Fracture toughness of irradiated and recovered vessel steels

This paper presents the fracture toughness measurements carried out on three vessel steels in an irradiated condition and after a post-irradiation recovery treatment. A statistical approach and the fracture parameters corresponding to two theoretical models of the fracture tests are used for evaluating toughness. Test results show that the neutron fluence gradually transforms the fracture behaviour of the vessel steels from ductile to brittle and seriously reduces their fracture toughness. The effectiveness of the recovery treatment, as evaluated from the toughness measurements, is confirmed, although the efficiency is not the same for the steels and depends on the evaluation parameter except in the case of almost complete recovery. The recovery effect increases with the received neutron fluence if the toughness values after treatment are compared with those in the irradiated condition rather than those in the as received condition.     

Fracture toughness of irradiated HT-9 for structural applications

The irradiation embrittlement sensitivity of HT-9 was investigated by conducting fracture toughness tests on compact tension specimens of HT-9 irradiated in Experimental Breeder Reactor (EBR) II to a fluence of 5.5 × 10²²n/cm². Test temperatures were 90, 205 and 450°C and the electropotential single specimen method was used. Results showed that the initiation fracture toughness of HT-9 is not affected significantly by test temperature, irradiation temperature and fluence. However, the tearing modulus increases substantially with increasing irradiation temperature but decreases with fluence in some cases. The effects of heat treatments on the fracture toughness are discussed also. Based on the fracture behavior described in terms of fracture toughness, the brittle fracture stress and the ultimate tensile stress, according to elastic-plastic fracture mechanics, a fracture toughness guideline is proposed.     

A review of the ASME low upper shelf toughness evaluation procedures for nuclear reactor pressure vessels

The ASME Section X1 Working Group on Flaw Evaluation has proposed criteria for the evaluation of reactor pressure vessel beltline materials which have an upper shelf energy less than 50 ft-lbs (69 J). These criteria have been assessed and applied to Linde 80 weld materials in recent investigations; this assessment and evaluation are described in the paper.

A key element in the evaluation procedure is the J_R curve for the relevant material. Recent experimental studies have demonstrated that the J_R curve is size dependent for some materials, in the sense that the J_R curve slope decreases with increasing specimen thickness. This paper assesses this experimental work and discusses it in the context of the integrity of nuclear reactor pressure vessels.     

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of . Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.     

Use of J–R curves in assessing the fracture behavior of low upper shelf toughness materials

The objective of this investigation was to evaluate the use of small specimen J–R curves in assessing the fracture resistance behavior of reactor vessels containing low upper shelf (LUS) toughness weldments. As required by the U.S. Code of Federal Regulations (10 CFR, Part 50), reactor vessel beltline materials must maintain an upper shelf Charpy V-Notch (CVN) energy of at least 50 ft-lbs (68 J) throughout vessel life. If CVN values from surveillance specimens fall below this value, the utility must demonstrate to the U.S. Nuclear Regulatory Commission (NRC) that the lower values will provide “margins of safety against fracture equivalent to those required by Appendix G of the ASME Boiler and Pressure Vessel Code”. This paper will present recommendations regarding the material fracture resistance aspects of this problem and outline an analysis procedure for demonstrating adequate fracture safety based on CVN values.It is recommended that the deformation formulation of the J-integral be used in the analysis described above. For cases where J-integral fracture toughness testing will be required, the ASTM E1152-87 procedure should be followed, however, data should be taken to 50% to 60% of the specimen remaining ligament. Extension of the crack growth validity limits for J–R curve testing, as described in E1152-87, can be justified on the basis of a “J-controlled crack growth zone” analysis which shows an engineering basis for J-control to 25% to 40% of the specimen remaining ligament. If J-R curve extrapolations are required for the analysis, a simple power law fit to data in the extended validity region should be used. The example analysis performed for low upper shelf weld material, showed required CVN values for a reactor vessel with a 7.8 inch (198 mm) thick wall ranging from 32 ft-lbs (43 J) to 48 ft-lbs (65 J), depending on the magnitude of the thermal stress component.     

Fracture toughness prediction for highly irradiated RPV materials: From test results to RPV integrity assessment

New fracture toughness data are represented for highly irradiated RPV materials that were obtained by testing standard compact specimens with thickness of 12.5 mm and 25 mm and pre-cracked Charpy specimens machined from the RPV decommissioned. Two advanced engineering methods, the Master Curve and the Unified Curve, are applied for treatment of the test results. Application of the dependence of fracture toughness K_JC on test temperature T predicted with the Master Curve and the Unified Curve methods on the basis of surveillance specimens testing is discussed for RPV integrity assessment when the reference K_JC(T) curve is recalculated to the crack front length of the postulated flaw that is considerable larger than thickness of surveillance specimens. The prediction of the K_JC(T) curve transformation caused by neutron irradiation is considered.     

Fracture toughness of China low activation martensitic (CLAM) steel at room temperature

The fracture toughness (J_IC) of China low activation martensitic (CLAM) steel was tested at room temperature through the compact tension specimen, the result is 417.9 kJ/m², which is similar to the JLF-1 at same experimental conditions. The microstructural observation of the fracture surface shows that the fracture mode is a typical ductile fracture. Meanwhile, the fracture toughness is also calculated on the basis of the fractal dimension and the calculated result is 454.6 kJ/m², which is consistent well with the experimental result. This method could be used to estimate the fracture toughness of materials by analyzing of the fracture surface.     

Irradiation effects on fracture toughness of four nuclear reactor pressure vessel submerged-arc welds

This study applies statistical analyses to fracture toughness results for four irradiated “current practice” submerged-arc welds and an A533 grade B class 1 plate. Charpy V-notch, tensile, and 25 mm thick compact specimens were irradiated at 288°C to neutron fluences of 0.7 to 2.0 × 10²³ neutrons/m² (>1 MeV). The plate material contained 0.14% Cu and 0.67% Ni. The four submerged-arc welds contained 0.04 to 0.12% Cu and 0.10 to 0.63% Ni. The plate material showed a Charpy V-notch impact transition temperature increase of 68°C, and a Charpy V-notch upper-shelf energy drop of 16%. The four submerged-arc welds showed smaller changes than the plate material did. The fracture toughness results from the 25 mm thick compact specimens showed approximately the same temperature shift as the Charpy V-notch results. The results imply that submerged-arc welds with both low-copper and low-nickel contents can exhibit essentially zero radiation embrittlement and that nickel can contribute to radiation embrittlement even when the copper content is low.     

Fracture toughness of Zr-2.5 wt% Nb pressure tubes

Measurements of fracture toughness of HT Zr-2.5 wt% Nb pressure tubes have been made by studying internally pressurizing (burst) test specimens and small bending test specimens. These tests were conducted from a viewpoint of the effects of hydrogen content, hydride orientation, temperature and crack configuration on the fracture thoughness K_c. Results of the experiments showed that K_c decreased with increasing hydrogen content, but is little affecting by hydrogen content at reactor operating temperature. The value of K_c can be quantitatively evaluated by RHC defined by radial hydride content (RHC) perpendicular to the tensile stress.     

Fracture toughness assessment of ferritic-martensitic steel in liquid lead-bismuth eutectic

The presence of micro-cracks at the surface of a ferritic-martensitic steel is known to favour its embrittlement by liquid metals and thus decrease the mechanical properties of the structural materials. Unfortunately, conventional fracture mechanics methods cannot be applied to tests in liquid metal environment due to the opaque and conducting nature of the LBE. Therefore new methods based on the normalization technique for assessment of plain strain fracture toughness in LBE were examined. This paper discusses the assessment of the plain strain fracture toughness of T91 steel in liquid lead bismuth environment at 473 K, tested at a displacement rate of 0.25 mm min⁻¹ and makes the comparison with results obtained in air at the same temperature and displacement rate. Although there is a decrease of the fracture toughness by 20-30% when tested in LBE, the toughness of the T91 steel remains sufficient under the tested conditions.     

Fracture behavior of austenitic stainless steels irradiated in PWR

Fracture behavior of cold-worked 316 stainless steels irradiated up to 73 dpa in a pressurized water reactor was investigated by impact testing at −196, 30 and 150 °C, and by conventional tensile and slow tensile testing at 30 and 320 °C. In impact tests, brittle IG mode was dominant at −196 °C at doses higher than 11 dpa accompanying significant decrease in absorbed energy. The mixed IG mode, which was characterized by isolated grain facets in ductile dimples, appeared at 30 and 150 °C whereas the fracture occurred macroscopically in a ductile manner. The sensitivity to IG or mixed IG mode was more pronounced for higher dose and lower test temperature. In uniaxial tensile tests, IG mode at a slow strain rate appeared only at 320 °C whereas mixed IG mode appeared at both 30 and 320 °C at a fast strain rate. A compilation of the results and literature data suggested that IG fracture exists in two different conditions, low-temperature high-strain-rate (LTHR) and high-temperature low-strain-rate (HTLR) conditions. These two conditions for IG fracture likely correspond to two different deformation modes, twining and channeling.     

Fracture mechanics evaluations of neutron irradiated Type 321 austenitic steel

The effects of fast reactor irradiation at temperatures of ~ 230° C and ~ 400° C on the fracture toughness and associated strength changes, induced in solution treated Type 321 stainless steel have been characterised using instrumented impact test procedures. The studies cover irradiation exposures in the range 16 to 43 displacements per atom (dpa) and test temperatures of 23–500° C.Irradiation results in significant but not catastrophic reductions in fracture toughness, together with radiation hardening effects. Both the dose and test temperature dependence of the toughness changes are sensitive to irradiation temperature. Thus, whilst maximum toughness loss occurs at or below 16 dpa for both irradiation temperatures, the 400° C-irradiation condition is associated with subsequent saturation of the toughness change, whereas for 230° C-irradiation measurable but low on-going toughness degradation occurs up to 43 dpa. The fracture toughness characteristics correlate with fractographic observations which demonstrate retention of a predominantly ductile fracture mode after irradiation, but with dramatic refinement in the scale of microvoid coalescence associated with TiC precipitates. It is suggested that the fracture mechanism after irradiation is controlled primarily by the irradiation-induced precipitate distribution, and furthermore, that the maintenance of ductile fracture, and hence good toughness, up to high irradiation damage levels is a consequence of inhibition of incipient channel fracture processes by the TiC particles.The application of general yield fracture mechanics to calculate critical defect sizes for unstable fracture in fast reactor wrappers is illustrated. These assessments demonstrate the importance of considering net section yield as an alternative failure criterion in thin section components. Finally, the use of empirical equations as a design philosophy to predict irradiation-induced toughness changes is briefly considered.     

Fracture toughness of welded joints materials for main pipelines at Ignalina NPP

This paper deals with an investigation of mechanical and fracture toughness characteristics of welded joint materials used in Ignalina Nuclear Power Plant (NPP) reactor main circulating circuit (MCC) and steam pipelines. Basic metal of MCC group distributing header (GDH) steel 08Ch18N10T (Du-300), its weld metal welded by manual and automatic arc method using the wire SV-04Ch19N11M3 and electrodes EA-100/10U or EA-100/10T, this joint heat-affected zone metal and base metal of the main steam system—steel 16GS (DU-630) and its weld metal welded by manual arc method using the wire SV-08GS2 and electrodes UONI-13/55 were tested.Mechanical properties of welded joints materials—proportional limit (σ_pl), yield (σ_y) and ultimate (σ_u) strength, fracture stress (σ_f) and ductility (Z) (percent reduction of area) of the specimens were determined. Investigation of relative critical stress intensity factor for fixed thickness of the specimen and critical J-integral, J_IC, was performed.The probabilistic investigation of influence of the mechanical properties (σ_pl, σ_y, σ_u) onto fracture toughness characteristics and J_IC for tested materials by using linear regression model with three independent variables was performed.Research enabled to conclude that proposed multivariable regression model with 80% probability (confidence coefficient α = 0.05) has explained reasonably well the dependence of with σ_pl, σ_y, σ_u and it has shown the non-acceptability of probabilistic evaluation of the model with respect to J_IC.     

Structural characterization of swift heavy ion irradiated polycarbonate

Makrofol-N polycarbonate thin films were irradiated with copper (50 MeV) and nickel (86 MeV) ions. The modified films were analyzed by UV-VIS, FTIR and XRD techniques. The experimental data was used to evaluate the formation of chromophore groups (conjugated system of bonds), degradation cross-section of the special functional groups, the alkyne formation and the amorphization cross-section. The investigation of UV-VIS spectra shows that the formation of chromophore groups is reduced at larger wavelength, however its value increases with the increase of ion fluence. Degradation cross-section for the different chemical groups present in the polycarbonate chains was evaluated from the FTIR data. It was found that there was an increase of degradation cross-section of chemical groups with the increase of electronic energy loss in polycarbonate. The alkyne and alkene groups were found to be induced due to swift heavy ion irradiation in polycarbonate. The radii of the alkyne production of about 2.74 and 2.90 nm were deduced for nickel (86 MeV) and copper (50 MeV) ions respectively. XRD analysis shows the decrease of the main XRD peak intensity. Progressive amorphization process of Makrofol-N with increasing fluence was traced by XRD measurements.     

Assessment of toughness in long term service CrMo low alloy steel by fracture toughness and small punch test

Small punch test (SPT) is a miniature sample test technique which can evaluate in-service material properties with an almost non-destructive method. In this paper, the 2.25Cr1Mo steel samples serviced for 10 years in hydrogenation reactor (with temper embrittlement), 1.25Cr0.5Mo supper-pressure vapor pipe serviced for 14 years at 520 °C and several other low alloy steels have been studied by J_IC fracture toughness and SPT. The linear relationship between the small punch (SP) equivalent fracture strain and the fracture toughness of J_IC was created. The correlations applied to the experimental data indicated advantages of using SPT for the determining fracture toughness of in-serviced low alloy steels. Additionally, size affects the fracture pattern. Small punch samples of small size show dimple fractures whereas large fracture toughness samples show quasi-cleavage fractures.     

Fracture mechanics concepts for evaluation of irradiated reactor pressure vessel material behavior

Within the German research program Forschungsvorhaben Komponentensicherheit (FKS), irradiation experiments were performed with ferritic reactor pressure vessel (RPV) steels and welds. The materials cover a wide range of chemical composition and initial toughness to achieve different susceptibility to neutron irradiation. Different neutron flux was applied and the neutron exposure extended up to 8×10¹⁹ cm⁻². The change in material properties was determined by means of tensile, Charpy impact, drop-weight and fracture mechanics tests, including crack arrest. The results have provided more insight into the acting embrittlement mechanisms and shown that the fracture mechanics concept of the Code provides in general an upper bound for the material which can be applied in the safety analysis of the RPV.     

X-ray characterization of surfaces irradiated with highly charged ions

Highly charged ions (HCI) approaching, touching or penetrating dielectric surfaces extract many electrons of the solid leading to the formation of permanent surface modifications. The ions which capture the electrons in their outermost shells form hollow atoms which emit X-rays during their decay to the ground state. In this paper one presents experiments showing that these X-rays) allow diagnosing the electric nature of the surfaces. HCI while modifying the structure of surfaces may then also be used to diagnose these changes on line or off line.     

Specific low temperature release of Xe from irradiated MOX fuel

A particular low temperature behaviour of the ¹³¹Xe isotope was observed during release studies of fission gases from MOX fuel samples irradiated at 44.5 GWd/tHM. A reproducible release peak, representing 2.7% of the total release of the only ¹³¹Xe, was observed at ∼1000 K, the rest of the release curve being essentially identical for all the other xenon isotopes. The integral isotopic composition of the different xenon isotopes is in very good agreement with the inventory calculated using ORIGEN-2. The presence of this particular release is explained by the relation between the thermal diffusion and decay properties of the various iodine radioisotopes decaying all into xenon.