Experimental validation of the helical steam generator model in the TASS/SMR code

The transients and setpoint simulation/system-integrated modular reactor (TASS/SMR) code has been used to identify the safety margin of a 65-MWt advanced integral reactor and to evaluate its design performance. Although, the code has been verified by using simplified and analytical problems as well as a reliable system code, its validation has not been fully established. This paper deals with a validation of the TASS/SMR code by using two kinds of separate effect tests related to heat transfer at a helically coiled steam generator. The heat transfer experiments were performed by using a full-scale prototype of the steam generator cassette of the advanced integral reactor and a scaled-down steam generator cassette. Analytical results show that the TASS/SMR code predicts the thermal hydraulic parameters, including the system pressure and fluid temperature at the primary and secondary sides of the steam generator cassette, and the heat transfer rate through the steam generator cassette well. The validation results in this study show that the TASS/SMR code is applicable for heat transfer calculations related to the helically coiled steam generator of the advanced integral reactor.     

Development and steady state level experimental validation of TASS/SMR core heat transfer model for the integral reactor SMART

An advanced integral pressurized water reactor (PWR) of a small size (330 MWt) is being developed by the Korea Atomic Energy Research Institute (KAERI). The purposes of the reactor are a sea water desalination and an electricity generation. To enhance its safety, many advanced design concepts are introduced such as a passive residual removal system and a low power density core. For the safety validation of the designed reactor, a system analysis code named TASS/SMR, was developed. TASS/SMR code uses a one dimensional node/path modeling for the thermal hydraulic calculation and point kinetics for the core power calculation. The code also has specific models for the developed integral reactor, such as a helical tube heat transfer model and a passive residual heat transfer model. One of the important models for the safety or performance calculation is the core heat transfer model. The core heat transfer model of TASS/SMR was developed to meet the requirements of the 10 CFR 50 appendix K EM model as well as the realistic models. The developed model was validated with experimental data. The results show that the model predicts the heat transfer phenomena in the reactor core with a reasonable conservatism.     

Design of the advanced accumulator for the pressurized water reactor

With the increased requirement for nuclear power generation as an effective countermeasure against global warming, Mitsubishi has developed the advanced pressurized water reactor (APWR) by adopting a new component of the emergency core cooling system (ECCS), a new instrumentation and control system, and other newfound improvements. The ECCS introduces a new passive component called the advanced accumulator which integrates both functions of the conventional accumulator and the low-pressure pump without any moving parts. The advanced accumulator uses a new fluidics device that automatically regulates flow rates of injected water in case of a loss of coolant accident (LOCA). This fluidics device is referred to as a flow damper. This paper describes the design method of the flow damper and the standpipe.     

Identification of initiating events for pressurized water reactor accidents

In order to aid operators in identifying the different initiating events as defined in the Final Safety Analysis Report (FSAR), we develop a novel identification procedure. The procedure is based on the monitoring of three key system parameters in a pressurized water reactor (PWR), i.e., the pressure, the average temperature, and the temperature difference of the hot-leg and cold-leg of the reactor coolant system. By monitoring the system thermal state diagram in a pressure–temperature space, an operator can easily identify what initiating event is taking place while a static point in the diagram starts to move. The event data pool is first established by storing the transient analysis results for events of different types using the optimal estimated RELAP5 model. Since the variation ranges of system key parameters at a specific time represent the specific character for each initiating event, the identification procedure can easily determine which cases in which the event data pool can be fitted to on-line data using only variation range comparison without complex calculations. This identification method is believed to be able to help the plant operator to identify the different events and then execute the Emergency Operating Procedure more effectively.     

Dynamic analysis of a pressurized water reactor

An apparent malfunction in a pressurized water reactor system has been investigated using fluctuation analysis. Both frequency-domain and time-domain analyses have been used and the results obtained by the two methods have been compared. The recording was performed by a relatively simple, cheap system giving high recording precision and the analyses were performed on an IBM 370 digital computer. It is shown that, while considerable information can be derived from frequency-domain analyses, a misinterpretation can occur in some cases. Time-domain correlation, as normally performed, was not very informative. However, time-domain correlation on bandwidth-limited time-series proved to be very valuable and could remove the misinterpretation of the frequency-domain analyses. The bandwidth limitation was performed by digital filters.     

Experimental investigations of the flow patterns in the hot leg of a pressurized water reactor

For the simulation of loss of coolant accidents in nuclear power plants, the flow patterns are predicted by using experimental results from small sized plants which usually have been achieved for fully developed flows. Experimental investigations in a large sized plant have indicated that these flow pattern maps are not fully applicable to the specific geometric properties of nuclear power plants. Therefore, we have conducted experimental investigations for cocurrent two-phase flow in the hot leg. For the experimental investigations a large sized experimental set-up has been constructed, which represents the hot leg of a pressurized water reactor at the scale of 1:1.7. To distinguish between the influence of the size of the plant and the influence of the elbow and the steam generator simulator on the flow pattern, the experimental investigations have been conducted in two steps. First, the flow in the horizontal part of the hot leg has been investigated without connecting the elbow to the plant. The flow regimes have been detected by visual observation. The experimental results are compared to those obtained for smaller pipe diameters and longer pipe lengths. Second, the 50° upwards inclined elbow and the steam generator simulator are added to the horizontal pipe and their influence on the flow patterns is investigated.     

Experimental study on the air/water counter-current flow limitation in a model of the hot leg of a pressurized water reactor

An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25 m × 0.05 m) and 2.59 m, respectively, whereas the inclination angle of the riser is 50°. The flow was captured by a high-speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Countercurrent flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels measured in the separators. The counter-current flow limitation is defined as the maximum air mass flow rate at which the discharged water mass flow rate is equal to the inlet water mass flow rate.From the high-speed observations it was found that the initiation of flooding coincides with the formation of slug flow. Furthermore, a hysteresis was noticed between flooding and deflooding. The CCFL data was compared with similar experiments and empirical correlations available in the literature. Therefore, the Wallis-parameter was calculated for the rectangular cross-sections by using the channel height as length, instead of the diameter. The agreement of the CCFL curve is good, but the zero liquid penetration was found at lower values of the Wallis parameter than in most of the previous work. This deviation can be attributed to the special rectangular geometry of the hot leg model of FZD, since the other investigations were done for pipes.     

Experimental validation of the thermal-hydraulic code SACATRI

A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.     

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.     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.     

Control rod drop hydrodynamic analysis for a pressurized water reactor

During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.     

A comparison between code calculation and blowdown experiments simulating a loss-of-coolant accident in a pressurized water reactor

Within the reactor safety programme of the EURATOM Joint Research Centre at Ispra the transient heat transfer phenomena during depressurization are experimentally investigated under PWR conditions. The special closed loop DHT-1 essentially represents one subchannel and the upper and lower plenum of a pressurized water reactor. A test series simulating rupture in the hot leg of a primary cooling circuit was carried out. Pressure and test tube temperatures were measured at various rupture cross-sections. Independently from these experiments, a blowdown computer code was developed by the Groupement Atomique Alsacienne Atlantique (GAAA). The core part of this code allows calculation of the thermohydraulic history of the coolant within the core after a rupture in the primary cooling circuit. It has been checked with regard to the hypothesis and correlations applied; the experiments and calculations are compared.     

Steady-state and transient investigation of a small pressurized water reactor ACPR50S for different ATFs based on Bamboo-C code

Small modular reactors have received widespread attention owing to their inherent safety,low investment,and flexibility.Small pressurized water reactors （SPWRs） have become important candidates for SMRs owing to their high technological maturity.Since the Fukushima accident,research on accident-tolerant fuels （ATFs）,which are more resistant to serious accidents than conventional fuels,has gradually increased.This study analyzes the neutronics and thermal hydraulics of an SPWR （ACPR50S） for different ATFs,Be O+UO₂-Si C,Be O+UO₂-Fe Cr Al,U ₃Si₂-Si C,and U ₃Si₂-Fe Cr Al,based on a PWR fuel management code,the Bamboo-C deterministic code.In the steady state,the burnup calculations,reactivity coefficients,power and temperature distributions,and control rod reactivity worth were studied.The transients of the control rod ejection accident for the two control rods with the maximum and minimum reactivity worth were analyzed.The results showed that 5%B-10 enrichment in the wet annular burnable absorbers assembly can effectively reduce the initial reactivity and end-of-life reactivity penalty.The Be O+UO₂-Si C core exhibited superior neutronic characteristics in terms of burnup and negative temperature reactivity compared with the other three cases owing to the strong moderation ability of Be O+UO₂ and low neutron absorption of Si C.However,the U ₃Si₂ core had a marginally better power-flattening effect than Be O+UO₂,and the differential worth of each control rod group was similar between different ATFs.During the transient of a control rod ejection,the changes in the fuel temperature,coolant temperature,and coolant density were similar.The maximum difference was less than 10?C for the fuel temperature and 2?C for the coolant temperature.     

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

1 Introduction With respect to the inherent safety of nuclear re- actors, application of passive systems/components including natural circulation phenomena as the main mechanism is intended to simplify the safety-related systems and to improve their reliability, to reduce the effect of human errors and equipment failures, and to provide more time to enable the operators to prevent or mitigate serious accidents. Natural circulation is the main mode of heat removal for removing decay heat from t…     

Reactivity control technique for a pressurized water reactor with an inventive TRISO fuel particle composition

In this research paper a reactivity control technique has been suggested for the conceptual design of a compact sized pressurized water reactor (PWR) core with an inventive tristructural-isotropic (TRISO) fuel particle composition. This conceptual design is a light water cooled and moderated reactor which utilizes TRISO fuel particles in PWR technology. The use of TRISO fuel in PWR technology improves integrity of the design due to its fission fragments retention ability. The fuel provides first retention barrier within fuel itself against the release of fission fragments that makes this design concept safer and environment friendly. The suggested TRISO fuel particle composition has a small amount of Pu-240 with 2.0 w/o in the place of U-238 which acts as reactivity suppressor. Reactor codes WIMS-D/4 and CITATION have been used for simulation and core design modeling. Results reveals that the amount of excess reactivity can be reduced significantly by using a small amount of Pu-240 in TRISO fuel which in turns reduces the number of gadolinia rods in the core required for excess reactivity control and completely eliminates the requirement of soluble boron system. Therefore the effective and optimal use of reactivity suppressor and burnable poison suppresses and flattens the core excess reactivity throughout the core life and hence number of control rods can be reduced without compromising on the shutdown margin.     

Identification of pressurized water reactor transient using template matching

Template matching, which is a pattern recognition method, was adopted to identify the transient in a pressurized water reactor (PWR). The transient data were generated using a plant simulation code, PCTran-PWR, and transformed into a feature vector sequence (FVS). The data set contained such FVS as the reference transients. To compare the FVS of the test transient and reference transient, a cost function was defined and dynamic programming was applied to obtain the minimum cost function value, which would indicate the degree of matching between the two transients. Considering the discrepancy between the real plant and the model to generate the transient data, the same test and reference transient may not be matched exactly. A dynamic threshold value was designed to determine if the test transient matched the reference transient of the data set. Experiments were performed and the results showed that the method was successful.     

Neutron noise monitoring of pressurized water reactor internal vibrations

Analytical and experimental results have shown that the neutron noise signals are typically the sum of a number of different noise sources. These can have significant interactions due to structural coupling and summation effects in the sensor. Analytical techniques have been developed to identify major neutron noise sources and to separate and account for some of the noise source coupling effects.

This work has demonstrated the use of various noise source models in neutron noise monitoring applications. Methods of identifying and separating the noise sources have been used to relate changes in the measured spectra to particular noise source properties. The noise source models can then be used to relate noise source properties to physical properties of the system. These techniques are used in routine surveillance applications and have provided proper evaluation of several trends and changes that have been observed in neutron noise monitoring programs.

All neutron noise measurements have shown small vibration amplitudes that are in agreement with results from preoperational measurements and analysis. Neutron noise monitoring is being continued on an optional basis in a number of plants as a means of monitoring core clamping and general long-term performance.