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.     

Study on mo steel for high temperature gas-cooled reactor material characterization of forged low Si mo steel

The Japan Atomic Energy Research Institute (JAERI) has carried out a series of research and development work related to the high temperature gas-cooled reactor (HTGR) and, accordingly the high temperature engineering test reactor (HTTR) will be constructed in the near future. As the reactor pressure vessel (RPV) material, Mo steel will be used. Material characterization tests have been carried out to evaluate the applicability of the Mo steel for the RPV and to prepare for the licensing. The present paper summarizes the fracture toughness behavior including K_Id and K_Ia, irradiation embrittlement susceptibility and degradation of steel due to the long term aging at high temperature of the forged low Mo steel. These tests reveal good fracture toughness which well meets the requirements of the ASME Code, low neutron irradiation embrittlement susceptibility, little embrittlement by long term aging and so on. The present test results demonstrate good applicability of forged low Mo steel to the RPV of HTGR.     

Brittle fracture analysis of the Kozloduy Unit 1 RPV

The safety of the RPV of the Bulgarian NPP Kozloduy Unit 1 was analysed within EC-financed contracts according to a pressurized-thermal-shock- (PTS-) procedure applied in Germany (Erve, M., Hertlein, R., 1991. Post SMiRT Seminar No 11, August 1991), considering the most relevant transients and taking into account the actual embrittlement in the core weldment. The paper reports on the main aspects of the PTS-procedure, determining the acceptable transition temperature (T_K^a-evaluation) to exclude brittle fracture, and compares the main results with the fluence related transition temperature (T_K^F) of the material got from sampling from the weldment concerned. Testing of the toughness properties by small size Charpy-V-notch specimens revealed only a small irradiation effect in comparison to the properties after the recovery annealing performed in 1989. This could be explained by the fact that only small values of Cu-content in the weld metal were confirmed, thus balancing the expected influence of the relatively high P-content. The main conclusion is: assuming a defect size of 10×60 mm, the evaluation shows, for KNPP 1 after the 18th cycle for the screening transient, a sufficient margin in the T_K^a-value to the actual material properties and—from the technical point of view—thus, recovery annealing is not necessary for the time being. Further embrittlement of the RPV will be covered by an additional surveillance program with samples accelerated re-irradiated in a Russian NPP. Proper operator actions during PTS events can further improve the situation with respect to loading of the RPV during transients, thus increasing the safety margins.     

Dynamic fracture toughness test and master curve method analysis of IAEA JRQ material

IAEA conducted a round-robin fracture test program to test and verify the Master Curve method. One of the materials selected for the round-robin is a A-533B1 plate designated as reference material JRQ. Unnotched Charpy-size specimens were fabricated and distributed to a number of testing laboratories. The three US Owners Groups received specimens for both Charpy impact and three-point bending tests to establish fracture toughness master curves. The B&W Owners Group elected to perform a dynamic fracture toughness test under a high loading rate using the JRQ specimens. The master curve method was successfully applied to numerous fracture toughness data sets of pressure vessel steels. Joyce [Small Specimens Test Technique, ASTM STP 1329, 1997, ASTM] applied this method to high loading rate fracture toughness data for A-515 steel and showed the applicability of this approach to dynamic fracture toughness data. This paper presents the data and the resulting reference temperature shift in the Master Curves from static to dynamic fracture data. Based on earlier PTS analyses performed in 1985, an appropriate T₀ shift value is selected for nuclear power plant applications. This shift in T₀ is compared with the temperature shift between K_Ic and K_Ia curves in ASME Boiler and Pressure Vessel Code.     

Measurement of cross sections for (n, 2n), (n, p) and (n, α) reactions on germanium isotopes induced by 14 MeV neutrons

Activation cross sections at the neutron energy about 14 MeV on germanium isotopes have been measured, employing the activation technique and γ-ray spectrometry. The data of the cross section are reported for the (n, 2n), (n, p) and (n, α) reactions. The neutron flux was determined using the monitor reactions ²⁷Al (n, α) ²⁴Na and the neutron energies were measured by the method of cross section ratios for ⁹⁰Zr (n, 2n) ⁸⁹Zr to ⁹³Nb (n, 2n) ^92mNb reactions. The measured results were compared with the other measurements.     

Evaluation of ductile–brittle transition temperature before and after neutron irradiation for RPV steels using small punch tests

Small punch (SP) tests were performed to evaluate the ductile–brittle transition temperature before and after a neutron irradiation of reactor pressure vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the conventional Charpy tests and the Master Curve fracture toughness tests in accordance with the American Society for Testing and Materials (ASTM) standard E1921. Small punch specimens were taken from a 1/4t location of the vessel thickness and machined into a 10 mm × 10 mm × 0.5 mm dimension. The specimens were irradiated in the research reactors at Korea Atomic Energy Research Institute Nuclear Research Institute in the Czech Republic at the different fluence levels of about 290 °C. Small punch tests were performed in the temperature range of RT to −196 °C using a 2.4 mm diameter ball. For the materials before and after irradiation, the small punch transition temperatures (T_SP), which are determined at the middle of the upper small punch energies, showed a linear correlation with the Charpy index temperature, T_41 J. T_SP from the irradiated samples was increased with the fluence levels and was well within the deviation range of the unirradiated data. However, the transition temperature shift from the Charpy test (ΔT_41 J) shows a better correlation with the transition temperature shift (ΔT_SP(E)) when a specific small punch energy level rather than the middle energy level of the small punch curve is used to determine the transition temperature. T_SP also had a correlation with the reference temperature (T₀) from the Master Curve method using a pre-cracked Charpy V-notched (PCVN) specimen.     

Use of the Master Curve methodology for real three dimensional cracks

At VTT, development work has been in progress for 15 years to develop and validate testing and analysis methods applicable for fracture resistance determination from small material samples. The VTT approach is a holistic approach by which to determine static, dynamic and crack arrest fracture toughness properties either directly or by correlations from small material samples. The development work has evolved a testing standard for fracture toughness testing in the transition region. The standard, known as the Master Curve standard is in a way “first of a kind”, since it includes guidelines on how to properly treat the test data for use in structural integrity assessment. No standard, so far, has done this. The standard is based on the VTT approach, but presently, the VTT approach goes beyond the standard. Key components in the standard are statistical expressions for describing the data scatter, and for predicting a specimens size (crack front length) effect and an expression (Master Curve) for the fracture toughness temperature dependence. The standard and the approach, it is based upon, can be considered to represent the state of the art of small specimen fracture toughness characterization. Normally, the Master Curve parameters are determined using test specimens with “straight” crack fronts and comparatively uniform stress state along the crack front. This enables the use of a single K_I value and single constraint value to describe the whole specimen. For a real crack in a structure, this is usually not the case. Normally, both K_I and constraint vary along the crack front and in the case of a thermal shock, even the temperature will vary along the crack front. A proper means of applying the Master Curve methodology for such cases is presented here.     

Irradiation embrittlement characterization of the EUROFER 97 material

The paper summarizes original results of irradiation embrittlement study of EUROFER 97 material that has been proposed as one candidate of structural materials for future fusion energy systems and GEN IV.Test specimens were manufactured from base metal as well as from weld metal and tested in initial unirradiated condition and also after neutron irradiation.Irradiation embrittlement was characterized by testing of toughness properties at transition temperature region - static fracture toughness and dynamic fracture toughness properties, all in sub-size three-point bend specimens (27 × 4 × 3 mm³). Testing and evaluation was performed in accordance with ASTM and ESIS standards, fracture toughness K_JC and K_Jd data were also evaluated with the “Master curve” approach. Moreover, J-R dependencies were determined and analyzed.The paper compares unirradiated and irradiated properties as well as changes in transition temperature shifts of these material parameters. Discussion about the correlation between static and dynamic properties is also given.Results from irradiation of EUROFER 97 show that this steel - base metal as well as weld metal - is suitable as a structural material for reactor pressure vessels of innovative nuclear systems - fusion energy systems and GEN IV. Transition temperature shifts after neutron irradiation by 2.5 dpa dose show a good agreement in the case of EUROFER 97 base material for both static and dynamic fracture toughness tests. From the results it can be concluded that there is a low sensitivity of weld metal to neutron irradiation embrittlement in comparison with EUROFER 97 base metal.     

The Euratom 5th Framework Programme Project FRAME (fracture mechanics based embrittlement); description of the project and first results

The objectives, the experimental research programme and the first results of the FRAME project are described in the paper. Altogether 26 different materials were irradiated in a single irradiation capsule and the Master Curve based T₀ shifts were measured for each material. In addition four materials were studied in the programme extension. Altogether approximately 700 fracture toughness specimens were tested in the project. The aim of FRAME was to start systematic development of Master Curve based embrittlement monitoring. Fracture toughness test is clearly superior to Charpy-V test but for historical reasons large databases, on which embrittlement trend curves can be based, exist only for the Charpy-V test. The test data validation procedure used in the project is described with some detail. No bias was noticed between the three main testing partners in the created data. First estimate for a model, which describes the dependence of embrittlement on the copper, phosphorus and nickel contents of the material, is given.     

Fracture assessment of a brittle RPV under cold loading

The oldest Swedish reactor is a boiling water reactor (BWR) with a vessel made of A302 Grade B with rather high Cu and Ni content. These elements have intensified the irradiation embrittlement in the beltline region so that RT_NDT of certain welds may exceed 100 °C at the end-of-life condition. A preliminary study of the fracture risk for the beltline region showed that the limiting loading case would be the cold over-pressurization of the reactor. The objective of this study was to develop a reliable methodology for fracture assessment of the aged reactor vessel under cold loading scenarios. The test program covered experiments on standard SEN(B) specimens and clad beams under uniaxial and biaxial loading. The test material was a reactor vessel steel prepared with a special heat treatment to simulate fracture toughness properties of the aged reactor. No significant effects of shallow crack and biaxial loading were observed on cleavage fracture toughness in different clad specimens. While the ASME K_Ic reference curve was shown to be overly conservative, the Master Curve methodology satisfactorily predicted the experimental outcomes of the test program. The Master Curve methodology indicated that a 20-mm deep surface crack was acceptable in the beltline region under a cold over-pressurization scenario. This value was three times greater than what a methodology based on the ASME K_Ic reference curve yielded.     

Isolation of Radioniobium from γ-lrradiated Molybdenum Oxide

A reverse-phase chromatographic isolation procedure for Nb from Mo and Tc is given. The separation is performed with TBP (tri-n-butyl phosphate) on celite column.

Molybdenum oxide was irradiated with 20MeV bremsstrahlung, which produces a mixture of ^93mMo, ⁹⁹Mo; ^91mTc; ^91mNb, ^95mNb, ⁹⁵Nb and ⁹⁶Nb. The irradiation was applied to a 6 N HC1 solution of the target. Separation was repeated twice, which resulted in isolation of the radioniobium with a decontamination factor of 10⁴ both for Mo and for Tc.

Decay analysis of the activity showed that radioniobium thus separated was composed of ^91mNb, ^95mNb, ⁹⁵Nb and ⁹⁶Nb. The γ-spectrum of ⁹⁶Nb was obtained by subtraction of the spectrum taken 119 hr after the end of irradiation from that of 74 hr.     

Activation cross sections for Os(n,p)Re, Os(n,p)Re and Os(n,n′)Os reactions from 13.5 to 14.8 MeV

Cross sections for (n,p) and (n,n′) reactions have been measured on osmium isotopes at the neutron energies from 13.5 to 14.8 MeV using the activation technique in combination with high-resolution gamma-ray spectroscopy. Neutrons were produced via the ³H(d,n)⁴He reaction using solid TiT. The neutron fluences were determined using the monitor reaction ⁹³Nb(n,2n)^92mNb. Data are reported for the following reactions: ¹⁹⁰Os(n,p)^190mRe, ¹⁹⁰Os(n,p)^190gRe, ¹⁹⁰Os(n,p)¹⁹⁰Re, ¹⁸⁸Os(n,p)¹⁸⁸Re and ¹⁹⁰Os(n,n′)^190mOs. Nuclear model calculations using the code HFTT, which employs the Hauser-Feshbach (statistical model) and exciton model (precompound effects) formalisms, were undertaken to describe the formation of the products. The cross sections were discussed and compared with experimental data found in the literature, with values of model calculations including the pre-equilibrium contribution, and with evaluation data of JEFF-3.1/A.     

Measurements of the Y(n,γ)Y cross-section in the neutron energy range of 13.5-14.6 MeV

The ⁸⁹Y(n,γ)^90mY cross-section has been measured at three neutron energy points between 13.5 and 14.6 MeV using the activation technique and a coaxial HPGe γ-ray detector. The data for the ⁸⁹Y(n,γ)^90mY cross-sections are reported to be 0.39 ± 0.02, 0.43 ± 0.02, and 0.38 ± 0.02 mb at 13.5 ± 0.2, 14.1 ± 0.1, and 14.6 ± 0.2 MeV incident neutron energies, respectively. The first data for the ⁸⁹Y(n,γ)^90mY reaction at neutron energy points of 13.5 and 14.1 MeV are presented. The natural high-purity Y₂O₃ powder was used as target material. The fast neutrons were produced by the T(d,n)⁴He reaction. Neutron energies were determined by the method of making cross-section ratios of ⁹⁰Zr(n,2n)^89m+gZr and ⁹³Nb(n,2n)^92mNb reactions, and the neutron fluencies were determined using the monitor reaction ⁹³Nb(n,2n)^92mNb. The results obtained are compared with existing data.     

Measurement of the activation cross section for the (p,xn) reactions in niobium with potential applications as monitor reactions

Excitation functions of the ⁹³Nb(p,n)^93mMo, ⁹³Nb(p,pn)^92mNb and ⁹³Nb(p,αn)⁸⁹Zr nuclear reactions were measured up to 17.4 MeV by the conventional activation method using the stacked-foil technique. Stacks were irradiated at different incident energies on the TR19/9 cyclotron at the Edmonton PET Centre. The potential of the measured excitation functions for use as monitor reactions was evaluated and tested by measuring activity ratios at a different facility. Single Nb foils were irradiated at incident energies in the range from 12 to 19 MeV on the TR19/9 cyclotron at Brookhaven National Laboratory. Results are compared with the published data and with theoretical values as determined by the nuclear reaction model code EMPIRE.     

Evaluation Methods of Resonance Absorption for System with Pellet Surrounded by Fuel Solution

Abstract

The reaction cross sections of ²⁷Al(n, p)²⁷Mg, ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn, ⁹⁰Zr(n, 2n)^89m+gZr and ⁹³Nb(n, 2n)^92mNb have been measured by the activation method in an energy range of 13.3–14.9 MeV using the intense D-T neutron source, FNS. Absolute flux was determined by the associated α-particle counting method incorporated with neutron spectra obtained from both a Monte Carlo calculation and a time-of-flight measurement. Corrections were extensively performed not only for the neutron flux determination, but also for the low energy neutron contribution to the reaction rates. The present data were compared with comprehensive evaluations as well as recent experimental data. The measured cross sections of ²⁷Al(n, a)²⁴Na, ⁵⁶Fe(n, p)⁵⁶Mn and ⁹⁰Zr(n, 2n)^89m+gZr are generally in good agreement within experimental errors with the values in both the JENDL Dosimetry File and IRDF-90. It is also shown that there are the overestimation of the cross sections for ⁹³Nb(n, 2n)^92mNb in the JENDL Dosimetry File, and the over- estimation and underestimation of the cross section for ²⁷Al(n, p)²⁷Mg in the JENDL Dosimetry File and IRDF-90, respectively.     

Pressure Recovery in Diffuser for Discharge Flows from Gas Centrifuges Rotating at High Mach Numbers

This paper deals with the irradiation effects on the thermal shock resistance Δ = σ_tk/Eα (σ_t : tensile strength, k: thermal conductivity, E: Young's modulus, α: thermal expansivity) and the thermal shock fracture toughness ?=K_Ick/Eα (K_Ic : fracture toughness value of mode I) in addition to other mechanical properties such as the diametral compressive strength and fracture toughness of two kinds of near isotropic graphite neutron irradiated at 750～1,000°C to a dose of (1.1 ～ 1.5)10²¹ n/cm² (>29 fJ). One of the graphite specimen is an isostatically molded graphite IG-11 with fine grain petroleum coke and the other is a binderless molded graphite HCB-18 which was prepared using mesophase pitch carbon with very fine grain size. These measurements are carried out by means of disk testing method developed by us. Results show that both the thermal shock resistance and the thermal shock fracture toughness of the two kinds of graphite after irradiation decrease considerably in contrast with increasing trends of the usual mechanical strengths.     

Ratio tests of dosimetry cross-sections for Nb(n,2n)Nb and Au(n,2n)Au to Al(n,α)Na reactions

Cross-section ratios of reaction ⁹³Nb(n,2n)^92mNb and ¹⁹⁷Au(n,2n)¹⁹⁶Au to the standard reaction ²⁷Al(n,α)²⁴Na have been measured in order to test the recently compiled dosimetry files: JENDL Dosimetry File and International Reactor Dosimetry File 1990 (IRDF-90). The experimental results for both reactions were consistent with the calculated ones based on IRDF-90 except for the ⁹³Nb(n,2n)^92mNb above 19 MeV.Both reactions, especially ⁹³Nb(n,2n)^92mNb, were recommended as new neutron monitors above 12 MeV because of their favorable characteristics from the decay-property and cross-section viewpoints.     

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.     

Inferring the temperature dependence of Beremin cleavage model parameters from the Master Curve

The temperature dependence of Beremin model parameters in the ductile-to-brittle transition region was addressed by employing the Master Curve. Monte Carlo simulation was performed to produce a large number of 1T fracture toughness data randomly drawn from the scatter band at a temperature of interest and thus to determine Beremin model parameters. In terms of the experimental data of a C-Mn steel (the 16MnR steel in China), results revealed that the Weibull modulus, m, decreases with temperature over the lower transition range and remains a constant in the lower-to-mid transition region. The Weibull scale parameter, σ_u, increases with temperature over the temperature range of investigated. A small sample may lead to a considerable uncertainty in estimates of the Weibull stress parameters. However, no significant difference was observed for the average of Weibull stress parameters from different sample sizes.