本文選題：活化腐蝕產(chǎn)物　切入點(diǎn)：CATE程序　出處：《華北電力大學(xué)(北京)》2017年博士論文　論文類型：學(xué)位論文

【摘要】：放射性源項(xiàng)關(guān)系反應(yīng)堆系統(tǒng)運(yùn)行、維修維護(hù)及退役等環(huán)節(jié),對(duì)輻射防護(hù)、個(gè)人和集體劑量以及安全分析有重大影響。水冷反應(yīng)堆中,結(jié)構(gòu)材料與冷卻劑接觸發(fā)生腐蝕,生成了較穩(wěn)定的氧化層,金屬離子穿過(guò)氧化層釋放進(jìn)入冷卻劑。輻照區(qū)的氧化層以及由冷卻劑攜帶進(jìn)入輻照區(qū)的金屬離子受中子輻照發(fā)生活化反應(yīng)成為放射性物質(zhì),冷卻劑中的放射性物質(zhì)在冷卻劑的攜帶下沉積到非輻照區(qū)形成了γ輻射場(chǎng),對(duì)電廠檢修維護(hù)及運(yùn)行人員構(gòu)成輻照危害。正常運(yùn)行工況下,壓水堆堆芯外90%的集體劑量是由與一回路冷卻劑接觸的管壁上沉積的活化腐蝕產(chǎn)物ACPs(Activated Corrosion Products)引起的。對(duì)于水冷聚變堆,不存在裂變產(chǎn)物,ACPs成為放射性的主要來(lái)源。無(wú)論壓水堆還是水冷聚變堆,ACPs對(duì)正常運(yùn)行工況下的ORE以及事故工況下的潛在放射性釋放都存在著重大影響,直接影響工作人員的照射劑量水平。對(duì)ACPs的研究是反應(yīng)堆事故分析、劑量與輻射防護(hù)優(yōu)化、放射性廢物管理等的重要技術(shù)基礎(chǔ),是反應(yīng)堆審查取證的重要環(huán)節(jié)。目前國(guó)內(nèi)外計(jì)算ACPs多數(shù)使用的是經(jīng)驗(yàn)?zāi)Ｐ秃桶虢?jīng)驗(yàn)?zāi)Ｐ?其應(yīng)用范圍非常有限,依賴于電廠運(yùn)行數(shù)據(jù)或試驗(yàn)數(shù)據(jù),模擬溫度、pH值等參數(shù)限制在一定范圍內(nèi)的變化,只適用于特定的堆型和工況;對(duì)放射性核素的種類和核反應(yīng)的種類有極大的限制,只能計(jì)算Co-58、Co-60、Fe-59、Cr-51、Mn-54等幾種放射性核素的核反應(yīng),不能滿足聚變堆高能中子輻照下多種材料的源項(xiàng)分析需求,也不能滿足事故瞬態(tài)下短壽命核素的計(jì)算需求;聚變堆獨(dú)有的脈沖運(yùn)行特點(diǎn)也對(duì)計(jì)算提出了新的要求。本論文開(kāi)發(fā)了基于經(jīng)典的經(jīng)驗(yàn)?zāi)Ｐ偷乃浞磻?yīng)堆主回路ACPs計(jì)算程序。對(duì)水冷反應(yīng)堆主回路ACPs的產(chǎn)生與遷移機(jī)理開(kāi)展研究,建立基于濃度差驅(qū)動(dòng)原理的機(jī)理模型,開(kāi)發(fā)了基于機(jī)理模型的水冷反應(yīng)堆主回路ACPs計(jì)算程序。脫離了對(duì)核電廠及試驗(yàn)回路的經(jīng)驗(yàn)系數(shù)的依賴,結(jié)合溶解度的計(jì)算成功實(shí)現(xiàn)了物質(zhì)遷移方向的自動(dòng)匹配功能,突破了以往程序?qū)Χ研图斑\(yùn)行工況的限制。借助課題組中的沉積試驗(yàn)及測(cè)量結(jié)果,根據(jù)對(duì)模型計(jì)算值和試驗(yàn)測(cè)量結(jié)果的分析,對(duì)沉積模塊進(jìn)行修正,成功實(shí)現(xiàn)了pH值對(duì)沉積行為的影響的模擬;對(duì)多種結(jié)構(gòu)材料進(jìn)行了不同運(yùn)行環(huán)境下的腐蝕行為模擬試驗(yàn),解決了聚變堆工況下腐蝕模型計(jì)算不準(zhǔn)確的問(wèn)題;引入EAF-2007數(shù)據(jù)庫(kù),為活化及衰變反應(yīng)提供核數(shù)據(jù),實(shí)現(xiàn)了計(jì)算任意放射性核素的功能;加入多種脈沖等效模塊,滿足不同計(jì)算需求及聚變堆型的要求,保證計(jì)算精度的同時(shí)可以大幅提高計(jì)算效率;添加點(diǎn)核積分模塊計(jì)算相應(yīng)的劑量率及職業(yè)照射ORE(Occupational Radiation Exposure),實(shí)現(xiàn)了活度濃度與劑量率的轉(zhuǎn)換。通過(guò)上述工作,克服對(duì)pH值變化范圍的限制,突破了以往程序?qū)Σ牧霞肮r、放射性核素種類的限制,直接給出γ劑量場(chǎng)使得計(jì)算結(jié)果更加直觀�；谝陨瞎ぷ�,開(kāi)發(fā)了適用于壓水堆和水冷聚變堆的ACPs計(jì)算分析程序CATE。為充分驗(yàn)證模型的正確性及程序的適用性,分別從試驗(yàn)驗(yàn)證和程序驗(yàn)證兩個(gè)角度選取了試驗(yàn)回路MIT-PCCL回路、水冷聚變堆ITER LIM-OBB回路和壓水堆秦山二期核電廠一回路進(jìn)行了模擬分析,并與公開(kāi)發(fā)表的文獻(xiàn)結(jié)果進(jìn)行了比對(duì)。計(jì)算結(jié)果均能與試驗(yàn)測(cè)量值和程序計(jì)算值保持在同一數(shù)量級(jí),在源項(xiàng)計(jì)算領(lǐng)域內(nèi)可以認(rèn)為計(jì)算結(jié)果是吻合的,從試驗(yàn)和程序的角度驗(yàn)證了模型的準(zhǔn)確性和結(jié)果的可靠性。水冷聚變堆的高溫高壓環(huán)境、產(chǎn)生的高能量中子會(huì)對(duì)結(jié)構(gòu)材料產(chǎn)生較強(qiáng)的腐蝕、活化作用,水冷聚變堆對(duì)結(jié)構(gòu)材料提出了更高的要求,結(jié)合我國(guó)已生產(chǎn)的多種低活化材料,應(yīng)用CATE程序首次實(shí)現(xiàn)了國(guó)際熱核聚變實(shí)驗(yàn)堆ITER(International Thermonuclear Experimental Reactor)環(huán)境下國(guó)產(chǎn)低活化材料及傳統(tǒng)奧氏體不銹鋼對(duì)水冷聚變堆ACPs影響的對(duì)比分析;當(dāng)前中國(guó)聚變工程試驗(yàn)堆CFETR(China Fusion Engineering Test Reactor)處于設(shè)計(jì)階段,ACPs源項(xiàng)的水平是其頒證的關(guān)鍵影響因素,可能對(duì)聚變堆設(shè)計(jì)和運(yùn)行有很大的影響,目前國(guó)內(nèi)尚無(wú)對(duì)CFETR的ACPs水平計(jì)算分析的研究工作,本文應(yīng)用CATE程序?qū)崿F(xiàn)了對(duì)CFETR包層回路的ACPs進(jìn)行計(jì)算分析。
[Abstract]:The radioactive source term relationship reactor system operation, maintenance and decommissioning and other sectors, have a significant impact on the radiation protection, analysis of individual and collective dose and safety. In water reactors, structural materials in contact with the coolant corrosion, formation of oxide layer is stable, metal ions through the oxide layer is released into the coolant. The oxide layer irradiated area and from the coolant carrying metal ions into the irradiated area irradiated by neutron activation reaction as radioactive substances, radioactive substances in the coolant in the coolant carrying deposition to the non irradiated area formed a gamma radiation field of power plant maintenance operation and maintenance personnel constitute radiation hazard. Under normal operating conditions, the collective dose 90% pressure the core is made of activated corrosion products ACPs deposition wall contact with a coolant on (Activated Corrosion Products) for the cause. Water cooled fusion reactor, there is no fission products, ACPs has become the main source of radioactivity. Both PWR or water-cooled fusion reactor under normal operating conditions, ACPs of ORE and the release of potential radioactive accident conditions have significant influence, directly affect the dose level of the staff. The study of ACPs reactor accident analysis dose, radiation protection and optimization, an important technical basis for management of radioactive waste, is an important part of the reactor. The review of evidence at home and abroad ACPs calculation is used mostly empirical and semi empirical models, the application range is very limited, depending on the power plant operation data and the test data, the simulation of temperature, pH value and other parameters change limit in a certain range, only suitable for specific reactor types and conditions; types of radionuclides and nuclear reactions are extremely limited, can only calculate the Co-58 Co-60, Fe-59, Cr-51, Mn-54 and other types of nuclear reactions of radionuclides, can not meet the high energy neutron irradiation of various materials in fusion reactor source analysis needs, can not meet the computing needs of short life nuclide transient accident; fusion pulse operation characteristic and puts forward new requirements for the calculation. This paper developed a calculation program the main circuit of water-cooled reactor ACPs empirical model based on the classical research. And the migration mechanism of water cooled reactor main loop of the ACPs was established based on the principle of differential drive mechanism model, a computer program is developed the main circuit of ACPs water cooled reactor based on mechanism model. From the experience dependent coefficient on the nuclear power plant and the test circuit. Automatic matching with the calculated solubility of the successful implementation of the migration direction, broke through the limitations of previous procedures on the reactor type and operating condition. By the research group The deposition of test and measurement results, according to the analysis of the calculated value and the experimental results of the model, the deposition module is modified, the successful implementation of the simulation effect of pH value on the deposition behavior of various structural materials; the corrosion behavior of different operation environment simulation test, solves the corrosion condition of fusion reactor model the problem of inaccurate calculation; introduction of the EAF-2007 database to provide data for the activation and nuclear decay reaction, the calculation of arbitrary radionuclides; adding a variety of pulse equivalent module, to meet the different computational requirements and fusion type requirements, and ensure the calculation accuracy can greatly improve the computational efficiency; calculation of dose rate and irradiation ORE corresponding occupation add the point kernel integral module (Occupational Radiation Exposure), the activity concentration and dose rate conversion. Through the above work, to overcome the pH value range In the limit, breaking the previous program of materials and conditions, radioactive nuclide types, directly given dose field makes the result more intuitionistic. Based on the above work, developed for PWR and water-cooled fusion reactor ACPs analysis the applicability of CATE. program is correct and full program verification model. From the test and verification procedures two angle test circuit MIT-PCCL circuit, ITER circuit and LIM-OBB pressurized water cooled fusion reactor Qinshan two nuclear power plant in a loop are simulated, and the results and the published literature were compared. The calculated results are with the experimental values and the calculated values remain in the same magnitude, that calculation results were consistent with the source term calculation domain, the reliability model and the accuracy of the results is verified by the test program and point of view. Water cooled fusion reactor The environment of high temperature and high pressure, high energy neutrons produced by strong corrosion of structural materials, activation, water-cooled fusion reactor has put forward higher requirements for structural materials, combined with a variety of production in China has low activation materials, application of CATE ITER for the first time the International Thermonuclear Experimental Reactor (International Thermonuclear Experimental Reactor) under the environment of domestic low activation of contrast material and traditional austenitic stainless steel pile of ACPs impact on water cooled fusion analysis; current China Fusion Engineering Test Reactor (CFETR China Fusion Engineering Test Reactor) in the design stage, ACPs source level is the key factor affecting the certification, may have a great impact on the fusion reactor design and operation, at present there is no domestic level of CFETR ACPs calculation and analysis of the research work, the application of CATE program on CFETR clad ACPs gauge circuit Calculation analysis.

【學(xué)位授予單位】：華北電力大學(xué)(北京)
【學(xué)位級(jí)別】：博士
【學(xué)位授予年份】：2017
【分類號(hào)】：TM623.2

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