Scram Signal Research Articles

Vibration analysis for predictive maintenance and improved reliability of rotating machines in ETRR-2 research reactor

Abstract In this work, both hardware and software modifications in a typical research reactor protection system (RPS) is proposed. The reactor cooling pumps are tripped based on vibrations safety signals of the pumps while the reactor SCRAM signal is initiated based on low flow rate and pressure drop across the reactor core which is a direct result of pumps trip. The main objective of this work is to develop reactor SCRAM signal based of core cooling pumps vibration signals. The early shutdown of the reactor based on pumps vibration signals is of significant importance not only in cooling the decay power of the reactor core after shutdown but also to prevent pumps failure. In the hardware model, the core cooling pumps vibration signals are feed to RPS to initiate reactor SCRAM signal. In the software model, a modular artificial neural network (ANN) is used in modeling the vibration monitoring of the research reactor (ETRR-2). The input and the output signals of the vibration transducer are used as a source data for training the neural network model. The type of the network used in this methodology is the supervised Multilayer Feed-Forward Neural Networks with the back-propagation (BP) algorithm. Vibration analysis programs are used in research reactors (RRs) to identify faults in machinery, plan machinery repairs, and keep machinery functioning for as long as possible without failure. The vibration severity limits are determined based on the International Organization for Standardization (ISO) 10816. The ANNs were designed using two different methods; one is by using hardware application contained two out of three voting and dynamic modules for trip signal by using ANNs. The current model classifies the vibration signals into five ranges low, good, satisfactory, unsatisfactory, and unacceptable vibration. The ANN is trained to detect the signal and vote to take the correct and safe action. The results demonstrate that the ANN can help in taking predictive actions for the safe core coolant pumps operation.

Simulation of Partial and Complete Loss of Flow Accidents for GPWR using ATHLET Code

In Nuclear power plants, simulation of the thermal hydraulic systems at loss of flow, accidents is commonly used to perform safety and licensing analyses through regulatory and safety authorities since they are considered as design base accidents. In this paper, a simulation of a thermal-hydraulic model of Germany Pressurized Water Reactor (GPWR) is developed, during loss of flow accident (LOFA) scenarios, such as: trip of one, two, three and four of the Main Coolant Pumps, at operation of 100% reactor power. The developed thermal hydraulic model is formed using ATHLET3.1A code with a point reactor kinetic to predict the behavior of the reactor through the thermal hydraulic major parameters such as power, mass flow rate, primary and secondary pressures, max clad surface and fuel temperatures. All reactivity feedbacks are taken into consideration in the developed simulation and the scram signal of the protection system depends on the mass flow rate. The results show that the investigated ATHLET model can precisely simulate, the steady state and the transient behavior of GPWR during many LOFA accidents. Moreover, the results show that there is good agreement between ATHLET model and RELAP5 code results are nearly in agreement with in previous literatures

Control rod drop-time reduction in typical swimming pool research reactors

Control rods play an important role in the safety of nuclear reactors, due to the fact that in the emergency situations, they are to be inserted into the reactor core and scram the reactor secure and fast. As far as safety concerns, drop-time of control rods should be minimized as much as possible. In this paper as a case study, TRR (Tehran Research Reactor) data is chosen as a typical swimming pool reactor and drop-time of control safety rod is to be minimized. Present fork type control rods are more prone to entanglement than former tube type. Methods to ease this problem are considered. Several approaches are studied and proposed in this work aimed to decrease the drop-time of control rods especially in case of emergency. Changing control rod shape into a more aerodynamic form, helps to reduce drag force and thus decreasing drop-time in an aqueous medium. Further improvement is achieved by provision of several holes in the head section. As fluid dynamic computations suggest, drop-time may reduce as much as 6% due to geometry modification. Focusing on electronic part, improved design on magnet (Coil and its core together) could reduce the lag time between scram signal issuance till control rod release. Using magnet with modified air gaps showed that existing drop-time from about 800 ms, could be reduced by as much as 34%. Moreover, by using amorphous core instead of existing one, drop-time is expected to reduce from 800 ms by as much as 30%.Measurements showed that existing electromechanical circuitry of relays, has an inherent 3 ms delay which is not so much compared to 800 ms total scram time. Therefore, working on this part is not a big deal with respect to other modifications and therefore is disregarded.

Primary pump coast-down characteristics analysis in lead cooled fast reactor under loss of flow transient

The primary pump coast-down characteristics research is very important to ensure reactor safety during power failure events in the lead cooled fast reactor (LFR) design. The primary pump coast-down Flow Halving Time (FHT) effects on a 10 MWth lead-bismuth cooled research reactor with critical and sub-critical operation modes were analyzed under loss of flow transient in this paper. Comparison of different coast-down FHT effects on critical and sub-critical reactors was conducted. Moreover, other relevant parameters those may affect the coast-down FHT requirements (e.g. clad temperature limitation and temperature trigger threshold) were analyzed as well. The results showed that: Increasing the primary pump coast-down FHT can decrease the clad peak temperature in lead cooled fast reactor under LOF transient. Both unprotected and protected loss of flow (ULOF and PLOF) transients in critical reactor and ULOF transient in sub-critical reactor were insensitive to coast-down FHT, while PLOF transient in sub-critical reactor was found to be sensitive to the coast-down FHT. The coast-down FHT effects on sub-critical reactor was bigger than critical reactor in the early phase of LOF transient. The primary pump coast-down FHT had little effects on the critical reactor safety under LOF transients. The clad peak temperature of the sub-critical reactor under ULOF transient was the only one that exceeded the temperature limitation among these four transients. Raising the clad temperature limit of the sub-critical reactor contributes little to reactor safety under ULOF transient. Measures should be taken to avoid the happening of serious accident, such as ULOF transient in sub-critical reactor. Scram signal setup (e.g. core outlet temperature trigger threshold) would not enhance the safety of lead cooled critical and sub-critical reactors and would not change the coast-down FHT requirements under LOF transient.

Three Design Basis Accidents’ Analysis on the HTR-10GT

The study simulated the design basis accidents (DBAs) sequences of the HTR-10GT core with THERMIX. When a DBA happens, the protection system will receive a scram signal which shall lead to active measures to shut down the reactor following it. In the paper, three typical DBA cases were studied. They include an accident induced by station blackout, a case caused by the withdrawal of one control rod out of the core by a mistake, and a case resulting from an earthquake, respectively. The simulation results illustrate that the fuel peak temperatures in the core during these accidents are 1066°C, 1201°C and 1067°C, respectively. It is shown that the HTR-10GT has a good safety characteristic.

Validation studies and interpretation of the Oskarshamn-2 1999 stability event with SIMULATE-3K

The OECD/NEA recently launched an international benchmark on a combined feedwater transient and stability event that occurred at the Swedish Oskarshamn-2 (O2) nuclear power plant (NPP). The primary benchmark objective is to assess advances in coupled neutronic/thermal–hydraulic codes for simulations of challenging transients including the appearance of unstable power oscillations. The Paul Scherrer Institut (PSI) is participating in this benchmark in order to enlarge the validation basis of its advanced stability analysis methodology currently under development for Swiss B WRs and based on the state-of-the-art SIMULATE-3K (S3K) code. This paper presents the development, optimization and validation of a S3K model for the O2 benchmark, including the analysis of the entire February 1999 event as well as additional stability tests considered as blind cases. With the optimized model, the S3K solution is compared to available benchmark event data both for at steady-state and transient conditions. For the latter, the qualitative as well as quantitative behavior of S3K results is compared and discussed in relation to the experimental observations. The modeling aspects found in this context to mostly affect the S3K ability to reproduce the event are also presented. As a result, the S3K model indicates that if the reactor scram had not been introduced, the observed diverging oscillations would reach a maximum amplitude before decaying back into a stable state. However, it is also found that the core could instead have evolved into limit-cycle oscillations if a stabilization of the feedwater flow and temperature had occurred just before the scram signal. Finally, the S3K model optimized for the event analysis is subsequently validated in a blind mode approach against several stability tests conducted before and after the event. Considering that the uncertainties in measured decay ratio values are not known and could be relatively large, the obtained results can on a qualitative basis be considered as satisfactory.

RETRAN analysis results of feedwater pump trip transient for Lungmen ABWR Plant

The RETRAN model of Lungmen ABWR was used to simulate one feedwater pump trip (FWPT) transient of the Lungmen start-up test program. The purpose of this test is to verify the capability of one surviving Turbine Driven Reactor Feedwater Pump (TDRFP) plus a Motor Driven Feedwater Pump (MDRFP) to continue operating the reactor stably following the incident. There are three major control systems implanted in Lungmen RETRAN model (LRM), which include Recirculation Flow Control System (RFCS), Steam Bypass and Pressure Control System (SBPCS), and Feedwater Control System (FWCS). The reactor water level margin with respect to the low level scram setpoint in the transient is monitored to confirm whether the acceptance criteria has been satisfied, which depends on the responses of the control systems to the FWPT transient. The analysis result of base case at 100% rated power/100% rated core flow with automatic start of MDRFP demonstrates that the acceptance criteria are met, which shows that the water level still sustains ample margin of 30 cm above the low level setpoint, and the reactor does not scram. To get more insight into the function of MDRFP, a set of sensitivity studies with the assumption of unavailable MDRFP, and with a different initial condition which extended to the maximum allowable core flow of 111% rated at rated power, was conducted to verify the superior capability of power coastdown due to the RIPs runback logic under the lowest load line, and also the delay time of the Reactor Internal Pumps (RIPs). Finally, it is concluded that FWPT transient without start of MDRFP eventually actuates the low level scram signal, even with the lowest load line at rated power as the initial condition.

Reactivity Accident in a High Temperature Gas-Cooled Reactor Due to Inadvertent Withdrawal of Control Rod

Abstract Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese pebble-bed modular high temperature gas-cooled reactor (HTR-PM) with single module power of 250 MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper (e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not) and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature, and xenon concentration are studied and compared in detail between these different cases. The calculating results show that the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way so that the temperature limits of the fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.

Design, Development, Testing and Qualification of Diverse Safety Rod and Its Drive Mechanism for a Prototype Fast Breeder Reactor

Prototype fast breeder reactor is U–PuO2 fueled sodium cooled pool type fast reactor and it is currently under construction at Kalpakkam, India. Prototype fast breeder reactor is equipped with two independent fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms, and neutron absorbing rods. The two shutdown systems of prototype fast breeder reactor are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of prototype fast breeder reactor are called as diverse safety rods (DSRs) and their drive mechanisms are called as diverse safety rod drive mechanisms (DSRDMs). DSRs are normally parked above active core by DSRDMs. On receiving scram signal, the electromagnet of DSRDM is de-energized and it facilitates fast shutdown of the reactor by dropping the DSR into the active core. For the development of prototypes of DSR and DSRDM, three phases of testing, namely, individual component testing, integrated functional testing in room temperature, and endurance testing at high temperature sodium, were done. The electromagnet of DSRDM has been separately tested at room temperature, in furnace, and in sodium. Specimens simulating the contact conditions between electromagnet and armature of DSR have been tested to rule out self-welding possibility. The prototype of DSR has been tested in flowing water to determine the pressure drop and drop time. The functional testing of the integrated prototype DSRDM and DSR in aligned and misaligned conditions in air/water has been completed. The performance testing of the integrated system in sodium has been done in three campaigns. During the third campaign of sodium testing, the performance of the system has been verified with 30 mm misalignment at various temperatures. The third campaign has qualified the system for 10 years of operation in reactor. This paper presents design, development, testing, and qualification of the prototype DSR and DSRDM. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual and detailed design features are explained with the help of figures. Details on material of construction are given at appropriate places. Test plans and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are briefed. Brief explanation of test setups and typical test results are also given.

Analysis of a BWR Turbine Trip Experiment by Entire Plant Simulation with Spatial Kinetics

An analysis of a boiling water reactor turbine trip was performed with the THYDE-NEU code. In spatial kinetics, reactivity was not used since the three-dimensional transient diffusion equation was solved with the implicit direct integration method. The plant was treated as a closed coolant system, and hence, it was necessary to cope with thermal-hydraulic behaviors at pressures as low as the atmospheric pressure. At low pressures, nonlinearity of the thermal-hydraulic equation is enhanced, and hence, a thermal nonequilibrium model is required. To satisfy the measured initial pressure distribution within the reactor, it was necessary to have the moisture separator model and to account for a reversible pressure drop at a junction with a flow area change. Among the parameters in THYDE-NEU is γ in the thermal nonequilibrium model in addition to C1 and C2 regarding the manner in which to express the coolant density used in the table look-up of cross sections. For a pair of C1 and C2, it is possible to find parametrically a value of γ, namely, γC, so that THYDE-NEU can reproduce the experimental fact that the core-averaged local power range monitor output RAPRM reached 0.95 at 0.63 s to generate a scram signal. One of the calculations with γC was compared with the experiment. It was shown that the spatial kinetics results are sensitive to the temporal behavior of the bypass valve opening. Among the assumptions in use, those to be scrutinized before further performing sensitivity calculations were indicated.

In-vessel Type Control Rod Drive Mechanism Using Magnetic Force Latching for a Very Small Reactor

A highly reliable control rod drive mechanism driven by an electric motor installed inside the reactor vessel (INV- CRDM) for a very small reactor has been designed. The INV-CRDM contributes to the compactness and simplicity of the reactor system, and can eliminate the possibility of a rod ejection accident. In the design, a new type of latch mechanism using an electromagnetic force to directly connect both of the shafts, one of which was the motor driven shaft and the other the control rod driving shaft, was applied so as to make the INV-CRDM very compact. The cable supplying current remained stationary, even when both of the shafts was moving. The required functions of the latch mechanism are to maintain an adequate latching force for the control rod shaft to move within a stroke of 370 mm, and to release the shafts in a shorter time than 0.2 s after a scram signal is received. A functional test with a model that approximately simulated the design was conducted to test the latching force and de-latching at room temperature. The test showed that the latching force increased with the current of the magnet coil, as did the de-latch time. The post-test analysis with a finite element analysis code revealed that the clearance between the two shafts greatly affected the latching force. With the same analysis method, the design analysis of the latch mechanism at a high temperature condition of 300°C was conducted, and it was confirmed that the latch mechanism contained enough latching force.

Cobalt irradiation box ejection accident of ETRR-2

The new Egyptian Test and Research Reactor Number 2 ETRR-2, MTR type, is now under operational tests. It has a main central irradiation channel for the purpose of Co 60 isotope production with an intended rated capacity of 50 000 Ci per year. The reactivity introduced in the reactor due to accidental ejection of the Co 60 irradiation box (CIB) should be discussed. This reactivity insertion accident (RIA) may be fast or slow with maximum reactivity worth 2.9428 $. The CIB may move with constant speed or variable acceleration according to its initial speed and the applied forces. This results in a linear, parabolic or sinusoidal motion, which in turn affects the reactivity insertion rate (RIR). The present work analyzes this type of perturbation during normal operating conditions: 22 MW full power and 1900 kg s −1 forced core cooling flow. The work serves as a part of the safety evaluation process applicable to similar MTR cores. The RIA code TRANSP20 is developed for this study. It simulates various types of RIR, fast or slow resulting from different CIB ejections. Scram signal due to power, period, inlet and outlet temperatures, or temperature difference is expected to activate the shutdown system. The work presents five case studies, two for fast ejection and three for slow. The transient behavior of the reactor during this is illustrated. The results show that the reactor can withstand slow ejection if the scram is available. However, for fast ejection the scram system does not prevent the clad temperature from exceeding safety limits. Recommendations to prevent or mitigate this accident are highlighted.

96/04998 Neutron scattering from elemental uranium and thorium

The RETRAN model of Lungmen ABWR was used to simulate one feedwater pump trip (FWPT) transient of the Lungmen start-up test program. The purpose of this test is to verify the capability of one surviving Turbine Driven Reactor Feedwater Pump (TDRFP) plus a Motor Driven Feedwater Pump (MDRFP) to continue operating the reactor stably following the incident. There are three major control systems implanted in Lungmen RETRAN model (LRM), which include Recirculation Flow Control System (RFCS), Steam Bypass and Pressure Control System (SBPCS), and Feedwater Control System (FWCS). The reactor water level margin with respect to the low level scram setpoint in the transient is monitored to confirm whether the acceptance criteria has been satisfied, which depends on the responses of the control systems to the FWPT transient.The analysis result of base case at 100% rated power/100% rated core flow with automatic start of MDRFP demonstrates that the acceptance criteria are met, which shows that the water level still sustains ample margin of 30 cm above the low level setpoint, and the reactor does not scram. To get more insight into the function of MDRFP, a set of sensitivity studies with the assumption of unavailable MDRFP, and with a different initial condition which extended to the maximum allowable core flow of 111% rated at rated power, was conducted to verify the superior capability of power coastdown due to the RIPs runback logic under the lowest load line, and also the delay time of the Reactor Internal Pumps (RIPs). Finally, it is concluded that FWPT transient without start of MDRFP eventually actuates the low level scram signal, even with the lowest load line at rated power as the initial condition.

Reducing Scram Frequency by Modifying Reactor Setpoints for a Westinghouse Four-Loop Plant

Several scram setpoints were analyzed for the purpose of reducing scram frequency in a Westinghouse four-loop plant. The results showed that the low-low steam generator (SG) level setpoint can be eliminated when reactor power is 50% or less during a loss of heat sink (LOHS) event. (The LOHS is the basis of this setpoint.) Without this setpoint, the reactor can still scram safely on either high pressurizer pressure or high pressurizer level without lifting the safety valves. The scram signal on the low SG level in coincidence with the signal from a mismatch of steam flow and feedwater flow can also be removed with no adverse effect on safety. This setpoint has never been included in the safety analysis. The results also showed that the power level above which the reactor should be scrammed when there is a turbine trip can be raised from its current value of 10% power to 50% when the condenser is available.

A New Reactor Safety Circuit for Low-Power-Level Operation

In the operation of nuclear reactors at low-power levels, one of the primary instrumentation problems is that the statistical fluctuations of reactor neutron population are accentuated by conventional log-count-rate and differentiating circuits and can cause frequent spurious scrams unless long time constants are incorporated in the circuit. Excessive time constants may introduce undesirable delay in the circuit response to legitimate scram signals. This paper develops the concept of a count doubling-time monitor which generates a scram signal if the number of counts from a pulse type neutron detector doubles in a given period of time. The paper demonstrates the theoretical relation between count doubling time and asymptomatic periods. A practical circuit to implement the function is described.