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.     

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.     

Drop testing conducted to benchmark the Shippingport reactor pressure vessel package safety analysis

The decommissioned Shippingport reactor pressure vessel and its integral neutron shield tank were transported from Shippingport, Pennsylvania, via barge to Richland, Washington, for burial in the Hanford Site radioactive waste disposal area. To ensure that the reactor pressure vessel/neutron shield tank assembly could be shipped safely without undue risk to the public or the environment, the reactor pressure vessel/neutron shield tank assembly was certified by the U.S. Department of Energy as a type B package. A safety analysis report for packaging was prepared in accordance with U.S. Department of Energy requirements to provide the technical basis for the U.S. Department of Energy certification. The reactor pressure vessel/neutron shield tank package is a monolithic structure of lightweight concrete and steel. Its estimated weight is 844 t (930 tons). To substantiate multidimensional inelastic analyses, a series of 11 drop tests was conducted on 7 benchmark models from heights of 30.5 cm (1 ft), 9.14 m (30 ft), and 13.7 m (45 ft). Technical evaluation and correlation of the test data were performed in conjunction with the structural analysis and assessment of the package. This paper provides a comprehensive discussion on the benchmark drop models and specific drop tests and also addresses the results obtained from comparing technical data with analytical data.     

Three-dimensional rupture analysis of a prestressed concrete pressure vessel including creep effects

This paper describes briefly the results obtained from a nonlinear analysis up to rupture of a PCRV taking into account creep effects. This analysis aims mainly at evaluating the influence of the redistribution of stresses due to the rheological behavior of concrete on the rupture pressure of a PCRV. First, the method of nonlinear analysis for creep and rupture is described briefly. The mathematical model, of a general application, is based on the finite element method, utilizing the isoparametric elements. Nonlinearities are introduced by the use of iterative techniques. It allows us to predict, within satisfactory limits, the behavior of massive prestressed concrete structures loaded up to destruction. The use of classical parameters for definition of the physical characteristics of materials for formulating the constitutive laws makes the model particularly interesting for practical applications. The analysis of the PCRV for a gas-cooled fast reactor developed by the Swiss Federal Institute for Reactor Research is also presented. This PCRV has large cavities in its walls to house direct cycle gas turbines and other mechanical equipment. First, the creep analysis is carried out taking account of the envisaged construction schedule and the loading during the testing period and finally during the normal exploitation of the PCRV. Proceeding from the state of stress obtained as such the rupture analysis is carried out for a sudden increase of internal pressure and temperature gradient due to a hypothetical accident.     

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.     

Three-dimensional analysis of flow characteristics on the reactor vessel downcomer during the late reflood phase of a postulated LBLOCA

A three-dimensional CFD analysis has been performed on the flow characteristics in the reactor vessel downcomer during the late reflood phase of a postulated large-break loss-of-coolant accident (LBLOCA), in order to validate the modified linear scaling methodology that was applied in the MIDAS test facility of Korea Atomic Energy Research Institute. The vertical and circumferential velocity similarities are numerically tested for the 1/1 and 1/5 linear scale models for the APR1400 reactor vessel downcomer. The effects of scale on flow patterns, pressure and velocity distributions, and the impinging jet behavior are analyzed with the FLUENT code. In addition, a simplified half cylinder model with a single emergency core cooling (ECC) nozzle is numerically tested to investigate the scale effect on the spreading width and break-up of ECC water film. The qualitative and quantitative results indicate that the 1/5 modified linear scale model of the reactor vessel downcomer would reasonably preserve the hydrodynamic similarity with APR1400.     

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.     

Positron affinity for precipitates in reactor pressure vessel steels

The sensitivity of positron annihilation spectroscopy to irradiation-induced precipitates in reactor pressure vessel steels is discussed in the light of recent positron affinity and lifetime calculations. Carbide and nitride precipitates are found to trap positrons only if they contain metal vacancies. Copper precipitates are also attractive to positrons but they are probably detected through annihilation at the precipitate-matrix interface. These findings are related to available experimental data.     

Deformation analysis of a 900 MW reactor pressure vessel head by means of holographic interferometry

During pressure build-up in a 900 MW reactor pressure vessel, the head of the vessel was holographed. It will be shown how a maximum of information can be extracted from the hologram using computer generated interferograms. Based on a trial and error method the deformation assumption for the head is altered until a best correlation is reached between computer generation and experiment.     

An evolution of understanding of reactor pressure vessel steel embrittlement

This paper attempts to summarize the lifetime contributions of Prof. G. Robert Odette to our understanding of the effects of neutron irradiation on reactor pressure vessel steel embrittlement. These contributions span the entire range of phenomena that contribute to embrittlement, from the production and evolution of fine scale features by radiation damage processes, to the effects of this damage microstructure on mechanical properties. They include the development and application of unique and novel experimental tools (from Seebeck Coefficient to Small Angle Neutron Scattering to confocal microscopy and fracture reconstruction), the design and implementation of large multi-variable experimental matrices, the application of multiscale modeling to understand the underlying mechanisms of defect evolution and property change, and the development of predictive methodologies employed to govern reactor operations. The ideas and discoveries have provided guidance worldwide to improving the safety of operating nuclear reactor pressure vessels.     

PISC full scale reactor pressure vessel validation of non-destructive examination

The PISC III Programme involves validation of techniques and procedures and, within this programme, evaluation has now started on the ability to discriminate service induced defects from indications produced by fabrication defects in A 508 Class 2 material when sensitive techniques are used.Action No. 2 of PISC III: Full Scale Vessel Testing is designed for the performance demonstration of three groups of inspection procedures:

• - Mechanized ASME type procedures with variable recording level and complementary techniques

• - Industrial full ISI procedures (mechanized);

• - Several detailed evaluation procedures (generally mechanized) based on advanced techniques to be used on defective areas detected by usual inspection.

These procedures, typical for ISI in most of the cases, are applied in four situations which could be typical of old and new LWR pressure vessels:

• - vessel material and welds containing important service and fabrication defects but mixed with base material defects and small welding defects;

• - nozzle to shell welds with typical service defects, often well isolated and distant from other defective areas in rather clear material and/or welds;

• - nozzle inner radius defects;

• - artificially heat and unbranched fatigue defects in the test blocks assembled to simulate a PWR pressure vessel wall portion.

The paper summarizes the PISC II programme results which stress the characteristics of capable NDT techniques, in opposition to material characteristics like acceptable base material defects. It describes the full scale pressure vessel components available to conduct the PISC III exercise with improved ultrasonic techniques.     

Embrittlement recovery due to annealing of reactor pressure vessel steels

The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.     

The coolability limits of a reactor pressure vessel lower head

Configurations II and III of the ULPU experimental facility are described, and results from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Additionally, with Configuration III, we examine the effect of a channel-like geometry created by the reactor vessel thermal insulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related to the observed two-phase flow regimes.     

Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Refined Monte Carlo analysis of the H.B. Robinson-2 reactor pressure vessel dosimetry benchmark

Refined analysis, based on use of the Monte Carlo code MCNPX-2.4.0, is presented for the “H.B. Robinson-2 pressure vessel dosimetry benchmark”, which is a part of the Radiation Shielding and Dosimetry Experiments Database (SINBAD). First, the performance of the Monte Carlo methodology is reassessed relative to the reported deterministic results obtained with DORT. Thereby, the analysis is accompanied by a quantitative evaluation of the optimal energy cut-off value for each of the in- and ex-vessel dosimeters that were employed. Second, a more realistic definition of the neutron source is implemented than proposed in the benchmark. Thus, the current procedure for power-to-neutron-source-strength conversion, as also for explicitly considering the burnup-dependent fuel assembly-wise average fission neutron spectrum, is found to affect the calculated values significantly.     

The prestressed cast iron reactor pressure vessel (PCIPV)

The concept of the prestressed cast iron reactor pressure vessel (PCIPV) emerges from the utilization of cast iron in the design of radiation and thermal shields. The principles of construction are explained using a model which is at present being assembled. Salient differences between the proposed vessel concept and a prestressed concrete reactor pressure vessel (PCPV) are discussed.     

Suggested toughness/temperature relations for reactor pressure vessel steels

Relations are suggested for the means and standard deviations of three toughness measures for reactor pressure vessel steels: static initiation, dynamic initiation, and arrest. All of the relations are of the form: K_Ix = K_LS{1 + exp[(T − [RT_NDT + δT])/T_O]}, where K_Ix is the toughness measure of interest, K_LS is the lower-shelf toughness, T is the temperature, RT_NDT is the reference transition temperature, δT is a temperature shift, and T_O is a temperature which characterizes the breadth of the transition. The mean of K_LS differs for initiation and arrest and its standard deviation accounts for variation within a single heat. The mean of δT differs for all three toughness measures and its standard deviation accounts for heat-to-heat variability. However, it is shown that a value of T_o = 33.2°C can be used for all of the toughness measures. Finally, the lower bound curves of the ASME Boiler and Pressure Vessel Code are shown to represent toughness levels of low probability.     

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.