Reactivity Insertion Accident Research Articles

Adoption of a fast dynamic model to simulate transient operations of MNSR

After developing a thermal–hydraulic model, it should be coupled to a point kinetic model to completely simulate transient behavior of a miniature neutron research reactor (MNSR). In this study introduction and validation of such a dynamic model which could be used as computing source of a micro-simulator, is presented. A time dependent point kinetic model in association to a simple control algorithm and reactivity feedbacks calculation (poisons and temperature) is adopted. In order to validate the model, it is compared against MNSR operational data in different scenarios such as normal fixed power operations, and reactivity insertion operations, and post shutdown scenarios. Result indicates that the coupled model could estimate the critical rod position for different power levels with an error less than 5%. Although the control algorithm is simple, it could predict the actual trend of control rod position accordance to the actual data. The coupled model simulates different reactivity insertion accident similar to the reactor operations in such a way that the error between estimated power peak, peak time, and long-term power is less than 5%.

Multiphysics analysis of a metal hydride moderated megawatt heat pipe reactor with burnable poisons

With the development of nuclear energy, microreactors have received increasing interest among researchers in recent years. In this paper, a megawatt heat pipe reactor with metal hydride moderators and burnable poisons is proposed. The hydrogen stability of the reactor under accident conditions, including reactivity insertion accidents, loss of power conversion unit heat sink accidents, and heat pipe failure accidents are analyzed. In this work, Gd2O3 is introduced as a burnable poison in the form of mixing with the UO2 fuel. According to the results of the burnable poison design, the 0.1% mass fraction of Gd2O3 is selected as the burnable poison loaded in the core. Safety analysis indicates that the introduction of burnable poisons can be beneficial during the positive reactivity insertion accident as it can reduce the excessive reactivity at BOL, thus reducing power peak and preventing hydrogen dissociation in ZrH1.4 rods. However, during the loss of PCU heat sink accident, ZrH1.4 rods will dissociate regardless of the presence of burnable poisons, whereas YH1.7 rods show better hydrogen stability. Moreover, in the event of the heat pipe failure accident, ZrH1.4 rods are more susceptible to dissociation than YH1.7 rods. As a result, the YH1.7+BP core is a better choice compared with other designs proposed in this paper as it provides a relatively high temperature margin.

Transient analysis of the safety characteristics on a super carbon-dioxide cooled micro modular reactor

The supercritical CO2 (S-CO2) demonstrates great potential as a coolant for nuclear reactors due to its low critical pressure and temperature. It offers improved thermal efficiency and decreased requirements for structural materials. By incorporating core physics models and updated thermal–hydraulic models, the SYST-SCO2 code developed in prior research has been enhanced and utilized for conducting transient safety analysis on the micro modular reactor (MMR) proposed by the Korea Advanced Institute of Science and Technology (KAIST). The analyzed transients include the Unprotected Loss of Flow Accident (ULOFA), Unprotected Reactivity Insertion Accident (URIA), Unprotected Loss of Heat Sink (ULOHS), Unprotected Over Cooling (UOC), and the Protected Loss of Flow Accident (PLOFA). It is found that the MMR experiences the exceeding of coolant outlet temperature limit of 676℃ in the ULOFA for over 155 s. With the reactor shutdown system put into operation, the reactor core can be maintained to operate with the fuel pellet, cladding and coolant temperatures below their limits respectively in the LOFA. Enough safety margin can be left for the MMR reactor core in URIA, ULOHS and UOC transients. Among the safety limits for the fuel and coolant, the core outlet temperature limit is the easiest to reach in the analyzed accidents.

Multiphysics fuel performance modeling of dispersed TRISO-coated particle fuel plate under long-time normal condition and accident conditions

TRISO-coated nuclear fuel has been extensively studied these decades due to the reliable accident tolerance performance. Based on the ABAQUS implicit finite element solver, the thermal–mechanical coupled simulations of the TRISO-particles embedded plate fuel were carried out. Multi-physical effects were incorporated into the simulation, including the plastic deformation, irradiation swelling, creep and gap heat transfer. Results show that the dispersed TRISO-coated particle fuel plate has robust thermal–mechanical performance under long-time normal condition. The maximum fuel temperature is 668 K and the cumulative failure probability for the SiC layer is 1.85 × 10−4. Under reactivity insertion accident and loss of coolant accident conditions, this fuel has reliable anti-accident capacities due to TRISO layer’s protection. Additionally, the porosity effect on thermomechanical performance of this fuel was analyzed.

Experimental investigation of gas-phase migration behaviors in a tight lattice bundle under two-phase non-equilibrium flow

During pressurized water reactor (PWR) operations, the two-phase flow may occur in local subchannels, such as the reactivity-insertion accident. The gas phase will migrate laterally to the surrounding subchannel, whose flow characteristics are still not clear but are very important for safety analysis. This paper conducted an experimental investigation on the gas-phase migration across the gap under two-phase non-equilibrium flow inside a vertical coupled subchannel tight lattice (P/D = 1.06). The subchannel injected with air is called the donor channel, and the other is called the receptor channel. 36 flow conditions were obtained (0.07 m·s−1 < 〈jg〉 < 1.02 m·s−1 and 0.93 m·s−1 < 〈jl〉 < 1.86 m·s−1), containing bubbly, cap-bubbly, and slug flow. In each case, four flow positions were measured (Z/Dh = 28.95, 57.90, 86.86, and 115.81). Results show the tight lattice bundles strongly restrict the gas-phase migrations across the gap. For bubbly flow, bubbles cross the gap in the modality of bubble clusters. The poor gas-phase migration was attributed to the bubbles' phase-interface tension and the lift force's action in the receptor channel. For cap-bubbly flow, the gas phase migrates across the gap by the coalescence and expansion of cap-shaped bubbles. The structure of the cap-shaped bubbles in the receptor channel is stable and does not break into small bubbles. Thus, the flow is still the cap-bubbly flow in the receptor channel as though the gas content is low. For slug flow, the gas-phase migrations are stronger than that of the bubbly and the cap-bubbly flows. As the flow fully develops, Taylor bubbles reach distribution symmetrically in both subchannels at Z/Dh = 115.81. Characteristics analyses indicate the two-phase non-equilibrium flow of this study promotes the bubbles’ aggregation under cap-bubbly flow. It will reduce void fraction under the two-phase non-equilibrium flow compared with that under the two-phase equilibrium flow for the same two-phase flux. Thus, bubble-size factors were recommended for consideration in the development of the two-phase flow model.

Transient thermal–hydraulic analysis of the thermionic space reactor TOPAZ-II

Space nuclear reactor system is of great significance in aerospace applications such as deep-space exploitation and exploration, high-power earth-orbiting satellites. A transient reactor system analysis code, RESYS, is developed for advanced non-light water reactor safety analysis recently. In this paper, the model of the thermionic space nuclear reactor TOPAZ-II including the thermionic fuel elements, thermionic energy conversion system, reactor core, coolant loop and radiator is established with RESYS code. The point reactor equation with six-group delayed neutron is utilized to calculate the fission power. The reactivity effect of various structural components and the reactivity insertion effects of control drums are considered. The thermionic emission model of cesium thermionic converter is based on the phenomenological model established by Rasor. Then, the steady state operation and three transient accidents including reactivity insertion accident (RIA), partial loss flow accident (LOFA), load resistance shortcircuit accident (LRSCA)are simulated. In the steady-state condition, the output electric power is about 5.8 kW with thermoelectric conversion efficiency of 5.0%. The results of steady calculation agree well with the design values. And in the three transient accidents, the moderator introduces a large temperature reactivity feedback which makes the reactor fission power increasing. The results indicate that the TOPAZ-II reactor system has a sufficient safety margin during these accidents. In addition, the calculation results preliminarily demonstrate the capability of RESYS code in modeling and simulation of thermionic space reactor.

Research on safety characteristics of heat pipe reactor with semiconductor thermoelectric conversion system

The heat pipe reactor exhibits notable characteristics of inherent safety and a compact configuration, rendering it a favored choice for space and ocean reactor designs. This study focuses on investigating the safety characteristics of the SAIRS-C space heat pipe reactor by employing a self-developed TAPIRS-D code under three selected typical accidents. The obtained results demonstrate that the heat pipe reactor exhibits inherent safety features, enabling it to sustain normal operation even in the event of partial heat pipe failure(2 out of 60 heat pipes fail) and partial heat sink loss(cooling rate decreasing from 100% to 10% in 10 s). The reactivity-insertion accident(0.358$ is introduced into the core) is found to exert the most significant influence on the heat pipe reactor. However, the crucial parameters of the reactor during the accident, such as fuel temperature, remain within the established safety limits. The outcomes of this research not only contribute to a comprehensive understanding of safety characteristics of heat pipe reactor, but also provide a crucial guidance for further design enhancements.

Methods for Predicting the Minimum Temperature of the Outage Loop and the Maximum Power Caused by the Low-Temperature Coolant

When the feedwater valve at the outage loop of the floating nuclear power plant leaks, thermal stratification occurs in the steam generator. It causes lower water temperature in the outage loop. The extent of hazard of this phenomenon cannot be directly determined by the existing measurement parameters, which poses a threat to the operational safety of the reactor. Therefore, this study adopts two routes: data-driven combined with safety analysis system (DSAS) and mechanism model-driven combined with safety analysis system (MSAS), to propose the prediction methods for the minimum temperature of the outage loop and the maximum power caused by the low-temperature coolant. Then, the actual data are used to verify these methods and the prediction results under different initial conditions are analyzed. The results show that both the DSAS method and the MSAS method can predict the minimum temperature of the steam generator in the outage loop and the maximum power when the outage loop is put into operation, but the DSAS method has better performance under certain conditions. These methods can provide guidance to the operators to avoid reactivity insertion accident.

Conceptualization of the micro research reactor cooled by heat pipes (MRR-HP), part-II: safety analyses

ABSTRACT The safety of the novel Micro Research Reactor cooled by Heat Pipes (MRR-HP) design is explored. This is achieved by developing a 3D thermal conduction model using STARCCM+ software, besides, coupling it with the 3D neutronics model for reactor dynamics. The MRR-HP has a unique design characteristic, which is the implementation of dual-sets of heat pipes (HPs) in the monolithic-block: Water-HPs for operation, and Potassium-HPs for accident/post-accident scenarios. Water-HPs cool the reactor effectively, allowing for maximum fuel temperature (K). However, water-HPs may suffer from: Loss-of-heat-Sink-Accident(LOSA), which may occur due to blackout, since it requires forced convective cooling in the condenser. Non-Condensable-Gases-Generation-Accident (NCGGA), due to radiolysis. These events cause significant melting in the monolithic-block and fuel rods after 3 and 6 hours, respectively. This paper proves the reliability of potassium-HPs as a passive countermeasure against LOSA and NCGGA, while natural convective cooling in the condenser is applied. Cold-state no-scram Reactivity-Insertion-Accident (RIA) scenarios with an insertion of 2.35$ are investigated, showing that no partial or significant melting occurs, due to the negative reactivity feedback of the MRR-HP. Besides, Potassium-HPs are essential in removing the decay heat, for warm-state RIA, after water-HPs dry out. This paper proves the robustness of the MRR-HP and encourages its utilization for remote operation.

Research on the Thermal Stratification Phenomenon at the Secondary Side of a Steam Generator in an Isolated Loop

In normal operation, there is no significant thermal stratification phenomenon in the tube bundle area of the steam generator of the nuclear power plant. However, this phenomenon is detected in marine nuclear power plant isolation loops in the presence of feedwater leakage. Therefore, an experiment was designed to verify the presence of thermal stratification, and the influence factors were analyzed by the RELAP5/MOD3.2 program. The results show that thermal stratification is mainly influenced by the initial water level at isolation, the initial pressure at isolation, the total leakage mass, and the internal circulation of the steam generator. Leakage flow has little effect on it. This phenomenon causes the coolant temperature at the primary side of the steam generator in the isolated loop to be lower than the temperature measured in the reactor inlet pipe or outlet pipe. This study aids the operator in determining the real coolant temperature at the primary side of the steam generator in the isolated loop and prevents reactivity insertion accidents when the isolated loop is put into operation.

CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors

CTF is a thermal hydraulic (T/H) subchannel tool that has been extensively developed over the past ten years as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program. The code was selected early in the CASL program for support of high-impact challenge problems that were found to be relevant to the nuclear industry and its currently operating fleet of pressurized water reactors (PWRs), including issues such as departure from nucleate boiling (DNB), crud-induced power shifts (CIPSs), and reactivity-insertion accidents (RIAs). By incorporating CTF into the multiphysics Virtual Environment for Reactor Application (VERA) core simulator software developed by CASL, CTF has become the primary means of providing fluid and fuel thermal feedback, as well as T/H figure-of-merits (FOMs) in large-scale reactor simulations. With the goal of solving industry challenge problems, CASL placed great emphasis on developing high-quality, high-performance, validated software tools that offer higher fidelity than what is currently possible with current industry methods. In support of this effort, CTF was developed from a research tool into an nuclear quality assurance (NQA-1)–compliant, production-level software tool that is capable of addressing the stated challenge problems and goals of CASL. This paper presents a review of the major technological achievements that were realized in developing CTF over the past decade of the CASL program and presents an overview of the code solution approach and closure models.

Three-dimensional transient analysis of Xi’an pulsed reactor by the coupled neutronics and thermal-hydraulics code

In this paper, an integrated code is proposed to simulate the transient thermo-neutronic behavior of the Xi'an Pulsed Reactor (XAPR) core during reactivity insertion accidents (RIA), which is developed by coupling the neutron transport kinetics code DAISY and a parallel-channel thermal-hydraulics code based on the internal (master–slave) coupling scheme. The few-group parameters adopted in the simulations including the seven-group cross-sections and six-group delayed neutron precursors are calculated by the NECP-X code. The XAPR pulse operation data are first employed to validate the ability of the coupled code to perform dynamic analysis, and then transient characteristics of different parameters such as core power and fuel temperature due to the different form and magnitude of reactivity insertion has been further simulated and analyzed. The numerical results indicate that the coupled code is correct and suitable for the transient calculation of reactivity insertion in XAPR and is able to capture the main physics behind fast RIA transients.

Development of transient thermal-hydraulic analysis code for SCO2-cooled reactor coupled with Brayton cycle and its application

Supercritical carbon dioxide (SCO2) direct cycle gas fast reactor system has attracted lots of attention from different institutions worldwide. Most of the research on SCO2-cooled reactors put emphasis on the design and optimization of core parameters, but few on detailed thermal-hydraulic analysis for SCO2-cooled reactors coupled with Brayton cycle. To supplement the lack of research in this area, a program based on Modelica language is established for transient simulations of SCO2-cooled reactor system. Then, the results of developed Modelica model are compared with code results obtained by GAMMA+, FROBA-SCO2 and SACOS-SCO2, which achieves good agreement. Based on the developed models and codes, thermal-hydraulic characteristics of SCO2-cooled reactor coupled with simple Brayton cycle under the conditions of reactivity insertion accident are studied. According to the validation and transient simulations for different reactivity insertion accidents, Modelica model developed in this paper shows great ability in getting more detailed transient performance of coupled SCO2-cooled reactor system under different working conditions.

Finite element based fuel performance investigation of U3Si2–FeCrAl design under normal and RIA conditions

U3Si2–FeCrAl system has been a promising Accident Tolerant Fuels (ATFs) candidate for its higher thermal conductivity, higher uranium density and lower oxidation rate. As a part of wide efforts to analyze the thermo-mechanical behaviors of the proposed ATFs, a finite element-based fuel performance code called BEEs, developed based on MOOSE (Multiphysics Object-Oriented Simulation Environment) framework, has been extended to analyze the behaviors of U3Si2–FeCrAl under reactivity insertion accidents (RIA). In this work, CABRI-REP tests were simulated to validate the modeling capacity of BEEs for RIA firstly. Then, with the fuel behavior models applicable to U3Si2 and FeCrAl implemented in BEEs, the thermo-mechanical response of U3Si2–FeCrAl to postulated RIA conditions were evaluated and compared with UO2-Zircaloy, UO2–FeCrAl and U3Si2-Zircaloy systems. The results show that U3Si2–FeCrAl system can reduce the fuel temperature (about 400–600 K) and delay gap closure during normal operation. The performance of cladding in U3Si2–FeCrAl system is comparable with UO2-Zircaloy and UO2–FeCrAl systems. However, the fuel central temperature of U3Si2–FeCrAl system is found to increase (about 110–180 K when RIA occurs at zero burnup, and about 170–190 K when RIA occurs after normal operation) compared to systems with UO2 fuel due to the lower volumetric heat capacity of U3Si2, which can lead to a threat to fuel security in RIA conditions, especially taking the lower melting point of U3Si2 (U3Si2 is ∼1938 K and UO2 is ∼3138 K) into account.

Advanced Couplings and Multiphysics Sensitivity Analysis Supporting the Operation and the Design of Existing and Innovative Reactors

Codes and methods are subjected to a continuous update process for answering the regulatory requirements concerning the long-term operation of existing reactors and new concept deployment. In this continuous improvement process, new generation codes are developed for supporting industrial applications and the long-term strategy. In this paper, attention is devoted to selecting codes under development in France for lattice and core steady-state and transient calculations. These codes, APOLLO3® and CATHARE3, have been selected for carrying out the activities of the H2020 CAMIVVER Project oriented to the 3D-multiphysics couplings improvements. Multiscale and multiphysics solutions are key topics to keep competitiveness and answer to newer industrial needs in plant operations, licensing, and safe operating envelope requirements. The paper presents an overview of the activities performed by Framatome to support the definition of benchmarks exercises proposed in the CAMIVVER Project. Small core configurations subjected to a Reactivity Insertion Accident (RIA) are presented with associated preliminary results. To open the discussions toward the development of Best-Estimate Plus Uncertainties (BEPU) solutions, the URANIE statistical platform was used for sensitivity analysis over different configurations. These preliminary results are also presented.

Inherent Safety Characteristics of Lead Bismuth Eutectic-Cooled Accelerator Driven Subcritical Systems

An accelerator driven subcritical system (ADS) is a new nuclear energy system which could not only produce clean energy but also incinerate nuclear waste. In this paper, inherent safety analysis of an ADS is performed with neutronics and thermal-hydraulics coupled code named ARTAP. Five typical accidents are carried out, including the cases of proton beam interruption, transient overpower, reactivity insertion, loss of flow, and loss of heat sink. The transient simulations are performed in the average channel and the hottest channel of the fuel pin in the ADS core. The simulation results for beam interruption show that the highest temperature of the pellet is in the middle of the fuel element in the average channel, while the peak temperature of the cladding is in the top of the fuel element. After the beam is interrupted for 20s, the maximum temperature drops at the fuel center, the cladding inner surface, and the outlet coolant in the hottest channel are 644.46K, 162.27K, and 136.42K respectively. For transient overpower accidents with the increase of beam intensity, the maximum temperature of the fuel and the cladding are below the safety limit. Concerning the reactivity insertion accident, it is found that the ADS has good inherent safety and its margin of criticality safety is large. The calculation results for loss of flow show that the power drop is small due to low sensitivity of the subcritical core to negative reactivity feedback, and the maximum temperature of the cladding reaches 1726K, which means the fuel element would rupture. However, the power and the temperatures of fuel, cladding, and coolant could decrease quickly to the safety level after the proton accelerator is cut off under a loss of flow accident. The results also show that the peak temperature of the cladding is lower than the safety limit under a loss of heat sink accident. The present simulation results reveal that the ADS has a remarkable advantage against severe accidents. It also implies that its inherent safety characteristics could ensure reactor shutdown by cutting off the proton beam after accidents occur.

Operation and safety analysis of space lithium-cooled fast nuclear reactor

As the power requirements for deep space exploration continue to increase, space nuclear power systems are imperative. In this paper, a set of models are established for the lithium-cooled space reactor combined with the Stirling engine. The space lithium-cooled reactor system is modeled, and the result of the steady-state is checked with the maximum relative error of 13.3%. Moreover, the characteristics under the unprotected reactivity insertion accident (URIA) and the unprotected loss of heat sink accident (LOHA) are obtained and analyzed. The results showed that: a) the solar heat flux cause the radiator temperature to fluctuate but has limited impact, b) under LOHA, the temperature of hot spot decreases rapidly from 1436 K to 1423 K, verifying the inherent safety of the system, c) under URIA, the hot spot temperature rises to 1546 K within 160 s. This work provides a solid basis for the design and analysis of space nuclear systems.

Comparative Safety Analysis of Accelerator Driven Subcritical Systems and Critical Nuclear Energy Systems

The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly different from those of the critical nuclear energy system (CNES). In this work, the safety characteristics of ADS are analyzed and compared with CNES by a developed neutronics and thermal-hydraulics coupled code named ARTAP. Three typical accidents are carried out in both ADS and CNES, including reactivity insertion, loss of flow, and loss of heat sink. The comparison results show that the power and the temperatures of fuel, cladding, and coolant of the CNES reactor are much higher than those of the ADS reactor during the reactivity insertion accident, which means ADS has a better safety advantage than CNES. However, due to the subcriticality of the ADS core and its low sensitivity to negative reactivity feedback, the simulation results indicate that the inherent safety characteristics of CNES are better than those of ADS under loss of flow accident, and the protection system of ADS would be quickly activated to achieve an emergency shutdown after the accident occurs. For the loss of heat sink, it is found that the peak temperatures of the cladding in the ADS and CNES reactors are lower than the safety limit, which imply these two reactors have good safety performance against loss of heat sink accidents.

Investigation of Control Characteristics for a Molten Salt Reactor Plant under Normal and Accident Conditions

A molten salt reactor (MSR) has unique safety and economic advantages due to the liquid fluoride salt adopted as the reactor fuel and heat carrier fluid. The operation scheme and control strategy of the MSR plant are significantly different from those of traditional solid-fuel reactors because of the delayed neutron precursors drift with the liquid-fuel flow. In this paper, a simulation platform of the MSR plant is developed to study the control characteristics under normal and accident conditions. A nonlinear dynamic model of the whole system is built in the platform consisting of a liquid-fuel reactor with a graphite moderator, an intermediate heat exchanger and a steam generator. A new control strategy is presented based on a feed-forward and feedback combined scheme, a power control system and a steam temperature control system are designed to regulate load changes of the plant. Three different types of operation conditions are simulated with the control systems, including transients of normal load-follow operation, a reactivity insertion accident and a loss of flow accident. The simulation results show that the developed control system not only has a fast load-follow capability during normal operation, but also has a good control performance under accident conditions.

Static and transient analyses of Advanced Power Reactor 1400 (APR1400) initial core using open-source nodal core simulator KOMODO

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.