压水堆核电厂选址假想事故初步研究

简要介绍了核电厂选址假想事故的发展过程,比较了基于RG 1.183和RG 1.195选址假想事故源项的计算假设,结合AP1000和CPR1000两种堆型计算了选址假想事故源项,同时结合某核电厂址计算了对公众造成的辐射影响。计算结果表明:1)参照RG 1.183计算假设,AP1000和CPR1000核电厂公众受照剂量最大的两小时分别为事故后1.25~3.25 h和0.7~2.7 h;2)无论参考RG 1.183还是RG1.195计算假设,CPR1000对公众造成的辐射后果要小于AP1000;3)无论是AP1000还是CPR1000,参照RG 1.183比RG 1.195计算得出的选址假想事故源项对公众造成的辐射后果均较小。     

RG 1.183修订对核电厂设计基准事故(DBA)放射性后果分析的影响研究

RG 1.183(核电厂设计基准事故(DBA)放射性后果评估所用源项分析导则)规定了核电厂DBA放射性后果分析过程中应遵循的设计原则、假设条件和验收准则等重要内容。自2000年首次出版RG1.183(0版)以后,美国核管会(NRC)仍致力于导则技术内容的研究,在导则完善方面开展了大量工作。2009年,基于上述研究成果,NRC起草了RG 1.183的修订稿(DG-1199),随后根据业界反馈意见对DG-1199进行了修订并拟在此基础上对原有的RG 1.183进行升版。考虑到RG 1.183对DBA放射性后果分析的深远影响,本文以AP1000核电厂作为参考电厂,对RG 1.183(修订版)中更新的内容开展了合理性评估以及影响分析,从而排除该导则的更新对目前的DBA放射性后果分析造成的影响和冲击的可能性。     

碘吸附器吸附效率对核动力厂应急控制中心工作人员受照剂量影响研究

为保证核动力厂应急控制中心在发生放射性大量释放事故时的可居留性,为其设计和安装了应急通风过滤系统,碘吸附器是该通风系统的主要碘过滤装置。碘吸附器的工作原理决定了其吸附效率受工作环境的温度、相对湿度、进风碘浓度等因素影响。针对某核动力厂应急控制中心设计特征,研究了在RG1.183DBALOCA和S3事故源项下,碘吸附器的吸附效率与室内工作人员接受的有效剂量、甲状腺当量剂量的对应关系,进行了线性拟合,给出了拟合系数,可用于事故后果快速剂量估算。     

利用可选择源项分析SGTR事故放射性后果的研究

介绍了可选择源项的基本假设和剂量计算的基本方法,采用一体化核电厂安全分析程序以及美国NRC RG 1.183中定义的放射性源项和方法,评估了900 MW级核电厂发生蒸汽发生器传热管破裂(SGTR)事故后的放射性后果,并计算了主控室、非居住区边界和低人口密度区外边界的剂量值。将计算结果与剂量准则进行比较,其结果完全在可接受的范围内。     

压水堆核电厂蒸汽发生器传热管道破裂事故源项的计算分析

当压水堆核电厂发生事故后,带有放射性的核素会通过破损处释放到环境中,从而危害核电厂周边环境及相关人员的安全,因此对事故后释放到环境中的放射性源项分析,对于核电厂的辐射防护具有重要意义。本文根据事故发生的频率以及后果严重程度,选取蒸汽发生器传热管破裂事故(Steam Generator Tube Rupture,SGTR)进行分析。事故分为事故前碘尖峰释放和事故并发碘尖峰释放两种事故工况,建立事故后放射性核素迁移和扩散计算模型,同时使用先进压水堆AP1000参数进行计算验证,并重点关注惰性气体和挥发性核素碘在环境中的放射性活度。计算结果显示:使用文中计算模型计算的放射性源项与设计源项比较一致,在两种工况下,惰性气体的释放活度与设计源项吻合较好,但碘的释放活度有明显差别。     

ACP100非居住区边界剂量环境影响评估

介绍了ACP100发生选址假想事故后,对非居住区边界剂量环境影响的评估方法。基于RG1.183建立了ACP100选址假想事故后释放到环境的源项计算模型,并结合厂址的气象条件采用PAVAN程序得到非居住区边界剂量,研究了裂变产物释放方式、照射时间、自然去除、厂址气象条件和源项对非居住区边界剂量的影响。结果表明:上述因素对剂量均有较大的影响,确定ACP100非居住区的边界应考虑这些因素。     

利用可选择源项分析大破口失水事故的放射性后果

阐述了应用可选择源项分析设计基准事故放射性后果的基本方法,并以900MW核电厂为研究对象,利用一体化安全分析程序分析大破口失水事故的放射性后果,包括主控室、非居住区边界和低人口密度区外边界的剂量计算,并与美国核管会(NRC)管理导则1.183中的剂量准则相比较,结果均在可接受值之内。     

AP1000剂量等效131I比活度运行限值优化研究

AP1000技术规格书中一回路剂量等效¹³¹I比活度运行限值是基于事故分析假设确定的,同时影响到放射性流出物的排放。根据AP1000原技术规格书中一回路剂量等效¹³¹I比活度运行限值计算出的SGTR事故厂外剂量和气态放射性流出物的排放都不满足GB6249—2011的要求。针对该问题,从碘尖峰释放机理、碘尖峰现实倍率、运行限值和条件以及放射性废物处理系统处理能力等方面进行了研究。计算结果表明,可将一回路剂量等效¹³¹I比活度的碘尖峰运行限值优化到1.11×106 Bq/g以满足GB 6249—2011的要求,同时不会对核电厂运行造成制约。     

AP1000核电厂安全壳内气溶胶自然去除分析

AP1000等非能动压水堆核电厂依靠自然的原理清除事故后安全壳气空间内的放射性气溶胶,可靠性较高,但对其进行分析较为复杂。事故后安全壳内气溶胶的主要运动形式有凝聚、重力沉降、扩散泳及热泳等,本文研究确定了合适的机理模型、假设条件和主要参数等,完成了AP1000核电厂的分析。分析结果表明,AP1000核电厂LOCA后,主要气溶胶去除机制中扩散泳贡献最大,其次是热泳和重力沉降;安全壳内气溶胶自然去除系数约为0.4~0.9h~(-1),堆芯裸露5h后变化较小;基于RG1.183源项、包络大气弥散因子及本文给出的安全壳气溶胶自然去除系数,计算得到的LOCA后厂外及主控室人员所受剂量可满足10CFR50中规定的限值要求。     

SGTR事故并发碘尖峰研究

结合实际电厂运行经验数据,提出了一种新的蒸汽发生器传热管破裂事故并发碘尖峰分析方法,并对新方法的合理性和保守性进行了分析。最后,将现有方法和新方法运用到实际电厂进行案例分析,结果表明,新方法得到的剂量结果可以满足我国国标的剂量验收准则。     

Potential impacts to operating nuclear power plants due to application of revised source term

TID-14844 was promulgated in 1962, and more than 30 years later there has been a big change of the US NRC's regulatory position in using accident source term for radiological assessment following a design basis accident (DBA). To replace the instantaneous source term of TID-14844, the time-dependent source term of NUREG-1465 was introduced in 1995, which represents the accident source term enveloping all light water reactor plants. In the meantime, the radiological acceptance criteria for reactor site evaluation in 10 CFR Part 100 were also revised. In particular, the concept of a total effective dose equivalent (TEDE) has been incorporated in accordance with the radiation protection standards set forth in revised 10 CFR Part 20. Subsequently, the publication of Regulatory Guide 1.183 and the revision of the Standard Review Plan 15.0.1 followed in 2000, which provided the licensee of a operating nuclear power reactor with the acceptable guidance of applying the revised source term. The guidance allowed the holder of an operating license issued prior to 10 January 1997 to voluntarily revise the accident source term used in the radiological consequence analyses of DBA. Depending on its type of application, there were suggested full and selective applications. Whether it is full or selective, based upon the scope and nature of associated plant modifications being proposed, the actual application of the revised source term to an operating plant is expected to give a large impact on its facility design basis. Prior to its actual implementation of design modifications, it is necessary to identify and analyze the potential impacts of each type of application and to derive the considerations taken in each application. In this paper, the experiences and lessons learned from its application to Ulchin Unit 3&4 are evaluated and presented.     

Radiological habitability evaluation of the control room of the high-temperature reactor pebble-bed module: post-accident dose assessment and sensitivity studies

The radiological habitability of a control room is important for nuclear emergency response, which is also a licensing prerequisite for nuclear power plants. It is determined by both atmospheric relative concentrations and doses received via different pathways. However, most recent studies have focused only on the former, which may not be adequate. The present study therefore investigates the radiological habitability and its sensitivity to different parameters in the high-temperature gas-cooled reactor pebble-bed module power plant. For three typical accidents, the study estimates the body, thyroid and skin doses received via different pathways using the Nuclear Regulatory Commission recommended ARCON96 and dose calculation method in RG 1.195. To make a realistic evaluation, the latest design and site-specific information, including the unique accidental source term, are collected and used as input parameters. The evaluation results reveal that the total dose of different pathways in the control room is far below the limit, which demonstrates the effectiveness of the current design. The inhalation exposure is the dominant pathway, and iodine and caesium are the primary contributors of the inhalation dose. The particle filter removal fraction is the most influential parameter in an accident which the activities of metal radionuclides are high.     

核电厂技术规格书冷却剂放射性比活度运行限制条件确定方法研究

本文通过对西屋标准电厂技术规格书中一回路冷却剂放射性比活度限值的研究,揭示了限值制定的背景及含义,有助于对技术规格书中相关规定的深入理解和后续的执行。通过AP1000电厂与西屋标准技术规格书的比较可知,AP1000电厂技术规格书用剂量等效¹³³Xe比活度限值替代了西屋标准技术规格书中的总放射性比活度限值,剂量等效¹³¹I比活度和¹³³Xe比活度限值均基于设计基准0.25%燃料包壳破损率计算得到,屏蔽设计、三废处理系统设计和事故后果分析等所采用的源项是一致的。最后,结合国内标准要求给出了可以对技术规格书中碘尖峰时比活度限值进行调整以及剂量等效¹³¹I和¹³³Xe比活度限值可以根据0.25%燃料破损率计算数据进行调整的建议。     

Ground deposition assessment of radionuclides following a hypothetical release from Ghana Research Reactor-1 (GHARR-1) using atmospheric dispersion model

With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. There is a challenge today in assessing radiological dose from nuclear reactor using a more reliable computer tool in addressing the released radionuclide to the atmosphere and ground effectively and to compute the dose rates. As such the dealing of atmospheric dispersion of radionuclide release from a nuclear facility has become very imperative. This has enhanced the idea of revisiting the safety features of the existing nuclear plants and particularly research reactors. One of such kind of research reactors whose safety is of concern now is the 30 kW Ghana Research Reactor-1 (GHARR-1) which uses a Highly Enrich Uranium (HEU) fuel. In connection with conversion of GHARR-1 from HEU fuel to the use of Low Enrich Uranium (LEU) fuel; assessment of a postulated radiological dose from possible radionuclides released using computer technology is essential. An effective computer model which is based on a reliable atmospheric transport and dispersion theory can help address such drawbacks. Atmospheric dispersion modeling and radiological safety analysis were performed for a postulated accident scenario of the HEU fuel of the GHARR-1 core. The simulation was performed using a reliable health physics atmospheric dispersion code called HotSpot. The HotSpot code which employs a Gaussian plume technique was used to perform the atmospheric transport modeling which was then applied to determine the ground deposition of radionuclides and to estimate the Total Effective Dose Equivalent (TEDE) of release radionuclides. The source term was generated from an inventory of peak radioisotope activities released by using the Oak Ridge isotope generation code ORIGEN-2. The adopted methodology used was based on the predominant site-specific meteorological data. Some selected radionuclides were evaluated to prove whether their release may have radiological effect on the public. Nonetheless, prudence requires assessing the effect on the public during such events. The results indicate that the maximum ground deposition value of 1.5E-04 kBq/m² occurred at 96 m distance and the maximum TEDE value of 1.9E-02 mSv occurred at 93 m from the reactor. It was observed that the values were far below the NRC acceptable limit of the 0.1 rem (1 mSv) for the public in a year even in the event of worse accident scenario.     

Iodine Removal Tests for BWR Containment Spray by Large Scale Facility

Iodine removal tests for a BWR containment spray were carried out with large-scale JAERI Model Containment Test Facility under LOCA simulated conditions. The tests consisted of two groups: “gas-phase based” tests mainly to obtain the initial iodine removal rate by the spray and “liquid-phase based” tests to obtain the iodine partition coefficient at equilibrium state. It was shown that the degree of iodine removal was largely influenced by pH-value of spray water. The results were discussed with calculated results by a code MIRA-PB using a dose reduction factor for the airborne iodine.     

拟建桃花江AP1000核电站LOCA 131I源项分析

针对核电厂事故工况下放射性物质的大气弥散问题,运用CALPUFF空气质量模型,模拟了桃花江核电厂冷却剂丧失事故(LOCA)工况下典型气载放射性物质131I的大气弥散过程,并对计算结果进行辐射剂量的估计,结果表明:1)事故开始后数小时内,源下风向8 km左右,高程与释放源有效高度相当,且海拔明显高于上风向海拔的地形区域,极易形成131I地面空气积分浓度峰值。2)三种化学形态的碘中,元素碘最易沉积。计算区域内地面沉积浓度与空气积分浓度呈现相同的分布规律。3)131I内照射造成的最大剂量当量比外照射高4个数量级,因而事故情况下防止放射性物质从呼吸道、口腔、伤口及皮肤进入人体,能极大降低131I的辐射剂量当量。     

Atmospheric dispersion modeling for an accidental release from the Pakistan Research Reactor-1 (PARR-1)

Atmospheric dispersion modeling and radiation dose calculations have been performed for a postulated accidental airborne radionuclide release from the Pakistan Research Reactor-1 (PARR-1) appropriate to a power upgrade to 10 MW. Estimates of releases for various radionuclide groups are based upon US-NRC regulatory guide 1.183. Committed Effective Doses (CEDs) to the public at various downwind distances were calculated using a health physics computer code “HotSpot” developed at the Lawrence Livermore National Laboratory, University of California, USA. The doses were calculated for various atmospheric stability classes, viz., Pasquill categories A–F with site-specific averaged meteorological conditions. The meteorological data on atmospheric stability conditions, mean wind speed and the frequency distribution of wind direction based on data collected near the reactor site have also been analyzed and are presented here.The results indicate that a person located within a downwind distance of about 500 m from the reactor would receive more than the permissible CED under the analyzed severe accident scenario. Analysis of one typical year of wind data indicates that the predominant wind direction is East–North East (ENE), which occurs at the site for more than 15% of the time.