Intermediate size pressure vessel programme (ZB 1)

One of the focal points in the discussion about the safety of nuclear power plants is the integrity of the reactor pressure vessel.In order to prove its integrity tests are in progress in an underground test facility of the main power station in Mannheim with an intermediate size vessel from the research programme “Integrity of Components”. Patches of A 533 B and modified A 508 B material were welded into the vessel ZB 1, the test temperatures are approximately 70 and 290°C. The main goal of the tests is to measure the behaviour of artificial and natural flaws during static hydrotests and simulated operational (cyclic) conditions.In the first half of the research programme the objective is to produce a crack growth of some centimetres by cyclic loading between a variable minimum pressure and a maximum pressure of about 24 MPa. The total number of load cycles will be approximately 30 000.In the second half of the tests the vessel will be loaded by a number of pressure cycles which correspond to the loading a reactor pressure vessel experiences during 40 years of operation.During the static and cyclic loading acoustic emission monitoring is being made by German and American laboratories.This paper presents details of the vessel, the test loop, results of the nondestructive examinations conducted to quantify the crack depths and results of the acoustic emission monitoring.     

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

Within the framework of the 6 month WANO program, small samples were cut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime—the ductile-to-brittle transition temperature (DBTT) in the initial state (T_ko) and the phosphorus and copper contents which affect the radiation stability of steel—were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objectives of the study.Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C.The estimated value of T_ko, for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel.The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of T_f obtained from tests before annealing (212 °C) is about 40 °C higher that the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of T_ko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of T_ko = 50 °C.The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.     

Calculation of neutron fluence in the region of the pressure vessel for the history of different reactors by using the Monte-Carlo-method

Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture.     

Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.     

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.     

Radiation Service Life of the Vessel Metal in Operating VVER Reactors

The current status of the prediction of radiation embrittlement of the vessel material in first- and second-generation VVER reactors is analyzed. The radiation service life of the vessel of each type of reactor is determined by factors due to the special features of the operating regime of the reactor and the chemical composition of the vessel metal. A method of monitoring the state of the material of first-generation reactor vessels is examined. The method is based on extracting and studying samples of a metal from the inner surface of the sample. The main problems of monitoring the state of the metal in VVER-440/213 and VVER-1000 vessels are analyzed. It is indicated that adjustments must be made in the normative relations which are currently used for predicting radiation embrittlement of vessel material. The most important questions concerning reactor dosimetry for VVER vessel material are illuminated.__________Translated from Atomnaya Energiya, Vol. 98, No. 6, pp. 460–472, June 2005.     

Experience of irradiation embrittlement in the Loviisa reactor pressure vessels

The water gap between the wall and the core of the RPV (Reactor Pressure Vessel) in a VVER-440 plant is small compared with typical Western type LWR₅. The neutron fluence on the RPV wall is, consequently, much higher in a VVER-440 plant. In older VVER-440 plants the material of the RPV, especially the horizontal core weld, contains so much impurities (P- and Cu-content) that the irradiation embrittlement has become a problem. On bases of fracture mechanics analyses in Loviisa, IVO has been forced to make several measures to ensure safe operation of the plants. According to IVO's current understanding, both plants may be in operation for the design life without annealing of the RPVs.     

The effect of ligaments on the reinitiation of fracture at the tip of a crack arrested during a hypothetical thermal shock event in a water-cooled reactor pressure vessel

During a hypothetical thermal shock event involving a water-cooled nuclear reactor pressure vessel, a crack can propagate deep into the reactor vessel thickness by a series of run-arrest-reinitiation events. Within the transition temperature regime, crack propagation and arrest in pressure vessel steels is associated with a combination of cleavage and dimpled rupture processes, the dimpled rupture regions being contained within ligaments that are normal to the crack plane and parallel to the direction of crack propagation. The present paper models the effect of ligaments on the reinitiation of fracture at the tip of an arrested crack, and the results of a theoretical analysis define the conditions under which ligaments might increase the reinitiation value above $\text{k}{}_{\mathrm{IC}}$, assuming that they fracture by a ductile rupture process. By comparing the predictions with experimental results for model vessels subject to thermal shock, it is shown that the ligaments, which are present at arrest, are unlikely to fail entirely by ductile rupture prior to the reinitiation of fracture at an arrested crack tip. Instead it is suggested that the ligaments fail by cleavage, whereupon they do not markedly affect the reinitiation $\text{K}$ value, which thus correlates with $\text{K}{}_{\mathrm{IC}}$.     

Verification of the VVER-440 pressure vessel neutron fluence calculation by activity measurement

To qualify the calculation methodology, measurements of neutron flux responses of a VVER-440/230 reactor pressure vessel have been carried out.The activity of shavings sampled out from the inner pressure vessel wall of Unit 1 of Kozloduy NPP after the 14th cycle has been measured. Calculation of the expected activity at the shaving positions has been carried out, taking into account the local power distribution. Comparison of calculated and measured activity values has indicated that the computed value for the fast neutron fluence is underestimated by up to 20%.     

Crack arrest in pressure vessel steels in a K-gradient and at low KIa levels

A simple theoretical model is used to examine the effect of the gradient of the crack tip stress intensity K on crack arrest in a nuclear reactor pressure vessel which is subject to a hypothetical thermal transient. Attention is focussed on the case where arrest occurs at the lower end of the transition temperature regime, when crack propagation and arrest are not accompanied by the formation of ductile ligaments. The analysis shows that the arrest K values depend on the gradient of K, and this leads to variability in the arrest values. In particular the arrest K value should be lower when the K gradient is positive than when it is negative; this prediction is reconciled with recent experimental results on crack arrest in model vessel tests.     

The viablility of the KIa procedure for predicting the arrest of a crack in a nuclear reactor steel pressure vessel

The current ASME Code procedure for predicting crack arrest in a nuclear reactor steel pressure vessel is based on a static linear elastic fracture mechanics analysis: a crack is presumed to arrest when the crack tip stress intensity factor K_I^ST falls below K_Ia, which is assumed to be a material property and is referred to as the arrest toughness. The viability of this procedure has been questioned since the theoretical justification, in the strictest sense, for this very simple K_Ia approach is based on the behaviour of a semi-infinite crack propagating in an unbounded solid due to the application of time-independent loads. Against this background, the present paper examines the effects of initial crack size and crack jump length on the viability of the K_Ia procedure. A theoretical analysis shows that the procedure should give accurate predictions of the crack length at arrest certainly if the crack jump length is less than twice the initial crack size.     

Validation of experimental and computational fracture assessment methods for flawed pressure components

Safety and integrity assessments of pressure boundary components require reliable knowledge of the material property values and the validated experimental and computational analysis methods. To improve the accuracy and validity of the experimental and computational fracture assessment methods, a four year Nordic research programme under the auspices of the Nordic Liaison Committee of Atomic Energy was initiated in 1985 and is now under completion. The main technical objective of the programme was to clarify how catastrophic failure can be prevented in pressure vessels and pipings.Experiments with small fracture mechanics specimens and pressure vessels were performed to validate the computational fracture assessment analysis. Two tests were conducted on a decommissioned full-scale chemical reactor pressure vessel from an oil refinery plant, and were extensively instrumented, e.g. by utilizing a 64-channel acoustic emission monitoring system. The scattering of their material property values were determined by numerous fracture mechanics samples. In addition, as a part of the experimental work, the reactor pressure vessel was repaired by welding after the first test. The repair was carried out without postweld heat treatment and welding was done by applying the temper-bead technique. Residual stresses were measured during and after welding.Different fracture assessment methods were developed and subsequently applied to the tested components. Inter-laboratory round robin programmes with the participation of several laboratories were arranged to examine elastic-plastic finite element calculations and fracture mechanics testing.     

Classification and characterization of materials by means of fracture mechanics parameters

Within the scope of the “Integrity of Components” research project a large number of heats of reactor pressure vessel steels representing a broad quality spectrum have been investigated. In this paper the conventional and fracture mechanics parameters of four typical materials are presented. The difficulties in determining the reference temperature for nil ductility transition and the fracture mechanics parameters in the transition and upper shelf region of the Charpy energy are discussed. The technique developed at MPA for the evaluation of a physically meaningful crack initiation parameter based on the size of the stretched zone ahead of the crack tip is described, and values are reported.     

Calculation results from radiation embrittlement of VVER pressure vessel at Kozloduy NPP

A calculational procedure for the evaluation of the transition temperature shift on the basis of neutron fluence has been applied for assessing the reactor pressure vessel (RPV) embrittlement and life time for a VVER-440/230. The calculated results are lower than the passport values, because the real fuel regimes, the low-leakage schemes and loadings with dummy cassettes have been taken into consideration in neutron fluence calculation. The temperature of the outer wall of the RPV has been measured. No significant deviation between the measurement and the data given in the reactor passport has been observed. This shows the correct application of the calculational procedure.     

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.     

Study of tearing behaviour of a PWR reactor pressure vessel lower head under severe accident loadings

In PWR severe accident scenarios, involving a relocation of corium (core melt) into the lower head, the possible failure mode of the reactor pressure vessel (RPV), the failure time, the failure location and the final size of the breach are regarded as key elements, since they play an important part in the ex-vessel phase of the accident.Both the LHF and OLHF experiments as well as the FOREVER experiments revealed that initiation of the failure is typically local. For the case of a uniform temperature distribution in the lower head, crack initiation occurs in the thinnest region and for the case of a non-uniform temperature distribution, it initiates at the highest temperature region. These experimental results can be modelled numerically (but more accurately with 3D finite element codes). The failure time predictions obtained using numerical modelling agree reasonably well with the experimental values.However, the final size of the failure is still an open issue. Analyses of both the LHF and OLHF experimental data (as well as of that from the FOREVER experiments) do not enable an assessment of the final size of the breach (in relation with the testing conditions and results).Indeed, the size of breach depends on the mode of crack propagation which is directly related to the metallurgical characteristics of the RPV steel. Small changes in the initial chemical composition of the vessel material can lead to different types of rupture behaviour at high temperatures. Different rupture behaviours were observed in the LHF and OLHF experiments using the SA533B1 steel. Similar observations were previously noticed during a CEA material characterization programme on the 16MND5 steel. To determine crack propagation and final failure size, 3D modelling would thus be needed with an adequate failure criterion taking into account the variability in behaviour of the RPV material at high temperatures.This paper presents an outline of the methodology being used in a current research programme of IRSN, in partnership with CEA and INSA Lyon. The aim is to model crack opening and crack propagation in French RPV lower head vessels under severe accidents conditions. This programme was initiated in 2003 and is made up of five main sections, namely an inventory of the different French PWR lower head materials, metallurgical investigations to better understand the cause of mechanical behaviour variability that is observed and related to material microstructure, Compact Tension (CT) testing of specimens to characterize the tear resistance of the material, validation of the modelling using experiments on tube specimens and the development of a new failure criterion for the 3D finite element models.     

Cyclic and transient thermal loading of the HDR reactor pressure vessel with respect to crack initiation and crack propagation

In a series of thermal loading tests at the HDR reactor pressure vessel – thermal stratification, cyclic thermal shock and pressurized thermal shock – the methods applied in safety analysis had to become qualified by a continuous intercomparison of calculated results and experimental data. Above all the complex boundary conditions of the HDR-tests offer a close approximation to the original components, so that they provide a real assessment of the transferability.The results of the thermal mixing tests indicated that during cold water inflow into the RPV longitudinal strains build up in the cylindrical wall which dominate over that in circumferential direction.During the cyclic thermal fatigue tests incipient crack formation in the cladding as well as the behaviour of crack propagation in the cladding and in the base material was analyzed.In the pressurized thermal shock tests, the nozzle region and the cylinder wall in the incipient crack condition were loaded by long cooling streaks. Even in the aggravated loading condition as the result of a routed cold water streak no remarkable indications of crack growth were noticed.In both cases, cyclic and pressurized thermal shock loading, the expected crack propagation was overpredicted by the fracture mechanical methods used.The non-destructive examination methods used were able to locate all of the cracks but they mostly overpredicted the actual crack depth.     

Heat flux distribution from a volumetrically heated pool with high Rayleigh number

Experimental results are presented on the heat flux distribution at the boundaries of volumetrically heated pools at high enough Rayleigh numbers to be directly relevant to the problem of retention of a molten corium pool inside the lower head of a reactor pressure vessel. The experimental facility, named COPO, is a 2-dimensional “slice”, Joule-heated and geometrically similar in shape (torispherical at 1/2-scale) to the lower head of a VVER-440 reactor. The results show that: the heat flux on the side wall (vertical portion) is essentially uniform; the downward heat flux strongly depends on position along the curved wall; and average fluxes on the side in the downward direction are in agreement with existing correlations, but somewhat underestimated in the upward direction. For the shape considered, the heat flux along the lower curved wall seems to be independent of the presence and extent of the liquid pool (contained by the vertical sidewalls) portion above it.