氦-3对Zr0.85Ti0.15Co合金吸氚特性影响研究

通过铀床向Zr_0.85Ti_0.15Co床转移氚过程中,研究氦-3对Zr_0.85Ti_0.15Co合金吸氚特性的影响,并结合Ortman模型分析氦-3影响机理。结果表明：氦-3对床体的屏蔽效应明显影响粉末的吸附性能,但屏蔽效应并不能完全抑制Zr_0.85Ti_0.15Co的吸氚反应,覆盖现象严重时其仍能在较低的速率持续吸氚直至吸附完全;氦-3浓度较低时,吸氚完成时间主要由氦-3浓度影响,而氦-3浓度较高时,主要由初始氚分压影响;氦-3对Zr_0.85Ti_0.15Co吸氚的屏蔽过程分为整体流阶段和扩散散流阶段,Ortman模型可较好的预测整体流阶段的氦-3屏蔽压强。     

压水堆核电站氚的辐射风险分析

氚是压水堆核电站常见的一种放射性核素,主要以HTO(氚化水)的形式存在。在正常运行情况下,对工作人员产生的辐射危害主要是存在于空气中的HTO被吸入人体后产生的内照射。由于HTO与水的物理化学性质完全相同,其进入空气的主要途径是随水的蒸发,由于核电站开放性系统很少,氚的辐射风险主要在乏燃料水池和换料大修的反应堆水池周围。在一定条件下,它能够产生的最大辐射风险——即空气中的最大浓度是在该条件下水-汽相的氚达到平衡后空气中的浓度。该文通过物理化学的方法分析了水中氚的浓度、水温及空气中氚的平衡浓度的关系,并得到了三者之间的理论关系式。空气中氚的平衡浓度随水温和水中氚的浓度的升高而升高。分析表明,当水温为30℃时,只有当水中的氚的浓度高达28GBq/m~3时,达到平衡后空气中氚的浓度才接近1DAC(导出空气浓度)的水平。而由于厂房通风系统的运行,进风中湿度的存在,空气中氚的实际浓度要远低于其平衡浓度。加上压水堆核电站开放性系统水中氚的浓度一般不高,因而氚可能产生的辐射风险水平是有限的。大亚湾核电站在1997年的大修中对空气中氚的浓度及工作人员尿氚的实际监测结果也支持这一论断。基于这一分析,文章对氚的辐射风险防护提出了建议。     

CFBR-II堆最大假想事故的环境影响分析

根据国外快中子增殖反应堆型最大假想事故的确定原则,计算分析了中国快中子增殖反应堆(CFBR-II堆)的最大假想事故的规模及源项。根据对释放模式、大气扩散模式和个人有效剂量来源的分析,在保守假设的情况下,计算得到距堆厅典型位置的公众和工作人员的个人有效剂量。与国家标准进行比较,发生最大假想事故时,对公众和工作人员不需要采取服碘、隐蔽和撤离等应急措施,事故应急处于应急待命水平。     

西安脉冲堆对环境的影响

应用厂址周围的人口与食谱调查资料以及反应堆参数 ,计算了该堆及其同位素生产线在正常运行及事故工况下厂址控制区边界的最大个人有效剂量当量和80km范围内的集体有效剂量当量。计算结果表明 :在正常运行时 ,厂址控制区边界最大个人有效剂量当量为6 0×10 -3mSv/a ,80km范围内的集体有效剂量当量为0 35人·mSv/a。反应堆最大假想事故事故下 ,所致厂址限制区外(500m)公众最大全身有效剂量当量为2 1×10-2mSv ,甲状腺剂量为3 8mSv ;事故持续30天后 ,80km范围内的集体有效剂量当量为0 14人·Sv(全身)和97人·Sv(甲状腺) ;正常运行工况和最大事故期间对本地区环境的影响都是可以接受的     

中国氦冷固态增殖剂实验包层系统氚输运初步分析

氚输运分析是开展中国氦冷固态增殖剂实验包层系统安全分析及未来聚变堆氚自持运行的重要研究内容之一。基于氚输运理论和固态增殖剂包层系统设计,利用FDS凤麟核能团队开发的聚变系统氚分析程序TAS,构建了固态增殖剂包层系统氚输运分析系统动力学模型。该模型氚输运结果与文献报道的吻合得很好,误差小于6%,验证了模型的正确性。针对中国氦冷固态增殖剂实验包层系统氚输运问题进行了两种计算方法(稳态、脉冲模式)的初步分析,获得了氚提取系统、氦气冷却系统回路氚分压,实验包层模块冷却流道、窗口室内氚提取系统和氦气冷却系统回路材料中氚滞留量,窗口室内氚提取系统和氦气冷却系统回路氚日渗透量等数据。最终对比结果显示,脉冲模式分析方法能够实时地跟踪源项的快速变化,更符合中国氦冷固态增殖剂实验包层系统实际运行情况。窗口室内氦气冷却系统回路材料中氚滞留量占到日产氚量的31.3%,因此需要在这些氚滞留损失严重的部位考虑适当的阻氚措施。     

CFBR-Ⅱ堆最大假想事故的环境影响分析

根据国外快中子增殖反应堆型最大假想事故的确定原则,计算分析了中国快中子增殖反应堆(CFBR-Ⅱ堆)的最大假想事故的规模及源项.根据对释放模式、大气扩散模式和个人有效剂量来源的分析,在保守假设的情况下,计算得到距堆厅典型位置的公众和工作人员的个人有效剂量.与国家标准进行比较,发生最大假想事故时,对公众和工作人员不需要采取服碘、隐蔽和撤离等应急措施,事故应急处于应急待命水平.     

秦山第三核电厂氚内照射辐射防护

重水堆核电站辐射防护的重点是氚内照射.秦山第三核电厂借鉴国外同类型电厂的经验,结合本厂运行和管理特点,初步建立了电厂的氚内照射辐射防护管理模式.本文从设计、运行过程、作业过程、工作人员防护等方面介绍秦山第三核电厂对氚的控制和防护措施,从场所监测、人员氚内照射剂量监测等方面介绍氚监测方法;给出了2003～2006年电站氚内照射剂量的监测结果,并与国外同类型电站进行了比较.     

某压水堆核电厂辐射控制区氚化水蒸汽监测

压水堆核电厂一回路冷却剂中的部分氚会通过废液和废气排放系统排放至工作环境中。本文报道某压水堆核电厂辐射控制区气态氚的监测结果:运行期间气态氚浓度范围为<LLD～9.21×10² Bq/m³;大修期间为<LLD～3.14×10³ Bq/m³。监测结果显示,压水堆核电厂运行初期工作环境中氚浓度较低,工作人员在现场工作无需采取额外的防护措施以及进行氚内照射剂量监测。     

反应堆卸料燃料组件吊运跌落事故辐射安全分析

在反应堆卸料过程中,燃料组件吊运是关键操作之一,由于在燃料组件吊运时发生跌落事故的后果较为严重,需要预先对该事故进行详细分析,评价事故对工作人员、公众及环境的影响。本文从事故起因、事故进程、事故缓解措施、事故处理程序及事故后果等方面对燃料组件吊运跌落事故进行了描述、分析和评价。从事故分析可见,发生燃料组件吊运跌落事故时,由于~(85)Kr的逸出而造成工作人员受到浸没外照射的剂量不超过1.28×10~(-1)m Sv。工作人员对事故进行处理而受照的剂量最大为15 m Sv。从环境影响评价可知,事故致使厂址N方向距离厂址边界约1.5 km处村庄公众接受的个人有效剂量最大,为2.56×10~(-6)m Sv,10 km范围内公众集体有效剂量为3.75×10-2人·m Sv。     

基于双塔式氢氦/氚氦混合气的快速分离技术

氢氦/氚氦混合气快速高效地分离与回收是产氚工艺的关键技术之一。利用液氮温度下吸附剂对氢同位素与氦吸附能力的差异,搭建了双塔式固定床低温吸附-解吸分离氢氦/氚氦的实验装置,通过该实验装置开展了不同组成的氢氦、氚氦混合气的分离实验。实验结果表明:采用双塔式固定床低温吸附-解吸法可以实现不同组成的氢氦、氚氦混合气的快速分离,分离后氦中氢的体积分数低于1.2×10-5,氦的纯度可达到99.998 8%,吸附柱的脱氚率大于97%;双塔式固定床低温吸附-解吸法的解吸氢中含有较高浓度的氦,不适于回收氢;在氚氦分离中采用固定床低温吸附柱是可行的,可作为钯膜分离器的辅助单元。     

中国核动力研究设计院外围氚分布特征及所致公众年有效剂量评价

基于2014—2017年中国核动力研究设计院核基地外围环境空气中和水中氚的监督性监测数据,对周围关键居民组各种途径的待积有效剂量进行了粗略的估算。结果表明:地表水上游、下游和饮用水中氚活度浓度差异无统计学意义,表明地表水采样时间可能与液态流出物排放时间错位,或者排放口下游1 km处液态流出物中氚已稀释到本底水平;空气和雨水中氚活度浓度随距离核基地增加而减小。综合楼监测采样点附近居民组中成人、青少年、儿童、幼儿、婴儿经各途径的平均氚摄入量分别为1.52、1.44、1.05、0.681、0.562 kBq/a,待积有效剂量分别为0.027 4、0.026 1、0.024 2、0.021 1、0.026 9 μSv/a。成人组成员所致待积有效剂量最大为0.027 4 μSv/a,但此待积有效剂量也仅占评价剂量目标值(0.25 mSv/a)的1‰以下。由此可以得出,核基地的核设施在正常运行工况下,对核基地外围环境的影响很小。     

Cold fusion reaction products and their measurement

The major reaction products that have been possibly associated with cold fusion reactions are neutrons, protons, tritium, He-3, He-4, internal conversion electrons, and gamma radiation. The branching ratios and relative reaction rates for these products are examined for consistency with cold fusion experiments. Both theoretical calculations and experimental data are examined and presented. The He-4 plus internal conversion reaction has been proposed to explain the absence of neutrons or gamma rays in successful cold fusion experiments. However, this reaction is not favored, even in a deuterium-palladium system. Measurement of these reactions must be made carefully owing to the presence in the background of 2.2-MeV gamma rays, background tritium in heavy water, and neutrons from the photodisintegration of the deuterium from background radiation. These problems confronting cold fusion experiments are addressed.     

Doses from accidental releases of tritium and activation products into the atmosphere

In view of public acceptance and the licensing procedure of projected fusion reactors, the release of tritium and activation products during normal operation as well as after accidents is a significant safety aspect. Calculations have been performed under accidental conditions for unit releases of corrosion products from water coolant loops, of first wall erosion products including different coating materials, and of tritium in its chemical form of tritiated water (HTO). Dose assessments during normal operation have been performed for corrosion products from first wall primary coolant loop and for tritium in both chemical forms (HT/HTO). The two accident consequence assessment (ACA) codes UFOTRI and COSYMA have been applied for the deterministic dose calculations with nearly the same input variables and for several radiological source terms. Furthermore, COSYMA and NORMTRI have been applied for routine release scenarios. The paper analyzes the radioation doses to individuals and the population resulting from the different materials assumed to be released in the environment.D.T.I. Dr. Trippe Ing. GmbH, Karlsruhe.     

Safety aspects and environmental impact of an inertial confinement fusion reactor

Releases into the environment of radioactive materials contained in heavy ion fusion (HIF) reactor plants must be prevented by similar safety design concepts as they are applied to present fission converter (e.g. LWR's) and breeder reactors (LMFBR's). This study identifies significant safety aspects of inertial confinement fusion power plant concepts and relates them to the more familiar basis of knowledge about the safety and the hazards of other advanced nuclear power reactor systems such as the LMFBR. Assessments of doses to be expected after the release of tritium from HIF reactor plants — normally and accidentally — are performed and compared with dose limits and with doses resulting from facilities of the fission fuel cycle. Needs for safety related research and development specifically for inertial confinement fusion as well as for the modelling of the various exposure pathways due to released tritium are pointed out.     

工业脉冲X射线发生装置放射防护检测与评价方法研究

目的探讨利用热释光剂量计对工业脉冲X射线发生装置进行放射防护检测与评价的可行性。方法分别用热释光剂量计和AT1123型X/γ辐射剂量率仪测量脉冲X射线发生装置的辐射剂量,通过对测量结果的分析和比较验证热释光剂量计测量脉冲X射线的有效性,通过对周围环境中年累积剂量的估算来评价试验人员和公众的受照剂量是否满足辐射防护要求。结果1号脉冲X射线发生装置曝光4次,2号脉冲X射线发生装置曝光10次,热释光法测得不同距离处的累积剂量符合距离平方反比衰减规律,实际检测结果与理论推算值较为吻合。两个X射线发生装置现有的工作负荷下,工作人员和公众的年累积剂量均在辐射防护控制目标值内。结论热释光法得出检测结果与理论推算值较为吻合,能够满足工业脉冲X射线发生装置放射防护检测需求。     

Analysis of hydrogen isotopes absorption between liquid lithium and yttrium under dynamic conditions

The use of D–Li stripping reaction to generate multiple neutron is expected in the International Fusion Materials Irradiated Facility (IFMIF). Tritium generated by its side reaction needs to be recovered from liquid Li for safety. Y hot trap is expected as an effective purification system to recover tritium from the liquid Li loop. In IFMIF, since liquid Li circulates in a loop, it is important to trace tritium behavior in Y and liquid Li under dynamic conditions. In this study, the authors investigate the effectiveness of tritium recovery from liquid Li under dynamic conditions of Y hot trap. Hydrogen absorption in Li instead of tritium is investigated under dynamically fluidized conditions by stirring liquid Li with an arbitrary rotational rate. The experimental results prove that the hydrogen absorption rate from liquid Li to Y is independent of the rotational rate. This means that the rate-determining step is hydrogen diffusion in Y.     

Simulation of Tritium Mass Transfer in a Three-Loop Sodium-Cooled Nuclear Power System

The simulation of the mass transfer of tritium in the technological media of sodium-cooled fast reactors is studied. The computational model of tritium mass transfer for a three-loop nuclear power system is based on an analysis of the balance of hydrogen and tritium in the first and second loops.The following are calculated for the BN-600 and Phoenix reactors which are currently in operation: the concentration of hydrogen and tritium in the first two loops, the concentration of tritium in the protective gas and the steam-water medium in the third loop, and the tritium fluxes into the cold traps, the third loop, and the surrounding environment. About 2.2 TBq/(GW·yr) tritium flow into the atmosphere through the walls of loops with nominal system parameters, and about 3 TBq/(GW·yr) flows into the third loop. The cold traps in the first and second loops catch about 99% of the tritium produced.__________Translated from Atomnaya Energiya, Vol. 98, No. 3, pp. 175–182, March, 2005.     

氟盐冷却高温堆氚输运特性瞬态分析

氚是熔盐堆运行过程中的固有产物，具有强腐蚀性和渗透性，是限制熔盐堆技术发展的瓶颈问题之一。本文围绕氟盐冷却高温堆（FHR）中氚输运特性在事故工况下的瞬态响应开展研究，主要讨论在无保护反应性引入（URI）及无保护冷却剂入口过冷（UOC）事故下，熔盐堆一回路中的氚产率、石墨吸附量、熔盐溶解量、腐蚀与沉积反应以及氚向二回路的扩散等特性。研究发现，瞬态条件下氚输运特性较稳态时更为复杂多变，呈现出强烈的动态耦合特点，这对氚控设备的性能提出了更高的要求。计算表明，在URI和UOC事故下，氚向二回路的扩散速率均降低，不需投入额外的氚控安全设施。     

Activation product transport in a FLiBe-vanadium alloy-HT9 system

An assessment is made of the gamma radiation hazards likely to be found around a fusion reactor heat transfer and tritium breeding loop which employs a vanadium alloy for the blanket and first wall structure and the ferritic-steel HT9 for the remainder of the loop. The coolant/tritium breeding fluid is the molten metallic salt FLiBe. Since the radiation levels near the primary loop components are found to be less than 100 mR/h 3–5 days after shutdown after three years of continuous full power operation, limited hands-on maintenance could be allowed. The very short half-lives of the predominant corrosion products make this result possible and make such a system very attractive.     

核电站氚的排放量及浓度限值比较分析

氚是核电站运行过程中向环境中排放较大的放射性核素之一,控制核设施中氚的产生和排放量越来越引起人们的重视。本文通过分析核电站产生氚的主要途径,结合国际上的运行经验参数,对比分析了不同国家、不同堆型核电站氚的排放量和浓度限值。分析结果表明:三十年间,全球核电站流出物中气态氚的排放量显著高于液态氚,重水堆是各堆型核电站中氚排放的主要贡献者,也是氚排放所致公众剂量的主要来源。为了更加有效的控制氚的排放,法国等国家核安全监管机构根据电站的装机容量、排放工艺、堆型等制定了各自国家核电站氚的年排放总量限值;加拿大等国的监管机构根据剂量限值制定了导出排放限值,该值的优点是便于审查核电站正常运行时氚的排放量;其它核电国家则是以剂量限值的形式提出了氚的排放限值。