Design related aspects in advanced nuclear fission plants

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<ul><li><p>ar</p><p>sideeeptic sal aancliclart vend cin-b</p><p>of exposed material could be used as an advanced method for damage assessment in future nuclear</p><p>1. Introduction</p><p>The GENIV initiative has proposed six rture advanced nuclear plants: sodium fasttemperature reactor (VHTR), gas cooled fametal cooled fast reactor (LFR), molten ssupercritical water reactor (SCWR). The de</p><p>02 is gHTR arhereas</p><p>But even for these reduced conditions still quite signicant de-sign questions for structural materials have to be solved whichcaused code related activities [2,3]. Typical problems to be ad-dressed are:</p><p> Creep and stress rupture: data acquisition, extrapolation to500,000 h, negligible creep, cyclic softening, irradiation creep</p><p> LCF and creep: creepfatigue interactions (perfect material andcrack growth)</p><p>be shown also in a direct comparison between a 1/T-representation(e.g. LarsonMiller [7]) and a T-representation (e.g. a simplerelationship which was very successfully used some yearsago at Brown Boveri Turbomachinery Development in Switzerland[8]):</p><p> LarsonMiller Parameter:</p><p>log 10tR a logr3 b logr2 c log 10r d=T CLMP1</p><p>Journal of Nuclear Materials 409 (2011) 112116</p><p>Contents lists availab</p><p>cl</p><p>elsE-mail address: wolfgang.hoffelner@bluewin.chas envisaged in [1] often fail because of lacking materials or at leastmaterials data.</p><p>The development of process boundary conditions for the VHTRfrom 2002 (roadmap) to 2009 (NGNP) shown in Table 1 provides agood example how parameters can change within short time.</p><p>1/T. Although one would expect differences in the quality of thedifferent correlations, no such differences were found for grade91 and IN-617. Fig. 1 shows the results for IN-617.</p><p>This leads to the conclusion that several parameters areexpected to give comparable representations of the data. This cantal for safe construction and operation of plants. Ambitious plantconcepts like the VHTR with gas outlet temperatures of 1000 Cmore in concept phase. Valid design- and assessment codes are vi- 1Mo steel in [4]. Similar comparisons also exist for grade 91 [5] orIN-617 [6]. These parametrizations are based on iso-stress lineswhich are either considered to be linear in T or to be linear intems as they were envisaged in 20IV-roadmap [1]. Currently, SFR and Vtems for near-term-deployment, w0022-3115/$ - see front matter 2010 Elsevier B.V. Adoi:10.1016/j.jnucmat.2010.09.013power plants. 2010 Elsevier B.V. All rights reserved.</p><p>eactor concepts for fu-reactor (SFR), very highst reactor (GFR), liquidalt reactor (MSR) andscription of these sys-iven in the generatione considered as the sys-the others are much</p><p> Condition based monitoring, damage and life-time assessments Some of these topics will be discussed in the following sections.</p><p>2. Extrapolation of stress rupture data and creep</p><p>Extrapolation of stress rupture data is extremely important forcomponents which are intended to remain in-service for500,000 h. Many parametrizations of stress rupture data existwhich are aimed at representing stress rupture in a time to rupture(tR)-temperature (T)-stress (r)-space as summarized e.g. for 21=4 Crcan be assumed that irradiation creep is a matrix property. Finally it is shown that micro-sample testingDesign related aspects in advanced nucle</p><p>Wolfgang HoffelnerPaul Scherrer Institute, 5232 Villigen PSI, Switzerland</p><p>a r t i c l e i n f o</p><p>Article history:Available online 21 September 2010</p><p>a b s t r a c t</p><p>Important issues to be concreepfatigue, negligible crexamples from a martensibase superalloys. TraditionMiller Parameter or Monkma signicant inuence of cyfor grade 91. This is particuiour it is also possible to gearation (inelastic fatigue afraction rule. Results from</p><p>Journal of Nu</p><p>journal homepage: www.ll rights reserved.ssion plants</p><p>red for design of future reactors are: extrapolation of stress rupture data,, damage monitoring. The paper highlights some new developments takingteel (mod 9% Cr), oxide dispersion strengthened (ODS) steels and nickel-pproaches to extrapolation of (thermal) stress rupture data like LarsonGrant rule seem to be valid concepts also for advanced reactors. However,softening on creep rates and stress rupture data can be expected as shownly true for creepfatigue interactions. Based on cyclic stressstrain behav-ry good life-time predictions under creepfatigue with a strain range sep-reep ranges) technique which could replace the currently used linear lifeeam irradiation creep reveal no signicant inuence of dispersoid size. It</p><p>le at ScienceDirect</p><p>ear Materials</p><p>evier .com/locate / jnucmat</p></li><li><p>req</p><p>C w</p><p>350</p><p>e-st</p><p>cle,</p><p>ydrondu</p><p>ar MTable 1Changes in process parameters for a VHTR from the roadmap (2002) to current NGNP</p><p>2002 2009</p><p>Gas outlet temperature At least 1000 C NGNP 750option</p><p>Reactor pressure vessel 600 C Probably</p></li></ul>


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