矿山环境中氡及其子体吸入所致内照射剂量估算方法的应用

通过对矿山氡及其子体内照射剂量评价方法的回顾,研究适合于矿山空气环境中氡及其子体吸入所致内照射剂量的计算方法.并应用所研究出的这些方法,结合实测数据来讨论评价方法的优劣.结果表明,由氡及其α潜能浓度估算个人内照射剂量的方法比较适合于估算矿山空气环境中氡及其子体吸入所致内照射剂量;而由平衡当量氡浓度估算个人内照射剂量的方法,其误差比较大,适合于在矿山大范围辐射环境评价中用来确定关键子区与关键人群.     

确定内陆核电厂液态放射性流出物排放浓度限值基准的讨论

介绍了核电厂放射性流出物审管控制的原则,对液态流出物排放浓度的限制方式、导出浓度限值的剂量基准和模式进行了讨论,对确定内陆核电厂液态放射性流出物排放浓度限值提出了建议。     

关于核电站排放限值的讨论

本文对核电站排放限值可以采用的指标及其相互关系进行了讨论,认为核电站排放限值应以公众的个人剂量限值为基本限值,而以排放总量和浓度限值作为辅助指标。文中还介绍了各国的剂量限值和排放总量限值标准,并对此作了比较和评价。     

放射性废水的排放标准

评述了1977年以前一些国家和组织的放射性废水浓度和排放标准,讨论了ICRP第2号出版物计算MPC_W的方法。为了控制低放废水向环境排放,提供液体流出物处理的基础,应该考虑导出水浓度(DWC)。DWC的计算是基于公众的年摄入量限值(ALI)和年有效剂量当量限值。给出了计算的导出水浓度。     

核素氚导出空气浓度控制限值的发展及其应用研究

氚是核设施运行过程中释放的放射性物质之一,对人体的内照射危害不仅与其活度有关,而且与氚的化学形态、摄入途径密切相关.针对目前国内涉及氚的控制标准较多,氚化水(HTO)与元素氚(HT)的控制限值均存在不一致的问题,梳理分析了氚浓度控制限值的发展及相关标准的制定依据.根据我国现行辐射防护有关标准,结合国际原子能机构(Int...     

有关内照射剂量学的重要国际会议和ICRP出版物

本文介绍了有关内照射剂量学的重要国际会议，以及ＩＣＲＰ有关的出生物，回顾了内照射剂量学的发展过程，说明其发展经历了从制定最大容许浓度标准，到从器官平均剂量、局部旧及细胞微剂量去衡量内照射的辐射危害的过程。对内照射剂量学的发展方向提出了笔者的看法。     

一种估算氚的内照射剂量的方法

为了估算内照射剂量,国际放射防护委员会(ICRP)推荐了多种放射性核素的参考人的生物动力学模型及其参数.ICRP还建议,若受照个体的剂量接近或大于个人年剂量限值时,应尽可能多的获得受照个体内照射监测资料,以便更准确地估算剂量;当受照个体接受了促排等医学处理,应采取个体的代谢资料估算剂量.本文以事故性过量摄入氚化水,且接受了促排医学处理的内污染为例,通过尿氚浓度的大量监测结果分析,得出了该受照个体的代谢资料.按此代谢资料估算的待积有效剂量为25.5 mSv,与国际原子能机构IAEA最终给出的参考值25.8 mSv仅相差1%.为了比较,也按照ICRP推荐的参考人的代谢资料估算了待积有效剂量,其结果为38.6 mSv,与参考值相差了48%.     

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。     

确定核电厂气载放射性流出物排放限值的设计方法研究

为了使核电厂的气载放射性流出物排放所致环境公众辐射剂量满足法规标准要求,本文从辐射防护剂量限值、大气污染物扩散规律及核电厂源项特点等方面考虑,通过实例分析,探讨了确定核电厂气载放射性流出物排放限值(排放率、排放浓度、地面浓度限值)的设计方法,提出了限值的建议值,并与已有核电厂的气载流出物排放报警阈值进行了验证,确认了方法的合理性。     

铀尾矿池的环境污染与居民所受剂量的估算

本文对距离尾矿池500米的村庄作了调查,结果认为,该村农民所受剂量主要来自废水灌溉水稻田所致的外照射与食用污染粮食所致的内照射。室内氢子体浓度高于对照点,但总剂量不超过限制剂量当量。     

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。     

放射性核素的年摄入量限值及其应用

本文介绍了放射性核素ALl及DAC的计算原则、职业性工作人员和公众的ALI和DAC及其应用,并讨论了连续摄入及单次摄入情况下由体负荷量求累积有效剂量当量和摄入量的方法。     

Setting standards for the safe level of radioactive contamination of water and bottom sediments in bodies of water

In this paper, we propose standards which can be used to simply and reliably monitor the content of technogenic radioactive substances in rivers and other bodies of water. We introduce the concept of “maximum safe concentration” and “reference monitored concentration” of radionuclides in water, and also the “maximum concentration of radionuclides in bottom sediments.” To determine the maximum safe concentration of radionuclides in water, we have used the basic public health principle established in NRB-96: to not exceed the limit of an effective dose of 1 mSv/y for external and internal radiation exposure for a critical population group, considering overall water use. To define the degree of radioactive contamination of the bottom of bodies of water, we propose using the maximum concentration of radionuclides in bottom sediments: a concentration less than that for which the bottom sediments may be classified in the category of “low activity solid radioactive wastes” according to OSP-72/87. We offer a calculation scheme and numerical values for the proposed standards for some radionuclides most frequently encountered in practice and examples of their use. State Institute of Applied Ecology NPO Taifun. Translated from Atomnaya énergiya, Vol. 86, No. 5, pp. 398–407, May, 1999. Original article submitted December 30, 1998.     

Incident Angle Dependence of Residual Defect in Silicon Surface Barrier Detector

The fate of radionuclides produced in a light water moderator was investigated by assaying with a Ge(Li) detector sampled coolant water and the waste water produced by regeneration of resins, measuring the radiation dose rates along the surface of the resin column length with TLDs, and studying the sorption of radionuclides to resins in dynamic experiments. The buildup concentrations of short-lived radionuclides, produced by activation in the reactor core water and the amounts of radionuclides sorbed on the resin bed, as estimated from a compartment model, are reported. A method for estimating the production rate of soluble radionuclides with long half-lives in the core water has been developed. Changes in the radiation dose rates along the resin column length recorded during reactor operation and after shut-down, showed that the sorbed radionuclides did not migrate downward, as expected from the chromatography theory, rather they were decreased by the decay process. Moreover, the slopes of the dose rates on both sides of the resin surface were almost identical. These results show the behavior and distribution of radionuclides in the light water moderator of a research reactor.     

A priori risk-informed approach to ensure the compliance with discharge limits for liquid radioactive effluent to be discharged from APR 1400

ABSTRACT

In order to develop a methodology to guarantee the conformance to operational discharge limits for liquid effluent from a NPP at an early stage of design, a risk-informed approach was proposed and its applicability was verified for APR 1400. Existing methodology to calculate risk-based detection limit for a single radionuclide was improved by incorporating a new model to derive more realistic pathway dose factors. A new simple expression was also proposed to adjust risk-based detection limits for multiple radionuclides mixture if necessary. In addition, a new procedure to warrant the compliance with discharge limits by controlling detection limits of only a few principal radionuclides was established in accordance with risk-informed concept. Through case studies for APR 1400 to be commissioned at a hypothetical site, it was shown that calculated pathway dose factors are more realistic for majority of radionuclides. It also turns out that neither present detection limits nor unadjusted risk-based detection limits can be justified when the radionuclide composition is unknown, however further adjustment of detection limits or increasing additional dilution factor resolves the problem. Finally, ten principal radionuclides were identified and shown to be enough for liquid effluent control at APR 1400 from a risk-informed point of view.     

Determination of the Admissable Technogenic Radioactive Contamination of Water in Bays during the Operation of Nuclear Objects in Marine Engineering

The various irradiation pathways due to radioactive contamination of water in bays (estuaries) are analyzed and the dose factors, characterizing the irradiation pathways considered, are determined for an extensive set of radionuclides. A method is proposed for determining the control concentration of radionuclides in water in reservoirs using the above-mentioned dose factors. The method is based on taking account of the professional work performed by workers. Critical population groups are identified on the basis of the irradiation pathways considered, and estimates of the control concentration of radionuclides in the water of a bay are made for these groups. 2 tables, 6 references.     

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.     

Estimating internal dose coefficients of short-lived radionuclides in accordance with ICRP 2007 recommendations

For radiation protection in high-energy accelerator facilities, internal dose coefficients of short-lived radionuclides were estimated using the dosimetric methodology in accordance with the International Commission on Radiological Protection (ICRP) 2007 Recommendations. A computational program was developed for estimating the dose coefficients. The program was verified by confirming whether it could reproduce the dose coefficients provided by ICRP for intakes of representative radionuclides. In addition, the estimated dose coefficients of short-lived radionuclides were compared to the values generated by Dose and risk CALculation software (DCAL), which is based on a dosimetric methodology that is in accordance with the ICRP 1990 Recommendations, to discuss the reasons why the dose coefficients were changed by the revision of the dosimetric methodologies. The comparison revealed a decreasing trend of dose coefficients in the case of inhalation upon revision of the dosimetric methodologies. By contrast, in the case of ingestion, the dose coefficients tended to increase.     

Radiation safety norms for atomic energy

The evolution of the dose limit and the admissable concentration of radionuclides in air and water in regulatory documents is analyzed. The legitimacy of classifying liquid radioactive wastes on the basis of the intervention level for radionuclide concentrations in drinking water is discussed. The SP AS-99 admissable irradiation levels for the general population due to emissions for nuclear power plants are compared with the dose in regions containing nuclear and thermal power plants.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 299–308, October, 2004.     

乏燃料深钻孔处置安全试评价

为探讨乏燃料深钻孔处置方式的安全性,本文在已有处置概念设计基础上,以某乏燃料核素类型与活度作为输入源项,通过保守假设选取处置孔上方打井饮水景象开展安全评价计算,得出所致最大个人有效剂量为0.045 mSv/a,最大剂量出现在处置孔关闭后的8 200年,主要剂量贡献核素为I-129,该结果可为我国高放废物深钻孔处置的研发提供参考。