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The modified sum-peak method estimates radioactivity by using only the peak and the sum-peak count rates.To verify the modified sum-peak method,the dependence of the full energy peak efficiency on the source-to-detector distance in a high-purity germanium detector system was studied using a Geant4 Monte Carlo simulation.The effect of the dead-layer in the germanium crystal was estimated by reference to experiments on ²⁴¹Am and the relative efficiency of the detector.The peak efficiency dependence on the source-to-detector distance was compared between the simulation and measurements.The modified sum-peak method is discussed with respect to these peak efficiencies.

On line continuous liquid radioactive effluent monitoring

SHEN Fu

RADIATION PROTECTION. 2020, 40(6): 533-539.

Abstract ( 277 )

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In this paper,a radionuclide identification continuous liquid radioactive effluent prototype monitor based on NaI detector was developed based on indigenous patent methods for the measurement of the key gamma radionuclide such as ⁶⁰Co,¹³⁷Cs in nuclear facilities.The monitor had been tested for more than 500 hours.Results showed it is stable and reliable,and has the ability of radionuclide identification.Through the national level Laboratories verification of the efficiency and 500 hours test,the detection limit of continuous liquid radioactive effluent monitor was achieved around 0.088 Bq/L.The results showed that it improves the detection limit nearly two orders of magnitude lower and the weight decreased one order of magnitude lower.The device can also be applied to monitor key gamma nuclides activity concentration in drinking water.This study has important reference value for the detection of effluent.And the developed instrument could be easier for radiation controlling monitoring because of the lower installation conditions requirements than traditional instruments.

Response evaluation of a compact and highly efficient neutron diffractometer for compact accelerator neutron sources

Sho Imai, Kenichi Watanabe, Astushi Yamazaki, Sachiko Yoshihashi, Akira Uritani, Seiji Tasaki, Setsuo Satoh

RADIATION PROTECTION. 2020, 40(6): 540-544.

Abstract ( 209 )

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We fabricated the spheroid neutron diffractometer with 79 neutron detector rings using the transparent rubber sheet type Eu:LiCaAlF₆ scintillators and wavelength-shifting fibers.We confirmed that the fabricated detector shows a clear neutron peak and can discriminate neutron and gamma-ray events in a signal pulse spectrum.We additionally checked that the fabricated diffractometer can detect a neutron diffraction peak of ferrite powder at Kyoto University Accelerator-driven Neutron Source.Consequently,it can be expected that crystal structural analysis will be possible even by small accelerator neutron sources.

In-situ evaluation for activated concrete in accelerator facility with scintillation-type gamma-ray spectrometer

Go Yoshida, Hiroshi Matsumura, Koichi Nishikawa, Akihiro Toyoda, Yoshiharu Miyazaki, Kazuyoshi Masumoto, Hajime Nakamura, Taichi Miura

RADIATION PROTECTION. 2020, 40(6): 545-549.

Abstract ( 193 )

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The assessment of activated concrete is particularly difficult during the decommissioning of an accelerator facility.Destructive analysis by core boring is the only method of investigating the activity of concrete material.To address this problem,an in-situ and nondestructive analysis method was developed to determine γ-ray-emitting nuclides and their specific activities in the concrete walls and floor by using a portable germanium semiconductor detector.In this work,we examined a substitute for Ge detector to establish a simpler and more convenient method.As candidates,we focused on some scintillation type spectrometers,and the possibility of a substitute for a Ge detector was examined by both simulation and experiment.The detection limits were roughly estimated through Monte Carlo simulation for various scintillation crystals,and it was found that 1.5-inch LaBr₃,CeBr₃,and SrI₂ could distinguish the clearance level.It was confirmed that the 1.5-inch LaBr₃ could reproduce the calibration curve of the Ge detector in the experiment.The required thickness and length of the radiation shield for suppressing the background radiation during the measurement was also determined for the convenience of an actual decommissioning work.

Study on reducing the low detection limit of net count rate by low backgroundα, β measurement instrument from theoretical point of view

JIN Tao, SUN Yu, ZHANG Zhipeng, WU Yao, SONG Jigao, LI Junjie, LUO Yuanpan, PEI Min, ZENG Bo

RADIATION PROTECTION. 2020, 40(6): 550-555.

Abstract ( 171 )

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In this paper,the formula of the low detection limit of the net count rate in radioactive measurement is derived from the point of non-equality,the actually probability of making type II errors corresponding to the two typical low detection limit is calculated.In the absence of an approximation,the number of radionuclides that have decayed follow a binomial distribution.When the average of the net count rate is the low detection limit of the net count rate,the probability that the random value of the net count rate is less than the decision threshold is the probability value of making type II errors.Combining the number of radioactive nuclei at initial moment,the number of radioactive nuclei in the whole measurement time that has decayed and the probability of radioactive nuclides decay,the actual probability value of making type II errors can be calculated.It is found that the maximum practical probability value of the type II errors is 1 to 2 orders of magnitude less than the nominal value in most case.Therefore,the low detection limit can be modified,and an appropriate correction coefficient can be given,so that the actual probability value of making type II errors is close to and slightly less than the nominal value.Through calculation,it is found that the correction coefficient given by k_α=1.645 and k_β=1.645 can reduce low detection limit by 22 percent.

The performance analysis of large area portablesurface contamination survey meter

ZHANG Jing, HOU Lei, RAO Xianming, LIU Jinjin, DU Xiangyang, ZHANG Jia, QIAO Minjuan

RADIATION PROTECTION. 2020, 40(6): 556-560.

Abstract ( 147 )

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Aiming at a large area portable surface contamination survey meter used in measuring α,β radiation,and ensuring the performance of measurement of the meter,the main performance of self-made instrument was measured based on JJG 478—2016 and GB/T 5202—2008 standards,including the uniformity of detector,surface emission rate response,minimum detectable limit,response time and temperature stability.The results show that the surface emission rate response of alpha for ²⁴¹Am is around 30.4%,of beta for ⁹⁰Sr-⁹⁰Y is around 48.1%,and of beta for ³⁶Cl is around 45.5%.The minimum detectable limit of alpha for ²⁴¹Am is 0.15 Bq/cm²,of beta for ⁹⁰Sr-⁹⁰Y is 0.07 Bq/cm²,and of beta for ³⁶Cl is 0.09 Bq/cm².The response time is less than 4 s.The instrument can operate normally in the temperature range of -10 ℃ to 40 ℃.These results are superior to the requirements of related standards and achieve the purpose of preventing the spread of radiation contamination to ensure the safety of field staff.

Main performance test of large area gas-sealed proportional counter

QIAO Li, RAO Xianming, DU Xiangyang, REN Yi, ZHANG Jia, GUO Xirong, WANG Yanfei

RADIATION PROTECTION. 2020, 40(6): 561-564.

Abstract ( 158 )

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Shanxi Zhongfu Nuclear Instruments Co.,Ltd.(CIRNIC) has independently developed and produced various types of large area gas-sealed proportional counter of 40-600 cm².An experiment focusing on background count rate,plateau curve,surface emission rate response,coefficient of variation and other physical characters,on one of the device of which with 517 cm² effective area has shown that under the same testing environment and experimental conditions,background count rate is 24.5 cps,surface emission rate 50.4%,coefficient of variation 0.85%.

Discussion on short-lived radionuclide aerosol monitoring

MENG Dan, MA Tao, LI Jianwei, ZHANG Fuguo, CHANG Xiang, YANG Yi, YANG Liu, MA Yinghao

RADIATION PROTECTION. 2020, 40(6): 565-570.

Abstract ( 169 )

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Continuous radioactive aerosol monitors,or continuous air monitors (CAMs) are referred usually to the long-lived α/β nuclides (such as U,Pu and ¹³⁷Cs etc).Thus it has not considered the decay corrections of radioactive nuclides themselves during sampling and measuring.But it is important as well to monitor short-lived nuclides (such as ⁸⁸Rb,¹³⁸Cs and ¹⁸F,etc) in many nuclear factories including nuclear power plants where the nuclear reactors are to be operated.Therefore,the monitoring of short-lived nuclides is important.This paper discusses the problems about the monitoring of short-lived nuclides,and suggests the data processing methods that can be used for the real monitors to obtain reasonable monitoring results of short-lived nuclides.

Development of an on-line continuous aerosol monitorsuitable for high radon environment

MENG Dan, YANG Liu, MA Yinghao, CHANG Xiang, MA Tao, ZHANG Fuguo, YANG Yi, SHANG Jie

RADIATION PROTECTION. 2020, 40(6): 571-576.

Abstract ( 180 )

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Through the combined application of radon subtraction technology of particle size separation sampling,vacuum measuring mode,energy screening method and α/β ratio method,most of the interference of radon thorium and their daughters on radioactive aerosol monitoring has been deducted.And an on-line continuous aerosol monitor suitable for high radon environment was developed.It has been verified in operation under different radon concentrations and the effect is excellent.It is suitable for places with high radon concentration such as related nuclear facilities,and it provides an important guarantee for the radiation safety of professional personnel such as fuel leaking inspections.

Long-lived neutron-induced radioisotopes in OKTAVIAN facility concrete wall after 38 year-operation

Fajar Panuntun, Shingo Tamaki, Sachie Kusaka, Fuminabo Sato, Isao Murata

RADIATION PROTECTION. 2020, 40(6): 577-582.

Abstract ( 179 )

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An intense 14 MeV neutron source facility named OKTAVIAN was installed in the A15 building,Osaka University in 1981.Along the operation period,new radioisotopes with various half-life have been produced as neutron activation products in its concrete wall shield.In this work,we investigated the concrete wall in the heavy irradiation room of OKTAVIAN using gamma spectrometry method to discover the presence of radioisotope having large half-life value (long-lived radioisotope) as neutron activation products.Computational simulations were performed prior to measurement to predict the presence of long-lived radioisotopes by employing MCNP5 and FISPACT codes.A pre-calibrated Germanium detector with high energy resolution was employed to measure the concrete.Several long-lived activation products have been observed such as ¹⁵²Eu,⁵⁴Mn,⁶⁵Zn,²²Na and ⁶⁰Co.The activity of each radioisotope was derived after estimating the detector efficiency using MCNP5.As a result of the measurement and analysis,the followings are concluded:(1) Though presence of activation products represents radiological risk to everyone who performs an experimental activity in the irradiation room of the OKTAVIAN facility,the present result shows that past experiments were carried out safely without any significant additional exposure dose coming from the wall for the last 38 years.(2) The approximated total fluence of D-T neutrons to the wall was successfully estimated from the produced radioisotope,¹⁵²Eu,because it has the longest half-life of 13.5 years among the observed radioisotopes.(3) From the results of (1) and (2),it could be possible to estimate the total activity of the concrete wall in the OKTAVIAN facility,which is very essential and important information,because this would be very valuable for decommissioning or disposal of the facility in the future.

Assessment of internal dose from tritiumof Qinshan Nuclear Power Base

YANG Jie, LIAN Bing, ZHAO Yangjun, WANG Yan

RADIATION PROTECTION. 2020, 40(6): 583-586.

Abstract ( 217 )

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The monitoring results of tritium concentration in the surrounding residential areas of Qinshan Nuclear Power Base from 2014 to 2016 were analyzed.Based on the statistical environmental monitoring data,the radiation dose of the surrounding public caused by tritium emission in recent yeas in Qinshan Nuclear Power Base was evaluated.Evaluation results show that the average internal dose caused by tritium emission of Qinlian residents groups of infants,young children,teenagers and adults were 0.59,1.35,1.18 and 0.92 μSv/a respectively,which is far less than the prescribed effective dose of 0.25 mSv/a caused by radioactive effluent of nuclear power reactor release,which was stipulated by《Regulations for environmental radiation protection of nuclear power plant》(GB 6249—2011).Qinlian residential area is located 2.4 km from WNW azimuth of the Qinshan Nuclear Power Base.It can be seen that the tritium emission in Qinshan Nuclear Power Base is stable and controlled,and the environmental and public impact is acceptable.

Distribution characteristics of ⁷Be and ²¹⁰Pb in near surface aerosolsin Mianyang area and their tracing of O₃

LIU Yuhang, HUANG Ying, GU Chuan, DU Yuanyuan

RADIATION PROTECTION. 2020, 40(6): 587-592.

Abstract ( 148 )

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In order to understand the radiation environmental quality status in the air near the ground in Mianyang,grasp its changing trend,and to explore the source of near surface aerosol and its significance for tracing the concentration of near surface O₃,the aerosol samples were analyzed in detail from Anzhou District,Jiangyou City,Zitong county and Pingwu County of Mianyang city from March 2018 to February 2019.In general,the activity concentration of ⁷Be was higher in spring and autumn,and the lowest level was in summer,with an annual average of 1.90-2.13 mBq/m³,which was basically consistent with the global distribution characteristics of inland,mid latitude and low altitude areas.The annual average of ²¹⁰Pb activity concentration was 1.24-1.66 mBq/m³,which was the relatively high value of ²¹⁰Pb activity concentration on the global land.Through the correlation analysis and P value test of ⁷Be,¹⁰Pb and ⁷Be/²¹⁰Pb ratios in near surface aerosols and O₃,the results show that there is a very significant positive correlation between ⁷Be/²¹⁰Pb ratio and near surface O₃,which can be used as a good tracer for the source of O₃ in the near surface air.The source of O₃ in the near surface air in Zitong and Pingwu counties in Mianyang city is greatly affected by the vertical convection.

The impact of construction of port approaching industrial zone ondilution and diffusion of liquid effluent from a nuclear power plant

LI Ting, ZHU Jun, LIU Tuantuan, ZHANG Aiming

RADIATION PROTECTION. 2020, 40(6): 593-604.

Abstract ( 148 )

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The construction of the port approaching industrial zone has led to changes in the local tidal field of the water intake and outlet of nuclear power plant,which brings effect to the dilution and diffusion of liquid effluent.Taking Lianjiang Port in the Beibu Gulf of China as an example,by using two-dimensional numerical simulation calculation,the concentration distribution of ¹³¹I and ⁶⁰Co nearby the water intake and outfall during mid-tidal in summer was analyzed,with and without consideration of construction of the zone after the two units of the nuclear power plant went into operation.The results showed that before the construction of the zone,¹³¹I and ⁶⁰Co mainly followed the tidal movement,and the average enveloping area of the concentration of ¹³¹I and ⁶⁰Co diluted by 10 times is 0.9×10^-3 km² and 1.2×10^-3 km² respectively.The high concentration area is mainly distributed in the Anpu Bay where the outfall is.After the construction of the zone,the tidal current field only changed in the vicinity of the industrial zone.The eastward side of the zone was significantly larger than the west side and a recirculation zone was formed on the east side.The flow in the vicinity of the zone is weaker due to the jet flow.The flow close to the industrial zone is affected most,and the flow rate and direction are significantly changed.The average envelope area of the concentration of ¹³¹I and ⁶⁰Co diluted 10 times was 1.0×10^-3 km² and 3.0×10^-3 km² respectively.The average envelope area of the concentration of 1 000 times diluted was 109.21 km² and 265.41 km² respectively.The high concentration zone of ¹³¹I and ⁶⁰Co increased by 11.1% and 150% respectively after the construction,while the low concentration zone decreased by 12.5% and 3.2% respectively,indicating that the industrial zone is not conducive to the dilution and diffusion of liquid effluent.The results will provide reference for optimizing the intake and drainage schemes,avoiding floodplain discharge and local concentration accumulation and providing technical support in reducing the environmental impact of liquid effluent discharge.

Simulation of uranium migration in groundwaterfrom uranium tailings leachate

XIE Tian, LI Ting, ZHU Jun, SHI Yunfeng

RADIATION PROTECTION. 2020, 40(6): 605-612.

Abstract ( 175 )

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A uranium tailings dam in northern China was selected as the research object to collect uranium tailings samples in tailings dams,aquifer sand samples and water barrier clay samples.The isotherm adsorption law of U on soil samples and the leaching process of U under rainfall conditions were studied to obtain U adsorption and migration parameters in key strata and the release rule of source terms.Using FEFLOW6.2 software to establish a three-dimensional model of groundwater in the uranium tailings evaluation area,simulating U migration behavior and concentration distribution.The results show that the adsorption of U on sand and clay conforms to the linear isothermal adsorption model,the distribution coefficients k_d are 20.41 L/kg and 45.92 L/kg respectively.The equilibrium concentrations of U under acid rain leaching and deionized water leaching conditions during the experiment period were 0.83 mg/L and 0.79 mg/L,respectively.And the leaching rates were 46.07% and 20.92%,respectively.The simulation results show that after 30 years of migration,the maximum concentration of U in groundwater will reach 0.595 mg/L and the peak concentration migration distance is 36.44 m.After 50 years of migration,the maximum concentration of U in groundwater will reach 0.440 mg/L and the peak concentration migration distance is 42.93 m.

The management control and optimization of typicalradioactive hotspots in NPP

CHEN Chenxiang, ZENG Jinzhong, NI Wei, HUANG Gang, HE Junnan

RADIATION PROTECTION. 2020, 40(6): 613-618.

Abstract ( 144 )

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The management of radioactive hot-spots is one of the core projects of radiation protection management in NPPs.Based on years of practical experience at Qinshan phase Ⅱ,this article elaborates on the establishment of radioactive hot-spots inventory,analysis and processing system,and cooperative management and control mechanisms,and establishes an effective set of control and optimization system for radioactive hot-spots at PWR nuclear power plants.This system was applied in Qinshan phase Ⅱ,which effectively improved the power plant’s ability to control radioactive hot-spots,reduced the cost of hot-spots management,and improved the performance of radiation protection operations.Related methods and measures can provide reference for the improvement of radioactive hot-spots management in other NPPs.

Estimation of exposure dose for decontamination workers from contaminated soil at a nuclear decommissioning site in Korea

Sohyeon Lee, Dong-Kwon Keum, Hyo-Joon Jeong, In Jun, Kwang-Muk Lim, and Yong-Ho Choi

RADIATION PROTECTION. 2020, 40(6): 619-624.

Abstract ( 152 )

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Assessment of the exposure dose for workers is crucial to protecting workers from the radiological risk.This preliminary study estimates the potential radiological exposure for a soil remediation worker at a nuclear decommissioning site contaminated with Cs-137 in Korea,and then calculates the maximum workable soil concentration to comply with the occupational dose constraint of 20 mSv per year.The Korean characteristic data,detailed exposure scenarios for workers by the type of work,and relevant exposure pathways were used in the dose estimation.As a result,the most severe exposure-induced work type was identified as the excavator operation with an annual individual dose of 5.92×10^-5 mSv for a unit concentration of soil,from which the derived maximum workable soil concentration was 3.38×10⁵ Bq/kg.Furthermore,dose contribution by each exposure pathway was found to be decreased in the following order:external radiation exposure,soil ingestion,dust inhalation,and skin contamination.The results of this study are expected to be used effectively to optimize radiation protection for workers and establish appropriate work procedures for future site remediation.

Episodes of thoron exposure due to consumer products claiminghealth benefits of negative ions

Jaiki Lee

RADIATION PROTECTION. 2020, 40(6): 625-630.

Abstract ( 192 )

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In May 2018,the primetime news casted a shocking report saying that radon concentration on a certain model of bed mattress found to be as high as 2 200 Bq/m³.After a humble,the Nuclear Safety and Security Commission(NSSC) of Korea confirmed that significant amount of thoron gas is emanated from several mattress models marketed by a company claiming beneficial health effects of negative ions.Laboratory analysis showed that some internal fabric sheets of those mattresses contain high concentration of Th-232.It was revealed that the manufacture treated the material with so-called ‘negative ion powder’ procured from the market and NSSC found that its radioactive content is the monazite powder.Although measurements with reliable instruments resulted in somewhat lower values,the tentative but conservative estimates of doses to the users are still remarkable,ranging from a few to 14 mSv a year.Most of the affected models have been marketed from 2010 but earlier models,with lower thorium content,were supplied from 2006.As many as 88,000 mattresses have been produced.The manufacturer with help of the government,recalled all the affected models and separated the radioactive internals.A large amount of waste is waiting for the government decision on disposal method.Similar problems were identified in other consumer products including latex mattresses and pillows imported,hot pads,and several models of sanitary or health-aid goods.These episodes called for revisiting NORM control strategy in Korea.

Analysis on γ radiation variation of the spent fuel pool in PWR NPP

HU Yipeng

RADIATION PROTECTION. 2020, 40(6): 631-639.

Abstract ( 187 )

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⁵⁸Co is the key γ radiation source of activated corrosion products of PWR nuclear power plant.Influenced by pH value and temperature variation,the solubility of activated corrosion products containing ⁵⁸Co will change continuously.During a shutdown process in Fuqing NPP,the specific activity of ⁵⁸Co in primary coolant was found to increase continuously as the coolant temperature drops.Meanwhile after completing fuel unloading within an outage period,the water temperature raised,leading an increase of the ⁵⁸Co specific activity in the spent-fuel pool,of which the maximum surface γ dose rate was reaching about 10 times of the designed value.By analyzing the specific activity of ⁵⁸Co,γ dose rate level and temperature variation trend in the two cases,as well as the system operation records comparison,it can be confirmed that the increase of ⁵⁸Co specific activity for both cases is closely related to the solution temperature.Under acidic environment,the solubility of activated corrosion products containing ⁵⁸Co has a positive temperature coefficient within a certain temperature range,which means the solubility will increase with the temperature rise.After reaching the maximum value,the solubility,decreasing with the temperature raise,shows a negative temperature coefficient.According to the conclusion,the specific activity of ⁵⁸Co was successfully reduced by starting the spare cooling circuit for the spent-fuel pool,of which the tem-perature drops.Finally,the surface γ dose rate of the spent-fuel pool quickly returned to the normal level,avoiding additional radiation exposure for the subsequent fuel operators.This successful practice provided a new idea for inhibiting and removing ⁵⁸Co in the activated corrosion products of PWR nuclear power plants.

Radiation protection control for maintenance of BOSSweld of M310 nuclear power unit

XU Zhuoqun

RADIATION PROTECTION. 2020, 40(6): 640-646.

Abstract ( 157 )

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BOSS weld is the joint weld between pipeline and branch pipe base.The branching and diameter change at BOSS weld position in radioactive system pipeline of nuclear power plant will likely result in the deposition of high radioactive particles.In addition,due to the limited maintenance space,its related operations have a high radiation risk.Based on the main steps of BOSS weld maintenance,this paper analyzed the external radiation and radioactive contamination risks during the operation process,introduced practical and feasible measures for radiation protection,as well as maintenance experience feedback and good practice during the process,which will provide reference for radiation protection control of BOSS weld maintenance of M310 nuclear unit.

The preliminary development of emergency conditionassessment system for HPR1000

YANG Yapeng, FENG Zongyang, JIA Linsheng, WANG Ning, WANG Renze, LIU Yining

RADIATION PROTECTION. 2020, 40(6): 647-651.

Abstract ( 143 )

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For meeting with the requirements of preparedness and response in case of a nuclear emergency of HPR1000,a system used to assess the core damage status and to estimate the source term released to the environment must be developed.The development of Emergency Condition Assessment System for HPR1000 (ECAS-HPR1000) was introduced,including its software framework,assessment modules,platform and data interfaces,etc.The system adopts the JAVA language as the platform,and open source MySQL as the data management system,supporting Windows and Linux OS.There are five subsystems,including 1) basic data acquisition subsystem,real-time acquiring the reactor operation parameters from Distributed Control System (DCS),2) core damage assessment subsystem,which used to assess the core damage status and degree,3) source term estimation subsystem,calculating the amount of radioactive material released from the core into the primary coolant system,containment and environment,4) assessment results display subsystem,5) user authorization management subsystem.

Research and development of core damage assessment system forpressurized water reactor under emergency condition

JIA Linsheng, WANG Renze, YANG Yapeng, FENG Zongyang, WANG Ning, LIU Yining

RADIATION PROTECTION. 2020, 40(6): 652-656.

Abstract ( 229 )

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Core damage assessment is an important part of nuclear accident emergency assessment.Based on foreign literature and combined with domestic operation experience,core damage assessment of PWR is researched and the corresponding software system is developed.There are three methods for core damage assessment based on core uncovering time,online monitoring readings and sampling.Considering real-time demand of emergency,actual application of various power plants,and international experience,assessment based on online monitoring readings is adopted in this paper,which is mainly based on the core thermocouple readings and the containment radiation monitoring readings.

Research progress of solvent fire accident in spent fuel reprocessing plants

LIAN Yiren, SUN Hongchao, ZHANG Zhi, MENG Dongyuan, SUN Shutang, CHEN Lei, WANG Xuexin, XU Xiaoxiao, LI Guoqiang, ZHANG Jiangang

RADIATION PROTECTION. 2020, 40(6): 657-662.

Abstract ( 161 )

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With the wide usage of nuclear energy,more and more spent nuclear fuel would be produced and accumulated.The spent fuel reprocessing technology recycling the spent fuel is advanced,but the security of spent fuel reprocessing facility is an important premise for the development of the spent fuel reprocessing technology.The solvent fire as one of the design basis accidents,has been an important concern at home and abroad.In order to analyse the combustion behavior of solvent fire,diffusion and deposition behavior of radioactive aerosol,performance of high efficiency filter and so on,the United States,Japan and other countries have established the experimental facilities and the combustion of solvent solution was studied.In this paper,the experimental methods and research results of solvent fire accident are reviewed,and the existing problems and future research directions are put forward.

A review of total ionizing dose (TID) inducedshift of threshold voltage of MOSFETs

LIU Yining, WANG Renze, YANG Yapeng, WANG Ning, FENG Zongyang, JIA Linsheng, ZHANG Jiangang, LI Quoqiang

RADIATION PROTECTION. 2020, 40(6): 663-670.

Abstract ( 151 )

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In the aim of helping the development of robots used in radiological emergency preparedness and response,the TID effects on the threshold voltage shift(ΔV_th) of different kinds of MOSFET with different geometry and different scaling technology was compared.The different gate width and the length dependent between bulk CMOS process and NW MOSFET is noticed.And the TID effects on ΔV_th of several kinds of new devices such as Ge-channel and GaN channel MOSFETs and MOSFETs with new layout geometry are described which can be investigated more deeply.In addition,TCAD simulation is introduced to be used in both mechanism investigation and modeling verification.

Study on the prediction method of large break loss ofcoolant initial emergency condition

WANG Ning, WANG Renze, YANG Yapeng, FENG Zongyang, JIA Linsheng, LIU Yining, LIANG Boning

RADIATION PROTECTION. 2020, 40(6): 671-676.

Abstract ( 179 )

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Method for prognosis of severe accident response progress initiated by large break loss of coolant (LBLOCA) was established,based on transient analysis for M310 reactor.The primary loop is rationally simplified based on characters of LBLOCA.The mass and energy conservation equations were solved approximately and the classical formula were used to prognose emergency condition progression.The active core area was divided into 4 radial rings and 10 axial levels,i.e.40 cells.Experimental correlations were used to calculate heat transfer in the core and the cladding temperature was obtained.Then the core condition can be judged according to the cladding temperature.A code for prognosis of LBLOCA initiated emergency condition could be developed based on the method established in the paper.

Nondestructive high-sensitivity measurement method for activation estimation in accelerator room concrete

Hiroshi Matsumura, Go Yoshida, Akihiro Toyoda, Kazuyoshi Masumoto, Hajime Nakamura, Taichi Miura, Koichi Nishikawa, Kotaro Bessho, Kimikazu Sasa, Tetsuaki Moriguchi, Fumiyoshi Nobuhara, Yoko Nagashima

RADIATION PROTECTION. 2020, 40(6): 677-682.

Abstract ( 320 )

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This study established a method for easily and quickly estimating the specific activity produced in the concrete walls and floors of accelerator rooms during long-term operation of accelerator,for advanced zoning of activated/non-activated areas in planning the decommissioning of an accelerator.We propose a new,highly sensitive method for nondestructively estimating the specific activity in concrete that can be applied to activation zoning.In this method,instead of direct determination of the specific activities of important long-half-life radionuclides for decommissioning,such as ¹⁵²Eu and ⁶⁰Co,we determine the specific activities of short-half-life radionuclides,²⁴Na and ⁵⁶Mn,in situ to obtain neutron flux.The obtained neutron flux and accelerator operation history yield the specific activities of ¹⁵²Eu and ⁶⁰Co for the advance zoning of activated/non-activated concrete.This method is a powerful long-term prediction tool for concrete activation.

The study on microwave treatment process for spent resin

GAO Chao, AN Hongxiang, GUO Xiliang, GAO Shuai

RADIATION PROTECTION. 2020, 40(6): 683-690.

Abstract ( 168 )

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Microwave treatment process for spent resin utilizes microwave penetration ability and body heat characteristics.The water and organic components in spent resin will be removed,so as to change the spent resin from organic to inorganic state.Compared with the original spent resin,the drying and ashing effect of microwave will greatly reduce the volume and weight.The product from this technology is convenient for subsequent processing.Based on the study,the technological parameters of microwave drum drying and microwave ashing of spent resin are determined.

Monitoring of gamma radiation level at Shanxi radioactivewaste repository from 2013 to 2018

WANG Liping

RADIATION PROTECTION. 2020, 40(6): 691-695.

Abstract ( 109 )

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To strengthen the safety management of radioactive waste repository,and to ensure the environmental radiation safety,this article analyzed the monitoring results of environmental gamma radiation dose rate in radioactive waste repository from 2013-2018 in Shanxi province radioactive waste repository.Results showed that the environmental gamma radiation levels at the repository meet the "Technical requirements for siting,design and construction of radioactive waste repository for nuclear technology application (Trial)".The gamma radiation dose rate at 0.5 m above the top of sources pit cover within the repository building is less than 20 μGy /h.The gamma radiation dose rate at 0.2 m from the outer surface of the repository building wall is less than 2.5 μGy/h.The annual environmental radiation level within the fluctuation range of background,which did not produce radiation effects of on the surrounding environment.Radiation environmental quality is generally good.In addition,a robust security level in the repository area will provide a strong guarantee for promoting the use of nuclear technology and the safety,health and sustainable development of the province.

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