Safety analysis of neutron flux optimization in irradiation channels at the NUR research reactor

Prior to core reloading, planned power upgrading, or as a part of required analyses of past events, accurate safety evaluations should be carried out. Generally speaking, the content of a safety report has to be modified whenever a new type or design of fuel is to be used in a reactor core. As the existing plants have well established licensing procedures, including well founded analysis methods, the application of new analysis methods has to be thoroughly evaluated, with specific emphasis on their capability of producing results beneficial to reactor operation. The detailed study presented here was carried out so as to insure that the allowed operational safety limits of the NUR research reactor are not exceeded under any circumstances

Assessment of water hammer effects on boiling water nuclear reactor core dynamics

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients.

Analysis of the peach bottom 2 BWR turbine trip experiment by RELAP 5/3.2 code

This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to simulate the interactions between the generated steam line pressure wave propagation and the instantaneous core void distribution. An overall comparison shows good agreement between the code calculations and the experimental data. A series of sensitivity analyses were also performed in order to assess the code prediction features, as well as to identify uncertainties related to the adopted thermalhydraulic parameters used for the plant modelisation

Methodologies for Qualification and Application of Computational Tools for Accident Analysis in Nuclear Power Plants

Computer codes are widely used for NPP safety analysis within a wide set of purposes including licensing issues, safety improvement programs of existing NPPs, better utilization of nuclear fuel, higher operational flexibility, for justification of lifetime extensions, development of new emergency operating procedures, analysis of operational events, and development of accident management programmes. A safety key parameter of the evaluation and assessment of NPPs is closely related to the code ability of determining the time-space thermal-hydraulic conditions throughout the reactor coolant system and especially in the core region. At the beginning, the code development took place between the sixties and seventies where sets of conservative models are used

CATHARE Assessment of Natural Circulation in the PKL Test Facility during Asymmetric Cooldown Transients

Results of the CATHARE code calculations related to asymmetric cooldown tests in the PKL facility are presented. The test under consideration is the G2.1 experiment performed within the OECD/NEA PKL-2 project. It consists of carrying out a cooldown under natural circulation conditions in presence of two (out of four) emptied Steam Generators (SGs) and isolated on their secondary sides. The main goal of the current study is to assess the impact of a chosen cooldown strategy upon the occurrence of a Natural Circulation Interruption (NCI) in the inactive (i.e., noncooling) loops. For this purpose, three G2.1 test runs were investigated. The calculation results emphasize, mainly, the effect of the cooldown strategy, and the conditions that could lead to the occurrence of the NCI phenomenon

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

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

International Conference Nuclear Energy for New Europe 2003 Sensitivity Analyses of the Peach Bottom Turbine Trip 2 Experiment

ABSTRACT In the light of the sustained development in computer technology, the possibilities for code calculations in predicting more realistic transient scenarios in nuclear power plants have been enlarged substantially. Therefore, it becomes feasible to perform 'Best-estimate' simulations through the incorporation of three-dimensional modeling of reactor core into system codes. This method is particularly suited for complex transients that involve strong feedback effects between thermal-hydraulics and kinetics as well as to transient involving local asymmetric effects. The Peach bottom turbine trip test is characterized by a prompt core power excursion followed by a self limiting power behavior. To emphasize and understand the feedback mechanisms involved during this transient, a series of sensitivity analyses were carried out. This should allow the characterization of discrepancies between measured and calculated trends and assess the impact of the thermal-hydraulic and kinetic response of the used models. On the whole, the data comparison revealed a close dependency of the power excursion with the core feedback mechanisms. Thus for a better best estimate simulation of the transient, both of the thermal-hydraulic and the kinetic models should be made more accurate

Assessment of Natural Circulation Interruption during Asymmetric Cooldown Transients

This paper presents assessment results for the Natural Circulation Interruption (NCI) phenomenon during the cooldown phase in a nuclear Pressurized Water Reactor (PWR). This phenomenon could take place due to several circumstances such as an asymmetric cooldown after the loss of the forced primary flow. Under NCI conditions, the connection of the Reactor Coolant System (RCS) to the Residual Heat Removal System (RHRS) and the homogeneous boration of the RCS could be hindered. Moreover, at very low or no primary flow rates and an operating Safety Injection (SI) system, a Pressurized Thermal Shock (PTS) could occur in the reactor vessel due to cold fluid stratification in the loops. It is therefore important to understand the cause of loop flow stagnation and what could be the appropriate operator actions in order to avoid such phenomenon. The main goal of the current study is to assess the effect of a cooldown strategy upon the single-phase NCI occurrence. For this purpose, two scenarios with asymmetric cooling between the reactor cooling loops were investigated: the first one concerns a Feed Water Line Break (FWLB) combined with a Loss Of Offsite Power (LOOP), while the second one is limited to the LOOP (or any other transient leading to the loss of the forced primary flow). The analyses were carried out using the CATHARE 2/V2.5_1mod8.1 thermal-hydraulic system code. The calculation results emphasize, mainly, the effect of the cooldown rate and the opening strategy of the main steam Atmospheric Discharge Valve (ADV) upon the occurrence of the NCI phenomenon.JRC.F.5-Safety of present nuclear reactor

ANALYSIS OF ROD EJECTION ACCIDENT IN A RESEARCH REACTOR BY THE COUPLED CODE TECHNIQUE

ABSTRACT: This paper investigates the possibility to extend standard computer tools and methods, commonly used in the safety technology of nuclear power reactors, to research reactor safety analysis. A 3-D Neutron Kinetics ThermalHydraulic code (3D-NKTH), based on coupling PARCS and RELAP5/3.3 codes, was developed for a standard Material Test Reactor (MTR). The assessment of the model has been performed by comparison of steady state calculations against conventional diffusion codes and Monte Carlo code results. The model is applied for the analysis of a rod ejection accident. The comparison of the 3D-NKTH code, with conventional conservative research reactor tools showed that 3D-NKTH provided a more realistic course of the accident and did not require to define hot channel parameters. This approach could also open new frontiers in the safety analysis of research reactor such as setting realistic safety margin and adequate limits and operation conditions for optimal utilization of research reactors