(候选包壳)SCC and corrosion evaluations of the F-M steels for a supercritical water reactor

SCC and corrosion evaluations of the F/M steels for

a supercritical water reactor

Seong Sik Hwang

a,*

,Byung Hak Lee a ,Jung Gu Kim b ,Jinsung Jang

a

a Korea Atomic Energy Research Institute,150Deokjin-dong,Yuseong-gu,Daejeon,Republic of Korea b

Department of Advanced Materials Engineering,SungKyunKwan University 300Chunchun-Dong,Jangan-Gu,

Suwon 440-746,Republic of Korea

Received 20November 2006;accepted 14March 2007

Abstract

As one of the Generation IV nuclear reactors,a supercritical water cooled reactor (SCWR)is being considered as a candidate reactor due to its high thermal e?ciency and simple reactor design without steam generators and steam separators.For the application of a structural material to a core’s internals and a fuel cladding,the material should be evaluated in terms of its corrosion and stress corrosion cracking susceptibility.Stress corrosion cracking and general corrosion tests of ferritic–martensitic (F/M)steels,high Ni alloys and an oxide dispersion strengthened (ODS)alloy were performed.Stress corrosion cracking (SCC)was not observed on the fractured surface of the T 91steel in the supercritical water at 500,550and 600°C.As the test temperature increased,the ultimate tensile strength (UTS)and yield strength (YS)of T 91decreased,and a high dissolved oxygen level induced corrosion and low ductility.The F/M steels showed a high corrosion rate whereas the Ni base alloys showed a little corrosion at 500and 550°C.Corrosion rate of the F/M steels at 600°C test was up to three times larger than that at 500°C.A thin layer composed of Mo and Ni seems to retard the Cr di?usion into the out layer of the corrosion product of T 92and T 122.ó2007Elsevier B.V.All rights reserved.

PACS:82.45.Bb;28.41.Te;81.65.Mq;68.37.Hk

1.Introduction

As one of the Generation IV nuclear reactors,a super-critical water cooled reactor (SCWR)is being considered as a candidate reactor due to its high thermal e?ciency and simple reactor design without steam generators and steam separators [1,2].At above the supercritical condition of 374°C,22.1MPa,the supercritical water does not change its phase through out the reactor core outlet.There-fore,the high temperature coolant is e?ectively used at over a 40%thermal e?ciency [3–5].

An elimination of reactor components such as the steam generators and steam separators is another advantage of the SCWR when compared to normal LWRs [6].Various

di?erent properties from those at a normal liquid state at room temperature such as a dielectric constant,an ionic product and a heat capacity are of major concern for the application of critical water to industrial utilities [7,8].A critical step for this good feature to be attainable is to choose a proper structural material.For the application of a structural material to a core’s internals and a fuel clad-ding,the material should be evaluated in terms of its tensile strength at a high temperature,creep strength,corrosion and stress corrosion cracking susceptibility,radiation resis-tance,weldability,etc.[9].Among the quali?cation items,corrosion and stress corrosion cracking (SCC)tests of F/M steels have been performed in a supercritical water environment in the present work.

The present work aims at evaluating the corrosion behavior and the SCC behavior of F/M steels as the candi-date material for the SCWR.

0022-3115/$-see front matter ó2007Elsevier B.V.All rights reserved.doi:10.1016/j.jnucmat.2007.03.168

*

Corresponding author.Tel.:+82428682310;fax:+82428688696.E-mail address:sshwang@kaeri.re.kr (S.S.Hwang).

https://www.sodocs.net/doc/285931958.html,/locate/jnucmat

Available online at https://www.sodocs.net/doc/285931958.html,

Journal of Nuclear Materials 372(2008)

177–181

2.Experimental

The SCC behaviors of two F/M steels (T 91a,T 92)were evaluated in a supercritical water environment.Deionized water ($0.05l S/cm)of below 10ppb of dissolved oxygen and about pH 6.5was used as a test solution.SCC tests using a slow strain rate tester (SSRT)and U-bend speci-mens were carried out at 500,550and 600°C.The surface of the U-bend specimens was polished with up to a 0.3l m alumina powder before U shape bending.The SSRT spec-imens were ground with #600SiC emery paper on their gauge section followed by a cleaning with acetone.The SCC tests using the U-bend and the SSRT specimens were carried out simultaneously in the same autoclave.After the tests,a gauge section of the SSRT specimen was observed with a scanning electron microscope (SEM)to con?rm a SCC occurrence.The cross sections of the U-bend speci-mens were also analyzed with a SEM.

The general corrosion resistance of three groups of materials such as F/M steels (T 91a,T 91b,T 92,T 122),high Ni alloys (alloy 625,690,Incoloy 800H)and an ODS alloy (MA 956,a commercial 20%Cr ODS alloy)was evaluated in the same supercritical water environment as the SCC test.Table 1shows the chemical composition of the test materials.Test coupons shown in Table 2were immersed in a supercritical water environment,and test conditions of the corrosion rate measurement are summa-rized in Table 3.A cross section of the corrosion coupons was analyzed with a SEM and energy dispersive X-ray spectroscopy (EDS).

The corrosion and SCC test loop of the supercritical water environment designed for operation at 650°C,

30MPa is schematically shown in Fig.1.It consists of a pressure vessel of a 3.3liter volume capacity made of Hastelloy C-276,a make up water control loop,a SSRT control unit and a data acquisition module.

Table 1

Chemical compositions (wt%)of the alloys for the general corrosion tests Alloy Fe C Si Mn P S Cr Ni Mo ––V Nb W Ti Al N O Cu Others T 9la Bal 0.0840.4380.3630.0190.00088.8760.1020.928––0.1970.0810.0160.03510.081T 91b Bal 0.10.280.450.01

0.0038.370.21

0.9––0.220.080.02

0.0480.17T 92a Bal 0.070.4690.5––0.20.05 1.80.06T 92b Bal 0.110.180.430.0160.0028.910.120.47–

–0.190.06 1.670.0040.0430.003B T 122Bal 0.10.270.580.0170.00212.120.350.35–

–0.190.06 1.840.0140.070.830.003B MA 956Bal 0.020.040.10.010.00819.40.05––0.32 4.80.0220.230.020.52Inconel 625 4.520.020.170.10.01<0.00121.860.049.02–

– 3.620.270.20.080.20Co Inconel 69010.70.0020.390.280.010.00229.858.30.01–

–0.340.020.0320.010

0.001Ta Incoloy 800H

47.60.07

0.4

0.74

0.01<0.00119.4

31.1

–

–

0.32

0.23

0.1

0.05Co

Table 2

Test materials and dimensions of the coupons for the corrosion rate measurement Alloy class Alloy

Shape

Ferritic–

martensitic steels

T 91a,T 91b,T 92,T

122

High Ni alloys (superalloys)Inconel 625,Inconel 690,Incoloy 800H ODS alloy

MA 956

Table 3

Test conditions of the corrosion rate measurement Environments

Subcritical and supercritical water Temperature (°C)370,400,500,600Pressure (kgf/cm)25

O 2(ppb)

<10(deaerated)Test times (h)

200Inlet conductivity (l S/cm)<0.1

pH

Neutrality (6.4–6.9)Flow rate (cc/min)

<10

178S.S.Hwang et al./Journal of Nuclear Materials 372(2008)177–181

3.Results and discussion

3.1.Temperature dependency of a SCC

T 91steels did not show SCC in the SSRT test at 550,and 600°C,instead they exhibited an oxide ?lm rupture at the necking area as shown in Fig.2.Fournier et al.also performed a SCC test on alloy 690,and a ductile feature with a cracking of the oxide at the gauge section found in their experiment was similar to that found in this study [10].The specimens also failed in a ductile fracture mode at 500°C.As the test temperature increased,the ultimate ten-sile strengths (UTS)and yield strengths (YS)of the T 91alloy decreased as shown in Fig.3.

Strain rates at 500°C and 550°C were 1.5*10à7s à1and 3.0*10à7s à1,respectively.It has been reported that the maximum stress increased as the strain rate increased for alloy 600in a high temperature water [11].In the pres-ent test,alloy T 91showed lower UTS at a faster straining rate of 3.0*10à7s à1than at a slower straining rate of 1.5*10à7s à1.The lower UTS at the faster strain rate of alloy T 91seems to be due to the high test temperature of 550°C;the increase of the strain rate did not give rise to an increase of the https://www.sodocs.net/doc/285931958.html,pared with the UTS at 550°C,the low UTS at 600°C explains the dominant e?ect of the test temperature on the tensile properties.The elon-

gations were similar (about 20%)regardless of the test temperatures.

Alloy 91a tested in a 100ppb dissolved oxygen (D.O.)content showed a little lower UTS than that in the 10ppb D.O.as shown in Fig.4.An elongation,however,decreased a lot as the D.O.level was increased.This implies that the high D.O.content caused a high oxidation reaction on the surface of the test specimen,resulting in a low duc-tility of the material.

3.2.Material dependency of a SCC

Fig.5shows the material dependency of the tensile strength of the F/M steels in the SCW.Alloy T 92b showed a higher UTS than that of T 91a at 500°C at the same strain rate of 1.5*10à7s à1;399MPa for T 92b,349MPa for T 91a.It is believed that the addition of W and N to alloy 92B increased its UTS.It seems to be nec-essary to undertake systematic tests to understand the e?ects of minor elements.

U-bend specimens of T 91did not show any cracking at the apex of the U-bend regardless of the test temperature at 500°C and 550°

C.

Fig.2.Feature of a necking of T 91a tested at 500°C by using an SSRT.

Fig. 1.Corrosion and SCC test loop of the supercritical water environment.

S.S.Hwang et al./Journal of Nuclear Materials 372(2008)177–181

179

3.3.Material dependency of general corrosion

The F/M steels showed a corrosion rate as a form of a weight gain,and the steels showed a higher corrosion rate than the Ni alloys at the supercritical conditions as shown in Fig.6.The big di?erence in the corrosion rates between the two groups(F/M steels and Ni alloys)seems to be from a dissolution behavior of the main alloying elements which are Fe for the F/M steels,and Ni for the Ni alloys.The F/M steels showed a smaller corrosion rate at400°C.At the subcritical condition of370°C,the F/M steels showed a weight loss.On the other hand,alloy MA956,Incoloy 800H and the Ni base alloys showed a little weight change before and after the immersion test regardless of the test temperatures;subcritical or supercritical water.

A main factor for the general corrosion rate was the test temperature as shown in Fig.7.The corrosion rates of the F/M steels at600°C were about three times larger than those at500°C.The high Cr content of the alloy122seems to show a lower corrosion rate than that of the alloy T91, but it was less e?ective than the test temperature.

There was a small di?erence in the corrosion layer thick-ness or the corrosion rates of the T92and T122samples, which had di?erent Cr contents(9%and12%,respectively).

3.4.Corrosion product analysis

Oxide of the alloy T91was composed mainly of three layers as shown in Fig.8.There was a thin layer(L3on the micrograph)between layers2and4,which was enriched in Mo and Ni.A distinct element variation was measured in layers2and3;depletion of Fe with an enrich-ment of Cr was shown in layer2,depletion of Cr with an enrichment of Mo was detected in https://www.sodocs.net/doc/285931958.html,yer3enriched in Mo seems to retard an outward Cr di?usion.A source of the Mo in layer3seems to be the alloying element of T91 and the test vessel material of Hastelloy C276.An ion exchanger to purify the test solution was not installed for this corrosion test,so the Mo from the test vessel is consid-ered to be dissolved and deposited onto the corrosion

Fig.8.Micrographs of the oxide on the T91sample tested in a SCW at

550°C/25MPa(D.O.<10ppb).

180S.S.Hwang et al./Journal of Nuclear Materials372(2008)177–181

product.Alloy T92and alloy T122showed a similar ele-mental pro?le to T91.Was et al.reported a similar behav-ior of a layered structure of an oxide in type304L Stainless steel for a550°C test in SCW[5].

On the surfaces of the alloy MA956,alloy625,alloy 690,and Incoloy800H,thin oxides were formed.A small amount of oxide was formed on the alloy625surface,so the main alloying elements did not change at the surface. Another research group has also reported a similar result for the corrosion of alloy625[5,6].Fig.9shows a compar-ison of oxide thickness of MA956,Ni alloys and Incoloy 800H in the sub critical and super critical water.Summary of the corrosion product analysis is shown in Table4. 4.Conclusions

–No SCC was observed on the fractured surface of the T 91steel in the supercritical water at500,550and600°C.–As the test temperature increased,the UTS and YS of T 91decreased.

–High dissolved oxygen level induced corrosion and a low ductility of the F/M steel,T92b.

–Due to the addition of W,Alloy92b showed higher UTS than that of alloy91a.

–The F/M steels showed a high corrosion rate whereas the Ni base alloys showed a signi?cantly smaller corro-sion rate at500and550°C.

–Corrosion rate in supercritical water at600°C was up to three times larger than that of at500°C for the F/M steels.

–A thin layer composed of Mo and Ni seems to retard the Cr di?usion into the out layer of the corrosion product of T92and T122.

Acknowledgements

The research was sponsored by the Ministry of Science and Technology,Korea under the National Mid-and Long-term Atomic Energy R&D Program.The authors acknowledge V&M Tubes,France for providing of sample materials.

References

[1]K.Kataoka et al.,in:Proceedings of ICAPP03,Cordoba,Spain,4–7

May,2003.

[2]D.Squarer,T.Schulenberg,D.Struwe,Y.Oka,D.Bittermann,N.

Aksan,C.Maraczy,R.Kyrki-Rajama¨ki,A.Souyri,P.Dumaz,Nucl.

Eng.Des.221(2003)167.

[3]Y.Oka,S.Koshizuka,SCR-2000,6–8November2000,Tokyo,The

University of Tokyo,2000,p.1.

[4]J.McKinley,S.Teysseyre,G.S.Was,D.B.Mitton,H.Kim,J.-K.Kim,

https://www.sodocs.net/doc/285931958.html,tanision,GENES4/ANP2003,15–19September,2003,Kyoto, Japan,Paper No.1027.

[5]G.S.Was,S.Teysseyre,J.McKinley,in:NACE’s International

Conference,Corrosion2004,New Orleans,Paper No.04492,2004.

[6]S.Teysseyre,J.McKinley,G.S.Was,D.B.Mitton,H.Kim,J.-K.Kim,

https://www.sodocs.net/doc/285931958.html,tanision,in:Proceedings of11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems–Water Reactors,Stevenson,WA,10–14August,2003,Paper No.71221.

[7]P.Kritzer,SCR-2000,6–8November,2000,Tokyo,The University of

Tokyo,2000,p.1.

[8]Y.Watanabe,H.Abe,Y.Daigo,GENES4/ANP2003,15–19Septem-

ber,2003,Kyoto,Japan,Paper1183.

[9]J.Jang,W.S.Ryu,J.H.Hong,in:Proceedings of6th Conference on

Mechanical Behavior of Materials,Jeju,Korea,8November,2002,p.

425.

[10]L.Fournier, D.Delafosse, C.Bosch,Th.Magnin,in:NACE’s

International Conference,Corrosion2001,Houston,TX,Paper 01361,2001,p.1.

[11]R.N.Parkins,in:G.M.Urgiansky,J.H.Payer(Eds.),Stress Corro-

sion Cracking–The Slow Strain Rate Technique,ASTM-STP-665, 1979,p.5.

Table4

Summary of the corrosion product analysis on each alloy Alloys Oxide

thickness

Oxide layer

F/M steels(T 91,T92,T

122)$10l m

Thick

Mo Segregated,a boundary between Fe

and Cr oxide Outside of Mo segregated

layer:Cr depleted

Ni base

alloys.

Incoloy

800H Very thin

(less than

1l m)

Not analyzed by EDS

S.S.Hwang et al./Journal of Nuclear Materials372(2008)177–181181

超临界水冷堆燃料包壳候选材料的耐腐蚀性能

第34卷第4期2014年8月中国腐蚀与防护学报Journal of Chinese Society for Corrosion and Protection V ol.34No.4Aug.2014 超临界水冷堆燃料包壳候选材料的耐腐蚀性能 沈朝张乐福朱发文鲍一晨 上海交通大学核科学与工程学院上海200240 摘要：介绍了超临界水冷堆候选材料的相关腐蚀实验结果，并且讨论了每种候选材料在超临界水环境中的耐 腐蚀性能。根据当前研究结果可知，高Cr 含量的奥氏体不锈钢在超临界水中具有良好的抗腐蚀性能，因此其 最有可能成为超临界水冷堆燃料包壳材料。 关键词：超临界水冷堆燃料包壳腐蚀氧化膜 中图分类号：TG172文献标识码：A 文章编号：1005-4537(2014)04-0301-06 Corrosion Behavior of Candidate SCWR Fuel Cladding Materials SHEN Zhao,ZHANG Lefu,ZHU Fawen,BAO Yichen School of Nuclear Science and Engineering,Shanghai Jiao Tong University,Shanghai 200240, China Abstract:Corrosion performance of candidate clad materials for fuel of supercritical water-cooled reactor (SCWR)is reviewed with emphasis on that of four typical candidate alloys.According to the results presented in this paper,it is noted that the austenitic stainless steels with high Cr content show excellent corrosion resistance.Therefore,this kind of steels should be good candidate clad material for the fuel of SCWR. Key words:supercritical water-cooled reactor,fuel cladding,corrosion,oxide film 1前言超临界水冷堆(SCWR)是最有前景的第四代概念堆型之一。其堆内运行条件处于水的临界点(374℃，22.1MPa)之上，与轻水堆相比具有诸多优点，如其冷却剂为单相高焓态，SCWR 在结构上还省去了蒸汽发生器、汽水分离器和干燥器等结构，由于冷却剂的质量减少使得整个反应堆的体积减小，同时工作效率更高(~45%vs 33%)。相对于目前的压水堆(PWR)和沸水堆(BWR)，SCWR 的主要特点是其提高了堆内工作温度和压力。目前，关于SCWR 堆芯结构材料如包壳、栅格、冷却剂棒等的研发和选材是限制SCWR 进一步发展的关键问题。在所有的堆芯结构材料中，燃料棒包壳材料的 工作条件最为苛刻，在正常工况下其表面热点设计温度超过600℃[1-5]，而在瞬态时将会更高。而其它堆芯构件的工作条件类似或者好于燃料包壳，可以选用类似于燃料包壳材料作为其结构材料。因此，发展SCWR 燃料包壳用材是发展SCWR 最为关键的问题。 当前压水堆和沸水堆燃料包壳用材均为锆合金，其工作温度严格限制在360℃以下，因为当温度超过360℃时，其腐蚀速率大幅升高，且其机械强度会大幅下降。目前，大部分的锆合金在超临界水(SCW)中其腐蚀速率随时间呈现出线性增长关系[6]。因此，将锆合金应用在SCW 中的可能性很小。到目前为止，关于SCWR 包壳材料的技术要求还没有明确的给出，本文综述了SCWR 包壳候选材 料的腐蚀筛选实验结果，并且对影响材料在SCW 中腐蚀性能的因素进行了相关分析。 定稿日期：2013-06-21基金项目：国家重点基础研究发展计划项目(2007CB209802)资助作者简介：沈朝，男，1990年生，硕士生，研究方向为核材料腐蚀及 其水化学通讯作者：张乐福，E-mail ：lfzhang@https://www.sodocs.net/doc/285931958.html, DOI ： 10.11902/1005.4537.2013.123

【CN109652628A】一种核燃料包壳用FeCrAl合金及其制备和晶粒尺寸控制方法【专利】

(19)中华人民共和国国家知识产权局 (12)发明专利申请 (10)申请公布号 (43)申请公布日 (21)申请号 201910126790.0 (22)申请日 2019.02.20 (71)申请人 哈尔滨工程大学 地址 150001 黑龙江省哈尔滨市南岗区南 通大街145号哈尔滨工程大学科技处 知识产权办公室 (72)发明人 乔英杰　王鹏　王晓东　苏鸿全　 王坤　戚文　张晓红　 (51)Int.Cl. C21D 1/26(2006.01) C21D 6/00(2006.01) C21D 8/02(2006.01) C21D 9/00(2006.01) C22C 38/06(2006.01) C22C 38/18(2006.01) G21C 3/07(2006.01) (54)发明名称 一种核燃料包壳用FeCrAl合金及其制备和 晶粒尺寸控制方法 (57)摘要 本发明属于核燃料包壳技术领域，具体涉及 一种核燃料包壳用FeCrAl合金及其制备和晶粒 尺寸控制方法。本发明所述的用于核电反应堆燃 料元件包壳的FeCrAl合金材料，以下组分组成： Cr，Al，Fe，杂质，其中，Cr合金元素的总重量百分 比含量为11.00wt％～14.00wt％，Al合金元素的 总重量百分比含量为5.50wt％-6.50wt％。本方 法制备的FeCrAl合金具有较好的高温氧化性能 及抗腐蚀性能。通过对合金元素含量、加工工艺 的控制，使得FeCrAl合金的晶粒尺寸能够满足设 计需要。同时本发明在制备FeCrAl合金材料的方 法中采用低温热轧、高温再结晶热处理等工艺， 使合金材料在1200℃水蒸气下具有非常优异的 抗高温氧化性能，高温蒸汽氧化速率远远低于目 前商用核电包壳材料Zr - 4。权利要求书1页 说明书5页 附图5页CN 109652628 A 2019.04.19 C N 109652628 A

耐事故燃料包壳涂层材料研究现状

耐事故燃料包壳涂层材料研究现状 1 前言 福岛事故暴露了现有UO2-锆合金燃料形式在抵抗严重事故性能方面的不足。锆合金涂层是耐事故燃料包壳的技术方向之一。通过在锆合金包壳表面添加涂层，使传统锆合金包壳材料发挥更大的效能或能经受苛刻的使用环境，并延长其使用寿命。目前国际上研究的锆合金涂层主要包括以下方向：MAX相涂层、Si涂层、Cr涂层等。 2 MAX相涂层 2.1 发展现状 MAX相材料是继碳化硅陶瓷材料发展之后一种新型的三元陶瓷材料，其微观结构具有典型的层状特征，宏观特性兼具结构陶瓷和金属材料的性能优势，如良好的导热性和导电性，易于机械加工，密度小，抗热振动，不易弯曲，较低的热膨胀系数，兼具各向异性的力学性能和各向同性的热学性能[1]。代表性的MAX相材料包括TixSiCy、TixAlCy等。 结合MAX相涂层的优点，采用MAX相涂层技术的锆合金包壳，

在保证涂层完整性的前提下可以解决包壳的如下问题： 1）提高正常运行下的耐腐蚀性能，减少氧化和吸氢（减少氢化和脆化），以及氢化物再取向。 2）缓解严重事故的后果：提高了高温下包壳强度；通过减少包壳氧化速率和阻止蒸汽与锆合金的直接接触，显著减少事故下的产氢速率，缓解严重事故后果和延长反应堆应对时间。 3）改善流致振动导致的磨损。 美国Drexel大学围绕MAX相核材料正在开展一系列研究，如MAX 相材料的中子辐照损伤特性、氟盐环境和液态铅铋中的腐蚀、包壳管的制备、MAX相与核燃料界面反应特性等。西屋公司报告中指出Ti3AlC和Ti3SiC2三层陶瓷由于易加工、高韧性，均有可能作为燃料包壳材料，而且以上两种材料的导热性同其他包壳（锆合金、SiC基包壳、304不锈钢）相比较大。西屋公司的报告认为，对于升高温度下的安全裕量，Ti3AlC表现较好，仅次于SiC。法国、意大利、澳大利亚等也相继发表了一系列MAX相材料的离子辐照损伤行为研究成果，显示出该类材料具有优越的耐辐照损伤特性和高温自修复能力。但是，Ti3AlC材料没有相关工程应用经验，而且有较大的中子吸收截面（与不锈钢相近）。

用于超临界水堆燃料包壳的ODS铁素体钢的研究进展

第21卷第11期 2009年11月 钢铁研究学报 Journal of Iron and Steel Research Vol.21,No.11November 2009 基金项目:国家973重点基础研究发展计划资助项目(2007CB209800) 作者简介:何　培(19832),女,硕士生; E 2m ail :hepei0310@https://www.sodocs.net/doc/285931958.html, ; 修订日期:2009206227 用于超临界水堆燃料包壳的ODS 铁素体钢的研究进展 何　培,　周张健,　李　明,　许迎利,　葛昌纯 (北京科技大学材料科学与工程学院,北京100083) 摘　要:超临界水堆具有热效率高、系统简化、成本低等优点,成为第四代核反应堆中优先发展的堆型。ODS 铁素体钢由于其优异的高温力学性能和良好的抗辐照能力成为超临界水堆包壳最有希望的候选材料。旨在回顾 ODS 铁素体钢制造工艺,包括机械合金化参数的优化,热处理工艺的选择以消除力学性能上的各向异性。根据 超临界水堆包壳的服役条件,结合最新的实验数据,对ODS 铁素体钢的高温力学性能、在超临界水中的耐腐蚀性以及中子辐照稳定性进行了总结和展望。关键词:超临界水堆;氧化弥散强化;铁素体钢 中图分类号:TL35212 文献标识码:A 文章编号:100120963(2009)1120005207 Progress of Using Oxide Dispersion Strengthened Ferritic Steels as Fuel Cladding Materials in Supercritical W ater R eactor H E Pei ,　ZHOU Zhang 2jian ,L I Ming ,　XU Y ing 2li ,　GE Chang 2chun (School of Materials Science and Engineering ,University of Science and Technology Beijing ,Beijing 100083,China )Abstract :Supercritical water reactor (SCWR )is considered to be the most promising reactor among G en IV reac 2tors due to its higher thermal efficiency ,considerable system simplification and improved economics.ODS ferritic steels have been considered as one of promising cladding candidate materials for SCWR because of the excellent properties ,such as superior high temperature strength and outstanding swelling resistance.The aim of this paper is to review both the fabrication technology of ODS ferritic steels ,including the optimization of mechanical alloying parameters and thermal treatment methods for ameliorating the densification and deforming work induced mechani 2cal anisotropy ,and the evaluation of the high temperature mechanical properties ,corrosion resistance in SCW and neutron irradiation resistance of ODS ferritic steels according to the working conditions in SCWR.K ey w ords :supercritical water reactor ;oxide dispersion strengthened ;ferritic steel 能源问题日益成为世界发展所面临的共同危 机。核能是解决我国能源问题的重要途径之一。超临界水堆(Super Critical Water Reactor ,SCWR )作为第四代核能系统国际论坛(Generation ⅣInter 2national Forum ,GIF )提出的六种概念堆型中唯一的水冷堆,具有高效率、低消耗、低成本等优点。材料问题是目前SCWR 发展面临的两大技术难题之一[1]。反应堆元件包壳是反应堆中工况最苛刻的重要部件,面临着核燃料、高温高压、超临界水的腐蚀和强烈的中子辐照。根据2002年GIF 发布的SC 2 WR 技术报告,燃料包壳及堆内构件要求具有以下 特性[1]: (1)在工作温度范围(280～620℃,非正常情况 高达840℃ )具有高强度和耐腐蚀性; (2)低的应力腐蚀开裂(SCC )敏感性; (3)较低的中子吸收截面和吸收中子后的感生放射弱性; (4)中子辐射稳定性:低辐照肿胀、低辐照脆性、低活化; (5)易加工成型。

包壳材料介绍——试题

包壳材料介绍 1. 燃料包壳的作用是什么？ 保护燃料芯块不受冷却剂的侵蚀、避免燃料中裂变产物外泄，使冷却剂免受污染、保持燃料元件的几何形状并使之有足够的刚度和机械强度。 2. 燃料包壳处于什么工况？ 包容核燃料，承受高温、高压和强烈的中子辐照、包壳内壁受裂变气体压力、腐蚀、燃料肿胀、吸氢致脆和芯块包壳的相互作用等危害、包壳外壁受冷却剂压力、冲刷、振动和腐蚀以及氢脆等威胁。 3. 燃料包壳要求有哪些性能才能满足使用要求？ 核性能：小的中子吸收截面，辐照稳定性； 特别是热中子堆或用天然铀作燃料的反应堆，对包壳材料中子吸收截面的限制十分严格； 堆快中子堆，大多数元素的快中子吸收截面很小，选择材料的余地比较大。但对材料的稳定性及耐蚀性的要求更为突出； 通常选用截面小于1巴的金属为主要组分，吸收截面为数巴的元素作为合金化元素，截面在几十巴的杂质的含量限制在量级。 机械性能：足够的机械强度（高温强度） 化学性能：抗腐蚀性能、与冷却剂、裂变产物及燃料的相容性。 4. 常用的燃料包壳有哪些？ 可作为包壳材料和堆内结构材料的金属元素必须是低中子吸收截面的材料。根据它们的性能特点，各种材料的包壳用于不同的堆型。 如Al和Al合金用于低温水冷堆、压水堆中用Zr合金（如Zr-4，M5），BWR用Zr-2合金、Nb用于快中子堆。 5. 锆合金的合金化目的是什么？ 1、锆的性能很容易受杂质的影响； 2、高纯锆有良好的抗蚀性，但对纯度要求苛刻，价格昂贵，因此工程中多降低对原料纯度要求，通过合金化提高其抗蚀性和机械性能。 6. 锆合金的腐蚀特征有哪些？ 高温下的耐蚀性不足：360℃以上水中的耐蚀性差、燃料芯块与包壳的交互作用（PCI）及包壳的应力腐蚀破坏（SCC）。 1、均匀腐蚀 在高燃耗（50GWd/tU）下，氧化膜厚度增到50-60μm，伴生的应力易使氧化膜破裂或剥落，所以包壳管的水侧均匀腐蚀受到重视。 2、疖状腐蚀

(候选包壳材料)Corrosion and stress corrosion cracking in supercritical water

Corrosion and stress corrosion cracking in supercritical water G.S.Was a,*,P.Ampornrat a ,G.Gupta a ,S.Teysseyre a ,E.A.West a ,T.R.Allen b ,K.Sridharan b ,L.Tan b ,Y.Chen b ,X.Ren b ,C.Pister b a University of Michigan,Nuclear Engineering,1911Cooley,Ann Arbor,MI 48109,United States b University of Wisconsin,United States Abstract Supercritical water (SCW)has attracted increasing attention since SCW boiler power plants were implemented to increase the e?ciency of fossil-based power plants.The SCW reactor (SCWR)design has been selected as one of the Generation IV reactor concepts because of its higher thermal e?ciency and plant simpli?cation as compared to current light water reactors (LWRs).Reactor operating conditions call for a core coolant temperature between 280°C and 620°C at a pressure of 25MPa and maximum expected neutron damage levels to any replaceable or permanent core com-ponent of 15dpa (thermal reactor design)and 100dpa (fast reactor design).Irradiation-induced changes in microstructure (swelling,radiation-induced segregation (RIS),hardening,phase stability)and mechanical properties (strength,thermal and irradiation-induced creep,fatigue)are also major concerns.Throughout the core,corrosion,stress corrosion cracking,and the e?ect of irradiation on these degradation modes are critical issues.This paper reviews the current understanding of the response of candidate materials for SCWR systems,focusing on the corrosion and stress corrosion cracking response,and highlights the design trade-o?s associated with certain alloy systems.Ferritic–martensitic steels generally have the best resistance to stress corrosion cracking,but su?er from the worst oxidation.Austenitic stainless steels and Ni-base alloys have better oxidation resistance but are more susceptible to stress corrosion cracking.The promise of grain boundary engineering and surface modi?cation in addressing corrosion and stress corrosion cracking performance is discussed.ó2007Elsevier B.V.All rights reserved. 1.Introduction One of the most promising advanced reactor concepts for Generation IV nuclear reactors is the Supercritical Water Reactor (SCWR).Operating above the thermodynamic critical point of water (374°C,22.1MPa),the SCWR o?ers many advan-tages compared to state-of-the-art LWRs including the use of a single phase coolant with high enthalpy,the elimination of components such as steam gener-ators and steam separators and dryers,a low coolant mass inventory resulting in smaller compo-nents,and a much higher e?ciency ($45%vs.33%in current LWRs).Overall,the design provides a simpli?ed,reduced volume system with high thermal e?ciency.The challenge is provided by the substantial increase in operating temperature and pressure as compared to current BWR and PWR designs.The reference design for the SCWR [1,2]calls for an operating pressure of 25MPa and an outlet water temperature up to 620°C,Fig.1.Since supercritical water has never been used in nuclear power applications,there are numerous 0022-3115/$-see front matter ó2007Elsevier B.V.All rights reserved.doi:10.1016/j.jnucmat.2007.05.017 * Corresponding author.Fax:+17347634540.E-mail address:gsw@https://www.sodocs.net/doc/285931958.html, (G.S. Was). Journal of Nuclear Materials 371(2007) 176–201

第五章授课(53首端过程)解析

5.3 首端处理过程 ●“首端处理”这个术语最初是指料液在萃取分离之前，用沉淀、挥发等方 法去除某几个特定的裂片元素（如锆、铌、钌），以补充萃取净化之不足。 ●后来用于概括从乏燃料的接受到萃取上料为止的所有操作步骤。 ●所以，燃料组件的解体、包壳的脱除、燃料芯体的溶解和溶解液的调制等 统称为首端处理。 ●首端处理的目的： ?尽量去除燃料芯体以外的部分，即包壳材料及其它非燃料构件，使它们不参加到化学分离过程中去，以免影响化学反应，并且尽可能不进入高放射性废液，以减少废物处理量； ?进而将不同种类的核燃料加工成具有特定的物理和化学状态的料液，供后续分离工序使用，以便于用同一化工流程来处理各种不同型式的燃料元（组）件。 ?首端处理不仅要去除结构另件和包壳层，有时还要去除某些裂变产物（如131I、3H、85Kr、133Xe等）。 ●棒束型燃料组件和片组型燃料元件首先用机械切割方法分解，以卸除附属 的非燃料构件，如两端构件、定位格架、节流装置、组件盒等。 ●脱除包壳主要有三种方法： 1）化学脱壳法； 2）机械脱壳法； 3）机械化学结合法――切断浸取法（溶解燃料芯体而不溶解包壳）。 ●化学脱壳法是用化学试剂把金属包壳材料溶解掉而不溶解芯体，因为芯体 在这阶段溶解会造成不能允许的燃料损失。

●适当的溶剂是： ?对不锈钢可使用硫酸或王水； ?对锆合金可使用氟化物（NH4F+NH4NO3）； ?对镁合金可使用硫酸； ?对铝合金可使用氢氧化钠或加硝酸钠以减少氢气的产生。 ●虽然用化学脱壳法很容易处理任何形状的燃料元（组）件，但它有下列缺 点： 1）为了减轻腐蚀，设备需用昂贵的特种合金钢制造； 2）产生大量放射性废液（5-6m3/tU），不易处理到容许排放的水平而必须贮存起来，花费很大； 3）随包壳而溶解或脱落的铀、钚损失较大。 ●现在化学脱壳法仅用于生产堆和研究试验堆的铝包壳燃料元件。 ●机械脱壳法在英、法两国的石墨气冷堆镁包壳元件的脱壳中得到应用。?它是用机械方法在水下脱壳，用水作为屏蔽层，并可避免细粒飞扬。 ?切除元件的两端，再将包壳管用对称布置的三个铣刀按纵向切成长条，像剥香蕉皮那样剥离 ?然后将长条切成碎块以减少贮存体积。将金属铀芯棒表面的包壳碎屑刷净后，送去溶解。 ?这种方法要求水下工作的剥壳机能自动控制、耐水腐蚀和动作可靠，优点是高放射性废液较少。 ●又如快堆燃料组件为六边形不锈钢外壳，内为燃料棒，先切去端部接头、 充气室和解剖外套箱，把燃料组件拆散成较小的燃料棒束，尽量减少后续剪切的困难。