船用堆一旦发生放射性泄漏事故,放射性核素将通过各种途径释放至工作舱室,极有可能导致船员受到超剂量内、外照射,为科学估计与评价船员受照剂量、有效开展医学应急处置,通过分析反应堆堆芯放射性核素释放机理及污染舱室主要核素类型与分布,利用0rigen2程序计算某型压水堆不同运行功率连续运行不同时间时主要裂变产物积存量,利用Matlab软件对裂变产物积存量与运行时间的函数进行最小二乘法拟合,建立船用堆变工况运行条件下主要裂变产物积存量以运行功率和时间为变量的拟合函数,进而依托船上固设的碘-131探测装置确定污染舱室辐射场绝对分布,建立了船用堆放射性泄漏事故污染舱室辐射源项的反演技术方法,并利用Visual C~(++)编制程序实现了船用堆放射性泄漏事故船员剂量快速估算。

Marine nuclear power plant in the event of a radioactive spill , radionuclide release through a variety of channels to work compartments , is very likely to lead to the crew by the irradiation or external overdose , ex-posure doses for scientific estimation and evaluation of crew to effectively carry out the medical emergency re -sponse , continuous operation at different times through the analysis of the reactor core radionuclide release mechanism and pollution compartments radionuclide type and distribution , using Origen2 procedure to calculate the cumulative amount of major fission product at a certain type of pressurized water reactor , using matlab soft-ware to estimate the fission product cumulative amount as a function of the run time and the operating power by a least squares fit to establish marine nuclear power plant operating conditions , and relying on the onboard solid set the iodine-131 detection devices to determine the pollution compartments radiation field absolute di

利用轻质海绵为填充材料，采用小体元分割法制备混合标准源，完成了高纯锗γ探测器对放射性气体源效率的校准。采用活性碳低温吸附法从235 U的裂变产物中快速提取放化纯88Kr ，并制成气体密封源，采用上述校准的高纯锗γ探测器对其进行了实验测量。利用放射性暂时平衡原理，通过子核88 Rb的活度计算得到了88Kr的活度，进而计算出88Kr各γ射线的发射几率，其结果的不确定度与评价值相比明显降低。

The HPGe detector efficiency calibration of gas source was made by mixing the reference standard source ,and the filling material is lightweight sponge and the small voxel partition method was used . The 88 Kr of radiochemical purity from 235 U fission products was quickly extracted using activated carbon cryogenic absorption method ,and was made into a gas sealed source .The 88 Kr of radiochemical purity was measured by the calibrated HPGe detector . Using the radioactive temporary balance principle ,the 88Kr activity was deduced from the 88Rb activity ,which is the decay daughter of 88 Kr . The absolute branching ratios of 88 Kr were gotten through calculation .Compared with the evaluation value , the uncertainty of data decreases significantly .

加速器质谱( accelerator mass spectrometry, AMS)是基于加速器和离子探测器的一种高能质谱,属于一种同位素质谱( mass spectroscopy, MS),它克服了传统MS存在的分子本底和同量异位素本底干扰,因此同位素丰度灵敏度很高,对14 C( T1/2=5730 a)、10 Be( T1/2=1．5×106 a)和36 Cl( T1/2=3．0×105 a)等核素测量的丰度灵敏度均达10-15(传统MS的同位素丰度灵敏度最高为10-8).因此, AMS具有极其广泛的应用前景.简述AMS原理、技术和发展现状,介绍中国原子能科学研究院的AMS技术,及该技术在核科学与技术中的应用研究进展,包括长寿命核素半衰期的测量(如79 Se ),核反应微小截面的测量(如238U(n,3n)236U),长寿命谷区裂变产物核测量以及129I的AMS测量作为核设施监测、核环境与核应急检测的新方法等.

Belonging to the category of isotope mass spectrometry ( MS) , accelerator mass spectrometry ( AMS) is a high-energy mass spectrometry based on accelerators and ion detectors. AMS overcomes the molecular and isobaric background interferences extant in conventional MS, and therefore has an extremely high isotopic abundance sensitivity, which reaches 10 -15 ( isotopic abundance sensitivity of conventional MS is 10 -8 at highest) for measure-ment of nuclides such as 14 C( T1/2 =5 730 a) , 10 Be( T1/2 =1. 5 × 106 a) and 36 Cl( T1/2 =3. 0 × 105 a) . Accordingly, AMS has extremely broad application prospects. This paper introduces the principle, technique and development sta-tus of AMS and focuses on the introduction of CIAE’s AMS technique and research advances in its application in nu-clear science and technology, such as studies on measurements of the half-life of long-lived nuclides(79Se), small nuclear reaction cross sections(238U(n,3n)236U) and long-lived fission product n

液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 MW熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。

Compared with solid fuel reactors ,there are differences in physics for liquid fuel reactor .As for molten salt reactor (MSR) ,due to fuel flow in primary loop ,the delayed neutron precursors and fission product partly decay out of core , resulting in reactivity loss as well as heat generation in the primary loop .In this paper ,the critical dynamics and safety characteristics of MSR were investigated using Cinsf 1D code .Con‐sidering the loss of delayed neutrons under different fuel flow speeds at zero‐power ,the corresponding control rod positions were calculated under pump start and stop condi‐tions .Keeping reactor power at 2 MW ,the temperature and power were computed for the primary loop system . Finally , the pump stop accident was simulated from rated power 2 MW . After pump stop , the reactor power increases slightly due to the reduction of delayed neutron loss at initial time and then it decreases to approach the decay heat power level quickly .The temperature in

使用M ECLOR1.8.6程序对严重事故实验Phebus FPT3进行了模拟分析。通过建模计算，得到了严重事故过程中燃料棒的行为，氢气的产生，裂变产物的释放、迁移和沉降及安全壳的热工水力响应等相关数据。计算值与实验值的对比分析表明，燃料棒的行为、氢气产生的时间和趋势及安全壳的热工水力响应与实验值吻合良好。由于相应程序模型的限制，最终产氢的总量及裂变产物相关的计算值与实验值有所不同。其中，计算的氢气总量较实验值偏小，计算的裂变产物释放量和在安全壳中的沉降量大多较实验值稍高。此外，还利用快速傅里叶变换方法（FFTBM ）对整个建模计算进行了详细的定量化评估。

The severe accident experiment Phebus FPT3 was simulated and analyzed by using MECLOR1.8.6 .The fuel rod behavior ,the hydrogen production ,the release , transport and deposition of fission products ,and the thermo‐hydraulic condition in the containment were calculated .The comparison between calculation results and experi‐ment data show s that the rod behavior ,the hydrogen production time and trend ,and the thermo‐hydraulic condition in the containment fit quite well .But the total quantity of hydrogen production and the fission product relative data have some differences betw een the calculation results and experiment data ,because of some limits of the model in the code .The calculated total quantity of hydrogen production is smaller than that of the experiment ,and most of the calculation results about the release and deposition of the fission products are a little bigger than those of the experiment .Besides ,the accura‐cy quantification of the calculation was evalu