常用γ放射源的屏蔽计算及方法评价
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摘要
论文由四部分组成,分别为引言、原理、辐射屏蔽模型建立和计算、数据分析及结论。第一部分引言:介绍常见民用放射源屏蔽防护的重要性、分析国内外最优化辐射防护研究的现状和本研究工作的意义。第二部分原理:首先介绍了外照射防护的相关知识:辐射的来源、辐射防护原则、常用材料等;其次是有关辐射量基础知识的介绍,一些与剂量有关的物理量的概念以及不同单位之间的换算关系;最后是射线在物质中的减弱规律以及计算辐射屏蔽的经验公式。第三、四部分是论文的重点:模拟实际情景,建立常见γ放射源的使用、储存、运输过程中的屏蔽防护模型,运用经验公式对各种状况下所需的屏蔽厚度进行计算,并应用MCNP 4C程序进行模拟,依据模拟结果对两种经验计算公式的计算效果进行评价。
本工作通过MCNP 4C的模拟计算证实了衰减倍数法在计算屏蔽防护厚度的安全性和精确性都要优于半减弱厚度法,为实际工作中的常用放射源屏蔽防护最优化设计提供了依据,为应对可能发生的事件及发生事件后的事故应急处理提供参考。
In industry,agriculture,medicine and other civilian applications of nuclear technology,theγradiation sources which may lead to great radioactive accidents include: 60Co (cobalt 60)、137Cs (Cs 137)、192Ir (iridium 192); 60Co are widely used in food irradiation processing, radiation therapy and non-destructive testing; 137Cs is mainly used for food irradiation processing, radiation therapy, and industrial measurement; 192Ir is mainly used for industrial flaw detection and install treatment. These sources have different applications and activities. The activity of the sources used for radiation processing is large,while the activity of the sources used for install treatment, short-distance radiotherapy is relatively small. However, the pollution of the sources on environment and the action to the body are very strong. The effective radiation shielding is needed to ensure that the radiation dosages accepted by the staff and closed public must be under the dose limit. It is increasingly important to find out how to evaluate the radiation protective effect of commonγsources fast and accurately, and how to determine the way of radiation protection fast and accurately in response to nuclear terrorism, nuclear accidents and other unexpected events.
The paper has compiled the parameters for shielding calculation from publications of ICRP and other references. It makes the fast calculation of security shield thickness for commonγsources by using the empirical formula achieved. Aiming at the three common sources, including high activity 60Co, medium activity 192Ir and low activity 137Cs,to establish the storage,usage and transportation scenarios,we calculated the security thickness of radiation shielding materials used (lead,concrete,iron,water) by empirical formula. This paper applied the MCNP simulation to test and evaluate the protective effect contrasting with the empirical formula , and put forward optimization recommendations for protection programs. It provides references for ensuring the annual dose of the staff and the public below the does limit with the lowest cost, and coping with events which may occur or emergency response after events.
For the calculation of three scenarios: well storage,lead shielding and protective concrete walls of high activity 60Co sources,relative to the MCNP simulated values,if we use the semi-reduced thickness method to calculate the thickness,the dose equivalent rate of the observation point is above the national limit, and in this case the protection is not valid. If we use multiple attenuation method to calculate the thickness,the dose equivalent rate of the observation point is below the national limit,and in this case the protection is valid. Compared with MCNP simulating values, the minimum safe thickness of water and lead shields has minor deviation and we can directly use this method in fast calculation. For the minimum safe thickness of concrete shielding, compared with MCNP simulated values, the error of calculated value is larger,so it is necessary to correct the corresponding coefficient.
For the calculation of two scenarios: lead shielding and protective concrete walls of medium activity 192Ir sources,we consider the shielding of the work area, relative to the MCNP simulated values. If we use the semi-reduced thickness method to calculate the thickness,the dose equivalent rate of the observation point is above the national limit, and in this case the protection is not valid. If we use multiple attenuation method to calculate the thickness,the dose equivalent rate of the observation point is below the national limit,and in this case the protection is valid. For the shielding of the public area, the thickness calculated by multiple attenuation method and the semi-reduced thickness method is both larger than the minimum safety value,so the protection is valid. Comparing the two methods,in accuracy, the error of the thickness calculated by the semi-reduced thickness method is less than that of multiple attenuation method. The thickness of lead shielding protection calculated by the two methods both meets the safety conditions, but larger error. Therefore it is necessary to correct the corresponding coefficient.
For the calculation of two scenarios: lead shielding and iron shielding of low activity 137Cs sources, relative to the MCNP simulated values,for iron shielding,If we use the semi-reduced thickness method to calculate the thickness,the dose equivalent rate of the observation point is above the national limit, and in this case the protection is not valid. If we use multiple attenuation method to calculate the thickness, the dose equivalent rate of the observation point is below the national limit, and in this case the protection is valid. For lead shielding,the thickness calculated by methods is both larger than the minimum safety value. The dose equivalent rates of observation points are both below the national limit,so the protection is valid. Comparing the two methods, the error of multiple attenuation method is less,so this method can be used in fast calculation directly.
本工作通过MCNP 4C的模拟计算证实了衰减倍数法在计算屏蔽防护厚度的安全性和精确性都要优于半减弱厚度法,为实际工作中的常用放射源屏蔽防护最优化设计提供了依据,为应对可能发生的事件及发生事件后的事故应急处理提供参考。
In industry,agriculture,medicine and other civilian applications of nuclear technology,theγradiation sources which may lead to great radioactive accidents include: 60Co (cobalt 60)、137Cs (Cs 137)、192Ir (iridium 192); 60Co are widely used in food irradiation processing, radiation therapy and non-destructive testing; 137Cs is mainly used for food irradiation processing, radiation therapy, and industrial measurement; 192Ir is mainly used for industrial flaw detection and install treatment. These sources have different applications and activities. The activity of the sources used for radiation processing is large,while the activity of the sources used for install treatment, short-distance radiotherapy is relatively small. However, the pollution of the sources on environment and the action to the body are very strong. The effective radiation shielding is needed to ensure that the radiation dosages accepted by the staff and closed public must be under the dose limit. It is increasingly important to find out how to evaluate the radiation protective effect of commonγsources fast and accurately, and how to determine the way of radiation protection fast and accurately in response to nuclear terrorism, nuclear accidents and other unexpected events.
The paper has compiled the parameters for shielding calculation from publications of ICRP and other references. It makes the fast calculation of security shield thickness for commonγsources by using the empirical formula achieved. Aiming at the three common sources, including high activity 60Co, medium activity 192Ir and low activity 137Cs,to establish the storage,usage and transportation scenarios,we calculated the security thickness of radiation shielding materials used (lead,concrete,iron,water) by empirical formula. This paper applied the MCNP simulation to test and evaluate the protective effect contrasting with the empirical formula , and put forward optimization recommendations for protection programs. It provides references for ensuring the annual dose of the staff and the public below the does limit with the lowest cost, and coping with events which may occur or emergency response after events.
For the calculation of three scenarios: well storage,lead shielding and protective concrete walls of high activity 60Co sources,relative to the MCNP simulated values,if we use the semi-reduced thickness method to calculate the thickness,the dose equivalent rate of the observation point is above the national limit, and in this case the protection is not valid. If we use multiple attenuation method to calculate the thickness,the dose equivalent rate of the observation point is below the national limit,and in this case the protection is valid. Compared with MCNP simulating values, the minimum safe thickness of water and lead shields has minor deviation and we can directly use this method in fast calculation. For the minimum safe thickness of concrete shielding, compared with MCNP simulated values, the error of calculated value is larger,so it is necessary to correct the corresponding coefficient.
For the calculation of two scenarios: lead shielding and protective concrete walls of medium activity 192Ir sources,we consider the shielding of the work area, relative to the MCNP simulated values. If we use the semi-reduced thickness method to calculate the thickness,the dose equivalent rate of the observation point is above the national limit, and in this case the protection is not valid. If we use multiple attenuation method to calculate the thickness,the dose equivalent rate of the observation point is below the national limit,and in this case the protection is valid. For the shielding of the public area, the thickness calculated by multiple attenuation method and the semi-reduced thickness method is both larger than the minimum safety value,so the protection is valid. Comparing the two methods,in accuracy, the error of the thickness calculated by the semi-reduced thickness method is less than that of multiple attenuation method. The thickness of lead shielding protection calculated by the two methods both meets the safety conditions, but larger error. Therefore it is necessary to correct the corresponding coefficient.
For the calculation of two scenarios: lead shielding and iron shielding of low activity 137Cs sources, relative to the MCNP simulated values,for iron shielding,If we use the semi-reduced thickness method to calculate the thickness,the dose equivalent rate of the observation point is above the national limit, and in this case the protection is not valid. If we use multiple attenuation method to calculate the thickness, the dose equivalent rate of the observation point is below the national limit, and in this case the protection is valid. For lead shielding,the thickness calculated by methods is both larger than the minimum safety value. The dose equivalent rates of observation points are both below the national limit,so the protection is valid. Comparing the two methods, the error of multiple attenuation method is less,so this method can be used in fast calculation directly.
引文
[1]刘保昌.实用医学辐射防护学[M].北京:军事医学科学出版社,2004:79-82
[2]电离辐射源与效应.联合国原子辐射效应科学委员会(UNSCEAR) 1993年向联合国大会提交的报告和科学附件.北京:原子能出版社,1995.
[3]刘华.辐射防护最优化方法及其应用[J].核安全,2007,2(1):
[4]国际放射防护委员会.ICRP第9号出版物.刘增鼎译.北京.原子能出版社,1975
[5]国际放射防护委员会.ICRP第60号出版物.李德平译.北京.原子能出版社,1975
[6] ICRP . The Optimization of Radiological Protection . Broadening the Process.ICRP Publication 101.Part 2.Elsevier,2006
[7] ICRP.2007 Recommendations of the International Commission on Radiological Protection.Draft for consultation, Jan 2007
[8]张松柏,田东风,伍钧.核恐怖主义与民用放射源的安全[J].科学, 2006, 58(2):31
[9]方杰.辐射防护导论[M].北京:原子能出版社,1991
[10]中华人民共和国国家标准.放射卫生防护基本标准.GB4192—84,1984
[11]中华人民共和国国家标准.电离辐射防护与辐射源安全基本标准.GB18871-2002,2002
[12]罗家治,徐建一,沈建,等.常用辐射量与单位[J].实用测试技术,2001(1):42-43 [13李星洪.辐射防护基础[M].北京:原子能出版社,1982
[14]胡逸民.肿瘤放射物理学[M].北京:原子能出版社,1999
[2]电离辐射源与效应.联合国原子辐射效应科学委员会(UNSCEAR) 1993年向联合国大会提交的报告和科学附件.北京:原子能出版社,1995.
[3]刘华.辐射防护最优化方法及其应用[J].核安全,2007,2(1):
[4]国际放射防护委员会.ICRP第9号出版物.刘增鼎译.北京.原子能出版社,1975
[5]国际放射防护委员会.ICRP第60号出版物.李德平译.北京.原子能出版社,1975
[6] ICRP . The Optimization of Radiological Protection . Broadening the Process.ICRP Publication 101.Part 2.Elsevier,2006
[7] ICRP.2007 Recommendations of the International Commission on Radiological Protection.Draft for consultation, Jan 2007
[8]张松柏,田东风,伍钧.核恐怖主义与民用放射源的安全[J].科学, 2006, 58(2):31
[9]方杰.辐射防护导论[M].北京:原子能出版社,1991
[10]中华人民共和国国家标准.放射卫生防护基本标准.GB4192—84,1984
[11]中华人民共和国国家标准.电离辐射防护与辐射源安全基本标准.GB18871-2002,2002
[12]罗家治,徐建一,沈建,等.常用辐射量与单位[J].实用测试技术,2001(1):42-43 [13李星洪.辐射防护基础[M].北京:原子能出版社,1982
[14]胡逸民.肿瘤放射物理学[M].北京:原子能出版社,1999