On the thermal hydraulics of Magnox reactor pressure vessel insulation: Part 2. Numerical simulations of laminar forced and mixed convection in an insulation passage

The paper presents a computational study of laminar flow and heat transfer in a layer of the multiple plate insulation used to line the interior surface of ‘Magnox’ reactor pressure vessels. The flow passage consists of a 2 mm deep channel, which is plane on its upper surface and has raised ‘dimples’ on its lower surface. Under forced convection conditions strong flow ‘streaming’ is found in the unobstructed regions of a channel, and Nusselt numbers there rapidly attain full development. Between adjacent dimples heat transfer levels are characterized by single- and double-peaking distributions which are shown to be due to convective influences. The effect of buoyancy is examined in mixed convection flows with the passage oriented at various angles between horizontal and vertical. Complex non-monotonic trends of heat transfer enhancement and impairment are found, and these are also shown to be attributable to changing patterns of convection.     

Round robin analysis of pressurized thermal shock for reactor pressure vessel

A comparative assessment study is performed here for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). A round robin problem is proposed using the data available in Korea and all organizations interested in the PTS analysis are invited. The problems consisting of two transients and 10 cracks are solved and their results are compared to generate a reference solution that could serve as benchmarks for future qualification of analytical method. Nine participants from seven organizations responded to the problem and their results are compiled in this paper.     

Investigation on constraint effect of reactor pressure vessel under pressurized thermal shock

In recent years, the integrity of reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) accident has been treated as one of the most critical issues. Under PTS condition, the combination of thermal stress due to a steep temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the RPV wall. As a result, cracks on the inner surface of RPVs can experience elastic-plastic behavior that can be explained using the J-integral. In such a case, however, the J-integral may possibly lose its validity due to the constraint effect. The degree of constraint effect is influenced by the loading mode, the crack geometry and the material properties. In this paper, three-dimensional finite element analyses are performed for various surface cracks to investigate the effect of clad thickness and crack geometry on the constraint effect. A total of 36 crack geometries are analyzed and results are presented by the two-parameter characterization based on the J-integral and the Q-stress.     

Combined effect of thermal and irradiation embrittlement in the reactor pressure vessel CrNiMoV steel

Combined effects of segregation and irradiation embrittlement in reactor pressure vessel CrNiMoV steel were studied. The study deals with an analysis of conditions affecting the 15Ch2NMFA type CrNiMoV steel susceptibility and the development of microsegregation processes in connection with temper brittleness formed on repeated annealing cycles. Microstructural analysis and results of tensile and impact testing for all the treatment conditions are presented.     

Structural integrity assessment of the reactor pressure vessel under the external reactor vessel cooling condition

Since the suggestion of external reactor vessel cooling (ERVC), the effects of melting and cooling on the response of structural integrity of the reactor pressure vessel (RPV) under core melting accident conditions have been investigated. To investigate the initial behavior of RPV lower head and the effects of analysis conditions on the structural integrity of RPV, the transient analysis is utilized considering the transient state. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The temperature and stress analyses are performed for core melting accident by using ABAQUS. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. In addition, the results by transient analysis are compared with those by steady state analysis and the effects of analysis conditions on structural integrity are reviewed.     

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.     

Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

A joint pressure vessel integrity research programme involving three partners is being carried out during 1990–1995. The partners are the Central Research Institute of Structural Materials “Prometey” from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material.     

On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ∼7 × 10¹⁹ n cm⁻². This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.     

Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: an atom probe study

Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally-aged (˜100 000 h at 280°C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate and forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress relief heat treatment. The composition of cementite carbides aged for 100 000 h were compared to the Thermocalc™ prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.     

Experimental study on the application of acoustic emission during the hydraulic pressurization of a reactor pressure vessel

The investigations were aimed at demonstrating the state of the art of acoustic emission testing (AET) of reactor pressure vessels. The object under investigation was the large reactor pressure vessel of the MPA in Stuttgart, a boiling-water reactor pressure vessel, which was provided with a multitude of flaws in weld seams and in the base material. Six hydrostatic tests approximately up to the working pressure of a boiling-water reactor (71 bar) were carried out. In addition to the global multichannel locating technique, also local monitoring techniques were applied. Global location permitted a large number of different indications to be detected simultaneously. Not all of the known flaws did, however, show the expected number of AE events. On the other hand, it was possible to detect flaws previously unknown to the AE staff in some weld seams; these indications were confirmed by nondestructive testing. It was demonstrated that the locating accuracy of local monitoring using signal analysis was improved by a factor of 20 to 30 compared to global monitoring.     

Results of probabilistic fracture mechanics analyses for pressurized thermal shock transients in the reactor pressure vessel of BIBLIS-B

Probabilistic fracture mechanics investigations of the contribution of pressurized thermal shock transients to reactor pressure vessel failure probability of the reference plant for the German reactor safety study phase B, BIBLIS-B, are presented. The applied methods and the calculation model are discussed. The most important result of parametric analyses is that the postulated flaw distribution in the vessel has a dominant influence on the calculated conditional failure probability. With regard to the transient behavior the results show, that the temperature drop induced by the thermal shock has great influence on the conditional failure probability, whereas the decay rate of the temperature change has minor influence.     

Corrosion of vessel steel during its interaction with molten corium: Part 1: Experimental

This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheres during an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities and oxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium–specimen ingot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction.     

Reactor pressure vessel design of the high temperature engineering test reactor

The reactor pressure vessel (RPV) of the HTTR is 5.5 m (inside diameter), 13.2 m (inside height), and 122 mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr–1Mo steel is chosen as the material for RPV. The temperature reaches about 400 °C at normal operation. The fluence of the RPV is estimated to be less than 1 × 10¹⁷ n/cm² (E > 1 MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr–1Mo steel for the RPV; embrittlement parameters, J-factor and are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr–1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and , and nil-ductility transition temperatures, T_NDT, by drop weight tests, are shown. In-service inspection and results of R&Ds are also described.     

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.     

Calculation of thermoplastic stresses of a reactor pressure vessel with cladding from stress-relief annealing up to a thermal shock

The distribution of stresses in the cylindrical section of a reactor pressure vessel is evaluated for transient temperature and pressure loads. The calculation takes particularly into account the thermoplastic stresses of the cladding and the transition zone to the base material.The considered load history comprises stress-relief annealing, pressure test, start-up and shut-down of the reactor, followed by start-up, pressure drop with subsequent thermal shock.The stresses and strains are determined by the Prandtl-Reuss equations and the v. Mises flow condition depending on temperature. Creep effects are additionally included for stress-relief annealing.The problem is approximated by a second order time step method and the resulting changes in stress and strain are calculated using a shooting method. The results of the elastoplastic calculation for thermal shock differ from the linear calculation particularly in the cladding layer and its immediate vicinity.     

Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10¹⁹ neutrons per cm², are presented. Within this range, it was observed that increased by ≈500 nV/°C and a linear dependency was noted between and the temperature shift ΔT_41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.     

Effect of cladding on the stress intensity factors in the reactor pressure vessel

In general, reactor pressure vessels (RPV) are cladded with stainless steel to prevent corrosion and radiation embrittlement. The ASME Sec. XI specifies that a subclad crack which may be found during the in-service inspection must be considered as a semi-elliptical surface crack when the thickness of cladding is less than 40% of the crack depth. In order to refine the fracture assessment procedures for such subclad cracks, three-dimensional finite element analyses were applied for various subclad cracks embedded in the base metal. A total of 18 crack geometries were analyzed, and the results were compared with those for idealized semi-elliptical surface cracks for two different loading conditions, i.e. internal pressure and pressurized thermal shock. The resulting stress intensity factors for subclad cracks were 50–70% less than those for idealized surface cracks. It has been proven that the condition specified on the ASME Sec. XI is overly conservative for subclad cracks which are assumed to be surface cracks.     

Three-dimensional transient sealing analysis of the bolted flange connections of reactor pressure vessel

The failure of sealing system of the bolt flange connections is the primary failure mode of the nuclear reactor pressure vessel (RPV). For the safety and integrity of RPV, it is important to predict the sealing behaviour of the bolt flange connections under various loading conditions. Based on the finite element (FE) method for coupled thermal elastoplastic contact problems, a three-dimensional (3D) transient sealing analysis program of nuclear reactor pressure vessels is developed with the consideration of the non-linearity from both surface and material, transient heat transfer and multiple coupled effects. A contact correction approach is proposed to simulate the loading of the bolt connection under the condition of pre-stressing. An automatic pre-processing program is developed for FE modelling of RPVs. Using these programs, a 1:4 scaled model of a 300 MW RPV is analyzed under the loading conditions including pre-stressing, pressurization, heating and cooling. The computational results obtained are in a good agreement with the data of experimental tests. These programs are also successfully used in analyzing the full-scale model of the RPV in a nuclear power plant.