The COVASTOL program

The COVASTOL program has been developed jointly by CEA, Euratom and Framatome. The resulting code is based on fracture mechanics with propagation expressed in a probabilistic form. The spatial and size distributions for the defects have been obtained from 3 European manufacturers, a total of 338 meters of LWR pressure vessel welds being analysed. The final sizing takes into account the NDE reliability and efficiency. Loading of the vessels has been determined on the basis of statistical data; resulting stresses and stress intensity factors are computed at different times and at different depth. Paris law coefficients are also determined probabilistically together with toughness for which irradiation embrittlement is considered taking into account P and Cu concentrations. Calculations performed using the COVASTOL code and French PWR specifications show that the main rupture risk is due to a LOCA (large or small break). The rupture probability at 40 years varies from 2×10⁻⁸ to 3×10⁻⁹ depending on the safety injection water temperature (respectively 10°C and 20°C). The most dangerous defects are located within the first millimeters of the internal layer and the most sensitive weld is located at the belt line.     

Assessment of irradiation response of WWER-440 welds using samples taken from Novovoronezh unit 3 and 4 reactor pressure vessels

The results of the study on Novovoronezh unit 3 and 4 (NV NPP-3 and 4) reactor pressure vessel (RPV) radiation embrittlement measured using subsize impact specimens (5×5×27.5 mm³) fabricated from samples taken from the corresponding RPV walls are presented. The post-irradiation annealing effectiveness and the embrittlement kinetics of Novovoronezh unit 3 and 4 RPV welds under re-irradiation are discussed. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy (TT^10×10) and subsize impact (TT^5×5) specimens of trepans cut out from Novovoronezh unit 2 RPV are compared. A new relation between TT^10×10 and TT^5×5 has been developed.     

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.     

Socket weld integrity in nuclear piping under fatigue loading condition

The purpose of this paper is to evaluate the integrity of socket weld in nuclear piping under the fatigue loading. The integrity of socket weld is regarded as a safety concern in nuclear power plants because many failures have been world-widely reported in the socket weld. Recently, socket weld failures in the chemical and volume control system (CVCS) and the primary sampling system (PSS) were reported in Korean nuclear power plants. The root causes of the socket weld failures were known as the fatigue due to the pressure and/or temperature loading transients and the vibration during the plant operation. The ASME boiler and pressure vessel (B & PV) Code Sec. III requires 1/16 in. gap between the pipe and fitting in the socket weld with the weld leg size of 1.09 × t₁, where t₁ is the pipe wall thickness. Many failure cases, however, showed that the gap requirement was not satisfied. In addition, industry has demanded the reduction of weld leg size from 1.09 × t₁ to 0.75 × t₁. In this paper, the socket weld integrity under the fatigue loading was evaluated using three-dimensional finite element analysis considering the requirements in the ASME Code. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P = 0 to 15.51 MPa, and the thermal transient ranging from T = 25 to 288 °C were considered. The results are as follows; (1) the socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME operation and maintenance (OM) code. (2) The effect of pressure or temperature transient load on socket weld in CVCS and PSS is not significant owing to the low frequency of transient during plant operation. (3) ‘No gap’ is very risky to the socket weld integrity for the systems having the vibration condition to exceed the requirement specified in the ASME OM Code and/or the transient loading condition from P = 0 and T = 25 °C to P = 15.51 MPa and T = 288 °C. (4) The reduction of the weld leg size from 1.09 × t₁ to 0.75 × t₁ may induce detrimental effect on the socket weld integrity.     

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.     

Influence of the depth position on the neutron embrittlement of the VVER reactor pressure vessel steel 15Cr2MoV(A) consequences for the assessment of reactor safety

The dependence of the mechanical properties on the depth position in the unirradiated state and after irradiation up to neutron fluences of approximately 5 × 10¹⁸ and 70 × 10¹⁸ cm⁻² (E > 0.5 MeV) is tested on a forging made out of VVER 440 reactor pressure vessel (RPV) steel 15CrMoV. The near-surface position shows a higher strength and a lower transition temperature than the positions greater than 1/4 wall thickness. Irradiation does not change these differences in a significant manner. The testing of specimens from the 1/4 depth position within the surveillance programme, as normally laid down in the legal rules relating to nuclear power plants, results in a conservative safety assessment against brittle failure up to the EOL fluence. On taking into account fluence attenuation, the flux effect, etc., the toughness gradually increases from the inside to the outside of the wall after longer RPV operating times.     

Neutron embrittlement surveillance of the garigliano reactor vessel steel

The effects of neutron radiation on the pressure vessel of the Garigliano Nuclear Power Station have been analyzed on the basis of results of a reactor vessel material surveillance program of the plant. A high radiation embrittlement sensitivity was determined for the weld metal and for the A336 forging steel of the ring forging course just above the level of the fuel core. Both showed high copper and phosphorus contents, which accounted for the embrittlement sensitivity. The ring forging opposite the fuel core had a low copper and phosphorus content and revealed relatively low embrittlement. A neutron fluence of 6.3 × 10¹⁹n/cm² > 1 MeV was determined for the peak flux plane for 40yr of operation. However, the 40yr fluence for the ring forging at the top of the core level (3.5 × 10¹⁹n/cm² > 1 MeV) resulted in the highest final transition temperature because of the sensitivity of this steel. The measured Charpy-V shelf energy absorption values were plotted against yield stress values for comparable irradiations on the ratio analysis diagram (RAD). The analysis revealed that the pressure vessel steel properties would continue to degrade toward a condition of possible frangibility at the end of its life. This projection is based on an assumption of uniform embrittlement throughout the vessel wall thickness. Such uniformity does not exist; in fact, a sharp gradient exists in the steel such that ductility rises rapidly in the steel toward the outside wall as well as above and below the fuel core. Hence, because of this strong ductility gradient, the Garigliano reactor vessel should be able to operate safely over its intended design lifetime.     

核电厂一回路管座焊缝超声波检测仿真研究

小尺寸支管接头（BOSS）焊缝作为核电厂一回路压力边界的薄弱环节，对其有效监控是核电厂日常和在役大修的重点和难点。采用仿真技术、工艺试验和现场应用验证等方法，设计并验证了BOSS焊缝的超声波相控阵检测工艺，解决了核电厂日常和在役大修中BOSS焊缝的监督难点。并得到类似超声波相控阵检测工艺的设计和验证方法。      

Preservice and inservice requirements for containment structures in the United States - A status report

Assuring the lifetime integrity of containment structures for nuclear power plants is becoming increasingly important as existing design criteria are reexamined, as new requirements for containment inspection and testing are formulated, and as today's operating nuclear plants are growing older.The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code contains requirements for the design and construction and for the preservice and inservice requirements for nuclear power plant systems and components in the United States. Section III of the ASME Code contains the rules for design and construction of nuclear systems and components. Rules for the preservice examination, inservice inspection, system pressure testing, repair, modification and replacement of nuclear systems and components are contained in Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Compliance with the rules of the ASME Code in the United States is mandated by the federal government in Title 10, Part 50 of the Code of Federal Regulations (10CFR50).Section XI of the ASME Code contains separate rules for metal (Class MC) and concrete (Class CC) containments. Requirements for Class MC containments have been published in Subsection IWE, Requirements for Class MC Components of Light-Water Cooled Power Plants, of Section XI. Rules for Class CC containments are currently being developed and will be published in Subsection IWL, Requirements for Class CC Components of Light-Water Cooled Power Plants, of Section XI.First published in 1981, Subsection IWE has been adopted by a number of state jurisdictions in the United States and is presently being reviewed by the United States Nuclear Regulatory Commission. Federal regulations that will require mandatory compliance by nuclear plant owners are forthcoming. When implemented, the requirements in Subsection IWE and Subsection IWL will provide a reasonable and systematic basis for assuring the integrity of metal and concrete containment structures during their service lifetime.This paper presents an overview of the preservice and inservice requirements for containment structures in Section XI of the ASME Code with consideration of the practical factors that should accompany user compliance.     

Fracture toughness of narrow-gap welded joints in the nuclear pressure vessel steel 22 NiMoCr 37

The mechanical testing of narrow-gap welded joints in 100 and 200 mm thick sections of the steel 22 NiMoCr 37 has revealed that the weld metal, and not the heat affected zone (HAZ) or the weld metal-parent metal boundary. is the critical region. This modified gas-shielded welding process operates with a very low heat input of the order of 6.500 J cm⁻¹ pass⁻¹ and the combination of small diameter welding wires and high welding speeds contributes to the excellent joint properties in the as-welded condition.To investigate the effect of preheating and post-welding heat treatment on the mechanical properties of narrow-gap welds, tensile, notch impact, flat bend and fracture toughness test specimens were extracted from joints welded with the following conditions: (1) no preheating: no post-weld heat treatment; (2) no preheating: soaking at 300°C: (3) no preheating: stress-relief heat treatment at 600°C; (4) preheating 200–250°C; no post-weld heat treatment; (5) preheating 200–250°C; soaking at 300°C; (6) preheating 200–250°C; stress relief heat treatment at 600°C. Tensile testing at room temperature and at 250°C of round specimens oriented across the seam revealed the ultimate fracture to be always located in the base material remote from the welded zone. Although pores or slag inclusions had an influence on bend-test results of specimens in the as-welded condition, the results generally show failure free bends to 180°C with no evidence of cracking in the HAZ or at the fusion boundary.Using sharp-notched impact bend specimens with the notch located in the centre of the seam as well as in and across the HAZ, absorbed energy-test temperature curves have been determined for each welding condition. In comparison with the base material impact toughness, the weld exhibits superior toughness in the temperature range − 60 – 0°C, but yielded lower values at room temperature. After stress relieving at 600°C, the impact toughness of the weld reduced significantly, apparently due to precipitations occurring in the weld-metal microstructure. Test results from welded specimens with the no notch in the HAZ show this region to have superior notch impact toughness to the base material.Crack opening displacement (COD) specimens 45 × 90 × 380 mm with the fatigue crack located in the weld metal and in the HAZ were tested at 0 and 20°C using both the recommendation in BS DD 19: 1972 as well as acoustic emission measurements for the determination of COD values. For this method of fracture toughness testing it has been shown that the occurrence of a critical event must be clearly defined as corresponding to stable crack growth or alternatively to unstable crack propagation.     

Quantification of the reliability required of non-destructive inspection of PWR pressure vessels

This paper presents a method of quantifying the reliability required of non-destructive inspections of PWR pressure vessels. It gives a strategy for improving the effectiveness of ultrasonic non-destructive testing in assuring the integrity of a PWR vessel and allows targets of inspection reliability to be set in order to achieve the requisite level of vessel integrity. To do this the failure rate of PWR pressure vessels is predicted on the basis of a probabilistic fracture mechanics model. We use various models of the reliability of non-destructive inspection to discover the minimum level of reliability which is consistent with the desired integrity of the structure and to demonstrate how improvements can be made most effective.The reliability of inspection is usually modelled by a function giving the probability of leaving an unacceptable defect in the vessel. This function B(a) is really the “unreliability” of inspection and so 1 - B(a) gives the usual reliability. A reliable inspection is one which detects and correctly classifies defects according to some criterion usually based on size. A reliable inspection must use a technique which is intrinsically capable of detecting and sizing defects in the required size range and it must be reliably applied in practice.We find that, based on certain stated assumptions, that an inspection reliability of 80% of detecting and correctly sizing a defect of 15 mm through-wall extent yields a predicted failure rate of 10⁻⁷ per vessel year. The failure rate includes a frequency of a major accident such as a large loss of coolant (LOCA) of frequency 10⁻⁴ per vessel year. The predicted failure rate can be reduced to 10⁻⁸ per vessel year if the sizing accuracy of the technique is improved so that the chance of undersizing a 15 mm defect falls from 0.19 to about 0.01. However, the failure rate of the vessel is not predicted to decrease further with any subsequent improvement in sizing accuracy unless there is also an improvement in the asymptote of the reliability of inspection. This asymptote is due to factors beyond the capability of the technique such as, for example, human error.     

Shaking table tests on failure characteristics of base isolation system for a DFBR plant

The building of a demonstration fast breeder reactor (DFBR) plant in Japan is planed to be base isolated in the horizontal direction. To verify the seismic safety of the isolation system, a series of shaking table tests was conducted using a reduced scale model with three types of base isolation system, natural rubber bearing with steel damper (NRB+SD), lead rubber bearing (LRB), and high damping rubber bearing (HRB). The results of these tests showed NRB+SD, LRB and HRB were within the stable domain (not hardening) at 1×S2 (maximum acceleration 3.80 m s⁻²) input, and were nearly hardening at 2×S2 input. None of the rubber bearings broke at 3×S2 input, which was the design limit. All these bearings broke at over 4×S2 input.     

Results on research of templates from Kozloduy-1 reactor pressure vessel

The studies on the specimens manufactured from the templates cut out from the weld 4 of Kozloduy NPP Unit 1 reactor vessel have been conducted. The data on chemical composition of the weld metal have been obtained. Neutron fluence, mechanical properties, ductile to brittle transition temperature (DBTT) using mini Charpy samples have been determined. The phosphorus and copper content averaged over all templates is 0.046 and 0.1 wt.%, respectively. The fluence amounted up to 5×10¹⁸ n cm⁻² within 15–18 fuel cycles, and about 5×10¹⁹ n cm⁻² for the whole period of operation. These values agree well with calculated data. DBTT was determined after irradiation (T_k) to evaluate the vessel metal state at the present moment, then after heat treatment at the temperature of 475°C to simulate the vessel metal state after thermal annealing (T_an), and after heat treatment at 560°C to simulate the metal state in the initial state (T_k0). As a result of the tests the following values were obtained: T_k, +91.5°C; T_an, +63°C; and T_k0, 54°C. The values of T_k and T_an obtained by measurements were found to be considerably lower than those predicted in accordance with the conservative method accepted in Russia (177°C for T_k and 100°C for T_an). Thus, the obtained results allowed to make a conclusion that it is not necessary to anneal Kozloduy NPP Unit 1 reactor vessel for the second time. The fractographic and electron-microscopic research allowed to draw some conclusions on the embrittlement mechanism.     

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10⁻⁸ year⁻¹ which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10⁻⁸ year⁻¹, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.     

Development of an Ultrasonic Performance Demonstration System for Dissimilar Metal Welds in Korean Nuclear Power Plants

To work towards better reliability of ultrasonic testing systems, the Korean Ministry of Science and Technology (MOST) published Bulletin 2004-13 in June 2004. This was replaced by the Ministry of Education and Science Technology (MEST) Bulletin 2009-37 published in September 2009, which directed that all nuclear power plants in Korea shall implement performance demonstration of dissimilar metal welds. Meeting the MEST bulletin requirements and increasing the reliability of ultrasonic testing require the development of a dissimilar metal weld performance demonstration system for ultrasonic testing. The present paper describes the development of a performance demonstration system for dissimilar metal weld ultrasonic testing for nuclear power plants, which meets the requirements of ASME Code Section XI, Appendix VIII and MEST Bulletin 2009-37 and details some of the results obtained for the system.     

Probability-based design criteria for shear walls in nuclear power plants

The development of probability-based criteria for the design of reinforced concrete shear walls subjected to dead load, live load and in-plane earthquake forces in nuclear plants is described. These criteria are determined for flexure and shear limit states in a load and resistance factor design (LRFD) format. The flexure limit state is defined according to traditional principles of ultimate strength analysis, while the shear limit state is established from experimental results. Resistance factors for shear and flexure, load factors for dead and live load, and a load factor for effect of Safe-Shutdown Earthquake are determined for target limit state probabilities of 1.0 × 10⁻⁶ and 1.0 × 10⁻⁵ over a a period of 40 years. Comparisons among the proposed design criteria, ACI 349 and US NRC Standard Review Plan 3.8.4 are included.     

Reduced seismic loads for temporary conditions

A new methodology, developed under an EPRI Tailored Collaboration Project, to calculate and apply reduced seismic loads (RLSs) for evaluation of temporary conditions (TCs) in nuclear power plants using design-basis (DB) allowables is described. The methodology, which was submitted to Nuclear Regulatory Commission (NRC) through the Nuclear Energy Institute (NEI), calculates load reduction factors based on an allowed limit for time-averaged increase in seismic core damage frequency within the duration of a refueling cycle. For this allowable in the range 5×10⁻⁶ to 1×10⁻⁵ per reactor year, substantial reduction relative to DB seismic load is possible. The methodology is equally applicable to plants with and without seismic probabilistic risk analysis model.     

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.     

JR curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents in oxygen-containing high temperature water at 240°C

J_R curves of the low alloy steel 20 MnMoNi 5 5 with two different sulphur contents (0.003 and 0.011 wt.%) were determined at 240°C in oxygen-containing high temperature water as well as in air. The tests were performed by the single-specimen unloading compliance technique at load line displacement rates from 1 × 10⁻⁴ down to 1 × 10⁻⁶ mm s⁻¹ on 20% side-grooved 2T CT specimens in an autoclave testing facility at an oxygen content of 8 ppm and a pressure of 7 MPa under quasi-stagnant flow conditions.In the case of testing in high temperature water, remarkably lower J_R curves than in air at the same load line displacement rate (1 × 10⁻⁴ mm s⁻¹) were obtained. A decrease in the load line displacement rate as well as an increase in the sulphur content of the steel caused a reduction of the J_R curves. At the fastest load line displacement rate a stretch zone could be detected fractographically on the specimens tested in air and in high temperature water and consequently J_i could be determined. When testing in high temperature water, the J_i value of the higher sulphur material type decreases from 45 N mm⁻¹ in air to 3 N mm⁻¹, much more than that of the optimized material type from 51 N mm⁻¹ in air to 20 N mm⁻¹ at 1 × 10⁻⁴ mm s⁻¹.     

The properties of WWER-440 type reactor pressure vessel steels cut out from operated units

This paper presents the results of study on radiation degradation occurring in WWER-440 reactor pressure vessel (RPV) steel, using subsize impact specimens (5×5×27.5 mm³). The results of testing trepans and templates cut out from WWER-440 reactor pressure vessels are considered. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy and subsize impact specimens are compared. The relation between these two values is established.