Multidimensional two-phase flow regime distribution in a PWR downcomer during an LBLOCA refill phase

The multidimensional countercurrent two-phase flow regimes that occur in a pressurized-water reactor (PWR) vessel downcomer during the refill phase of a large-break loss-of-coolant accident are studied using a transparent 1/10 scale model of a PWR vessel. The various flow regimes and their distribution in the downcomer have been identified and mapped for a range of air-water flooding experiments. The two-phase flow patterns that are identified in the downcomer included various types of film flows, droplet flows, countercurrent churn flows and cocurrent flows depending on the flooding condition. Through observation of the two-phase flow dynamics it was deduced that the physical mechanisms associated with the flooding processes could be separated into a liquid entrainment process and a film flow reversal process. In addition to the above exercise, the effect of non-uniform injection of water into the downcomer via different combinations of cold leg was studied similarly by determining flooding curves and flow pattern maps. It was found that differences in the flooding characteristic were noticeable for various water inlet configurations when compared with the uniform injection case. The differences could be explained qualitatively in terms of the flooding mechanisms identified previously by examining the flow patterns in the downcomer for the non-uniform injection tests.     

Scaling for the ECC bypass phenomena during the LBLOCA reflood phase

As one of the advanced design features of the APR1400 (Advanced Power Reactor), a direct vessel injection (DVI) system is adopted instead of the conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is underway. In this paper, a new scaling method, using the time and velocity reduced “modified linear scaling law”, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in the PWR downcomer.     

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.     

Downcomer boiling phenomena during the reflood phase of a large-break LOCA for the APR1400

Downcomer boiling phenomena in a conventional pressurized water reactor has an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling (DOBO) is being progressed for the reflood phase of a postulated LBLOCA. Test facility was designed as a one side heated rectangular channel which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II(a)) measurement of the local bubble flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and a part of Phase-II(a) were introduced.     

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.     

Load carrying capacity of a reactor vessel head under a corium slug impact from a postulated in-vessel steam explosion

The load carrying capacity of the pressure vessel head to withstand an in-vessel steam explosion is investigated. Firstly, as a key problem, the impact of molten core material against the vessel head is studied by model experiments scaled down 1:10. Structural details are considered carefully. The results are converted to reactor dimensions using similarity theory. This approach was checked by simplified liquid-structure impact experiments in different scale. Secondly, the upward acceleration of molten core material is studied by computational models. As results the mechanical energies which the vessel head can withstand are presented.     

Aseismic study on the reactor vessel of a fast breeder reactor

This paper presents the results of a seismic study using an scale steel model and a scale plastic model which simulate the reactor vessel of a loop type Fast Breeder Reactor (FBR). The main purposes of this study are to confirm the structure/liquid interaction and the aseismic safety of the reactor vessel experimentally, and also to verify the validity of the seismic response analysis model of the prototype vessel.The characteristics of coupled vibration between the structure and liquid were clarified, and the approach of calculation model to aseismic design was worked out. And, the dip plate and other core internals were found to be effective in suppressing the liquid free surface oscillation.     

Hydrodynamic phenomena in the downcomer during flow rate transients in the primary circuit of a PWR

Mixing phenomena observed when the flow rate in a single loop of the primary circuit is changed can influence the operation of pressurized water reactor (PWR) by inducing local gradients of boron concentration or coolant temperature. Analysis of one-dimensional Laser Doppler Anemometry (LDA) measurements during the start-up and shutdown of pump on a single loop of the ROCOM test facility has been performed. The effect of a step change and a ramped change in the flow rate on the axial and azimuthal velocities was examined. Numerical simulations were also performed for the step change in the flow rate that gave quantitative agreement with the axial velocities. Phenomenological agreement was made on the turbulent kinetic energy; however, observed values were a factor of 2.5 less than the turbulent kinetic energy derived from the measurements.     

Debris interactions in reactor vessel lower plena during a severe accident II. Integral analysis

The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 °C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 Vessel Inspection Program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation.     

Static and stability analysis of the inner vessel of a fast breeder reactor

Liquid metal fast breeder nuclear reactors demand the usage of large sized thin shells for their reactor vessel components due to low operating pressure and high thermal load. Buckling is a very important aspect in the design of these vessels. In this article, analysis of the inner vessel of a typical 500 MWe fast breeder reactor is presented. Here, two different geometric configurations of the inner vessel are considered. One configuration is with the conical step joining the upper and the lower cylindrical portions, and the other is with the toroidal bottom joining the upper cylindrical part. The buckling strength of the vessel for both configurations are calculated and compared. Also, the effects of thermal load, initial geometric imperfection, geometric nonlinearity, etc. are investigated. The finite element method is used for analysis.     

Satisfying flow criteria in the reactor vessel of an atomic power plant

Special Design Office for Machinery, Institute of Machine Science. Translated from Atomnaya Énergiya, Vol. 71, No. 3, pp. 214–220, September, 1991.     

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.     

An analysis of containment failure by a steam explosion following a postulated core meltdown in a light water reactor

The Reactor Safety Study (WASH-1400) assessed the probability of containment failure via a steam explosion during a postulated core meltdown accident to be 10⁻². Large uncertainties were attached to this probability and research has continued to reduce the uncertainty.In this paper, we discuss the possible consequences of a steam explosion for a specific reactor system (Zion Nuclear Station—Pressurized Water Reactor). It is our opinion, based on the analysis performed, that generation of large mass missiles by the explosion is unlikely, while small mass missiles, although more likely would not pose a threat to the containment. We do not mean to imply that steam explosions can be disregarded during a postulated meltdown accident, but rather that emphasis should now be placed on how the explosion affects the overall core meltdown accident instead of causing a direct failure.     

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.     

Two-phase natural circulation flow of air and water in a reactor cavity model under an external vessel cooling during a severe accident

As a part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of the Advanced Power Reactor (APR) 1400, a Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale (HERMES-HALF) experiment has been performed by using the non-heating method of an air injection. This large-scale experiment uses a half-height and half-sector model of the APR1400. This experiment has been analyzed to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is very similar to the experimental results of the HERMES-HALF, in general. Increases in the water inlet area and the water level in the reactor cavity lead to an increase in the water circulation mass flow rate. The effects of an air injection mass flow rate and the water outlet area on the water circulation mass flow rate are dependent on the water inlet area size. As the water outlet moves to a lower position, the water circulation mass flow rate increases slowly.     

Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test

As part of the reactor dynamics activities of FZK/IRS, the qualification of a detailed 3D CFD model of a reactor pressure vessel is a key step in safety evaluations for improving predictive capabilities and acceptability of commercial CFD tools in reactor physics. The VVER-1000 Coolant Transient Benchmark, initiated by OECD, represents an excellent opportunity for validation. In this work a CFD model for the complete VVER-1000 reactor pressure vessel is presented. Due to computational limits simplifications of the core and of some other geometrical details are introduced. The simulated scenario is the heat-up of one coolant loop in case of the isolation of a steam generator while the reactor is operating at a low power level. Two transient runs with a first and second order approximation of the spatial discretization are performed. Unexpectedly, the first order method reveals better agreement with measured data.