Reactor Pressure Vessel Lower Head Research Articles

LIVE Experiments on Melt Behavior in the Reactor Pressure Vessel Lower Head

The main objective of the LIVE program at Karlsruhe Institute of Technology is to study the core melt phenomena both experimentally in large-scale three-dimensional (3D) geometry and in supporting separate-effects tests, and analytically using computational fluid dynamics (CFD) codes in order to provide a reasonable estimation of the remaining uncertainty band under the aspect of safety assessment. Within the LIVE experimental program several tests have been performed with water and with noneutectic and eutectic melts (mixtures of KNO3 and NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used to assess the correlations derived for the molten pool behavior.

LIVE-L4 and LIVE-L5L Experiments on Melt Pool and Crust Behavior in Lower Head of Reactor Pressure Vessel

The LIVE-L4 and LIVE-L5L experiments investigated the thermal-hydraulic behavior of the corium pool in the reactor pressure vessel lower head with the three-dimensional test vessel LIVE. The simulant material is a noneutectic binary mixture of 20% NaNO3-80% KNO3. Transient and steady-state parameters such as melt temperature and heat flux distribution through the vessel wall as well as crust formation characteristics were obtained. The two tests demonstrated that transient events like repeated melt relocation and change of decay power density facilitate crust deformation and change of crust thickness. Massive crust formation in a noneutectic melt pool leads to a change of melt pool composition and a decrease of melt-crust interface temperature. The melt temperature and heat flux at the same pool height and same power density can be roughly compared independent of heating history and initial melt pouring pattern. The dimensionless melt temperature as well as the dimensionless heat flux through the wall during the steady state are independent of power density if the pools have the same height. But, they are dependent on the pool height. For a low pool, the gradients with height of both melt temperature and heat flux through the vessel are larger than those for a high pool.

Characterization of PWR vessel steel tearing under severe accident condition temperatures

Abstract In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead to different types of rupture behaviour at high temperatures. The experimental programme has been conducted on various French RPV 16MND5 steels for temperatures ranging from 900 °C to 1100 °C. Comparisons between the tests performed on these various 16MND5 steels show that this approach is appropriate to characterize the difference in ductility observed at high temperatures. The aim of this experimental study is also to contribute to the definition of a tearing criterion by identifying, on the basis of CT results, the related material parameters at temperatures representative of the real severe accident conditions. This experimental campaign has been carried out in partnership with IRSN in the framework of a research programme whose purpose is to complete the mechanical properties database of 16MND5 steel and to model tearing failure in French RPV lower head vessels under severe conditions ( Koundy et al., 2008 ).

Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed ...

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part I: Physical processes, modeling and model implementation

This paper, and its companion paper [Tran C.T., Dinh, T.N. The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application. Progress in Nuclear Energy (companion paper), in preparation] document the development, validation and applications of a simulation platform for computationally-effective, sufficiently-accurate numerical predictions of core melt-structure-water interactions in the light water reactor lower head during a postulated severe core-melting accident. The centerpiece of this work is the Effective Convectivity Model (ECM) for description of energy splitting in a core melt pool. Built on the concept of characteristic velocities in Effective Convectivity Conductivity Model and supported by the key findings obtained from Computational Fluid Dynamics (CFD) simulations of turbulent natural convection, heat transfer and phase changes in volumetrically heated liquid pools, the ECM is refined and extended to three-dimensions and phase changes to enable simulations of melt pool formation and corium coolability in complex geometry such as a Boiling Water Reactor (BWR) lower plenum.

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application

The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper ( Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.

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.

Simplified modeling of a PWR reactor pressure vessel lower head failure in the case of a severe accident

In order to characterize the timing, mode and size of a possible lower head failure (LHF) of the reactor pressure vessel (RPV) in the event of a core meltdown accident, several large-scale LHF experiments were performed under the USNRC/SNL LHF program. The experiments examined lower head failure at high pressures (10 MPa in most cases) and with small throughwall temperature differentials. Another USNRC/SNL LHF program, called the OLHF program, has been undertaken in the framework of an OECD project. This was an extension of the first program and dealt with low and moderate pressures (2–5 MPa) but with large through wall temperature differentials. These experiments should lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all external-vessel events. The large quantity of escaping corium may lead to direct heating of the containment. This is an important severe accident issue because of its potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, numerical modeling was performed to simulate these experiments. This paper presents a detailed description of three of our numerical models used for the simulation. The first model is a simplified semi-analytical approach based on the theory of a spherical shell subjected to internal pressure. The two other methods deal with 2D finite element (2D-FE) modeling: one combines the Norton–Bailey creep law with a damage model proposed by Lemaitre–Chaboche while the other uses only a creep failure criterion but takes into account thermo-metallurgical phase transformations. One of the objectives of the simulation is the development of a simplified and valid analytical code that can be implemented in integral severe accident computer codes. The numerical results are consistent with the experimental measurements. The effect on the numerical results of the multiphase transformation of the shell material and of the two failure criteria used, one involving necking (Considère's criterion) and the other involving creep damage (Lemaitre–Chaboche) is discussed.

Neutron Radiography Flow Visualization of Liquid Metal Injected into an Empty Vessel and a Vessel Containing Saturated Water

Neutron radiography was used to observe the behavior of molten lead-bismuth alloy injected into a thin (10-mm) semicircular vessel, which was empty or contained saturated water. The fluid velocity distribution for the melt injected into the empty vessel was successfully measured using particle image velocimetry (PIV). The numerical analysis for this case using the FLOW-3D code revealed that the wall heat transfer attenuates the rotational speed of swirls that were formed during the melt injection. The melt-coolant interaction experiment demonstrated several features of molten fuel and coolant interactions in the reactor pressure vessel lower head. The violent vapor expansion occurred at the initial melt impact on the saturated pool water. A one-dimensional model predicted the observed behavior well by assuming the adiabatic expansion immediately after an instantaneous heat transfer during the first melt coolant contact. A crust was formed between the melt and coolant by the heterogeneous distribution of the coolant and pool melt. The convective velocity distribution in the pool melt was measured well by the PIV.

Steam explosion resistance of an internal core catcher of a nuclear reactor pressure vessel

This paper deals with a feasibility design of a core catcher structure that could mitigate the consequences of the core meltdown in a commercial light water reactor. This activity has been performed in the framework of the Concerted Action IVCRS (In Vessel Core Retention Strategies), financed by the European Union in order to assess the safety of the next generation of nuclear power plants. These plants will be designed to support the consequences deriving from a core meltdown accident and from the phenomena that accompany it. Steam explosion is considered as a potential risk in the hypothesis of a severe accident occurring in a Pressurised Water Reactor nuclear power plant. The loss of coolant, which can occur in the case of a pipe break, provokes the degradation of the core geometry, its coolability and then its melting. The molten core falls down in the vessel to the lower hemispherical part rapidly transferring its energy to the water remaining in the lower plenum, which vaporises. In order to mitigate the consequences of a severe accident, a core catcher device, able to contain and to cool the molten core, has been designed. The same structure is analysed here as energy dissipators to prevent the reactor pressure vessel lower head failing in the case of steam explosion. In this paper, emphasis is placed on the structural aspects of the problem. Thermal fluid dynamics is treated macroscopically. This conservative approach allows us to overcome the actual uncertainties in the heat transfers mechanism between molten core and water. The results of a simulation, conducted with a finite element code, shows that the implementation of an internal core catcher could prevent the Reactor Pressure Vessel lower head from failing.

Optimum Design for External Reinforcement to Mitigate Deteioration of a Nuclear Reactor Lower Head under Temperature Elevation

This paper is concerned with the optimum design for external reinforcement of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severs accident. The reis of the temperature and the internal pressure cause deterioration of the load carrying capacity and could cause failure of the RPV lower head. The deterioration of failure can be mitigated by the external cooling or the reinforcement of the lower head with additional structures. While the external cooling forces the temperature of an RPV to drop to the desired level, the reinforcement of the lower head can attain both the increase of the load carrying capacity and the temperature drop. The reinforcement of the lower head can be optimized to have the maximum effect on the collapse pressure and the temperature at the inner wall. Optimization results are compared to both the result without the reinforcement and the result with the designated reinforcement.

Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of ‘overpressure-molten pool structure’ when the vessel failure started at the ‘hot’ layers of the vessel. It was shown in this study that the processes in the molten pools reach a quasistationary state at 2000…3000 s after molten pool formation. Numerical results in this paper illustrate that the large creep deformations of the vessel lower head can lead to an appearance of the gaps between the vessel surface and the molten pool crust. It is obvious that the joint thermal and structural analyses are needed for the accurate tracing of the initial bounds of the vessel and molten pool during simulations.

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.

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.

Critical heat flux through curved, downward facing, thick walls

Experimental data are presented that provide a lower envelope on the critical heat flux distribution over the external surface of a hemisphere submerged in water. The experiment was designed and run so as to provide an essentially full-scale simulation of a reactor pressure vessel lower head. Thus the data, and the correlation derived from them, can be applied directly, and they are supportive of the important, severe accident management idea of retaining the core debris in the reactor vessel by external flooding (“in-vessel melt retention”).

Cooling of Core Debris Within the Reactor Vessel Lower Head

Under severe accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. Some accident situations could result in the transport of molten core debris to the lower plenum, as occurred to some extent ([approximately]20 tonnes) during the TMI-2 accident, boiloff of water in the lower plenum, and an inability to add water to the reactor coolant system (RCS). In this extreme set of circumstances, sufficient external reactor pressure vessel (RPV) cooling may be available to prevent failure of the RPV lower head and, thereby, retain the core debris within the vessel. Containment configurations like Zion would result in substantial accumulation of water around the lower parts of the reactor vessel for most accident sequences. The experiments which were performed in support of the Commonwealth Edison individual plant examination and accident management programs, are heat transfer tests designed to demonstrate that nucleate boiling is the dominant heat removal process from the outer surface of a simulated RPV lower head surrounded by typical reflectivemore » insulation used in nuclear power plants.« less

Thermal and stress analyses of the reactor pressure vessel lower head of the three mile island unit 2

Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head.