E. Altstadt

Coupled thermal structural analysis of LWR vessel creep failure experiments

Considering the hypothetical core melt down scenario for a light water reactor (LWR) the failure mode of the reactor pressure vessel (RPV) has to be investigated to determine the loadings on the containment. The failure of reactor vessel retention (FOREVER)-experiments, currently underway, are simulating the thermal and pressure loadings on the lower head for a melt pool with internal heat sources. Due to the multi-axial creep deformation of the vessel with a non-uniform temperature field these experiments are an excellent source of data for validation of numerical creep models. Therefore, a finite element (FE) model has been developed based on a commercial multi-purpose code. Using the computational fluid dynamics (CFD) module the temperature field within the vessel wall is evaluated. The transient structural mechanical calculations are performed using a new numerical approach, which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a three-dimensional array is developed where the creep strain rate is evaluated according to the values of the actual total strain, temperature and equivalent stress. Care has to be exercised performing post-test calculations particularly in the comparisons of the measured data and the numerical results. Considering the experiment FOREVER-C2, for example, the recorded creep process appears to be tertiary, if a constant temperature field is assumed. But, small temperature increase during the creep deformation stage could also explain the observed creep behavior. Such considerations provide insight and better predictive capability for the vessel creep behavior during prototypic severe accident scenarios.

Simulation of creep tests with French or German RPV-steel and investigation of a RPV-support against failure

Abstract Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be considered for a determination of the loadings on the containment. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed in which the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties is performed. For the consideration of the tertiary creep stage and for the evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in three levels: starting with the simulation of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called “tube-failure-experiments” are modeled: the RUPTHER-14 and the “MPA-Meppen”-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi5-5 RPV-steels, which are chemically nearly identical. Since these two steels show a similar behavior, it should be allowed to a limited extend to transfer experimental and numerical data from one to the other. After analyzing the FOREVER calculations, it seems to be advantageous to introduce a vessel support which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it is possible to install an absolutely passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.

Finite-element based vibration analysis of WWER-440 type reactors

Abstract A finite-element model describing the mechanical vibrations of the whole WWER-440 primary circuit was established to support the early detection of mechanical component faults. A special fluid–structure module was developed to consider the reaction forces of the fluid in the downcomer upon the moving core barrel and the reactor pressure vessel. This fluid–structure interaction (FSI) module is based on an approximated analytical 2D-solution of the coupled system of 3D fluid equations and the structural equations of motions. By means of the vibration model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated. It is shown that the FSI strongly influences those modes that lead to a relative displacement between reactor pressure vessel and core barrel. Moreover, by means of the model the shift of eigenfrequencies due to the degradation or to the failure of internal clamping and spring elements was investigated. Comparing the frequency spectra of the normal and the faulty structure, it could be proved that a recognition of such degradations and failures even inside the reactor pressure vessel is possible by pure excore vibration measurements.

Component vibration of VVER-reactors — diagnostics and modelling

Abstract Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have lead to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostics, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies.

Investigation on primary side oriented accident management measures in a hypothetical station blackout scenario for a VVER-1000 pressurized water reactor

Abstract As a consequence of a total loss of AC power supply (station blackout) at a VVER-1000 leading to unavailability of major active safety systems, the safety criteria ensuring the safe operation of the nuclear power plant would be violated and core heat-up with possible core degradation could occur. A dedicated accident management measure (primary side depressurization) can be applied to reduce the primary pressure and to activate the injection from the passive emergency core cooling systems (accumulators). The analyses presented in this paper are aiming at both a detailed investigation of the accident sequence, taking into account the depressurization of the primary circuit, and the possibilities to prevent or at least to mitigate a damage of the reactor core so as to gain additional time for taking necessary countermeasures. The analyses are performed using the codes ASTEC and ATHLET developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH).

Development and validation of the pressure surge computer code DYVRO mod. 3

Abstract Pressure fluctuations are generated in pipeline systems, when a fluid is rapidly accelerated or decelerated by fast closing or opening a valve, pump trip or start, breaking of pipes etc. TUEV NORD SysTec GmbH & Co. KG performs calculations of pressure surges in power plants and especially in nuclear power plants for many years. For this reason TUEV NORD has developed and qualified the pressure surge computer code DYVRO. This contribution describes in detail the actual code version DYVRO mod. 3 that has been modified with regard to the system of partial differential equations and the numerical scheme. Also the validation against representative experiments like the Simpsons experiment and an experiment at the Cold Water Hammer Test Facility CWHTF is shown. Results of DYVRO calculations are afterwards compared with results of own and external simulations with system codes (ATHLET and RELAP5) and with the pressure surge code WAHA. For this purpose the outcome of the EU project WAHALOADS that was supported within the 5th framework programme has been used.

Vibration analysis of the pressure vessel internals of WWER-1000 type reactors with consideration of fluid–structure interaction

Abstract The influence of fluid–structure interaction (FSI) on vibration modes is investigated using the finite element method. The method of modelling is verified by comparing the finite element results with the exact analytical solution of a simple fluid–shell test system. It is shown that the method of coupling between structural and fluid elements is important for the accuracy of the eigenfrequencies. The vibration modes of the reactor pressure vessel and its internals of a WWER-1000 type reactor are calculated. The FSI causes a considerable down shift of the shell mode frequencies of pressure vessel, core barrel and thermal shield. Some bending modes which exhibit a relative displacement between pressure vessel and core barrel or between core barrel and thermal shield are significantly affected too. Some simple analytical approaches to consider the FSI are discussed on the basis of the numerical results.

RPV Long Term Operation: Open Issues

This paper presents and describes key open issues which are being debated nowadays by experts in the field, and for which clarification is essential for a safe operation of the nuclear power plants during life extension. Notably: late blooming effects in low Cu steels; effects of Cu, Ni, Mn, and P on the irradiated microstructure and on hardening and embrittlement; use of material test reactor data for assessment in power reactors (including flux and spectrum effects); Master Curve versus Unified Curve and fracture toughness behavior of highly irradiated material; embrittlement in RPV zones out of the traditional beltline (“the expanding beltline”); embrittlement trend curves at high neutron fluence, where data are scarce; re-embrittlement after annealing.

Study on severe accidents and countermeasures for VVER-1000 reactors using the integral code ASTEC

Abstract The research field focussing on the investigations and the analyses of postulated severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises of the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. This paper is focused on the possibilities for accident management measures in case of severe accidents. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accidents, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operators actions and countermeasures in the preventive or mitigative domain are addressed.

Irradiation Damage and Embrittlement in RPV Steels Under the Aspect of Long Term Operation: Overview of the FP7 Project LONGLIFE

The increasing age of the European NPPs and envisaged lifetime extensions up to 80 years require an improved understanding of RPV irradiation embrittlement effects connected with long term operation (LTO). Phenomena which might become important at high neutron fluences (such as late blooming effects and flux effects) must be considered adequately in the safety assessments. Therefore the project LONGLIFE was initiated within the 7th Framework Programme of the European Commission. The project aims at: i) improved knowledge on LTO phenomena relevant for European reactors; ii) assessment of prediction tools, codes, standards and surveillance guidelines. In the paper, we give an overview of the project structure and the related tasks. Furthermore we present two examples for the experimental evidence of LTO relevant phenomena: the first example is related to the flux dependence of defect cluster formation in a neutron irradiated weld material. We have found that the size of the irradiation induced defects exhibits a flux effect whereas the mechanical properties are almost independent of the flux. The second example refers to the acceleration of irradiation hardening after exceeding a threshold fluence. This effect was observed by means of both small angle neutron scattering (SANS) and tensile testing for low Cu RPV steels irradiated at a temperature of 255 °C. These examples demonstrate that LTO irradiation effects have to be investigated in more detail to guarantee the applicability of the embrittlement surveillance guidelines beyond 40 years of operation.Copyright