Calculation of risk-based detection limits for radionuclides in the liquid effluents from Korean nuclear power plants

In order to review if present detection limits of radionuclides in liquid effluent from nuclear power plants are effective enough to warrant compliance with regulatory discharge limits, a risk-based approach is developed to derive a new detection limit for each radionuclide based on radiological criteria. Equations and adjustment factors are also proposed to discriminate the validity of the detection limits for multiple radionuclides in the liquid effluent with or without consideration of the nuclide composition. From case studies to three nuclear power plants in Korea with actual operation data from 2006 to 2015, the present detection limits have turned out to be effective for Hanul Unit 1 but may not be sensitive enough for Kori Unit 1 (8 out of 14 radionuclides) and Wolsong Unit 1 (9 out of 42 radionuclides). However, it is shown that the present detection limits for the latter two nuclear power plants can be justified, if credit is given to the radionuclide composition. Otherwise, consideration should be given to adjustment of the present detection limits. The risk-based approach of this study can be used to determine the validity of established detection limits of a specific nuclear power plant.     

基于MIKE21的滨海核设施液态流出物稀释扩散规律及排放优化的模拟研究

本文针对滨海核设施液态流出物海域排放问题,利用水动力软件M IK E21建立了排放海域的水动力和稀释扩散模型,模拟了液态流出物的放射性核素在潮汐作用下的稀释扩散规律,研究了潮型、潮位、排放时段等因素对放射性核素稀释扩散效果的影响.模拟研究发现,在潮差较大的落潮时段集中排放放射性核素,近岸放射性核素的浓度更低,更有利于液...     

Limiting Activities and Activity Concentrations in the Transport of Radioactive Materials

Abstract

The activity limits for Type A packages are determined by a set of criteria and hypothetical exposure scenarios known as the Q system. This system was first used to calculate activity limits for Type A packages for inclusion in the 19.85 IAEA regulations and the methodology has been described in the explanatory material for the regulations, IAEA Safety Series No 7. The new IAEA Basic Safety Standards incorporate revised dose coefficients for intakes of radionuclides. For the latest IAEA Transport Regulations, the Q system has been updated to take into account these more recent dosimetric data, and more comprehensive nuclear data. The updated Q system is described and some of the consequent changes in activity limits for Type A packages are illustrated. Member States of the European Union are subject to a Basic Safety Standards Directive, which contains radionuclide specific Exemption Values. The derivation of these values together With their application to transport is discussed.     

在线式放射性液态流出物监测仪研制

本文针对核设施中液态流出物关键伽马核素⁶⁰Co、¹³⁷Cs的测量,采用自主专利等多项技术设计样机,基于NaI(Sodium Iodide)闪烁体探测器的自动核素识别,开发了探测灵敏度更好、质量更轻便且满足通用的建筑承载能力的在线式液态流出物监测装置。设计加工集成的样机经过能量刻度、效率刻度、感兴趣区自动划分核素识别测试,并通过国家一级计量站校准测试,经过超过500多小时实验,其性能稳定可靠,具有核素识别能力,测试显示⁶⁰Co、¹³⁷Cs探测限小于0.088 Bq/L。对比国内传统监测技术,质量减轻接近1个数量级,探测灵敏度提升超过2个数量级,监测技术及其样机从技术能力而言也适用于饮用水伽马关键核素活度浓度监测。     

核设施液态流出物非介入式连续监测仪校准技术研究

在用谱仪测量样品活度时,准确地校准谱仪系统探测效率曲线是测量结果准确可靠的先决条件。在核设施液态流出物监测中,由于被测样品空间结构复杂,核素种类多且半衰期较短,通过制作标准液体源校准探测器系统的可行性低。本文讨论了几种核设施管道外非介入式液态流出物监测仪的校准方法：标准液体源法、数值积分法以及代表点法,选取其中的标准液体源法为初级方法,同时进行数值积分法和代表点法的实验,从相对偏差和不确定度两方面考虑,发现代表点法可作为适用有效的校准方法。     

Sweepout model implementation in RELAP5/MOD3.3 to improve RCS coolant inventory calculation during a LBLOCA

According to the experiments of the Upper Plenum Test Facility (UPTF) and advanced power reactor 1400 MW_e (APR1400), the sweepout in the downcomer has been identified to play an important role in depleting the core coolant inventory during a Large-Break Loss-of-Coolant Accident (LBLOCA). In order to identify the sweepout mechanism and to estimate the amount of coolant discharged by sweepout, the separate-effect test was carried out in the plate type test apparatus, which was scaled down to 1/5 of the size of the APR1400 downcomer. In addition, the sweepout model was developed by correlating the experimental data on the critical void height and the discharge flow rate at the break to the values of analytically derived non-dimensional parameters. This model was implemented in RELAP5/MOD3.3 to improve its calculation of coolant inventory loss during a LBLOCA. To validate the modified RELAP5/MOD3.3 by implementing the sweepout model, the sweepout separate-effect test was simulated by both the original and the modified RELAP5/MOD3.3. The original one predicted the different discharge flow rates according to the node size of the donor volume, and these flow rates were larger than those of the experiment. On the other hand, the modified one calculated the discharge flow rate and the critical void height much more similar to those of the experiment than the original model did. In the future, the improved RELAP5/MOD3.3 adopted in an integrated analysis system will support a more realistic thermal hydraulic analysis.     

压水堆核电厂液态流出物排放量统计方法研究

本文主要针对压水堆核电厂液态流出物排放的除³H和¹⁴C外的其余核素,从监测核素的种类、核素的探测限,以及小于探测限测量结果的统计等方面,比较分析我国与欧美国家的取样监测和统计要求,提出合理可行的改进建议,以更好反映我国运行压水堆核电厂液态流出物的排放现状。     

低放固体废物核素活度浓度上限值的推导

介绍了近地表处置设施在300 a监护期前及其以后的任何时间,公众个人及闯入者通过各种途径的受照剂量分别小于剂量限值时所要求的低放固体废物核素活度浓度上限值的推导方法及过程。以我国放射性废物近地表处置的基本安全要求为前提,并以遥田处置场和北龙处置场为对象,分析处置设施关闭后各景象的核素迁移过程和照射途径,建立各景象核素迁移的概念模型、数学模型,并计算各景象对人类产生的照射剂量。假设核素活度浓度与剂量之间呈线性关系,推导满足剂量准则下各景象各放射性核素的活度浓度上限值,选择最小的上限值,从而确定出低放固体废物各核素活度浓度上限值的量级。     

Computer Code System DSOCEAN for Assessing the Collective Dose of Japanese due to Radionuclides Released to the Ocean from a Reprocessing Plant

A computer code system DSOCEAN has been developed for assessing the collective dose of Japanese due to radionuclides released to the ocean from a spent nuclear fuel reprocessing plant. This computer code system uses a box model which represents the transfer of radionuclides between boxes of seawater into which the ocean around Japan is divided. The code system consists of a series of three interlinked main computer codes, which estimates the exchange rates of radionuclides between the boxes, the radionuclide concentrations in each box, and the collective dose from various exposure pathways, respectively.

By using DSOCEAN, two calculations were carried out to estimate the collective dose from a liquid effluent. One is associated with a routine release of radionuclides from a hypothetical reprocessing plant. The other is an application of the code system to disposal of liquid radioactive waste to the surface water of the ocean. The calculated results identified important radionuclides and exposure pathways.     

Development of passive flow controlling safety injection tank for APR1400

A vortex valve, called fluidic device, is to be installed inside a Safety Injection Tank (SIT) of Advanced Power Reactor 1400 MWe (APR1400) that passively controls an Emergency Core Cooling (ECC) water discharge flow rate without any moving part or any action of the plant operator. The fluidic device was designed, and its performance was evaluated by a series of repetitive experiments using VAlve Performance Evaluation Rig (VAPER), a prototypical full-scale test facility. The passive flow controlling SIT satisfied the major performance requirements of the APR1400 plant design, in view of peak discharge flow rate, pressure loss coefficient, and duration time of the ECC water discharge.     

核电厂液态流出物中63Ni的排放、剂量评估及监测方法研究

⁶³Ni是核电厂液态流出物中排放量较大的核素,在排放的裂变产物和活化产物中占比在10%以上,但我国核电厂目前对液态流出物中的⁶³Ni没有开展监测,没有相关监测数据的积累。本文在核电厂排放的⁶³Ni的来源分析及国外压水堆核电厂排放统计参考值调研的基础上,对⁶³Ni排放的剂量评估和测量方法进行了分析研究,并提出我国核电厂开展液态流出物中⁶³Ni监测的建议。     

风险指引的CPR1000核电厂LBLOCA分析方法初步研究

随着核电厂安全分析方法的不断发展，结合传统确定论分析与概率风险评价（PSA）的风险指引型安全分析方法逐渐引起安审当局和核电业主的广泛关注。本文基于国际上风险指引型分析方法在其他领域的应用现状，提出了风险指引的大破口失水事故（LBLOCA）分析方法，并重新评估了CPR1000核电厂的堆芯燃料包壳峰值温度（PCT）裕量。在PSA分析中，识别并量化了LBLOCA发生后可能发生的162个事件序列，并采用确定论现实分析方法（DRM）对筛选出的18个概率较大的事件序列进行了计算分析。然后通过期望值评估法和特定序列覆盖法对LBLOCA的PCT裕量进行了评估。结果表明，本文方法下LBLOCA的PCT裕量约为36~55 ℃，相比于传统的DRM裕量提升了16~35 ℃。     

Operating margin for radioactive effluent released to the environment from korean pressurized water reactors compared to the effluent control limits,obtained by monitoring the concentration of radioactive effluent with radioactive effluent monitors

Korean pressurized water reactors (PWRs) generally use radioactive effluent monitors for monitoring the concentration of radioactive effluents released to the environment. In this study, the operating margins for radioactive effluent monitors were analyzed to determine the levels of real-time concentration of effluents compared to effluent control limits (ECLs), the regulatory limits. The results show that the concentration of radioactive effluents released from Korean PWRs complied with the ECLs during the years 2012–2016. It was also found that outages at Korean PWRs did not impact the operating margins for radioactive effluent monitors; that is, there was no remarkable difference of the concentration of effluents between normal operation and maintenance periods. In terms of simultaneous effluent releases, the results demonstrate that exceeding the ECLs is unlikely to occur even under the hypothetical condition of coincident effluent releases from multiple discharge points at a Korean PWR.     

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.     

非人类物种辐射影响评价方法分析

本文对美国核能管理委员会（以下简称NRC）的RESRAD-BIOTA程序和欧盟推荐的ERICA程序进行了对比.通过设定相同的参考生物和核素浓度,对比分析了两个程序计算的差异性.结果表明,两种方法得到的生物受照总辐射剂量率在一个量级上,变化趋势相同,具有一定的通用性;但在不同核素对剂量的贡献份额上,两种方法的计算结果有一定的差异,因此可根据不同核电机型的实际源项选用适当的程序.     

压水堆核电厂常规岛液态流出物排放管理探讨

本文根据《核电厂放射性液态流出物排放技术要求》(GB 14587—2011)的基本要求,提出基于核电厂蒸汽发生器排污水放射性控制进行常规岛液态流出物排放管理的思路,并针对M310/CPR1000、WWER、AP1000和EPR四种堆型工程设计及存在的问题,通过研究提出我国压水堆核电厂常规岛液态流出物的排放管理改进建议,可供核安全监管和核电厂设计、运行管理参考。     

Scaling methodology for a reduced-height reduced-pressure integral test facility to investigate direct vessel injection line break SBLOCA

A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal-hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400 MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal-hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal-hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.     

关于新疆某些煤矿辐射水平审管限值的初步探讨

通过煤矿氡浓度及γ辐射剂量率限值的初步探讨,为编制科学和实用的煤矿辐射水平审管标准提供理论依据;为进一步加强对煤矿的辐射监管,促进煤炭开发利用事业的发展,保护环境、公众和工作人员健康,都具有十分重要的意义,也为提出伴生放射性矿的审管标准限值提供借鉴。     

秦山核电二期扩建工程废液排放系统的排放阈值分析

分析了核岛废液排放系统(TER)和常规岛废液排放系统(SEL)的废液排放在线监测阈值和槽式排放活度浓度,论述了废液排放系统的放射性总量控制和废液排放浓度的再线监测要求,最终分析确定了秦山核电二期扩建工程的废液排放浓度指标.