Downcomer Of Reactor Pressure Vessel Research Articles

Effect of cladding thickness on brittle fracture prevention of the base wall of reactor pressure vessel

In case of a loss of coolant accident, the cooling water is injected into the downcomer of reactor pressure vessel (RPV) and produces a pressurized thermal shock (PTS). Considering that the current cladding thickness is small, thermal shock and neutron irradiation are easy to cause brittle fracture of base material. The transient stress intensity factors of the crack tips at the base-cladding interface are obtained by the interaction integral method. Since the test of the crack propagation on the cladding-base interface is quite difficult and dangerous, the allowable internal pressure values obtained from the K-T curves are verified by XFEM in the thermo-mechanical coupling fields. By increasing the cladding thickness, the resulting temperature fields are compared with each other to demonstrate the different embrittlement effects caused by thermal shock. Then the elastic-plastic fracture mechanics (EPFM) method and the linear elastic interaction integral method are used to study the structural integrity at relatively low and high nil-ductility reference temperatures respectively. Furthermore, the influence of the crack size on the ultimate bearing capacity of the structure is quantitatively analyzed for various thicknesses of cladding and base wall. The size effect of crack decreases obviously with the increase of cladding thickness.

Boron concentration field measurement and mixing analysis in the annular downcomer of reactor pressure vessel

Boron concentration field measurement and mixing analysis in the annular downcomer of reactor pressure vessel

Thermal hydraulic and stress coupling analysis for AP1000 Pressurized Thermal Shock (PTS) study under SBLOCA scenario

Pressurized Thermal Shock (PTS) analysis for AP1000 under Small Break Loss of Coolant Accident (SBLOCA) was performed in this paper. The three-dimensional models of Direct Vessel Injection (DVI) line, Reactor Pressure Vessel (RPV) nozzles and downcomer were established. The mathematic models of three-dimensional thermal hydraulic and stress analysis were introduced. The numerical simulation of thermal-hydraulic mixing was carried out using Computational Fluid Dynamics (CFD) method under SBLOCA scenario. It was found that the temperature distribution in RPV downcomer depended on the injection velocity relative to cross-flow greatly. The RPV cooling center region was moving up with the increase of injection velocity. The temperature distribution was non-uniform along the circumferential orientation on RPV wall and the great temperature gradient was generated between the cooling center and other regions. The AP1000 RPV stress analysis was performed using Finite Element Analysis (FEA) method following thermal-hydraulic mixing study. Results show that the most critical zone was located in the DVI nozzle chamfering under SBLOCA transient. The stress was mainly induced by high temperature gradient and the maximum stress occured when the wall temperature has the largest reduce rate. This work is meaningful for the structure integrity study of AP1000 nuclear power plant.

Unsteady Single-Phase Natural Circulation Flow Mixing Prediction Using CATHARE Three-Dimensional Capabilities

Abstract Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal–hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

Validation of the Large Interface Method of NEPTUNE_CFD 1.0.8 for Pressurized Thermal Shock (PTS) applications

NEPTUNE_CFD is a code based on a 3D transient Eulerian two-fluid model. One of the main application targets is the two-phase Pressurized Thermal Shock (PTS), which is related to PWR Reactor Pressure Vessel (RPV) lifetime safety studies, when sub-cooled water from Emergency Core Cooling (ECC) system is injected into the possibly uncovered cold leg and penetrates into the RPV downcomer. Five experiments were selected for the validation, a selection reviewed by a panel of European experts. The dynamic models are validated with a co-current smooth and wavy air–water stratified flow in a rectangular channel with detailed measurements of turbulence and velocities. The condensation models are validated with a co-current smooth and wavy steam-water stratified flow in a rectangular channel with measurements of the steam flow rates. The dynamic models are validated in the situation of a jet impinging a pool free surface with two experiments dealing with a water jet impingement on a water pool free surface in air environment. Finally, all the models involved in the reactor conditions are validated with the COSI experiment. The calculations are done with the same set of Large Interface Method models and a RANS (k–ɛ) approach for turbulence. They substantiate the application of the tool to PTS studies.

Major findings from LBLOCA reflood tests using the ATLAS facility

A set of LBLOCA (large-break loss of coolant accident) reflood tests was performed in the first phase of the ATLAS (advanced thermal-hydraulic test loop for accident simulation) program. Their main objectives were to identify the major thermal-hydraulic characteristics during the reflood phase of a LBLOCA for APR1400 and to provide qualified data for APR1400 licensing. The ATLAS reflood test program could be divided into two phases (Phase-1 and Phase-2) according to the target period to be simulated. The Phase-1 tests were parametric effect tests for downcomer boiling in the late reflood phase of LBLOCA and the Phase-2 tests were integral effect tests for the entire reflood phase of LBLOCA. The experimental results from both Phase-1 and Phase-2 tests reproduced typical thermal-hydraulic trends expected to occur during the APR1400 LBLOCA scenario. A separate effect test was also performed under a low reflooding rate condition to provide data to validate the RELAP5 reflood models, and its experimental results showed a gradual reflooding in the core, a subsequent quenching of the core heater rods and the cooling of the reactor pressure vessel downcomer.

THERMAL-HYDRAULIC TESTS AND ANALYSES FOR THE APR1400'S DEVELOPMENT AND LICENSING

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

Evolution of Design Methodologies for Next Generation of Reactor Pressure Vessels and Extensive Role of Thermal-Hydraulic Numerical Tools

The thermal-hydraulic design of the first pressurized water reactors was mainly based on an experimental approach, with a large series of tests on the main equipment [control rod guide tubes, reactor pressure vessel (RPV) plenums, etc.] to check performance.Development of computational fluid dynamics codes and computers now allows for complex simulations of hydraulics phenomena. Provided adequate qualification, these numerical tools are an efficient means to determine hydraulics in the given design and to perform sensitivities for optimization of new designs. Experiments always play their role, first for qualification and then for validation at the last stage of the design. The design of the European Pressurized Water Reactor (EPR), jointly developed by Framatome ANP, Electricité de France (EDF), and the German utilities, is based on both hydraulics calculations and experiments handled in a complementary approach.This paper describes the collective effort launched by Framatome ANP and EDF on hydraulics calculations for the RPV of the EPR. It concerns three-dimensional calculations of RPV inlets, including the cold legs, the RPV downcomer and lower plenum, and the RPV upper plenum up to and including the hot legs. It covers normal operating conditions but also accidental conditions such as pressurized thermal shock in a small-break loss-of-coolant accident. Those hydraulics studies have provided much useful information for the mechanical design of RPV internals.

Restart of natural circulation in a PWR–PKL test results and s-relap5 calculations

At reduced primary water inventory in a pressurized water reactor (PWR) with heat transfer to the secondary by means of reflux-condensation (i.e. after a small-break loss-of-coolant-accident), condense with low boron concentration may accumulate in the primary system. The restart of natural circulation after the refilling of the primary can then lead to displacement of low borated water into the reactor pressure vessel and finally into the core, which can result in a recriticality of the reactor. The restart of natural circulation was investigated in the test facility PKL, which is operated by Framatome ANP GmbH (former Siemens/KWU) in Erlangen/Germany. The test results as well as the results of a post-test calculation using the computer code s-relap5, which are both subjects of this presentation, demonstrated that despite symmetrical injection of emergency core coolant (ECC), the restart of natural circulation takes place at different times with varying intensities in the individual loops. In a PWR, this would favor the mixing of water flows with different boron concentrations in the reactor pressure vessel downcomer, which again would reduce the risk of a recriticality transient.

Thermal hydraulics of PWRS with respect to boron dilution phenomena. Experimental results from the test facilities PKL and UPTF

Within the PWR safety analyses, attention has increasingly focused in recent years on boron dilution events which could potentially lead to reactivity transients. In Germany, the current discussion concerning boron dilution events is focused on small break loss of coolant accident (SB-LOCA) scenarios with reflux-condenser mode and restart of natural circulation (NC). The investigation of this topic was the subject of experiments performed in the Primärkreislauf (PKL) test facility and in the upper plenum test facility (UPTF). The two test facilities represent a typical western-type pressurized water reactor (PWR) and are/were operated by Siemens/KWU in Germany. While the restart of NC was investigated in the PKL system test facility (volume 1:145, height 1:1), the UPTF experiments dealt with the mixing of water flows with different boron concentration in the cold legs, reactor pressure vessel (RPV) downcomer and the lower plenum (all these components were full-scale models). The results from the PKL test facility demonstrate that in case of a postulated SB-LOCA (30 cm 2-break in the cold leg; two out of four high pressure safety injection pumps (HP-SIPs) available) NC does not start up simultaneously in all loops. This means, that plugs of non-borated condensate, that might have accumulated in the pump seal during reflux condenser mode of operation, would reach the RPV at different points in time. The UPTF tests showed an almost ideal mixing of water flows with different boron concentration in the RPV downcomer. The most important results from PKL and UPTF tests which concern boron dilution events following the restart of NC are presented in this paper.