Trip setpoint analysis for the reactor protection system of an advanced integral reactor

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria.     

Three-dimensional analysis of flow blockage accident in Tehran MTR research reactor core using CFD

In the present study, a 3D simulation of flow blockage accident which may occur in the coolant channels of a fuel assembly of Tehran research reactor (TRR) is investigated using CFD code. Consideration is given to the scenario in which partial blockage of hot channel occurs due to buckling of its fuel plates towards the inside. Governing conservation laws are solved using Control volume approach and pressure field is coupled to the velocity field through the SIMPLE algorithm. Flow convergence is considered when the residual for all flow variables are less than 10⁻⁵. The simulation is performed under four different obstruction levels of the nominal flow area, i.e., 0%, 20%, 50% and 70%. By solving momentum and energy equation in three channels with their fuel plates, it is found that heat transfer is substantially affected by channels flow field. In the blockage accident, decrease in flow rate of the obstructed channel decreases cooling capacity of the obstructed channel as a result of hydraulic resistance augmentation. The obtained results show that above the 50% blockage, critical phenomena will appear which may compromise the clad integrity. Moreover, in the 70% blockage scenario, the clad temperature in the obstructed channel reaches the value associated with nucleate boiling temperature at the operative pressure.     

Overpressure protection analysis of an advanced integral reactor

Overpressure protection analysis of KAERI's advanced integral reactor, which has been developed to verify the performance of the System integrated Modular Advanced ReacTor (SMART), has been performed using the Transients And Setpoint Simulation/Small and Medium Reactor (TASS/SMR) code. In the analysis, the loss of feed-water and the regulating bank withdrawal events on behalf of the decrease in the heat removal by the secondary system and the reactivity and power distribution anomalies are selected as the initiating events for the analysis because the highest peak pressures of the primary system occur during these events. Conservative assumptions and the various initial/boundary conditions have been applied to the overpressure protection analysis for the advanced integral reactor. Although the pressurization of the primary system occurs due to an unbalance between the power generation in the core and the heat removal through the steam generator, the peak pressures in the cases of using the loss of feed-water and the regulating bank withdrawal event as an initiating event are well below the acceptance criteria of 18.7 MPa, due to the reactor protection system and three pilot operated safety relief valves installed in the advanced integral reactor.     

Simplified numerical model for defining Ledinegg flow instability margins in MTR research reactor

Establishment of safety margins and the corresponding operating condition limits will ensure achievement of a safe operation of nuclear installations. For this purpose, several critical phenomena have been analyzed theoretically and experimentally and a great number of models and correlations are made available. Among these critical issues the well-known flow instability has been intensively investigated by several authors especially for nuclear power plants' (NPPs) operating conditions. However, limited published work is available for research reactor operation conditions. In general, the Whittle and Forgan correlation is widely used to define the margin to static flow instabilities in narrow parallel heated channels for research reactors.In the framework of verification and assessment of the capabilities of the RELAP5/Mod 3 system code to determine the onset of flow instability in research reactor conditions, a simple model based on steady-state equations adjusted with drift-flux correlations has been developed. The program is used to draw the pressure drop characteristic curves and to establish the conditions of the Ledinegg instability in a uniformly heated channel subject to constant outlet pressure. The model is assessed by using experimental data from a thermal hydraulic test loop by Siman-Tov and numerical results from RELAP5/Mod 3. The model presents acceptable estimation of the target mass flow that would induce flow instability and the latter could be then used to establish a conservative margin to the Ledinegg instability.     

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.     

Assessment of TAPINS code for application to thermal-hydraulic analysis of an integral reactor,REX-10

The thermal-hydraulic analysis program for integral reactor system (TAPINS) is a thermal-hydraulic system code developed by Seoul National University for transient analysis of an integral reactor, REX-10. Specialized for a fully passive integral pressurized water reactor, TAPINS adopts a one-dimensional four-equation drift-flux model for two-phase flows. It also consists of component models for the core, the helical-coil steam generator, and the steam-gas pressurizer. This paper presents the developmental assessment of TAPINS to validate its applicability to the thermal-hydraulic analysis of REX-10. Assessment problems are determined by taking into account thermal-hydraulic phenomena expected during design basis accidents of REX-10, including the loss-of-feedwater accident and the small-break loss-of-coolant accident. To confirm the predictive capability of TAPINS for these phenomena, the TAPINS model is validated against four sets of separate effects problems, including the pressurizer insurge test, the subcooled boiling experiment, the critical flow test, and the Edwards pipe problem. In addition, the calculation results of TAPINS are compared with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. From the validation results, it is demonstrated that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions.     

Safety analysis for an inadvertent control rod withdrawal event during a power operation of an advanced integral reactor

To identify a safety margin in the case of an inadvertent control rod withdrawal event of a 65-MW_t advanced integral reactor, safety analysis has been carried out by using the Transients And Setpoint Simulation/System integrated Modular Reactor (TASS/SMR) code. The diverse initial conditions, various reactivity insertion rates into a core, different combinations of a reactivity feedback and three different speed modes of a main coolant pump (MCP) have been considered to identify the effect of each parameter on a critical heat flux ratio (CHFR) and the initial condition resulting in the worst consequences from the viewpoint of the minimum critical heat flux ratio. The analysis results show that the worst consequences occur when a reactivity of 17.61 pcm/s is inserted into a core at an initial condition of a 45% initial core power, high coolant temperature at the core inlet position, low system pressure and a thermal design flow. It is also assumed that the least negative fuel and moderator temperature coefficients are applied. The safety parameters such as the minimum critical heat flux ratio and the system pressure are maintained within the safety limits and the reactor is safely transferred to a safe condition by a functioning of the safety systems of the advanced integral reactor.     

Thermohydraulic characteristics of the condenser in the natural-circulation loop of an MTR modular heavy-water reactor

Scientific Design Department, I. I. Polzunov Central Scientific-Research Institute for the Planning and Design of Boilers and Turbines. St. Petersburg Technical University. Institute of Theoretical and Experimental Physics. Translated from Atomnaya Énergiya, Vol. 72, No. 2, pp. 136–142, February, 1992.     

Fully-coupled transient simulation for RCS and SPCS of an integral reactor plant

Design features of SMART such as a built-in once-through steam generator (OTSG) and a close interaction between the feedwater flow rate and steam pressure controls leads to the necessity of fully-coupled transient analysis tools of the reactor coolant system (RCS) and the steam and power conversion system (SPCS) for the purpose of a plant control system development. A fully-coupled transient simulation tool, MMS/SMART, was developed to test the capability of the plant control system for the normal load-following event and the anticipated abnormal events. The MMS/SMART was composed of several interacting MMS modules with numerical data, each of which represented a component of the SMART plant and a control logic. The RCS and the SPCS with their control logics were modeled using default modules such as a pipe, pump and tank. The developed MMS/SMART was validated by using the scaled-down experimental data and the analysis result from the TASS/SMR code. A simulation result for the 100–50–100% load-following operation with a 25%/min rate shows that the feedwater flow rate and the steam pressure are controlled well as expected, except for small-amplitudes of steam pressure fluctuation at the lower power operating region. The loss of turbine load event was also simulated and the result shows that the plant can be operated stably with the steam bypass control system.     

A study of the decommissioning procedure of an activated structure through an evaluation of the decommissioning cost for a research reactor

As decommissioning of a research reactor and a nuclear installation requires a long period of time from the decommissioning preparation work to the site remediation, the management of the data generated during the entire period of decommissioning is one of the most important tasks. In particular, the data obtained from research reactor decontamination and decommissioning activities can be important resources securing the safety and economic feasibility for other research reactor decommissioning. The owner of the research reactor and nuclear power plant need to submit decommissioning plan to the regulatory body at the starting stage of the research reactor and nuclear installation decommissioning project. The cost plan for decommissioning and the method for assessing the amount of exposure to protect workers must be stated in the decommissioning plan.This paper introduces the DES (Decommissioning Engineering System) that can be able to manage the data generated in the process of decommissioning of the TRIGA research reactor, to calculate an amount of waste, to evaluate decommissioning cost after deriving unit work productivity factor, and to predict the decommissioning process in advance. To verify the usability of this system and data integrity through connections among the unit systems, it describes the process to calculate the decommissioning cost using the data generated in dismantling an activated bio-shielding concrete in the TRIGA research reactor.As a result of the experiment to calculate the decommissioning cost with the TRIGA research reactor structure, it was found that the calculations were done precisely without flaw as the purpose of the experiment. Therefore, the DES can not only be used for other research reactors decommissioning, but also it is expected to be applied to other research reactors in the future. As a decommissioning cost between an activated concrete and a non-activated concrete according to the method of the dismantling procedure was significantly different, a study regarding the dismantling procedure needs more research.     

Reactivity worth of the thermal column of a MTR type swimming pool research reactor using low enriched uranium fuel

The reactivity worth of the thermal column of a typical MTR type swimming pool research reactor using low enriched uranium fuel has been determined by modeling the core using standard computer codes. It was also measured experimentally by operating the reactor in the stall and open ends. The calculated value of the reactivity worth of the thermal column is about 14% greater than the experimentally determined value.     

Nuclear steam supply system with an integral nuclear reactor for a floating NPP

The paper presents the results of conceptual design of the integral reactor plant of enhanced safety for a small-size floating NPP which is transported to the operation site in the state most prepared for operation and which is designed for power supply to remote and not easily accessible areas. Schematic flow diagrams, design and layout of NSSS and reactor plant as a whole are presented, as well as basic specifications.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Performance evaluation of an advanced integral reactor against an anticipated transient without scram

Performance evaluation of KAERI’s advanced integral reactor against an anticipated transient without scram has been carried out with the transients and setpoint simulation/small and medium reactor code, by considering a decrease in the heat removal by the secondary system, a loss of offsite power and an inadvertent control rod withdrawal event as an initiating event. In a decrease in the heat transfer by the secondary system and a loss of offsite power, the reactor coolant system pressures can be maintained below 110% of the design pressure during the transition period due to the effect of the large negative moderator temperature coefficient. On the other hand, in an inadvertent control rod withdrawal event, the pressure of the reactor coolant system increases up to the ASME service level C stress limit due to a high reactivity insertion into a reactor core by the adoption of a boron free core concept. Therefore, a hardware installation against an anticipated transient without scram is essential to mitigate the consequences resulting from an inadvertent control rod withdrawal event. A diverse protection system, which is an independent and diverse reactor shutdown system that is initiated by the signals of a high core power or a high pressurizer pressure, is adopted in the advanced integral reactor. According to the reassessment results by considering the diverse protection system for a reactor shutdown, the diverse protection system is helpful in mitigating the consequences of an anticipated transient without scram.     

Application of seismic analysis methodology to small modular integral reactor internals

The fluid–structure interaction (FSI) effect should be carefully considered in a seismic analysis of nuclear reactor internals to obtain the appropriate seismic responses because the dynamic characteristics of reactor internals change when they are submerged in the reactor coolant. This study suggests that a seismic analysis methodology considered the FSI effect in an integral reactor, and applies the methodology to the System-Integrated Modular Advanced Reactor (SMART) developed in Korea. In this methodology, we especially focus on constructing a numerical analysis model that can represent the dynamic behaviors considered in the FSI effect. The effect is included in the simplified seismic analysis model by adopting the fluid elements at the gap between the structures. The overall procedures of the seismic analysis model construction are verified by using dynamic characteristics extracted from a scaled-down model, and then the time history analysis is carried out using the constructed seismic analysis model, applying the El Centro earthquake input in order to obtain the major seismic responses. The results show that the seismic analysis model can clearly provide the seismic responses of the reactor internals. Moreover, the results emphasize the importance of the consideration of the FSI effect in the seismic analysis of the integral reactor.     

A new integral method for solving the point reactor neutron kinetics equations

A numerical integral method that efficiently provides the solution of the point kinetics equations by using the better basis function (BBF) for the approximation of the neutron density in one time step integrations is described and investigated. The approach is based on an exact analytic integration of the neutron density equation, where the stiffness of the equations is overcome by the fully implicit formulation. The procedure is tested by using a variety of reactivity functions, including step reactivity insertion, ramp input and oscillatory reactivity changes. The solution of the better basis function method is compared to other analytical and numerical solutions of the point reactor kinetics equations. The results show that selecting a better basis function can improve the efficiency and accuracy of this integral method. The better basis function method can be used in real time forecasting for power reactors in order to prevent reactivity accidents.