1、文章编号:摇 1007鄄8827(2017)03鄄0193鄄12用于高温气冷堆的核石墨周湘文,摇 唐亚平,摇 卢振明,摇 张摇 杰,摇 刘摇 兵(清华大学 核能与新能源技术研究院,先进核能技术协同创新中心,先进反应堆工程与安全教育部重点实验室,北京 100084)摘摇 要:摇 自 1942 年首次在 CP鄄1 反应堆中使用以来,核石墨因其优异的综合性能,在核反应堆特别高温气冷堆中被广泛使用。 作为第四代候选堆型之一,高温气冷堆主要包括球床堆和柱状堆两种堆型。 在两种堆型中,石墨主要用作慢化剂、燃料元件基体材料及堆内结构材料。 在反应堆运行中,中子辐照使得石墨的相关性能下降甚至可能失效。 原材料
2、及成型方式对于石墨的结构、性能及其在辐照中的表现起到决定性的作用。 辐照中石墨微观结构及尺寸的变化是其宏观热力学性能变化的内在原因,辐照温度及剂量对于石墨的结构及性能变化起决定性作用。 本文介绍了高温气冷堆中核石墨的性能要求及核石墨的生产流程,阐述了不同温度及辐照条件下石墨热力学性能及微观结构的变化规律,并对当前国内外核石墨的研究现状及未来核石墨的长期发展如焦炭的稳定供应和石墨的回收进行讨论。 本文可为有志于研发用于未来我国商业化的高温气冷堆中的核石墨的生产厂家提供参考。关键词:摇 核石墨; 高温气冷堆; 辐照; 微观结构; 物理、力学及热学性能中图分类号: 摇TQ127. 1+1文献标识码:
3、摇 A基金项目:国家公派留学基金(201406215002);国家科技重大专项(ZX06901);清华大学自主科研项目(20121088038).通讯作者:周湘文,副教授,博士. E鄄mail: xiangwen tsinghua. edu. cnNuclear graphite for high temperature gas鄄cooled reactorsZHOU Xiang鄄wen,摇 TANG Ya鄄ping,摇 LU Zhen鄄ming,摇 ZHANG Jie,摇 LIU Bing(Institute of Nuclear and New Energy Technology of T
4、singhua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology,the key laboratory of advanced reactor engineering and safety, Ministry of Education, Beijing100084, China)Abstract: 摇 Since its first successful use in the CP鄄1 nuclear reactor in 1942, nuclear graphite has pl
5、ayed an important role in nucle鄄ar reactors especially the high temperature gas鄄cooled type (HTGRs) owing to its outstanding comprehensive nuclear properties. Asthe most promising candidate for generation IV reactors, HTGRs have two main designs, the pebble bed reactor and the prismatic re鄄actor. In
6、 both designs, the graphite acts as the moderator, fuel matrix, and a major core structural component. However, the me鄄chanical and thermal properties of graphite are generally reduced by the high fluences of neutron irradiation of during reactor opera鄄tion, making graphite more susceptible to failu
7、re after a significant neutron dose. Since the starting raw materials such as the cokesand the subsequent forming method play a critical role in determining the structure and corresponding properties and performance ofgraphite under irradiation, the judicious selection of high鄄purity raw materials,
8、forming method, graphitization temperature and anyhalogen purification are required to obtain the desired properties such as the purity and isotropy. The microstructural and correspond鄄ing dimensional changes under irradiation are the underlying mechanism for the changes of most thermal and mechanic
9、al properties ofgraphite, and irradiation temperature and neutron fluence play key roles in determining the microstructural and property changes ofthe graphite. In this paper, the basic requirements of nuclear graphite as a moderator for HTGRs and its manufacturing process arepresented. In addition,
10、 changes in the mechanical and thermal properties of graphite at different temperatures and under differentneutron fluences are elaborated. Furthermore, the current status of nuclear graphite development in China and abroad is discussed,and long鄄term problems regarding nuclear graphite such as the s
11、ustainable and stable supply of cokes as well as the recycling of usedmaterial are discussed. This paper is intended to act as a reference for graphite providers who are interested in developing nucleargraphite for potential applications in future commercial Chinese HTGRs.Key words:摇Nuclear graphite
12、; High temperature gas鄄cooled reactors; Irradiation; Microstructure; Physical, mechanical and ther鄄mal propertiesReceived date: 2017鄄02鄄26;摇 Revised date: 2017鄄05鄄13Foundation item: State Scholarship Foundation of China (201406215002); Chinese National S&T Major Project (ZX06901); Tsin鄄ghua Universi
13、ty Initiative Scientific Research Program (20121088038).Corresponding author: ZHOU Xiang鄄wen, Associate Professor. E鄄mail: xiangwen tsinghua. edu. cnEnglish edition available online ScienceDirect ( http:蛐蛐www. sciencedirect. com蛐science蛐journal蛐18725805 ).DOI: 10. 1016/ S1872鄄5805(17)60116鄄1摇第 32 卷摇
14、 第 3 期2017 年 6 月新摇 型摇 炭摇 材摇 料NEW CARBON MATERIALSVol. 32摇 No. 3Jun. 2017摇1摇 IntroductionThe phrase nuclear graphite began to be used atthe end of 1942 when the first nuclear fission occurredin the graphite moderated nuclear reactor CP鄄11.From the early 1960s, the United Kingdom, the Unit鄄ed States a
15、nd Germany began to develop high temper鄄ature gas鄄cooled reactors (HTGRs). Japan began theconstruction of a 30 MWthhigh temperature test reac鄄tor (HTTR) in 1991, which reached its first criticali鄄ty in 1998. In China, a 10 MW experimental hightemperaturegas鄄cooledreactor( HTR鄄10 )2, 3,whose design s
16、tarted in 1992 and construction com鄄menced in 1995, reached it criticality in the end of2000, and its full power in the beginning of 2003.Since the Fukushima accident in March, 2011, thepublic has paid more and more attention to the safetyof nuclear power. As a candidate reactor for the Gen鄄eration鄄
17、IV reactors, the construction of a 2伊250 MWhigh temperature gas鄄cooled reactor pebble鄄bed mod鄄ule (HTR鄄PM) with inherent safety is underway inShidao Bay, Rongcheng of Shandong province, Chi鄄na and is expected to complete in 20174. In both ofthe research and commercial HTGRs, the reactor re鄄flectors
18、and cores have been constructed by structuralgraphite components. Past designs represent two pri鄄mary core concepts commercially favored for HTGRs:the prismatic block reactor (PMR) and the pebble鄄bed reactor (PBR)2. In both of the HTGR conceptsthe polycrystalline graphite not only is a major struc鄄t
19、ural component which offers thermal and neutronshielding and provides channels for fuel and coolantgas, channels for control and safety shut off devices,but also acts as a moderator and matrix material forthe fuel elements and control rods and a heat sink orconduction path during reactor trips and t
20、ransients.The polycrystalline graphite exhibits significantimportance in HTGRs because of its outstanding nu鄄clear physical properties such as high moderating andreflecting efficiency, a relatively low atomic mass anda low absorption cross鄄section for neutrons, in addi鄄tion to high mechanical streng
21、th, good chemical sta鄄bility and thermal shock resistance, high machinabili鄄ty and light weight5. The following example illus鄄trates the importance of nuclear graphite in more de鄄tails. For the thorium high temperature reactor (TH鄄TR) in Germany with a power of 300 MWe, nearly400 000 kg of nuclear g
22、raphite has been used2. InChina, approximately 60 tons of graphite was used inHTR鄄103, and more than 1000 tons of nucleargraphite will be used in HTR鄄PM as the structural ma鄄terial and matrix graphite of pebble fuel elements4.The raw materials of matrix graphite of fuel elementsfor HTR鄄10 and HTR鄄PM
23、 such as natural flake graph鄄ite and artificial graphite powder are supplied by Chi鄄nese domestic providers6, 7. The behavior of the in鄄dividual fuel particles and the matrix graphite materialin which the particles are encased are not consideredhere. However, it should be noted that although thegrap
24、hite technology associated with the matrix graph鄄ite is related to that of the main structural graphitesuch as the moderator there are differences as non鄄graphitized materials and natural flake graphite areused in the matrix graphite. Because so far no quali鄄fied domestic nuclear graphite is availab
25、le, all thestructural nuclear graphite materials for HTR鄄10 andHTR鄄PM are imported from Toyo Tanso of Japan. InApril 2015, China Nuclear Engineering CorporationLtd ( CNEC) announced that its proposal for twocommercial 600 MWe HTGRs (HTR鄄600) at Ruijincity in Jiangxi Province had passed an initial fe
26、asibili鄄ty review. The HTR鄄600 is planned to start construc鄄tion in 2017 and for grid connection in 20218. In or鄄der to achieve the economy and security of supply,the structural nuclear graphite must be provided bydomestic providers in China in the future. Fortunate鄄ly, with the rocketing developmen
27、t of photovoltaic in鄄dustry in China, several Chinese companies haveemerged which can produce the fine鄄grained isotrop鄄ic, isostatic molded, high strength graphite in largescale. Some of the manufacturers with state鄄of鄄the鄄artgraphite manufacture capabilities should be chosen asthe potential candida
28、te providers of the structural nu鄄clear graphite for HTGRs based on qualification pro鄄grams. However, during the operation of a reactor,many of the graphite physical properties are signifi鄄cantly changed due to the high fast neutron doses.The physical, mechanical and chemical properties ofgraphite c
29、an be influenced negatively by irradiationinduced damage, which would lead to the failure ofgraphite components. In pebble鄄bed HTGRs such asHTR鄄PM in China, the core support graphite structureis particularly considered permanent, although it isexpected that certain high neutron dose components(inner
30、 graphite reflector) will be replaced during thewhole lifetime of the reactor. During the life time ofthe reactor, the reflector graphite would be subjectedto a very high integrated fluence of fast neutrons ofaround 3伊1022n/ cm2(E0. 1 MeV)9, 10. Therefore,the pre鄄irradiation and post鄄irradiation com
31、prehensiveproperties of nuclear graphite candidates must be thor鄄oughly examined and evaluated. Those properties ofnuclear graphite are strongly dependent on the extentof anisotropy, grain size, microstructural orientationand defects, purity, and fabrication method.In this paper, basic nuclear requi
32、rements of nu鄄491摇新摇 型摇 炭摇 材摇 料第 32 卷clear graphite are presented and the specifications suchas the manufacture, material properties with three pri鄄mary areas (physical, thermal and mechanical) andirradiation responses of nuclear graphite suitable forHTGRs are elaborated, which could be a reference
33、forthe potential providers who are anxious to develop thenuclear graphite for future commercial HTGRs of Chi鄄na. The long鄄term considerations such as those invol鄄ving the cokes and recycle for nuclear graphite are al鄄so discussed.2摇 Nuclear requirements of graphite forHTGRs2. 1摇 Fission reactions wi
34、th neutronsThe tremendous energy produced in HTGRs isfrom the fission of isotopes such as92U233,92U235,and94Pu239. Fission of a heavy element, with releaseof energy and further neutrons, is usually initiated byan impinging neutron. The fission of92U235can be de鄄scribed as:92U235+0n1寅 A+B + v*n+ 200
35、MeV(1)The average yield per fission of92U235is about2. 5 fast neutrons. The energy of neutrons releasedfrom the fission reactions can be described by a Max鄄wellian distribution, with an average value of approx鄄imately 2 MeV. The probabilities of the fission reac鄄tions initiated by neutrons (the cros
36、s section) are in鄄versely proportional to the velocity of the neutrons. Itis essential to slow down the “fast冶 neutrons yieldedby fission to “thermal冶 neutrons with lower energies( 0. 025 eV at room temperature), which corre鄄spond to a neutron velocity of 2. 2伊103m/ s. The slo鄄wing down process resu
37、lts principally from energytransfer during elastic collisions between the neutronsand medium which is commonly called “moderation冶and the non鄄absorbing medium where the moderationtakes place outside the fuel is termed “moderator冶.As is known, the nuclear fuel for HTGRs is common鄄ly a mixture of low鄄
38、enriched92U235and92U238. Oncethe moderated thermal neutrons return to the fuel,they are most likely to cause fission in the92U235, in鄄stead of being captured by92U238.2. 2摇 Nuclear requirements for a good moderatorThere are two very fundamental nuclear require鄄ments for any moderator in HTGRs. First
39、, it musthave a small cross section for neutron absorption.Second, fast neutrons must be effectively sloweddown to thermal neutrons over short distances andwithin few collisions in the moderator.Thus, theprobability of neutron absorbed by the moderator im鄄purities and,92U238or absorbing structural m
40、aterials ina reactor is reduced. Therefore, a good moderatormaterial should exhibit a high slowing down powerand low adsorption ability11, 12.Furthermore, themoderator material should be economically acceptableand compatible with the other materials used in thecore of reactor, and maintain physical
41、and chemicalstability against the bombarding neutrons.According to the fundamental consideration ofNewton爷s law, the more energy loss of neutron percollision takes place when the target nuclei have thelower atomic mass. The average logarithmic energychange per scattering, 孜, can be described as equa
42、tion(2):孜抑2A+2/3(2)Where, A is the atomic mass of the collision nu鄄cleus. Therefore, the moderator should be based onthe elements of low atomic mass, which actually lim鄄its the choice to elements of atomic number (Z) lessthan sixteen. The parameter 孜 is a good index of themoderating ability of a mat
43、erial, but it is also depend鄄ent on the chances of a scattering collision occurring(scattering cross section, 撞s).The slowing downpower (SDP), 孜*撞s, which takes the two aspects a鄄bove into consideration is frequently used to indicatethe ability of a moderator material to slow down neu鄄trons. However
44、, since the parameter SDP is inde鄄pendent of the neutron absorption property of a mate鄄rial, it alone is not enough to guarantee a particularmaterial to be a good moderator. Therefore, the ratioof the SDP to the absorbing cross section ( 撞a),which is referred to as the moderating ratio, should bea b
45、etter indication to evaluate a particular material as amoderator. A shortcoming of the moderating ratio isthat it does not take the density of a material into con鄄sideration. For example, some gases such as heliumhave a high value for the moderating ratio, but theyare of little use as moderators bec
46、ause of their lowdensity at normal pressures and temperatures. Accord鄄ing to the requirements of a good moderator men鄄tioned above, the choice of potential moderator candi鄄dates of practical use thus reduces to the four materialslisted in Table 1.Table 1摇 Moderating properties of candidatemoderator
47、materials13.ModeratormaterialsSDP(孜*撞s)(cm-1)ModeratingratioDensity(g cm-3)H2O1. 530721.000D2O0. 370120001.106Be0. 1761591.852Graphite0. 0641701.70鄄1.85591第 3 期ZHOU Xiang鄄wen et al: Nuclear graphite for high temperature gas鄄cooled reactors摇摇 摇The materials in Table 1 are either in liquid orsolid sta
48、tes, which present acceptable density proper鄄ties. It can be seen that the deuterium oxide (heavywater) has the highest value of moderating ratio. Butthe costs to separate the heavy water from the ordinarywater are high and the heavy water is likely to leakout. The initial capital cost and leakage m
49、ake鄄up costare too high to make the heavy water as the firstchoice for the moderator. Ordinary water is very easyto obtain with low cost and relatively unaffected byneutron irradiation. However, neutron absorption byhydrogen reduces the moderating ratio. As a result,enriched uranium fuel should be u
50、sed in water moder鄄ated reactor to achieve the required neutron economy.Beryllium is not a good moderator for HTGRs becauseit is expensive, hard to machine, and highly toxic.Finally, graphite is a good moderator, which offersan acceptable compromise between nuclear properties,cost and utility as a s
