Safety Of Reactor Operation Research Articles

Characteristics of heat transfer in corrosion products deposited on the surface of pressurized water reactor fuel elements

The occurrence of porous deposits on the core surface during the operation of a pressurized water reactor can seriously affect its heat transfer characteristics and increase the surface temperature of the cladding. Continued heat transfer deterioration causes subcooled boiling and boron enrichment within porous deposits, leading to axial power anomalies in the core, and the study of porous deposits is beneficial to improving the economics and safety of reactor operation. In this paper, a two-dimensional boiling unit model of a porous deposits was established using COMSOL software to couple heat conduction, darcy flow, solute transport and chemical reactions in a multiphysics field to reasonably predict the temperature field, velocity field, boric acid concentration distribution, and PH distribution within the porous deposits. On this basis, the influence of different parameters (porosity, boiling unit radius of porous deposits, heat flow density) on the maximum effective thermal conductivity, the pattern of thermal parameters (temperature, superheat, percentage of subcooled boiling, and saturation temperature) on the chimney boundary are analyzed. The results show that localized superheating, maximum boric acid concentration and subcooled boiling occur at the bottom of the porous deposits. There is a negative correlation between the porosity and the percentage of subcooled boiling area, and a positive correlation between the radius of the porous deposits and the heat flow density and the percentage of subcooled boiling area.

Experimental analysis of boron dilution transient flow mixing characteristics based on planar laser induced fluorescence measurement

In Reactor Pressure Vessels (RPV), flow mixing between high and low concentration coolants will cause uneven distribution of boric acid. Severe cases will lead to boron dilution transient and re-criticality safe operation accident. Therefore, accurate measurement of coolant flow mixing characteristics and boron concentration distribution in RPV and study of the mechanism affecting the mixing process are of great significance to improve reactor operation safety. Based on the structural characteristics of large advanced pressurized water reactor pressure vessel, a large-scale visual experimental device was built by 1:6 scale modular design. Optical path correction scheme is used to solve the problem of optical path distortion existing in PLIF technology, and improved ratio calibration method is used to reduce the experimental error of concentration field distribution measurement. Flow mixing process and boron concentration field distribution in pressure vessel under single loop conditions were obtained based on plane laser induced fluorescence technique (PLIF). The influence of Reynolds number on boron diffusion behavior was studied. The experimental results show that during single loop operation, the localized eddy current formed in the left area of the lower head will result in a high concentration retention zone and a V-shaped boric acid concentration distribution in the core. With the increase of the ratio of cold tube flow to injection flow, the time when the high concentration coolant reaches the core hardly changes with the boric acid solution flow rate. The mixing diffusion behavior of boron concentration is the best when the ratio of injection to Reynolds number in the cold tube is at a certain value.

Experiment and Modeling of Interfacial Area Concentration of Liquid Film in Upward Annular Two-Phase Flow

To accurately quantify the interfacial transfer terms in the two-fluid model, the reliable prediction of the interfacial area concentration (IAC) is crucial. The IAC in annular flow, especially the interface between the liquid film and gas core, is particularly important due to its relevance to critical heat flux and reactor operation safety. However, very few experimental and analytical studies have been performed that focus on the IAC of the liquid film in annular flow. In this work, the IAC of the liquid film is measured using a parallel-wire conductance probe for upward annular flow in a 25.4-mm one-dimensional pipe. A total of 25 flow conditions are measured with the range of superficial liquid velocity from 0.15 to 2.00 m/s and the range of superficial gas velocity from 10.0 to 29.6 m/s. The IAC radial profile is obtained from the liquid film time trace measured by the conductance probe, and the accuracy of this method is verified by flow visualization. The effects of the inlet gas and liquid flow rates on the characteristics of the IAC radial distribution as well as area-averaged IACs are analyzed. A new model is developed to predict the IAC radial distribution of the liquid film. The IAC profiles predicted by the model agree very well with the measured IAC profiles for typical annular flow conditions and have a reasonable agreement for the wispy annular flow conditions.

An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes

Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.

ANALYSIS OF THE PPF VALUE DEPENDENCE ON THE FUEL BURNUP

The RSG-GAS reactor has been operated in a safe and reliable manner for about 35 years since it commenced in operation in 1987 to serve radioisotopes production, NAA, neutron beam experiments, material irradiation, and reactor physics experimental activities as well as training. PPF value is necessary to determine by calculation because it is impossible to determine by experiment and also has a strong relation to the operation safety. The paper is intended to analyze the PPF values of the RSG-GAS reactor core as a function of burn up. The analysis is using WIMSD-5B/BATAN-3DIFF computer codes calculation. The result shows that the PPF values are significantly different for each burn-up or energy in MWD. The result also shows that the BATAN-3DIFF code accurately determines the PPF values of the RSG-GAS reactor core and supports the safety of reactor operation.

Analysis of load-following operation characteristics of liquid fuel molten salt reactor

As the IEA reports indicated that nuclear power owns a growing market share, which means that nuclear power plants (NPPs) must be able to respond quickly to grid fluctuations. A good load-following capability is the basic requirement for the flexibility and safety of reactor operation. In order to study the load-following capacity of the molten salt reactor, this study took the liquid fuel molten salt reactor (MSR) as the research object. The system model and power control model were established with RELAP5/MOD4.0 program, which has been implanted with relevant physical properties of molten salt and calculation formula. The response characteristics of the control system under step load change, linear load change and typical daily load-following operation were simulated and analyzed. The results show that MSR has a good load-following capacity under the designed control logic, the core power can respond to the load variation quickly, the power overshoot and temperature overshoot are small, and the operating parameters of the reactor are under a reasonable operating range. The simulation analysis provides the reference for subsequent safety analysis and power control of molten salt reactor.

Mass estimation method of loose parts based on function model

Abstract The nuclear main pump has a complex structure and contains a large number of components. With the continuous flow of the primary circuit coolant, there is a risk of loosening and falling off of the components in the reactor. The existence of loose parts in the reactor makes the reactor operation a safety hazard. Accurate mass estimation of the loose parts is conductive to quickly judge the degree of the failure of loose parts, provide effective suggestions for the subsequent maintenance of the reactor and ensure the safety of reactor operation. The existing mass estimation methods have problems of large errors and poor consistency. This article proposes a method for estimating the mass of loose parts by establishing a functional relationship model based on signal characteristic value. The characteristic value of the signal is constructed by extracting the Root Mean Square value and the main frequency of the shock signal, and then the functional relationship between the quality of the loose part and the characteristic value of the signal is established, which is used to estimate the mass of the loose part. The experimental results show that the method proposed in this paper has the characteristics of small estimation error and good consistency, which can meet the needs of practical engineering applications.

Visualizing primary cooling system risks with bowtie diagram for RSG GAS operator guidance

AbstractThe G.A. Siwabessy Multipurpose Reactor (RSG GAS) is a pool‐type research reactor. When the reactor operates, the heat released in the core is removed by the primary cooling system. It is necessary to evaluate this system, because failure can cause worst events such as the melting of reactor fuel and the release of radioactivity to the environment. Due to the possibility of radioactive material releases, safety issues are important to be evaluated. Several studies using event tree analysis (ETA), fuzzy fault tree analysis (FFTA) describe the results quantitatively. The varied educations of operators will provide various responses to quantitative results, and a need to create qualitative results that are simple and easy to understand using bowtie analysis that visualizes the risk scenario. Failure of this system causes loss of coolant accident (LOCA) and loss of flow accident (LOFA). The results show that the relationship between the consequences and causes of the risk occurrence LOCA and LOFA, and the controls used to prevent and reduce the hazard based on safety and emergency system at RSG GAS, can be depicted in a bowtie diagram. The bowtie diagram is simple and easy for an RSG GAS operator to carry out their duties and increase awareness regarding the safety of reactor operations.

Preliminary Study on the Remaining Life Estimation Method for Research Reactor Tank Liner

The reactor tank liner is one of the most crucial safety barriers in a research reactor as it retains the radioactive material released from the fuel during the accident condition. It also contains the primary coolant for fission heat removal. The integrity of the tank liner determines the service life of the research reactor. So far, the remaining life estimation of pressure vessels in nuclear power plants is more widely applied and established than that of the research reactor tank liner. Therefore, a study on the remaining life estimation method of the research reactor tank liner is needed to ensure the research reactor operation safety. This paper aims to preliminarily study several methods applied to estimate the remaining life of a research reactor tank liner. The preliminary study consists of a qualitative assessment and a quantitative assessment. The qualitative assessment aims to propose several techniques or methods applied in estimating the remaining life of the reactor tank liner. The quantitative assessment applies one of the remaining life estimation methods discussed in the previous assessment. Generally, the remaining life of the research reactor tank liner can be estimated using the theoretical method and the experimental method. The theoretical methods are applied by calculating the neutron fluence received by the tank liner or by analyzing the fracture mechanics using numerical modeling if the cracks or other defects exist. The calculation of atom displacement number (dpa), as a standard measure of the neutron-induced radiation damage of the materials, can support the neutron fluence calculation. The experimental method is conducted by measuring several parameters of the tank liner material, such as the corrosion rate or the mechanical properties. In the quantitative assessment, the remaining life estimation of the Kartini Reactor tank liner was performed by neutron fluence calculation method using MCNP6 computer code. The result shows that the maximum neutron fluence received by the tank wall is 2.950E+17 n/cm2 for 40 years operating period. By comparing the cumulative neutron fluence received for 40 years to the thermal neutron fluence limit value of 1.18E+23 n/cm2, the Kartini Reactor tank liner can still be used for the next 1.6E+07 operation years. The result of the quantitative assessment implicitly shows that the remaining life estimation of the tank liner needs to: 1) consider all defects experienced by the tank liner and all factors (e.g., thermal, radiation, chemical, cyclic loading) which affect the tank liner material condition, and 2) perform the combination of theoretical and experimental methods. For an open-pool type reactor, corrosion monitoring and corrosion rate measurement are essential to perform the remaining life assessment of the tank liner.

Power Measurement Methodologies for Pool Nuclear Research Reactors

Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 Triga nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 Triga reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of nuclear reactors using alternative methods for power monitoring.

Time History of Performance Parameters of WWR-K Reactor during Gradual Replacement of the Water Reflector by a Beryllium One

The water-cooled WWR-K research reactor has been operating since 2016 using low-enrichment uranium fuel. In order to maintain high levels of generated power and reactivity margin sufficient for a 21-day operation cycle, the water-cooled core reflector of neutrons was gradually, depending on the fuel burnup, replaced by a beryllium one in order to prevent decreasing of the reactor performance. A series of calculation studies are performed with a simulation of the core region of the WWR-K reactor using an MCNP transport code including determination of neutron flux density, reactivity, efficiency of the control rods of the control and protection systems, as well as kinetic parameters critical for ensuring safe reactor operation. The calculated curves of the principal performance characteristics of the WWR-K reactor are presented and analyzed in terms of their use in scientific research and their effect on the reactor operation safety.

EVALUATION OF EQUILIBRIUM CORE OPERATION OF THE RSG-GAS REACTOR

The Indonesian Multipurpose Reactor, RSG-GAS reactor will accomplish its first lifetime in December 2020. The reactor has been operated in safe and reliable manner for about 33 years since it commenced in operation in 1987 to serve radioisotopes production, NAA, neutron beam experiments, material irradiation, and reactor physics experimental activities as well as training. The paper is intended to evaluate its in-core fuel management that is the conformance between the theory and implementation of the equilibrium core. Evaluation of the reactor operation parameter was carried out for core numbers 91 – 100. The data show that excess reactivity, shutdown reactivity and control rod reactivity have no significant difference at each core. The result shows that the BATAN-FUEL accurately determine the equilibrium core and its fuel loading pattern.This in-core fuel management of the RSG-GAS reactor supports the safety of reactor operation.

Transient calculation on TRIGA 2000 of plate-type fuel element using RELAP5, EUREKA2/RR, and MTR-DYN codes

The TRIGA 2000 reactor Bandung is proposed to convert from rod-type fuel to plate-type fuel because there are no manufacturers in the world that still produce rod-type fuel. Calculating the safety parameters and their reliability related to reactor operation has become an important thing to do. The objective of this study was to perform the reactor transient calculation of TRIGA 2000 that uses plate-type fuel. The reactor core modification is based on the capability of the existing primary coolant system, without changing its flow rates. Thermal-hydraulics calculations related to reactivity insertion accident and loss of flow accident were analyzed using EUREKA2/RR, MTR-DYN, and RELAP5 codes. The chosen loss of flow accident scenario is a decrease in flow caused by a sudden stop of the pump power supply, while reactivity insertion accident is conducted by the withdrawal of control rods with maximum reactivity insertion of 32.33 ? 10?5 s?1. The calculated parameters are reactivity, reactor power, and temperature of the coolant, cladding, and fuel in the hottest channel position. In general, from a safety point of view, those computer codes were capable of performing the transient calculations with appropriate results. Based on the calculation results, during the transient condition of both reactivity insertion accident and loss of flow accident scenario, the reactor operation safety parameters do not exceed the allowable safety limit.

Research Progress of Reactor Thermal-Hydraulic Characteristics Under Ocean Conditions in China

Floating nuclear power plants are affected by sea wind and waves, which will produce various forms of movement, and cause changes in the thermal and hydraulic characteristics of the reactor core and threaten the safety of reactor operation. In response to the R&D and design requirements of floating nuclear power plants, the research progress on the thermal-hydraulic characteristics of reactors under ocean conditions in China are reviewed in this paper. The emphasis is put on flow heat transfer, bubble behavior, flow instability, and critical heat flux under ocean conditions. Research progresses as well as the issues that need to be focused on in the future research are discussed in detail.

ANALYSIS OF UNCONTROLLED REACTIVITY INSERTION TRANSIENT OF TRIGA MARK 2000 BANDUNG USING MTR PLATE TYPE FUEL ELEMENT

ANALYSIS OF UNCONTROLLED REACTIVITY INSERTION TRANSIENT OF TRIGA MARK 2000 BANDUNG USING MTR PLATE TYPE FUEL ELEMENT. Analysis of uncontrolled reactivity insertion is very important for the safety of reactor operations. Determination of melting point limit, critical heat fluxes and melting temperatures of cladding are the main objectives for most of these studies to determine whether fuel temperature can withstand the transient insertion of reactivity. In this study, uncontrolled reactivity insertion transient was carried out due to the withdrawal of control rods in nominal power of 1 MW and 2 MW. Analysis of reactivity transient was carried out using the WIMSD/5B and MTRDYN codes. The WIMSD/5B code is used to generate cross sections and the MTRDYN program is used for analysis under transient conditions. Based on calculations on the initial power of 1 MW and 2 MW with an insertion of reactivity of greater than 0.5 $/s the reactor operation is not safe because the fuel temperature exceeds the design limit. For reactivity insertion 0.5 $/s allows increased power can be stabilized by feedback reactivity. For 1 MW of nominal power, the maximum coolant temperature, cladding and fuel are 86.39 oC, 164.86 oC and 165.33 oC, respectively. For 2 MW of nominal power, the maximum coolant temperature, cladding and fuel are 89.09 oC, 176.96 oC and 177.602 oC, respectively. Based on calculation, It is concluded that the feedback mechanism can protect the fuel cladding from a local meltdown if reactivity insertion 0.5 $/s and the reactor is in nominal power of 1 MW and 2 MW.

Reactivity insertion accident analysis during uranium foil target irradiation in the RSG-GAS reactor core

Analysis of the steady-state and reactivity insertion accident is very important for the safety of reactor operations. In this study, steady-state and reactivity insertion accident analysis when the low enriched uranium foil target is irradiated in the reactor core has been carried out. The analysis is carried out by the best estimate method by using a coupled neutronic, kinetic, and thermal-hydraulic code, MTR-DYN. The MTR-DYN code is based on the 3-D multigroup neutron diffusion method. The cell calculations for the target are carried out by the WIMSD/5 and MTR-DYN code. After reactivity insertion, the coolant, fuel, and clad temperature are observed. The calculation results for the initial power of 1 W showed that the maximum temperature of the coolant, clad, and fuel are 49.76?C, 65.01?C, and 65.26?C, respectively. Meanwhile, when the reactivity insertion at the initial power of 1 MW, the maximum temperature of the coolant, clad, and fuel are 72.23?C, 140.79?C, and 141.97?C, respectively. Based on those calculation results during irradiation low enriched uranium foil target, the temperature in the steady-state and reactivity insertion accident does not exceed the allowable safety limit.

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR. Analysis of neutronic and thermal-hydraulics parameters of whole operation cycle is very important for the safety of reactor operation. During the reactor operation cycle, the position of the control rods will change due to reactivity changes. The purpose of this study is to determine the effect of control rods position on neutronic and thermal-hydraulics parameters in relation to the safety of reactor operation of the TRIGA 2000 reactor using silicide fuel of MTR plate type. Those parameters are power peaking factor, reactivity coefficients, and steady-state thermohydraulic parameters. Neutronic calculations are performed using a combination of WIMSD/5 and Batan-3DIFF codes and for thermal-hydraulics the calculations are done using WIMSD/5 and MTRDYN codes. The calculation results show that the reactivity coefficient values are negative for all control rod positions both at CZP and HFP conditions. The MTC value decreases when the control rod is inserted into the active core while the FTC value increases. The total ppf results and temperature in steady-state rise when the control rods are inserted of into the active core whereby the maximum value occurs at the position of the control rods of 20 cm from the bottom of the active core. The calculation results of ppf, reactivity coefficient, and thermal-hydraulics parameters lay below safety limits, indicating that the TRIGA 2000 reactor can safely use U3Si2-Al silicide fuel as a substitute fuel for cylindrical type fuel.Keywords: neutronic, thermal-hydraulic parameter, control rod effect, TRIGA 2000, silicide fuel.

Assembly-level analyses of an innovative long-life marine SMR loaded with accident tolerant fuel

A call for flexible power generation for a wide range of applications is gradually increasing, making Small Modular Reactor (SMR) a heat topic, especially for marine power. Unlike normal reactors, marine reactors demand a long lifetime up to 15 years without refueling, a high power-density and also a higher reliability and security. The standard PWR cannot achieve this goal. Here, a new 13 × 13 assembly loaded with Accident Tolerant Fuels (ATFs): U3Si2 – FeCrAl system, is proposed and analyzed on an assembly-level. U3Si2 is one kind of ATFs with high heavy metal density and thermal conductivity. It has a lower parasitic absorption. FeCrAl has a great radiation resistance, corrosion resistance and thermodynamic properties although with serious neutron penalties. It is currently the fastest researched accident tolerant cladding material. The fuel enrichment of this assembly is raised to 13%, prolonging the burnup to up to 95,000 MWD/tU, which means that the reactor can survive for more than 15 years at its rated power density. Using 15 B4C control rods can help achieve the beginning-of-life cool shutdown (K < 0.95). The power of the assembly is well distributed during the whole burnup. In consider of the inherent safety of reactor operation, the reactivity coefficients, especially Fuel Temperature Coefficient (FTC) and Moderator Temperature Coefficient (MTC) are computed and proved to be negative values, forming a negative feedback effect. The preliminary design is well-pleasing and it may be a better choice for future marine SMRs.

Estimation of the Reactivity Temperature Coefficients of the AP-1000 Reactor by MCNPX Code

The estimation of temperature coefficients of reactivity is a very important task because of its relationship to the safety of reactor operation. The temperature coefficient of reactivity includes the Moderator Temperature Coefficient (MTC), Doppler temperature coefficient (DTC) and the isothermal temperature coefficient (ITC). The present work investigated the temperature coefficients of reactivity for the advanced first core of AP1000 PWR. The calculations were carried out using Monte Carlo code MCNPX and the Evaluated Neutron Data File library, ENDF/B-VII.1 (ENDF71x), cross section data. The calculated results have been compared with the same results of KENO and VERA-CS codes. The study was investigated at a boron concentration of 1300 ppm. The MCNPX values of the MTC, DTC, and ITC were observed to be: (-1.27), (-1.48), (-2.75) pcm/F respectively. The results showed the compatibility between the MCNPX and KENO and VERA codes. Also the results demonstrate that the MCNPX code showed good estimates of Temperature Coefficients of Reactivity of AP-1000 PWR. The results can be used as reference for reactor physics studies.

Steady State Temperature Distribution Investigation of HTR Core

Reactor operation safety is highly related to fuel temperature parameter in the reactor core. To maintain the fuel in good integrity (no crack or melting), fuel temperature should be continuously within licensing criteria. Fuel temperature parameter is one of safety parameters in nuclear reactor operation. Fuel temperature is determined by local heat flux. High heat flux generation will cause a change in heat equilibrium and consequently result a change in fuel temperature. In the present study, steady state temperature distribution investigations of pebble bed for HTR-10 have been performed. The calculation is performed by using VSOP’94 code to do the calculation by which the core is divided into 50 components to represent the positions of various material compositions and to model fuel into 5 layers. The analysis is carried out based on the input data of reactor parameters, core specification, and fuel specifications as well as layer data of HTR-10. The results of this simulation in HTR-10 core at steady state indicated that the maximum temperature is 969.9°C at the solid material and 1031.3 °C in the fuel at the steady state condition. These temperatures are much lower than that of the maximum safety margin of fuel pebble, i.e. 1600°C, therefore, it can be concluded that TRISO fuel is able to contain all radioactive fission products.