Comparison of Plant Response of Forsmark 3 Nuclear Power Plant to Forsmark 1 Type of Transient (25th July 2006), Using the HAMBO BWR Simulator: Thermal-Hydraulic Aspects

Modernisation of control rooms of the nuclear power plants has been a major issue during the last few years. With this as a basis, the BWR plants in Sweden and Finland funded, in co-operation with the Halden Project, an experimental HAMBO BWR simulator project based on the Forsmark 3 plant in Sweden. VTT Energy in Finland developed the simulator models for HAMBO with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator and its performance have been described in other publications [1, 2]. On July 25th 2006 there was a short circuit at Forsmark 1 nuclear power plant when manoeuvring equipment in the 400kV-switch yard. Due to the short circuit, the plant suffered an electrical disturbance that led to scram and failure of two out of four diesel generators. The purpose of the study carried out at VTT in 2007 was to assess the capabilities of the HAMBO BWR simulator to handle Forsmark 1 type of events in different nuclear power plants (Forsmark 3 in this case). The Forsmark 1 incident showed (among other things) that the intention to protect certain components (in this case the UPS-system) can in certain situations affect negatively to the safety functions. It is concluded that most of this type of BWR transients may be simulated to a certain extent using the existing HAMBO- and APROS- models. A detailed modelling of the automation and electric systems is required sometimes if the complex interplay between these systems and the process is to be predicted reliably. The modelling should be plant specific and level of detail should be assessed case-by-case (i.e. what kind of transient is in question, what systems are available, what is the main purpose of the analyses etc.). The thermal-hydraulic models of the APROS-code seem to replicate well the real behaviour of thermal-hydraulic process provided that there is enough information about the transient in consideration.

RELAP5/MOD3 Code Verification Through PWR Pressure Vessel Small Break LOCA Tests in OECD/NEA ROSA Project

The Japan Atomic Energy Agency (JAEA) started OECD/NEA ROSA Project in 2005 to resolve issues in the thermal-hydraulic analyses relevant to LWR safety through the experiments of ROSA/LSTF in JAEA. More than 17 organizations from 14 NEA member countries have joined the Project. The ROSA Project intends to focus on the validation of simulation models and methods for complex phenomena that may occur during DBEs and beyond-DBE transients. Twelve experiments are to be conducted in the six types. By utilizing the obtained data, the predictability of codes is validated. Nine experiments have been performed so far in the ROSA Project to date. The results of two out of these experiments; PV top and bottom small-break (SB) LOCA simulations are studied here, through comparisons with the results from pre-test and post-test analyses by using the RELAP5/MOD3.2 code as a typical and well-utilized/improved best estimate (BE) code. The experimental conditions were defined based on the pre-test (blind) analysis. The comparison with the experiment results may clearly indicate a state of the art of the code to deal with relevant reactor accidents. The code predictive capability was verified further through the post-test analysis. The obtained issues in the utilization of the RELAP5 code are summarized as well as the outline of the ROSA Project.

Analysis of a Loss of Feedwater Case at the Ringhals-3 NPP Using RELAP5/PARCS Coupled Codes

This paper deals with the development and validation of a coupled RELAP5/PARCS model of the Swedish Ringhals-3 pressurized water reactor against a Loss of Feedwater transient, which occurred on August 16, 2005. At first, the stand-alone RELAP5 and PARCS models are presented. All the 157 fuel assemblies are modeled in individually in both codes. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, because of the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions. Capabilities of the RELAP5 code were more challenged in this transient, where the influences of the feedback from the neutron kinetic side were also taken into account in the analysis. The calculated values of the parameters show good agreement with the measured data.

Post-Test Analysis of an Integral Effect Reflood Test for an APR1400 Large-Break LOCA Scenario

Post-test analysis was performed on an integral effect reflood test, the ATLAS Test No. 9, for an APR1400 Large-Break LOCA (LBLOCA) scenario by using the MARS code. Integral effect tests on the reflood period of a large break LOCA were performed by using the ATLAS facility to help understand the thermal-hydraulic phenomena during the reflood period of a large-break LOCA for APR1400 and for resolving the current safety issues for the APR1400 licensing on the downcomer boiling phenomenon. The present ATLAS Test No. 9 is one of the integral effect reflood tests for investigating the thermal-hydraulic characteristics during an entire reflood period to provide reliable data to help validate the LBLOCA analysis methodology for APR1400.

Heat Transfer Boundary Conditions in the RELAP5-3D Code

The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones.

Digital I&C Operating Experience in the US

The Electric Power Research Institute (EPRI) is conducting research in cooperation with the Nuclear Energy Institute (NEI) regarding Operating Experience of digital Instrumentation and Control (I&C) systems in US nuclear power plants. The primary objective of this work is to extract insights from US nuclear power plant Operating Experience (OE) reports that can be applied to improve Diversity and Defense in Depth (D3) evaluations and methods for protecting nuclear plants against I&C related Common Cause Failures (CCF) that could disable safety functions and thereby degrade plant safety. Between 1987 and 2007, over 500 OE events involving digital equipment in US nuclear power plants were reported through various channels. OE reports for 324 of these events were found in databases maintained by the Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO). A database was prepared for capturing the characteristics of each of the 324 events in terms of when, where, how, and why the event occurred, what steps were taken to correct the deficiency that caused the event, and what defensive measures could have been employed to prevent recurrence of these events. The database also captures the plant system type, its safety classification, and whether or not the event involved a common cause failure. This work has revealed the following results and insights: - 82 of the 324 “digital” events did not actually involve a digital failure. Of these 82 non-digital events, 34 might have been prevented by making full use of digital system fault tolerance features. - 242 of the 324 events did involve failures in digital systems. The leading contributors to the 242 digital failures were hardware failure modes. Software change appears as a corrective action twice as often as it appears as an event root cause. This suggests that software features are being added to avoid recurrence of hardware failures, and that adequately designed software is a strong defensive measure against hardware failure modes, preventing them from propagating into system failures and ultimately plant events. 54 of the 242 digital failures involved a Common Cause Failure (CCF). - 13 of the 54 CCF events affected safety (1E) systems, and only 2 of those were due to Inadequate Software Design. This finding suggests that software related CCFs on 1E systems are no more prevalent than other CCF mechanisms for which adherence to various regulations and standards is considered to provide adequate protection against CCF. This research provides an extensive data set that is being used to investigate many different questions related to failure modes, causes, corrective actions, and other event attributes that can be compared and contrasted to reveal useful insights. Specific considerations in this study included comparison of 1E vs. non-1E systems, active vs. potential CCFs, and possible defensive measures to prevent these events. This paper documents the dominant attributes of the evaluated events and the associated insights that can be used to improve methods for protecting against digital I&C related CCFs, applying a test of reasonable assurance.

Numerical Analysis of Heat Transfer Test of Supercritical Water in a Tube Using the Three-Dimensional Two-Fluid Model Code

The three-dimensional two-fluid model analysis code ACE-3D is developed in Japan Atomic Energy Agency for the thermal design procedure on two-phase flow thermal-hydraulics of light water-cooled reactors. In order to perform thermal hydraulic analysis of SCWR, ACE-3D is enhanced to supercritical pressure region. As a result, it is confirmed that transient change in subcritical and supercritical pressure region can be simulated smoothly using ACE-3D, that ACE-3D can predict the results of the past heat transfer experiment in the supercritical pressure condition, and that introduction of thermal conductivity effect of the wall restrains fluctuation of wall.

Static Instability in Supercritical Parallel-Channel Systems

An analytical study of static instability in parallel channels at supercritical pressure is conducted. Until now, primarily the density-wave type has been investigated and reported. This paper derives an analytical expression for the static instability in parallel channels, which lends useful insight. This topic that would be of immense value to the new reactor designs that aim to use supercritical light water on the primary side. The finding is that static instability would unlikely be a problem in a supercritical pressure reactor because of the low inlet core temperature that it would be encountered.

Numerical Simulation of Boiling Two-Phase Flow in Tight-Lattice Rod Bundle by 3-Dimensional Two-Fluid Model Code ACE-3D

Two-fluid model can simulate two phase flow less computational cost than inter-face tracking method and particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system in order to simulate complex shape channel flow. In this paper, boiling two-phase flow analysis in a tight lattice rod bundle is performed by ACE-3D code. The parallel computation using 126CPUs is applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. At height z = 0.5 m, void fraction in the gap region is higher in comparison with that in center region of the subchannel. However, at height of z = 1.1m, higher void fraction distribution exists in center region of the subchannel in comparison with the gap region. The tendency of void fraction to concentrate in the gap region at vicinity of boiling starting point, and to move into subchannel as water goes through rod bundle, is qualitatively agreement with the measurement results by neutron radiography. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight lattice rod bundle with no lift force model (neglecting lift force acting on bubbles) is also performed. From the comparison of numerical results, it is concluded that the effects of lift force model are not so large on overall void fraction distribution in tight lattice rod bundle. However, higher void fraction distribution in center region of the subchannel was not observed in this simulation. It is concluded that the lift force model is important for local void fraction distribution in rod bundles.

Hydrogen Distribution in a Nuclear Reactor Containment

The management of hydrogen in nuclear reactor containment after LOCA is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. Accurate turbulence modelling is important to predict the concentration distribution correctly. The turbulence models which were most commonly cited in the literature for modelling buoyancy driven flows were assessed for their suitability and it was found that the buoyancy modified Standard k-ε model is adequate for the purpose by comparing with some experimental data available in the literature. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian nuclear reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. Results of all these cases have been presented in this paper.