Effect of reverse cyclic loading on the fracture resistance curve in C(T) specimen

Fracture resistance (J–R) curves, which are used for elastic–plastic fracture mechanics analyses, are known to be dependent on the cyclic loading history. The objective of this paper is to investigate the effect of reverse cyclic loading on the J–R curves in C(T) specimens. The effect of two parameters was observed on the J–R curves during the reverse cyclic loading. One was the minimum-to-maximum load ratio (R) and the other was the incremental plastic displacement (δ_cycle/δ_i), which is related to the amount of crack growth that occurs in a cycle. Fracture resistance tests on C(T) specimens with varying the load ratio and the incremental plastic displacement were performed, and the test results showed that the J–R curves were decreased with decreasing the load ratio and decreasing the incremental plastic displacement. Direct current potential drop (DCPD) method was used for the detection of crack initiation and crack growth in typical laboratory J–R tests. The values of crack initiation J-integral (J_I) and crack initiation displacement (δ_i) were also obtained by using the DCPD method.     

The fracture toughness measured on sensitized 304 stainless steel in simulated reactor water

The environmental conditions chemically equivalent to BWR primary water, e.g. 288°C, 0.2 ppm O₂ and/or 98°C, air-saturated, were found to influence considerably the in-water fracture toughness values of furnace-sensitized Type 304 stainless steel.Notched compact tension and three point bend specimens sampled from two heats of standard materials (0.06% C) showed significant reduction in dJ/da values reflecting consistently the effects of loading rate, temperature, dissolved oxygen concentration and degree of sensitization. In particular the crack enhancement with lowering the loading rate was significant. The effect became apparent with dJ dt at and below 1× 10−1 kg·mm/mm²/min (1.6 × 10 J/m²/s) in the typical BWR environment.Based on the results, it is suggested that a critical consideration is needed on the significance of such an environmental effect in the LWR structural safety evaluation, in particular that the probability of instable fracture at the “rings” of sensitized material near welded joints is subject to reviewing.     

CROSS SECTIONS AND RATE COEFFICIENTS FOR EXCITATION OF THE 1s2s2pPJ→ 1s2s2pPJ′FINE-STRUCTURE TRANSITIONS IN BERYLLIUM-LIKE IONS BY HEAVY PARTICLE IMPACT

Cross sections for excitation of the 1s²2s2p³P^o_J→ 1s²2s2p³P^o_J′fine-structure transitions in beryllium-like ions by proton, deuteron, triton, and α-particle impact have been calculated using a close-coupled impact-parameter method. This technique includes the effects of dipole coupling to the nearby triplet 2p², 2s3s, and 2s3dconfigurations by means of a polarization potential. We consider the ions C III, N IV, O V, Ne VII, Mg IX, Al X, Si XI, S XIII, Ar XV, Ca XVII, Ti XIX, Cr XXI, Fe XXIII, and Ni XXV. Excitation rate coefficients have also been calculated from the cross sections for a range of temperatures.     

New observations on the crack growth rate of low alloy nuclear grade ferritic steels under constant active load in oxygenated high-temperature water

Within the scope of reactor safety research attempts have been made over several decades to determine corrosion-assisted crack growth rates. National and international investigations have been performed on both an experimental and an analytical basis. A compilation of internationally available experimental data for ferritic steels exhibits a scatter of crack growth rates of up to 5 decades. This was one of the reasons for commencing further experimental investigations focused on the evaluation of corrosion-assisted crack growth rates. These experimental studies were performed under constant, active, external load on 2T-CT specimens of the materials 20 MnMoNi 5 5 with 0.009 and 0.020% S (similar to A508 Cl.3), 22 NiMoCr 3 7 with 0.006% S (similar to A508 Cl.2) and 17 MnMoV 6 4 with 0.017% S. The tests were carried out in deionized oxygenated high-temperature water (240°C; 0.4 and 8.0 ppm O₂). For K_I values up to 60 MPa m^1/2, the experimental results showed no significant dependence between corrosion-assisted crack growth rates and the stress intensity factor, the oxygen content of the medium or the sulphur content of the steel. Here it is important to note, that in this K_I region the high crack growth rates after the onset of cracking due to loading are decreasing and finally come to a standstill after a short period of time as compared with operational times of plants. Consequently, the determination of crack growth velocities as corrosion-assisted crack advance divided by the test duration, so far practised worldwide, results in wrong crack growth rate values in the above-mentioned range of loading up to 60 MPa m^1/2. Based on a test duration of 1000 h, the average crack growth rates are below 10⁻⁸ mm s⁻¹ for K_I ≤ 60 MPa m^1/2. When applied to a single start-up and service period of one year, this would formally lead to an average crack growth rate of 2·10⁻⁹ mm s⁻¹ (equivalent to 0.06 mm per year). At K_I values between 60 and 75 MPa m^1/2 the average corrosion-assisted crack growth rates increase significantly. It can be observed experimentally that the crack propagates during the whole period of the test. Consequently the calculation of crack growth velocities as corrosion-assisted crack advance divided by the test duration as mentioned earlier can be applied as a first estimate. Finally, for K_I values ≥ 75 MPa m^1/2 high crack growth rates up to 10⁻⁴ mm s⁻¹ can be observed. In this region the average crack growth rates are also in quite good agreement with a theoretically based crack growth model.     

Ductile–brittle transition of polycrystalline iron and iron–chromium alloys

Fracture toughness of polycrystalline Fe, Fe–3%Cr and Fe–9%Cr was measured by four-point bending of pre-cracked specimens at temperatures between 77 K and 150 K and strain rates between 4.46 × 10⁻⁴ and 2.23 × 10⁻² s⁻¹. For all materials, fracture behaviour changed with increasing temperature from brittle to ductile at a distinct brittle–ductile transition temperature (T_c), which increased with increasing strain rate. At low strain rates, an Arrhenius relation was found between T_c and strain rate in each material. At high strain rates, T_c was at slightly higher values than those expected from extrapolation of the Arrhenius relation from lower strain rates. This shift of T_c was associated with twinning near the crack tip. For each material, use of an Arrhenius relation for tests at strain rates at which specimens showed twinning gave the same activation energy as for the low strain rate tests. The values of activation energy for the brittle–ductile transition of polycrystalline Fe, Fe–3%Cr and Fe–9%Cr were found to be 0.21, 0.15 and 0.10 eV, respectively, indicating that the activation energy for dislocation glide decreases with increasing chromium concentration in iron.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

Ion irradiation induced nanocrystal formation in amorphous Zr55Cu30Al10Ni5 alloy

Ion irradiation can be used to induce partial crystallization in metallic glasses to improve their surface properties. We investigated the microstructural changes in ribbon Zr₅₅Cu₃₀Al₁₀Ni₅ metallic glass after 1 MeV Cu-ion irradiation at room temperature, to a fluence of 1.0 × 10¹⁶ cm⁻². In contrast to a recent report by others that there was no irradiation induced crystallization in the same alloy [S. Nagata, S. Higashi, B. Tsuchiya, K. Toh, T. Shikama, K. Takahiro, K. Ozaki, K. Kawatusra, S. Yamamoto, A. Inouye, Nucl. Instr. and Meth. B 257 (2007) 420], we have observed nanocrystals in the as-irradiated samples. Two groups of nanocrystals, one with diameters of 5–10 nm and another with diameters of 50–100 nm are observed by using high resolution transmission electron microscopy. Experimentally measured planar spacings (d-values) agree with the expectations for Cu₁₀Zr₇, NiZr₂ and CuZr₂ phases. We further discussed the possibility to form a substitutional intermetallic (Ni_xCu_1−x)Zr₂ phase.     

Effect of irradiation by 140 Mev Ag ions on the optical and electrical properties of polypropylene/TiO2 composite

Changes in the optical, structural, dielectric properties and surface morphology of a polypropylene/TiO₂ composite due to swift heavy ion irradiation were studied by means of UV–visible spectroscopy, X-ray diffraction, impedance gain phase analyzer and atomic force microscopy. Samples were irradiated with 140 MeV Ag¹¹⁺ ions at fluences of 1 × 10¹¹ and 5 × 10¹² ions/cm². UV–visible absorption analysis reveals a decrease in optical direct band gap from 2.62 to 2.42 eV after a fluence of 5 × 10¹² ions/cm². X-ray diffractograms show an increase in crystallinity of the composite due to irradiation. The dielectric constants obey the Universal law given by ε α f ⁿ⁻¹, where n varies from 0.38 to 0.91. The dielectric constant and loss are observed to change significantly due to irradiation. Cole–cole diagrams have shown the frequency dependence of the complex impedance at different fluences. The average surface roughness of the composite decreases upon irradiation.     

Flow structure and bubble characteristics of steam–water two-phase flow in a large-diameter pipe

The flow structure and bubble characteristics of steam–water two-phase upward flow were observed in a vertical pipe 155 mm in inner diameter. Experiments were conducted under volumetric flux conditions of J_G<0.25 m s⁻¹ and J_L<0.6 m s⁻¹, and three different inlet boundary conditions to investigate the developing state of the flow. The radial distributions of flow structure, such as void fraction, bubble chord length and gas velocity, were obtained by horizontally traversing optical dual void probes through the pipe. The spectra of bubble chord length and gas velocity were also obtained to study the characteristics of bubbles in detail. Overall, an empirical database of the multi-dimensional flow structure of two-phase flow in a large-diameter pipe was obtained. The void profiles converged to a so-called core-shaped distribution and the flow reached a quasi-developed state within a relatively short height-to-diameter aspect ratio of about H/D=4 compared to a small-diameter pipe flow. The PDF histogram profiles of bubble chord length and gas velocity could be approximated fairly well by a model function using a gamma distribution and log–normal distribution, respectively. Finally, the correlation of Sauter mean bubble diameter was derived as a function of local void fraction, pressure, surface tension and density. With this correlation, cross sectional averaged bubble diameter was predicted with high accuracy compared to the existing constitutive equation mainly being used in best-estimate codes.     

An evaluation of containment inerting and air dilution systems as methods for post-accident hydrogen control in BWRS

Hydrogen control is important in post-accident situations because of possibilities for containment rupture due to hydrogen deflagration or detonation. Post-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control failure rate (air dilution: 8.3 × 10⁻²−1.1 × 10⁻²; nitrogen inerting: 1.3 × 10⁻²−2 × 10⁻³). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify ‘unidentified’ leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations-related disadvantages.     

Constraint effects on fracture toughness of impact-loaded, precracked Charpy specimens

Impact-loaded, precracked Charpy specimens often play a crucial role in irradiation surveillance programs for nuclear power plants. However, the small specimen size B = W = 10 mm limits the maximum value of cleavage fracture toughness J_c that can be measured under elastic—plastic conditions without loss of crack tip constraint. In this investigation, plane strain impact analyses provide detailed resolution of crack tip fields for impact-loaded specimens. Crack tip stress fields are characterized in terms of J—Q trajectories and the toughness-scaling model which is applicable for a cleavage fracture mechanism. Results of the analyses suggest deformation limits at fracture in the form of b > MJ_c/σ₀, where M approaches 25–30 for a strongly rate-sensitive material at impact velocities of 3–6 m s⁻¹. Based on direct comparison of the static and dynamic J values computed using a domain integral formulation, a new proposal emerges for the transition time, the time after impact at which interial effects diminish sufficiently for simple evaluation of J using the plastic η factor approach.     

Fluence dependence of the surface roughness of InP after bombardment

InP(1 0 0) surfaces were sputtered under ultrahigh vacuum conditions by 5 keV ions at an angle of incidence of 41° to the sample normal. The fluence, , used in this study, varied from 1 × 10¹⁴ to 5 × 10¹⁸ cm⁻². The surface topography was investigated using field emission scanning electron microscopy (FE-SEM) and atomic force microscopy (AFM). At the lower fluences ( 5 × 10¹⁶ cm⁻²) only conelike features appeared, similar in shape as was found for noble gas ion bombardment of InP. At the higher fluences, ripples also appeared on the surface. The bombardment-induced topography was quantified using the rms roughness. This parameter showed a linear relationship with the logarithm of the fluence. A model is presented to explain this relationship. The ripple wavelength was also determined using a Fourier transform method. These measurements as a function of fluence do not agree with the predictions of the Bradley–Harper theory.     

Effect of loading rate upon fracture behavior of ferritic steel under large scaled yielding

Dynamic loading to ferromagnetic materials and large scaled yielding result in peak or valley and non-linear curve, respectively, on the Direct Current Potential Drop (DCPD) versus Crack Opening Displacement (COD) plots, which make it difficult to determine the crack initiation point. In this work high intensity of current up to 100 A was applied to the specimens of SA106Gr.C ferritic steel and the crack growth behavior was directly monitored by a high speed camera to obtain the crack initiation point. The effects of loading rate up to 1200 mm min⁻¹ upon the fracture resistance were explored. As the results, it has been shown that, although no substantial difference was seen in the load–COD plots, the crack initiation and then J_i and J–R curve were quite sensitive to the loading rate. That is, under the loading rate of 300 mm min⁻¹ the material showed the worst fracture resistance than under static loading and even under the higher loading rates of 600 and 1200 mm min⁻¹. Also applying the high speed camera and high current source have been proved to be an effective way to find out the accurate crack initiation point and to compensate the pulse of DCPD due to the ferromagnetic effect.     

Fission product release and its environment impact for normal reactor operations and for relevant accidents

The radioactive concentration in the primary loop and the radioactive release for both normal operations and accidents for the HTR-10 are calculated and presented in the paper. The coated-particle fuel is used in the HTR-10, which has good performance of retaining fission products. Therefore the radioactive concentration in the primary loop of the HTR-10 is very low, and the amount of radioactive release to the environment is also very small for both normal operation and accident conditions. The radiation doses to the public caused by radioactive release for both normal operations and accidents are given in the paper. The results show that the maximum individual effective dose to the public due to the release of airborne radioactivity during normal operations is only 1.4×10⁻⁴ mSv a⁻¹, which is much lower than the dose limit (1 mSv a⁻¹) stipulated by Chinese National Standard GB8703-86. For depressurization accident and water ingress accident, the maximum individual whole-body doses to man are only 7.7×10⁻² and 2.0×10⁻¹ mSv, thyroid doses only 1.7×10⁻¹ and 1.1 mSv, respectively. They are much lower than the prescribed minimum of emergency intervention level (whole-body dose: 5 mSv, thyroid dose: 50 mSv) for sheltering measures stipulated by the Chinese Nuclear Safety Criterion HAD002/03. The conclusion is that the environmental impact is very small for normal operations and accidents for the HTR-10, and the requirements stipulated in the Chinese Nuclear Safety Criterions are satisfied perfectly.     

Intergranular stress corrosion cracking susceptibility of sensitized type 304 steel in 561 K pure water under γ-ray irradiation

Intergranular stress corrosion cracking (IGSCC) of sensitized type 304 stainless steel has been investigated in 561 K water under γ-ray irradiation at a flux of 2.6 × 10³ C kg⁻¹ h⁻¹ by slow-strain-rate tensile tests. The IGSCC susceptibility was enhanced by γ-ray irradiation in water containing 8 ppm dissolved oxygen (DO). The DO dependence of the IGSCC susceptibility was observed in the water under γ-irradiation. Although slight IGSCC susceptibility was observed even in deaerated water (less than 1 ppb DO) under γ-ray irradiation, the susceptibility was completely suppressed by injection of hydrogen into the water. The enhancement of IGSCC susceptibility seems to be related to the formation of H₂O₂ in high temperature water by radiolysis under γ-ray irradiation and the H₂O₂ formation rate is markedly decreased by hydrogen injection.     

HTGR core auxiliary cooling system availability and reliability

Using fault tree techniques, a quantitative estimate is made to predict both the start-up availability and operational reliability of the core auxiliary cooling system (CACS) of an HTGR following the postulated, simultaneous occurrence of a design basis depressurization accident (DBDA) and the complete loss of main loop cooling (LOMLC). The effects of a postulated, concurrent loss of offsite power are considered as well. Several potential common mode failures are identified. The limited availability of data presents a problem to numerical evaluation and estimates of uncertainty are at best crude. To provide a basis for measure of this uncertainty, the fault trees were solved using, on a consistent basis, either ‘optimistic’ failure rates, ‘pessimistic’ failure rates, or mean values (the geometric mean).Generally, about 80% of the failure rate data was larger than the ‘optimistic’ value, while only 20% was larger than the ‘pessimistic’ value. Predicated on a variety of assumptions, many of which may be unduly pessimistic, the CACS unavailability following a postulated DBDA and LOMLC has been estimated to be between 4 × 10⁻⁷ and 3 × 10⁻⁵ for the 2000 MW (th) HTGR and between 5 × 15⁻⁷ and 5 × 10⁻⁵ for the 3000 MW (th) HTGR. At the end of 300 hr, the estimated probability that the CACS will not leave sufficient core cooling capacity varies between 9 × 10⁻⁵ and 4 × 10⁻² for the smaller plant and 3 × 10⁻⁴ and 6 × 10⁻² for the larger plant. If it is further postulated that offsite power is concurrently lost, then the estimated mean unavailability at start-up is 3 × 10⁻³ for the 2000 MW (th) plant. The estimated mean probability that the CACS of the smaller plant will not be available at start-up and not be operational after 300 hr is 8 × 10⁻².     

Vapour void fraction in subcooled flow boiling

Experimental data on steam void fraction and axial temperature distribution in an annular boiling channel for low mass-flux forced and natural circulation flow of water with inlet subcooling have been obtained. The ranges of variables covered are: mass flux = 1.4 × 10⁴−1.0 × 10⁵ kg/hr m²; heat flux = 4.5 × 10³−7.5 × 10⁴ kcal/hr m²; and inlet subcooling = 10–70°C. The present and literature data match well with the theoretical void predictions using a four-step method similar to that suggested by Zuber and co-workers.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

Time dependence of J-integral

Tests performed within the framework of earlier RWTÜV projects together with results obtained elsewhere with regard to the time dependence of fracture mechanics data show that time effects reduced the toughness of materials, according to the nature of the test (extremely slow load rate or hold times with sustained load).

Reduction in toughness has an effect on the following:

&#x02022; - decrease in critical material data (J₀, δ_i)

&#x02022; - levelling off of the crack resistance curve J = J(Δa) and in consequence a decrease of tearing modulus.

This tendency is confirmed quantitatively by recent test results. These tests were performed with the material 15 Mn Ni 63 at room temperature with hold times under sustained load and according to the appropriate standards (without hold times). The tests show that hold times cause additional stable crack growth. The resulting J − Δa curve is lower and less sloping than the curve obtained in a standardized test. The time effect should be taken into account in a safety analysis.     

Wear behavior of graphite studies in an air-conditioned environment

The wear performance of graphite used in the high-temperature gas-cooled reactor (HTR-10) was researched. The wear mechanism, worn surface and wear debris were analyzed under SEM. Under test conditions, the wear rate was 2.27×10⁻⁷ g/m for surface contact, and 1×10⁻⁶ g/m level for line contact. The main wear mechanisms of graphite were groove and fatigue. The projected area of wear debris followed the logarithm normal school, giving most wear debris as a small sphere and large flake debris as only a small part.