ITER 试验包层模块活化计算与环境安全分析

运用聚变堆放射性计算程序FDKR 和衰变链数据库AFDC-DLIB, 计算了ITER 中国氦冷固态增殖试验包层模块设计中活化产物的放射性、衰变余热和潜在生物危害因子BHP 值。计算结果表明: 对于试验包层模块来说, 在500MW 聚变功率下运行一年, 停堆时的总放射性、余热和BHP 值分别为2. 10× 10¹⁶B_q 、5. 06 × 10^{- 3}MW 和 68. 6km³•( kW) ^{- 1}。结果表明: 中国氦冷实验包层模块不存在突出的安全问题。     

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用三维蒙特卡罗中子-光子输运程序MCNP和放射性计算程序FDKR,对中国ITER氦冷固态试验包层模块(CH HCSB TBM)进行了初步的三维活化分析。计算结果表明：TBM设计中活化产物的放射性、余热和潜在生物危害因子(BHP)主要来自结构材料;在500MW聚变功率下辐照0.53年,停堆时CH HCSB TBM的总放射性、余热和BHP分别为1.182×1017Bq、2.463×10-2MW和5.651×103km3.kW-1。在计算时,通过自动接口程序,实现了MCNP与FDKR之间的自动连接及数据处理。     

FEB—E氚循环系统的计算机模拟

对聚变实验增殖堆（ＦＥＢ）工程概要设计的氚燃料循环构造了一个动态子系统模型，研制了模拟氚燃料循环系统的计算机程序ＳＷＩＴＲＩＭ，计算运行一年后１０个子系统中的氚投料量和整个推系统总的氚投料量，这对预示起动一个聚变热功率的１５０ＭＷ量级的实验增殖堆所需的最少初始氚投料量有参考价值，计算结果表明，要求的最少初氚贮备量除了与燃料气体净化系统和同位互分离系统中氚的平均逗留时间有关外，还与包层液态锂中提取氚     

Activation and afterheat analysis for CH HCSB TBM

1 Introduction ITER will play a very important role in first integrated blanket testing in fusion environment. Some DEMO blanket relevant technology issues, especially safety requirements and environment impacts will be demonstrated during the ITER Test B…     

Application of the program LISE to fusion-evaporation

A new fusion-evaporation model LisFus for fast calculation of fusion residue cross sections has been developed in the framework of the code LISE. This model can calculate very small cross sections quickly compared to programs using the Monte Carlo method. Such type of the fast calculations is necessary to estimate fusion residue yields. Using this model, the program LISE now has the possibility of calculating the transmission of fusion residues through a fragment separator. It is also possible to use fusion residue cross sections calculated by the program PACE, which has been incorporated in the LISE package. The code PACE is a modified version of JULIAN—the Hillman-Eyal evaporation code using a Monte Carlo code coupling angular momentum. A comparison between PACE and the LisFus model is presented.     

球形托卡马克聚变嬗变堆中子学设计

基于对球形托卡马克ST聚变堆的研究,提出了ST聚变嬗变堆的设计概念。对堆芯参数作了初步选择,确定了一组适合于嬗变包层的堆芯参数供中子学计算和结构设计参考,给出了在以嬗变次锕系元素（MA）核废物为目标的一维中子学计算结果。     

核用结构材料SIMP钢低活化特性的分析与比较

在中国聚变工程实验堆(CFETR)水冷陶瓷增殖(WCCB)包层的设计条件下,对水冷包层应用SIMP 钢,使用蒙特卡罗中子输运程序MCNP 与欧洲研制的材料活化计算程序FISPACT 耦合计算,分析了SIMP 钢结构材料的放射性比活度、衰变热、接触剂量率和辐照损伤等。通过与EUROFER-97、F82H 等多种低活化钢的对比发现,SIMP 钢在多个活化结果上达到国际上认可的低活化铁素体/马氏体钢(RAFM)的低活化特性。因此,SIMP 钢可作为未来聚变堆包层的候选结构材料。     

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The activation properties of SIMP steel have been simulated under the conditions of the water cooled ceramic breeder (WCCB) blanket for China Fusion Engineering Test Reactor (CFETR) by coupling of the MCNP (Monte Carlo N-Particle Transport Code) and FISPACT activation code. The simulation results of specific activity, decay heat, contact dose rate and radiation damage of SIMP steel have been fully compared with other RAFM steel such as EUROFER-97, F82H and so on. The comparisons show that the SIMP has good activation performance corresponding to low activation performance of RAFM. It is conclusion that the SIMP can be chosen as one candidate structural material used for future fusion reactor blanket.     

Modified WKB transmission for fusion

In the optical model (OM) approach for fusion, absorption of flux occuring beyond the barrier position is presented in detail at low energies. It has been shown that the OM transmission can be well approximated as a sum of the WKB transmission and a long range absorption (LRA) contribution. Owing to absence of LRA, the fusion predictions of coupled channel codes based on transmission approach like the CCFUS code, do not agree with the predictions of complete coupled reaction channel (CRC) calculations based on OM approach using the code FRESCO. The CCFUS code with a modified transmission which includes LRA contribution is shown to be consistent with the CRC results using FRESCO. The static deformation of the colliding nuclei strongly influences the fusion imaginary potential and therefore the deep sub-barrier fusion cross sections.     

CFETR 水冷陶瓷包层时间相关产氚特性初步分析

水冷陶瓷包层是中国聚变工程实验堆(CFETR)的三种候选包层概念之一。基于CFETR水冷陶瓷包层的一维中子学模型,通过蒙特卡罗输运模拟程序MCNP和活化计算程序FISPACT的耦合计算,经三维转换系数修正,分析了CFETR水冷陶瓷包层时间相关产氚特性。结果表明,当CFETR运行因子为0.5,聚变功率为200MW时,水冷陶瓷包层在运行5年、10年、20年后,氚增殖率(TBR)的降低都不显著,但是年产氚剩余量的降低很明显。此外,产氚包层内初始时刻TBR对产氚特性的影响也很大。     

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水冷陶瓷包层是中国聚变工程实验堆(CFETR)的三种候选包层概念之一。基于CFETR水冷陶瓷包层的一维中子学模型,通过蒙特卡罗输运模拟程序MCNP和活化计算程序FISPACT的耦合计算,经三维转换系数修正,分析了CFETR水冷陶瓷包层时间相关产氚特性。结果表明,当CFETR运行因子为0.5,聚变功率为200MW时,水冷陶瓷包层在运行5年、10年、20年后,氚增殖率(TBR)的降低都不显著,但是年产氚剩余量的降低很明显。此外,产氚包层内初始时刻TBR对产氚特性的影响也很大。     

聚变实验增殖堆FEB-E放射性废物处置指标的计算

应用中子输运程序ＢＩＳＯＮ３．０、增殖堆放射性计算程序ＦＤＫＲ、剂量率计算程序ＤＯＳＥ完成了聚变实验增殖堆ＦＥＢ－Ｅ的放射性、核废物特性及废物处置额定容量（ＷＤＲ）的计算。结果表明,在停堆以后几周内,ＦＥＢ－Ｅ设计的经一壁和包层结构材料满足１０ＣＦＲ６１Ｃ级核废物处置额定容量的要求。对包层中的重要锕系元素＾２３２Ｕ、＾２３７Ｎｐ的含量也作了计算分析。     

实验混合堆FEB设计的远距离维修指标计算

动用ＢＩＳＯＮ,ＦＤＫＲ和ＤＯＳＥ程序,计算了实验混合堆ＦＥＢ在额定功率下运行一年后装置的第一壁,包层和屏蔽层的远距离维修指标ＲＭＲ值和有关结果。结果表明,需在现有屏蔽设计基础上适当增加铅屏蔽层厚度,才能满足美国核管理委员会的１０ＣＦＲ２０法规规定的辐射安全要求。计算结果可供实验混合堆的工程设计,安全运行及维修操作设计参考。     

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In part one published in the last issue, the tritium retention and the total tritium inventory in PFC materials of FEB-E fusion reactor had been calculated. The tritium well depth, tritium well time during the FEB-E fusion reactor start-up and initial operation phase had been obtained. In this part, how to improve tritium recovery efficiency in the ITER TBM solid breeder blanket with using purge gas has been discussed. Some new innovative schemes for reducing tritium retention and improving tritium recovery efficiency are proposed. Such as, sponge mechanism based on deuterium saturated PFC materials; deuterium and beryllium co-deposition layer created on first wall surface; SPB scheme for enhancing tritium recovery efficiency of purge gas in ceramic breeder blanket based on the electrical polarization rotations catalyzing isotope exchange rate enhancement resulted from applied low frequency electric-field, of Li4SiO4 grain and purge gas molecular particles and so on, are explored.     

聚变堆氚系统设计中的一些重要问题研究(Ⅱ)

在文献[1]中,计算了FEB-E 聚变堆PFC 材料内的氚滞留量、堆系统总的氚投料量、启动运行开始阶段的氚坑深度和氚坑时间大小。这里将讨论在ITER 的TBM 氚增殖包层内固体氚增殖剂中的氚如何高效率地被载氚气体带出并且以高效率地提取回收。本部分将进行创新的探索性研究并且提出某些减少氚滞留量和改善氚提取回收效率的新方案,例如：基于氘饱和的海绵效应;第一壁表面建立氘和铍的伴同沉积层;基于在低频外电场作用下载氚气分子和硅酸锂颗粒电极化旋转催化同位素交换速率的增强载氚气提取氚效率“SPB 方法”。     

磁化套筒惯性聚变一维集成化数值模拟

磁化套筒惯性聚变(magnetized liner inertial fusion,MagLIF)结合了传统磁约束聚变和惯性约束聚变的优点,理论上可以显著地降低聚变实现的难度,具有极大的应用潜力.以研究MagLIF中的关键问题为目标,建立能够综合考虑磁化、预加热、套筒内爆、聚变反应、端面效应、磁通压缩等多种复杂机制在内的集成化物理模型,特别是通过引入流体喷射模型,使得可以在一维计算条件下考虑具有二维特性的端面损失情况,并额外考虑Nernst扩散项对磁通损失的影响.在此基础上编写实现一维集成化MagLIF数值模拟程序MIST(magnetic implosion simulation tools),与FP-1装置(2 MA,7.2μs)上铝套筒内爆实验结果的对比验证了程序磁流体模块的正确性;将聚变模块纳入后与国外同类程序LASNEX和HYDRA计算结果进行整体比较,所得数值结果总体接近,主要差异体现在燃料温度的计算上,对可能影响的原因进行了简要分析.所建立的集成化模型与程序将为未来开展MagLIF聚变实验研究提供坚实的理论基础和重要工具.     

CFETR水冷陶瓷增殖剂包层中子学分析

基于中国聚变工程实验堆(CFETR)水冷陶瓷增殖剂(WCCB)三维中子学模型,应用蒙特卡罗输运程序MCNP5和IAEA聚变评价核数据库FENDL2.1,完成了WCCB中子学性能分析。研究了在200MW、500MW、1.0GW、1.5GW聚变功率下中子壁载荷(NWL)、氚增殖率(TBR)、核热沉积以及包层材料的辐照损伤。结果显示,目前WCCB包层核分析结果满足CFETR设计要求。     

中国ITER固态实验包层模块活化计算分析

根据中国ITER实验包层模块CH HCSB TBM设计结构建立了TBM径向一维模型。运用燃耗计算程序BISON、聚变堆放射性计算程序FDKR和衰变链数据库AFDC-DLIB,计算了TBM设计中不同停堆时间活化产物的放射性、余热沉积、潜在生物危害因子BHP和辐射剂量,并根据计算结果初步得出相应的变化规律。计算结果表明,对于HCSB TBM来说,在500MW聚变功率下运行一年,停堆时的总放射性、余热、BHP分别为4.587×10¹⁵Bq, 9.009×10^-3MW和90.73 km³•kW^-1,后板处辐射剂量3.278×10⁵ mSv•h^-1。结果表明,CH HCSB TBM不存在突出的放射性环境安全问题。     

The reactions of 12C with 12C AT 293 MeV

Energy spectra, angular distributions, and elemental yield distributions have been measured for products Z = 1?9 produced in the reactions of ¹²C on ¹²C. A total reaction cross section 1170_?100⁺¹⁷⁰ mb was determined from the measured elemental cross sections and the principle of charge conservation. This total reaction cross section is about 250 mb less than the geometric cross section and agrees with the Glauber-model calculations of DeVries and Peng. The experimental energy spectra, angular distributions, and yield distributions were compared with those from model calculations for the statistical decay of the products of fusion and of incomplete fusion reactions. For both types of calculations, a modified version of the code LILITA was used. By comparing the data to model calculations, an upper limit of 75 mb for the fusion cross section was determined. That limit corresponds to an upper limit of L_crit for fusion of 10? in the sharp-cutoff approximation. The dominant reaction mechanisms appear to be incomplete fusion processes.     

基于垂直磁场压缩的EAST等离子体初步研究

 基于等离子体参数随大半径/小半径变化的规律,分析了电流平顶段垂直磁场与相关等离子体参数的关系。垂直磁场的非线性部分用于分析非感应电流驱动效应,包括自举电流效应。此外,对于给定的等离子体电流,推导了平顶段垂直磁场与线平均密度之间的关系,并进一步研究压缩后的等离子体。基于EAST第41195次放电的数据分析表明,垂直磁场强度的增大可以使得等离子体温度、密度、βp和自举电流份额获得提升,为等离子体参数高参数,特别是βp提供了一种可能性参考。