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1 EUROPEAN COMMISSION nuclear science and technology Properties of Irradiated Stainless Steels for Predicting Lifetime of Nuclear Power Plant Components (PRIS) Contract No: FIKS-CT Final report (summary) Work performed as part of the European Atomic Energy Community's R&T Specific Programme Nuclear Energy, key action Nuclear Fission Safety, Area: Operational Safety of Existing Installations Directorate-General for Research 2004 Euratom

2 PARTNERS U. EHRNSTEN, P. KARJALAINEN-ROIKONEN VTT (Finland) A. JENSSEN Studsvik Nuclear AB (Sweden) J. LAPENA CIEMAT (Spain) M. LUDWIG Framatome ANP Erlangen (Germany) P. OULD, P. SCOTT Framatome ANP Paris (France) S. VAN DYCK SCK-CEN (Belgium) H. WESTERMARCK Westinghouse Atom AB (Sweden) II

3 FOREWORD The EC project FIKS-CT entitled "Properties of Irradiated Stainless Steels for Predicting Lifetime of Nuclear Power Plant Components" (PRIS) started in October The project was carried out under the Euratom 5 th Framework Programme by a consortium of seven members, each with assigned tasks as part of the overall project. The members came from the following companies: CIEMAT (Spain), Framatome-ANP SAS (France), Framatome- ANP Gmbh (Germany), SCK-CEN (Belgium), Studsvik Nuclear AB (Sweden), VTT (Finland), Westinghouse Atom AB (Sweden). The objectives of this project were to generate materials properties data for irradiated austenitic stainless steels of LWR internals as a function of fluence. This report, which is the final report of the PRIS project, has been prepared by Pascal OULD and Peter SCOTT (Framatome-ANP SAS) with the contribution of the following co-authors: Jesus LAPENA (CIEMAT), Ulla EHRNSTEN (VTT), Michael LUDWIG (Framatome-ANP GmbH), Steven Van DYCK (SCK-CEN), Anders JENSSEN (Studsvik Nuclear AB) and Henrik WESTERMARK (Westinghouse Atom AB). III

4 IV

5 ABSTRACT The in-core internal structures of LWR, mainly fabricated from austenitic stainless steels, receive very high neutron fluences during the life of a nuclear power plant (maximum dose around 100 dpa for 40 years of operation of a PWR). Numerous studies, performed on stainless steels irradiated usually in fast breeder reactors, have shown that neutron irradiation induces modifications of the material microstructure and mechanical properties (e.g. reduction of the ductility and of the fracture toughness). These studies have also shown a certain amount of swelling of the austenitic stainless steel in case of very high neutron doses. However, few data are available on the effect of irradiation on austenitic stainless steels with the neutron flux spectrum and irradiation temperature of LWRs. The objective of the PRIS project was to generate materials property data for irradiated austenitic stainless steels of LWR internals as a function of fluence that could be used for the structural integrity analyses. The material properties data investigated are the tensile and fracture toughness properties as well as the microstructural changes caused by neutron irradiation. The irradiated austenitic stainless steels studied came from the in-core internals of PWR and BWR plants. The present report compiles and analyses all the results obtained in the PRIS project. This paper presents and discusses the main results of the different tasks performed in the frame of the PRIS project as follows: - Tensile and microstructure (FEG-STEM analysis) characterisation of a highly irradiated Type 20 CW 316 stainless steel flux thimble tube from Ringhals 2 - Tensile characterization of Type 304L and 316L stainless steel control rod blades from Barsebäck 1 and Olkiluoto 2 irradiated to doses up to ~3 dpa (for the Barsebäck 1 materials) and 8.7 dpa (for Olkiluoto 2 materials) - Validation of fracture toughness test procedure using sub-size specimens. This task was performed using an unirradiated "model material", having mechanical properties approaching, as far as possible, those of an irradiated austenitic stainless steel. Fracture toughness tests and a Finite Element (FE) simulation of the J-R curves have been carried out. From this validation programme the sub-size 20 Side Groove SE(B) specimen was chosen for fracture toughness testing of irradiated Type 304L and 316L stainless steel control rod blades. - Fracture toughness characterization of the annealed Type 304L and 316L stainless steel control rod blades irradiated in the BWR plants Barsebäck 1 and Olkiluoto 2 - Fracture toughness characterization of the Type 20 Cold Worked 316 stainless steel flux thimble tube irradiated in the Ringhals 2 PWR plant. V

6 KEYWORDS Austenitic stainless steels Fracture toughness Irradiation Tensile properties Microstructure Tearing resistance ABBREVIATIONS BWR CMOD CW EDS FBR FEG STEM LWR PLT PWR SFE TEM UTS YS ε total, E tota ε u, E unif. ε rupt R of A σ flow, σ f RT Boiling Water Reactor Crack Mouth Opening Displacement Cold Worked (material) Energy Dispersive Spectroscopy Fast Breeder Reactor Field Emission Gun Scanning Transmission Electron Microscope Light Water Reactor Pin Loading Tension Pressurized Water Reactor Stacking Fault Energy Transmission Electron Microscope Ultimate Tensile Strength Yield Strength (0.2 offset yield strength) Total elongation Uniform elongation Deformation at rupture Reduction of Area Flow stress Room temperature VI

7 TABLE OF CONTENTS 0. REFERENCES X 1. INTRODUCTION 1 2. MICROSTRUCTURE AND MECHANICAL PROPERTIES OF IRRADIATED STAINLESS STEELS 2 3. VALIDATION PROCEDURE FOR FRACTURE TOUGHNESS DETERMINATIONS USING SUB-SIZE SPECIMENS Model material Fracture toughness of the model material 8 4. FRACTURE TOUGHNESS DETERMINATION FOR IRRADIATED STAINLESS STEELS USING SUB-SIZE SPECIMENS Material and test procedures Results TEARING RESISTANCE PROPERTIES OF A FLUX THIMBLE TUBE IRRADIATED TO HIGH NEUTRON DOSE Material and test procedures Results RECOMMENDATIONS 16 VII

8 LIST OF TABLES Table 1: Tensile properties of the flux thimble tube of Ringhals 2 3 Table 2: Unirradiated and irradiated tensile properties of the control rod blade materials from OL2 5 Table 3: Tensile properties of the control rod blade materials from Barsebäck 1 (irradiation temperature 288 C) 5 Table 4: Tensile properties at RT of the warm rolled AISI 304 plates 8 Table 5: Fracture toughness test matrix for the warm rolled type 304 model material 9 Table 6: J 1c, dj/a and K J evaluated for irradiated type 304L and 316L control rod blades of Barsebäck 1 12 Table 7: J 0.2 evaluated for the irradiated Type 304L and 316L control rod blades of Olkiluoto 2 12 Table 8: Fracture toughness results from the PLT tests on 20 CW 316 stainless steel flux thimble tube [2] 15 VIII

9 LIST OF FIGURES Figure 1: Composition profile across a grain boundary of the cold work type 316 thimble tube irradiated to 22 dpa (a) and 65 dpa (b) [3] 2 Figure 2: Y.S. and UTS data for the irradiated thimble tube material compared to data from the literature [1] 3 Figure 3: Tensile properties at RT and ~ 300 C of irradiated 304L and 316L control rod blades of Olkiluoto 2 and Barsebäck and of irradiated CW 316 thimble tube of Ringhals 2 6 Figure 4: Schematic view of BWR control rod blades 11 Figure 5: Samples cut (red lines) from the Type AISI 304L (a) and AISI 316L (b) control rod blades of Barsebäck 1 and Olkiluoto 2 [4] 11 Figure 6: J 0.2 evaluated for the irradiated Type 304L and 316L control rod blades of Olkiluoto 2 and Barsebäck 1 13 Figure 7: Schematic of the PLT specimen [2] 14 Figure 8: Sketch of the fixture and loading system for the PLT testing of thimble tube material [2] 14 IX

10 0. REFERENCES [1] Jenssen A. and Jakobsson R. "Mechanical properties, hardness and microstructure of a flux thimble tube irradiated in a PWR Work Package 2, Studsvik's part of deliverable D3 and D4." Studsvik Report, STUDSVIK/N(K)-02/106., October 24, 2002 [2] Jenssen A., Gigoriev V. and Jakobsson R. "Fracture toughness determination of a flux thimble tube irradiated to high dose in a PWR Work Package 2, deliverable D5." Studsvik Report, STUDSVIK/N(K)-04/049., April 02, 2004 [3] Nenonen P. and Karjalainen-Roikonen P. " TEM study of high fluence thimble tube material" VTT Research Report NO TU , April 02, 2004 [4] Ehrnstén U. and Karjalainen-Roikonen P. "Tensile tests, hardness and microstructure of irradiated AISI 304L and AISI 316L material from OLkiluoto 2 NPP" VTT Research Report NO TU , June 03, 2004 [5] Van Dyck S. "Metallographic and mechanical characterisation of LWR irradiated components BWR control blade handle materials PRIS project, work package 2, D3&D4" SCK.CEN Report. SCK.CEN-R-3829, March 2004 [6] Pelli R., Ehrnstén U. and Wallin K." Literature survey of unirradiated model material for the validation of fracture resistance procedure using sub-size specimens" VTT Research Report NO TU , March 16, 2004 [7] Kemppainen P., Ehrnstén U. and Karjalainen-Roikonen P. "Preparation of a model material" VTT Research Report NO TU , February 13, 2004 [8] Lapena J., Serrano F.J, Perosanz G., de Diego G. and Serrano J; "PRIS Project Properties of irradiated stainless steels for predicting lifetime of nuclear power plant components" CIEMAT Report DFN/ME-14/SP-03, June [9] Planman T., Wallin K. and Karjalainen-Roikonen P. "Fracture resistance analysis of cold-worked AISI 304 model material" VTT Research Report NO TU , February 10, 2004 [10] Ludwig M. "Report on FEM-calculations of different specimen types and sizes" Framatome-ANP Gmbh Report N NGTM/2004/en/0052 Rev. A, April [11] Ludwig M. "Material testing for the PRIS project to generate input data for Finite Element calculations" Framatome-ANP Gmbh Report N NGTM/2004/en/0051 Rev. A, March [12] Van Dyck S. "Fracture resistance determination of LWR irradiated components BWR control blade handle materials PRIS project, work package 5". SCK.CEN Report. SCK.CEN-R-3839, March [13] Grigoriev V., Jakobsson R. and Jenssen A. "Application of Pin-Loading Tension (PLT) technique for fracture toughness evaluation of a flux thimble irradiated in a PWR - Development of test procedure." Studsvik Report, STUDSVIK/N04/067, May 3, X

11 1. INTRODUCTION The in-core internal structures of LWR, which are mainly fabricated from austenitic stainless steels, are subjected to very high neutron fluences during the life of a nuclear power plant (maximum dose around 100 dpa for 40 years of PWR operation). Numerous studies, performed on stainless steels, irradiated mainly in fast breeder reactors, have shown that neutron irradiation induces very significant modifications of the material microstructure and mechanical properties (e.g. decrease of ductility and of the fracture toughness). These studies have also shown a certain amount of swelling of the austenitic stainless steel in case of very high neutron doses. However, few data are available on the effects of irradiation of austenitic stainless steel with the neutron flux spectrum and irradiation temperature typical of LWR. Moreover, cracking of some components has been detected in the internal structures of LWRs (e.g. in the core shrouds of BWR and in baffle-former bolts in PWR). These observations show the need for representative metallurgical data (microstructure, mechanical and toughness properties) on austenitic stainless steel irradiated in LWR conditions. The objective of the PRIS project was to generate materials property data for irradiated austenitic stainless steels typical of LWR internals as a function of neutron dose that could be used for the structural integrity analyses. The data consists of tearing resistance properties, tensile properties and information on microstructure evolution for austenitic stainless steels irradiated in LWRs. The irradiated austenitic stainless steels studied came from in-core internals of PWR and BWR plants. The PRIS project was divided into seven tasks named "work packages" (WP1 to WP 7), each of them being realized by one of the 7 partners. The main aims of the overall PRIS project work are briefly presented hereafter. Tearing resistance, mechanical properties characterisation and microstructure examinations were performed on a PWR flux thimble tube irradiated to high neutron doses (up to 65 dpa). The tearing resistance tests in this case were done using a pin loading tensile (PLT) technique. In order to develop a test methodology for validated tearing resistance values of irradiated austenitic stainless steels tested using sub-sized specimens, a programme was conducted using a "model unirradiated material". This material was chosen to have mechanical properties as close as possible to that of the irradiated materials tested. Tearing resistances were determined using "standard" and "sub-sized" specimens. Finite Element calculations were also performed on "standard" and "sub-sized" specimen geometries to support the analysis. Finally, tearing resistance tests were carried out using "sub-sized" specimens, on irradiated stainless steels machined from control rod blades of BWRs irradiated up to 3 dpa (Barsebäck 1) and up to 8.5 dpa (Olkiluoto 2). Tensile tests and microstructural examinations were also performed for these two materials. The objective of the present report, which corresponds to the "work package" WP6, is to compile and analyse all the results obtained in the PRIS project. 1

12 2. MICROSTRUCTURE AND MECHANICAL PROPERTIES OF IRRADIATED STAINLESS STEELS PWR FLUX THIMBLE Properties A ~15 cold drawn type AISI 316 flux thimble tube from the Ringhals 2 was investigated. The nominal outer diameter of the tube was 7.62 mm and the wall thickness 1.26 mm. Three regions of the flux thimble tube exposed to different irradiation conditions were investigated : irradiation temperature and doses ranging from 290 to 315 C and 22 to 65 dpa respectively. A detailed presentation of the microstructures and mechanical properties of the archive and irradiated materials are given in reference [1] Microstructural feature Optical and Field Emission Gun Scanning Transmission Electron Microscope (FEG-STEM) [3] were carried out. The optical micrographs of all the various material conditions examined did not show any apparent differences between the un-irradiated and the irradiated conditions. Slip bands visible on the etched cross section of the un-irradiated material are consistent with the cold worked structure of the material. A low density of inclusions was observed ; EDS analysis has shown that they are rich in Al, Ca, Mg, Si and Ti. FEG-STEM analyses revealed, as expected, the presence of Frank-loops defects and small voids of constant size (1-2 nm) -which were probably helium bubbles- the density of which increasing with dose. The voids were observed in the grains, but not at the grain boundaries. Small precipitates of a new phase were also observed which increased with dose. Local EDS chemical analyses and compositions profiles have shown segregation of Ni and Si to grain boundaries and depletion of Cr (Figure 1) Fe 22 dpa Cr 22 dpa Ni 22 dpa Fe 65 dpa Cr 65 dpa Ni, 65 dpa Concentration, el, Ni, Fe Concentration, el, Cr Concentration, el, Ni, Fe Concentration, el, Cr Distance from GB, nm Distance from GB, nm a b Figure 1: Composition profile across a grain boundary of the cold work type 316 thimble tube irradiated to 22 dpa (a) and 65 dpa (b) [3] Tensile properties The tensile properties of the flux tube thimble irradiated in a PWR to high doses of ~30 and 65 dpa at temperature of about 300 C (see Table 2) are in agreement with results reported in the literature for similar irradiations doses and temperature with either thermal reactor or fast reactor fluxes (Figure 2). In particular, the yield (YS) and tensile strengths (UTS), which 2

13 increased significantly with the dose, have similar values at 30 dpa and at 65 dpa. The ductility decreased with increasing dose but there is no further evolution of the ductility from 30 to 65 dpa. In fact the saturation of the tensile properties is obtained at a dose significantly lower than 30 dpa Fractographic examinations revealed ductile tearing behaviour with dimples; some brittle-like zones were observed only at the higher dose (~ 65 dpa) when tested at room temperature (RT). It is suggest that these zones could be related to channel fracture or could be due to strain-induced martensite enhanced by the neutron irradiation and hydrogen embrittlement. This point remains to be clarified. Table 1: Tensile properties of the flux thimble tube of Ringhals 2 Test Temp. C RT 320 C Dose dpa Irrad. Temp. C YS* MPa UTS MPa ε u ε total ε rupt. YS MPa UTS MPa ε u ε total ε rupt ~ ~ Yield strength (MPa) This study RT This study 320 C Fukuya et al. (Ref. 5), 320 C Furutani et al. (Ref. 2), 320 C Shogan et al. (Ref. 3), RT Shogan et al. (Ref. 3), 360 C Conermann et al. (Ref. 4), RT Conermann et al. (Ref. 4), 320 C Ultimate tensile strength (MPa) This study, RT This study, 320 C Fukuya et al. (Ref. 5), 320 C Furutani et al. (Ref. 2), 320 C VTT Type 316L (Ref. 6), RT VTT Type 316L (Ref. 6), 288 C Shogan et al. (Ref. 3), RT Shogan et al. (Ref. 3), 360 C Conermann et al. (Ref. 4), RT Conermann et al. (Ref. 4), 320 C Dose (dpa) Fluence (dpa) Figure 2: Y.S. and UTS data for the irradiated thimble tube material compared to data from the literature [1] 3

14 Note: Vickers Microhardness measurements (HV 0.05) were also performed on the flux thimble tube. The results are given in detail in [1]. The hardness increased with dose in agreement with the evolution of the tensile properties and the irradiation temperature: the hardness being slightly higher for the same fluence when the irradiation temperature was lower BWR CONTROL ROD MATERIAL OF OLKILUOTO 2 AND BARSEBACK Presentation Austenitic stainless steel materials of type AISI 304L and 316L were taken from control rod blades of Barsebäck Unit 1 (B1) and Olkiluoto Unit 2 (OL2) BWR nuclear Power Plants. The irradiation doses are ranging from 2.7 to 3.5 dpa for the B1 materials and dpa for the OL2 materials A detailed presentation of the material and of the microstructures and mechanical properties of archive and irradiated materials are given in references [4] and [5] Microstructural features Optical micrographies were taken of polished and etched cross sections of the un-irradiated and irradiated materials of the Type 316L and 304L control rod blades of OL2 [4]. The microstructures of the irradiated material are similar to those of the un-irradiated materials on this micrometer measurement scale. Micrographs were taken of the type AISI 304L and 316L control rod blades materials from Barsebäck Unit 1 (B1) un-irradiated and irradiated [5]. For these materials, there were also no obvious differences of microstructure at this level of magnification between the unirradiated and the irradiated materials Tensile properties Tensile properties were determined for the un-irradiated and irradiated control rod blade of B1 and OL2 [4] [5] (see Tables 2 and 3 and Figure 3). The tensile properties of the annealed type 304 and 316 stainless steel control rod blades irradiated to ~3 dpa (for the Barsebäck 1 materials) and 8.7 dpa (for the Olkiluoto 2 materials) showed an increase in YS and UTS and a decrease in ductility. These results suggest that the saturation of tensile properties may occur around 5 to 10 dpa. These properties are in agreement with those given in the literature for annealed austenitic stainless steel irradiated at temperatures close to 300 C The relative increases of YS and UTS with dose were quite similar for the two grades of stainless steels studied (Types 304 and 316). The fracture surfaces of the tensile specimens tested at RT and 288 C showed only ductile features with dimples. 4

15 Table 2: Unirradiated and irradiated tensile properties of the control rod blade materials from OL2 304 L 316 L Dose dpa Test Temp. C RT 288 C Irrad. Temp. C Orient. YS* MPa UTS MPa ε u ε total YS* MPa UTS MPa L T L L T L ε u ε total Table 3: Tensile properties of the control rod blade materials from Barsebäck 1 (irradiation temperature 288 C) Test Temp. C RT 288 C 304 L 316 L Dose dpa YS MPa UTS MPa ε u ε total R of A YS MPa UTS MPa ε u ε total R of A Note: Vickers hardness measurements with a 0.2kg load (HV0.2) were performed on the irradiated 304L and 316L control rod blade materials of Barsebäck 1. The hardnesses of the two materials were similar and increased with irradiation (~150 HV0.2 on unirradiated state and ~ 326 HV0.2 on irradiated state) as expected [5]. 5

16 Yield strength (MPa) L - B1 (RT) 304L - B1 (288 C) 316L -B1 (RT) 316L- B1 (288 C) 304L -OL2 (RT) 304L -OL2 (288 C) 316L -OL2 (RT) 316L - OL2 (288 C) CW 316 Ringhal2 (RT) CW 316 Ringh. 2 (~300 C) Dose (dpa) UTS (MPa) L - B1 (RT) 304L - B1 (288 C) 316L -B1 (RT) 316L- B1 (288 C) 304L -OL2 (RT) 304L -OL2 (288 C) 316L -OL2 (RT) 316L - OL2 (288 C) CW 316 Ringhal2 (RT) CW 316 Ringh. 2 (~300 C) Dose (dpa) L - B1 (RT) Uniform Elongation () L - B1 (RT) 304L - B1 (288 C) 316L -B1 (RT) 316L- B1 (288 C) 304L -OL2 (RT) 304L -OL2 (288 C) 316L -OL2 (RT) 316L - OL2 (288 C) CW 316 Ringhal2 (RT) CW 316 Ringh. 2 (~300 C) Total Elongation () L - B1 (288 C) 316L -B1 (RT) 316L- B1 (288 C) 304L -OL2 (RT) 304L -OL2 (288 C) 316L -OL2 (RT) 316L - OL2 (288 C) CW 316 Ringhal2 (RT) CW 316 Ringh. 2 (~300 C) Dose (dpa) Dose (dpa) Figure 3: Tensile properties at RT and ~ 300 C of irradiated 304L and 316L control rod blades of Olkiluoto 2 and Barsebäck and of irradiated CW 316 thimble tube of Ringhals 2 6

17 3. VALIDATION PROCEDURE FOR FRACTURE TOUGHNESS DETERMINATIONS USING SUB-SIZE SPECIMENS 3.1. Model material The fracture toughness of irradiated austenitic stainless steels can be obtained using classical fracture mechanics test specimens but usually, to save valuable material or when the dimensions of the irradiated material are too small, sub-size specimens have to be utilized. This also facilitates radiation precaution measures during specimen manipulation. In order to validate a procedure giving fracture toughness values on irradiated materials using sub-size specimens, an experimental programme was carried out on a "model material". The "model material" was an unirradiated material chosen to represent as far as possible the behaviour of an irradiated austenitic stainless steel from the point of view of the mechanism of fracture. It is obvious that the microscopic fracture mechanism of an irradiated austenitic stainless steel cannot be simulated using an unirradiated analogue because the microstructure and the defects created by irradiation cannot be obtained in unirradiated material. However, in fracture toughness testing, the material criteria used to obtain valid toughness values are only based on macro-plastic deformation characteristics i.e. the stress-strain behaviour of the material. It has been recommended in this study [6] to use model material having "σ f " values and fracture toughness properties close to that of the irradiated austenitic stainless steels within a recommended range of ± 20 for σ f and ± 30 for toughness. As a first step, three types of materials were retained as possible candidates for the model material: - Cold worked type 304 or 316 austenitic stainless steels, - Precipitation hardenable austenitic stainless steel of Type A286 - Precipitation hardenable martensitic stainless steel of Type 17-4PH. Based on the literature survey, the yield and ultimate strengths of Type stainless steels irradiated at LWR temperature seemed to saturate after about 3-5 dpa to a level of about 900 MPa; the fracture toughness decreased with increasing dose to saturate also around 3-5 dpa to values of about 50 kj/m² and even lower (but not the possible impact of void swelling at very high doses). The tensile and fracture toughness properties of the steels selected as possible model material were compared to the tensile properties of the irradiated Type 304L and 316L control rode blades from OL2 and to the range of toughness values expected from the literature survey for austenitic stainless steels irradiated at LWR temperature to a dose around 8 dpa. From these comparisons it was decided to select a cold work type 304 or 316 austenitic stainless steel as the model material. Therefore a 40 warm rolled (at about 200 C to avoid deformation martensite) AISI 304 was fabricated and characterized. The description of the fabrication and properties of the model material is given in detail in reference [7]. The Type 304 stainless steel chosen was a hot rolled plate of 49.2 mm thick produced at the Avesta Polarit Degerfors steel mill in Sweden. The YS and UTS obtained had RT average values of 720 and 810 MPa respectively (see table 4). 7

18 The tensile properties of these plates have flow stresses in the range of the flow stresses expected for the model material. Based on the criteria set up, the model material covers therefore materials with flow stresses ranging from 600 to 950 MPa. Consequently the fracture toughness validation programme on this model material only covered the case of irradiated stainless having flow stress values equal to or higher than that of the model material. Table 4: Tensile properties at RT of the warm rolled AISI 304 plates Plate 0.2 YS UTS N MPa MPa E uniform E total R of A Flow stress MPa Fracture toughness of the model material The test matrix of the fracture toughness tests performed on the model material by the different laboratories is given in table 5. Tests at CIEMAT, SCK-CEN and VTT In order to examine the different tests parameters which may affect ductile tearing resistance properties, 2 types of specimens were studied: SE (B) and C(T) specimens. Fracture toughness tests performed on 20 S.G C(T) specimens with different size (1/4T C(T )to 1T C(T) and on sub-size 20 S.G SE(B), 8 x10 x 55 mm 3, have shown that the subsize 20 S.G SE(B) specimen gives comparable J- a values to those obtained with the 20 S.G 1T C(T) specimens within the ASTM capacity limit or even beyond it [8-9]. It must be emphasized that the studied warm rolled stainless steel was prone to some splitting crack growth mechanism at RT perpendicular to the fracture plane. This feature is probably the reason why the toughness properties measured were lower at RT than at 288 C. Due to this unexpected behaviour the RT data must be used with care for validation purposes. The toughness values obtained at elevated temperatures were regarded as more reliable, their fracture being essentially by a ductile tearing mechanism without the splitting observed at RT. 2D finite element calculations of the fracture toughness tests have been performed to evaluate the validity of the fracture toughness results [10-11].The FE simulation showed that the reduced thickness of the sub size SE(B) specimen was sufficient to obtained the same J- a values as those obtained using the standard specimen thickness. 8

19 From this validation study, the sub-size 20 S.G SE(B) specimen was chosen to perform fracture toughness tests on the Type 304 and 316 stainless steel control rod blades irradiated in BWR plants. Tests at Studsvik Nuclear Fracture toughness tests on the model material were performed on C(T) specimens with very small thicknesses from 4mm to 0.6 mm at Studsvik (cf. [9]). The analysis performed suggests that the J-calculation methods applied by Studsvik Nuclear and VTT are not fully consistent. The high fracture resistance values on some specimens could result, for example, from a shear fracture mode which can be encountered with thin sheets [9]. Thus, it was not possible to use these results in the analysis of the model material. Table 5: Fracture toughness test matrix for the warm rolled type 304 model material Specimen Test Test Specimen Type Size (mm) Side grooves () Temp. ( C) Method* Number of tests F7- SE(B) 8 x10 x RT M-S, UC 7 VTT A1-1T C(T) standard 20 RT UC 10 VTT E3- SE(B) 8 x10 x PD 5 VTT A2-1T C(T) standard PD 5 VTT CR25 (T-S) SE(B) 8 x10 x RT UC 1 SCK-CEN CR26- (T-S) SE(B) 8 x10 x RT UC, PD 3 SCK-CEN CR29 (T-S) SE(B) 8 x10 x UC, PD 1 SCK-CEN CR37- (T-S) SE(B) 8 x10 x RT UC, PD 2 SCK-CEN CT T C(T) standard UC 11 CIEMAT CT T C(T) standard UC 5 CIEMAT CT T C(T) B = UC 4 CIEMAT CT T C(T) standard M-S, UC 19 CIEMAT CT T C(T) standard UC 5 CIEMAT CT T C(T) B = UC 8 CIEMAT M1-M2 0.5 T C(T) W/B =17/4 0 RT UC or PD 2 Studsvik M3 0.5 T C(T) W/B =17/4 33 RT UC or PD 1 Studsvik M4-M5 0.5 T C(T) W/B =17/ RT UC or PD 2 Studsvik M6-M7 0.5 T C(T) W/B =17/1.2 0 RT UC or PD 2 Studsvik M8-M9 0.5 T C(T) W/B =17/0.6 0 RT UC or PD 2 Studsvik * M-S: Multi-specimen test UC: Single-specimen technique, crack length measured by unloading compliance PD: Single-specimen technique, crack length measured by Direct Current Potential Drop. Lab. 9

20 Summary of the experimental results of the analysis on the model material The fracture toughness test results from VTT, SCK-CEN and CIEMAT on the model material have shown the following main conclusions: - The 20 S.G SE(B) of size 8 x10 x 55 mm 3 give comparable J - a values to those obtained with the 20 S.G 1T C(T) specimens tested within the ASTM defined capacity limits and even beyond it (as might be anticipated for a work hardening material; however caution is required for extending this conclusion to work-softening irradiated materials because of loss of constraint by the formation of intense slip bands Channel fracture). - The ~20 SG 1/2T C(T) and 1/4T C(T) specimens gave similar fracture toughness values -The unloading compliance and load normalisation J-R curves were both applicable for characterising the ductile tearing behaviour of austenitic stainless steels. Both the load normalisation J-R and key-curve data agreed acceptable with those measured by the unloading compliance or PD methods. - The J-R curves obtained with the partial unloading compliance technique were also in relatively good agreement with the multiple-specimen J-R curves at least for moderate crack extension with a 2mm. Nevertheless, the multiple-specimen method was seen to give the more reliable results. - The test procedures used by VTT, SCK-CEN and CIEMAT gave consistent J-R results at room temperature. -The unloading compliance method tended to underestimate the true crack length, up to 35 with the SE(B) specimens and 20 with the C(T) specimens. -The studied cold rolled stainless steel grade was prone to a splitting-type selective crack growth mechanism, which could here significantly complicated the interpretation of the ductile tearing test data measured on this material around room temperature. Normal, dimple-type ductile fracture was expected and observed at 288 o C. -The room temperature data should be applied with care for validation purposes due to the exceptional fracture behaviour that occurred at room temperature. The results measured at elevated temperatures could be used for the present application. 10

21 4. FRACTURE TOUGHNESS DETERMINATION FOR IRRADIATED STAINLESS STEELS USING SUB-SIZE SPECIMENS 4.1. Material and test procedures Material Ductile tearing resistance tests have been performed by SCK-CEN [12] and VTT on irradiated type AISI 304 L and 316 L control rod blades taken from Barsebäck 1 and Olkiluoto 2 BWR plants. Figure 4presents a schematic view of a control rod blade and Figure 5 shows the zones taken for the fracture toughness tests. Figure 4: Schematic view of BWR control rod blades a Figure 5: Samples cut (red lines) from the Type AISI 304L (a) and AISI 316L (b) control rod blades of Barsebäck 1 and Olkiluoto 2 [4] b 11

22 Test procedure The Fracture toughness tests were performed by SCK-CEN on the control rod material from Barsebäck and by VTT on the material from Olkiluoto 2. The tearing resistance tests were carried out using the 20 S.G. sub-size, 8 x 10 x 55 mm 3, SE(B) specimens. SCK-CEN performed tests at RT, 150 and 288 C on the AISI 304L material and at 288 C on the AISI 316L material (from Barsebäck 1). The J-R curves were obtained using the partial unloading method and also the reversing DC potential drop technique. The final crack extensions were measured after an oxidization treatment in air at 600 C; these measured crack extensions were used for correction of the crack extension deduced from compliance or potential drop [5]. VTT performed fracture toughness tests on the Olkiluoto 2 materials at RT and at 288 C. The J-R curves were obtained by the load normalization procedure Results Presentation The J 1c and dj/da obtained by SCK-CEN with the partial unloading compliance tests results and taking the experimental slope of the blunting line are given in table 6. The J 0.2 values deduced, from the fracture toughness test results of VTT, using a slope of 4x σ flow are given in Table 7. Table 6: J 1c, dj/a and K J evaluated for irradiated type 304L and 316L control rod blades of Barsebäck 1 Material Dose Test Temp. J Q (kj/m²) dj/da (MPa) K J (MPa.m½) 30 C AISI 304L dpa 150 C C AISI 316L 2.7 dpa 288 C Table 7: J 0.2 evaluated for the irradiated Type 304L and 316L control rod blades of Olkiluoto 2 Material Dose Test Temp. J 0.2 *(kj/m²) AISI 304L AISI 316L 8.3 dpa 8.7 dpa 23 C 288 C 23 C 288 C * Calculated with a blunting line slope of 4x σ flow

23 The J 0.2 values estimated for the AISI 304L and 316L control rod blade materials of Barsebäck 1 and of Olkiluoto 2 irradiated respectively to ~3 dpa and to dpa are given in Figure 6. It can be observed that the estimated J 0.2 at 288 C of the Olkiluoto 2 materials are consistent with the results at the same test temperature on the Barsebäck 1 material. There is therefore no a significant effect of dose between 2 and 8.7 dpa on the J 0.2 values L -23 C 316L -288 C 304L-23 C 304L-288 C 500 J0.2 (kj/m²) Dose (dpa) Figure 6: J 0.2 evaluated for the irradiated Type 304L and 316L control rod blades of Olkiluoto 2 and Barsebäck 1 Summary of the test result analysis The fracture toughness tests performed on Type 304L and 316L control rod blades of Olkiluoto 2 and Barsebäck 1 irradiated from ~2 to 8.7 dpa yielded high toughness values. These tests were carried out using the sub-size 20 S.G. sub-size, 8 x 10 x 55 mm 3, SE(B) specimens. The high J values obtained were beyond the validity limit given in the ASTM fracture toughness test standard. However it seems that the criteria for allowable J values given in the standards are too conservative and consequently the J values at crack initiation obtained here can be used as reasonable estimates of the J 1c values. This is supported by the validation work performed on the model material. The toughness results obtained at 288 C for doses around 2-3 dpa are in good agreement with the data from the literature. For higher doses of dpa, very few data have been found in the literature; the only results found correspond to temperatures of irradiation and testing of ~400 C. The upper bound of these literature data is consistent with the lower bound J 1c (288 C) values obtained in the present study. The results of the present study do not show a significant difference between the estimated J 1c values obtained at ~2 dpa and those obtained at dpa. The fracture surfaces showed a ductile tearing aspect. However, in the zone of the fatigue precrack, some intergranular fracture was observed. The reason(s) for these intergranular cracks remains to be investigated. 13

24 5. TEARING RESISTANCE PROPERTIES OF A FLUX THIMBLE TUBE IRRADIATED TO HIGH NEUTRON DOSE 5.1. Material and test procedures In order To evaluate the fracture toughness of the highly irradiated flux thimble tube, a special test technique was applied. In this test procedure, the fracture toughness was performed on a portion of the tube. Side grooving and pre-fatigue cracking was performed on the tube specimen before the ductile tearing test (Figure 7). The sketch of the Pin loading tension (PLT) test system is shown in Figure 8. The detailed analysis of these tests is given in reference [2]. The procedure used is given in ref. [13] Figure 7: Schematic of the PLT specimen [2] Figure 8: Sketch of the fixture and loading system for the PLT testing of thimble tube material [2] 14

25 5.2. Results The toughness parameters obtained are summarised in Table 8. Table 8: Fracture toughness results from the PLT tests on 20 CW 316 stainless steel flux thimble tube [2] Dose Test Temp. J (Pmax) J 0.2 K J dj/da Specimen dpa C kj/m² kj/m² MPa.m. ½ MPa Fa RT Fb Da 39 RT Db Aa RT Ab The toughness values of the flux thimble tube irradiated in a PWR up to 65 dpa are lower than those given in reference [83-84] for 20 CW type 316 stainless steels. This result can be due to the fact that the published data are from fast neutron reactors and the irradiation temperatures are usually higher than those of the present study. However it seems that the low fracture toughness obtained on the irradiated flux thimble tube of this study could also be explained by the relatively low fracture toughness of the material in the unirradiated condition. The fracture surface of the irradiated flux thimble tube tested at 320 C showed ductile tearing in agreement with published data. For the test at RT, fractographic data showing intergranular facets have not been found in the literature for comparison. It seems, nevertheless, that the brittle features appearing on the specimens tested at RT could be due to a channel rupture mechanism. The reason(s) for the intergranular zones observed on the fracture toughness tests at RT for the high irradiation dose of 65 dpa remains to be explained; one of possible mechanism could be low temperature hydrogen embrittlement. However, since the fracture resistance properties are higher at room temperature than at 320 C (table 8), it seems that this intergranular fracture mode has only a small effect on fracture properties. 15

26 6. RECOMMENDATIONS From this study, it appears that austenitic stainless steels irradiated in BWR at doses up to 9 dpa have fracture toughness properties which remain relatively high (J kj/m² at 288 C). To pursue proper size requirements for irradiated stainless steels, a validation programme using a wider range of model materials of high quality would be desirable. Moreover, it would be also interesting to perform some fracture toughness tests with thicker standard specimens (e.g. with 0.5T or 1T C(T) specimens) machined from components having the appropriate thickness, in order to obtain toughness properties data which are valid to accepted standards of today on these irradiated austenitic stainless steels. Such tests would give additional data that would show that sub-size specimens give accurate fracture toughness properties of irradiated austenitic stainless with high toughness values. However, it must be pointed out that large thickness components undergo non-uniform irradiation conditions (temperature and fluence). This could be a difficulty for the interpretation of the results. Fracture surfaces of the tensile and fracture toughness tests on the irradiated stainless steels in this study have shown some brittle-like and intergranular facets particularly at low temperature. This should be investigated in more detail to try to clarify the reasons for the appearance of these fracture features. 16