Unprotected Loss Of Heat Sink Research Articles

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.

Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Abstract Feedback reactivity caused by core deformation is one of the inherent safety features in a sodium-cooled fast reactor (SFR). To validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code, named Super-COPD, we have been conducting the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of balance of plant (BOP)-302R and BOP-301 as one of the representative issues in Experimental Breeder Reactor-II (EBR-II), a pool-type experimental SFR in the United States. During the transient of the ULOHS tests, the reactor power decreased to the decay heat level due to the negative reactivity caused by the radial expansion of the core. Developing the evaluation models of the reactivity feedback is essential for predicting the plant behavior in unprotected events, such as ULOHS events, and the numerical analyses modeling the feedback reactivity were conducted. By comparing the numerical results and the experimental data, the increment temperature profiles of the core inlet and the decreasing reactor power profile calculated by Super-COPD were comparable with experimental data. The applicability of the evaluation models for the reactivity feedback was indicated during the ULOHS event. In addition, through the sensitivity analyses on the cold pool model, it was found that the modification of the plenum model of the cold pool to take into account the thermal stratification was required to improve the core inlet temperature profile.

Research on core power maximization method of natural circulation lead-bismuth cooled fast reactor

To improve the inherent safety and cost-effectiveness of lead-bismuth cooled fast reactors, the SPALLER-100 reactor designed by the University of South China has been selected as the research object to determine the maximum power it can produce. This is a multi-objective, complex, multi-dimensional, nonlinear, and constrained optimization problem. To maintain the transportability, material durability, and long-term operation stability of the reactor core and ensure safety under accident conditions, three steady-state limitations and three accident limitations are proposed. The platform used to calculate the maximum neutronic power produced by the reactor at different core heights has been built using Latin hypercube sampling and the Kriging proxy model. Meanwhile, the cooling power of the reactor at different core heights is calculated by considering its natural circulation capacity. Finally, a design scheme is obtained that meets the requirements of neutronic and thermal-hydraulic assessments, while producing maximum power. Consequently, during the entire life-cycle of SPALLER-100, a safety analysis of three typical accident scenarios (unprotected loss of heat sink, unprotected transient over power, and unprotected coolant inlet temperature undercooling) is performed using a Quasi-Static Reactivity Balance (QSRB) approach. The results show that the platform used to calculate the maximum neutronic power exhibits high accuracy, and that the design scheme with maximum power is safe and economical. Overall, this study can provide reference ideas for designing natural circulation reactors that can maximize power output.

Improvement of a design of the molten salt fast reactor

This study provides an improved design of molten salt fast reactors to reduce the maximum magnitude velocity in the core. The results indicated that the improved design reduces the magnitude velocity with acceptable hot spot temperature where the maximum magnitude velocity is nearly equal to 4.59 m/s and the temperature of the hot spot is greater than the outlet temperature by 9°. The performed analyses are used to adjust the amount of fissile material started for keff = 1. Neutron flux and delayed Neutron distribution study where the position of the maximum Neutron flux is in the center and the position of the maximum thermal flux is in a fertile blanket. Simulations of unprotected loss of heat sink, unprotected loss of flow, unprotected transient overpower, and Unprotected over-cooling are performed and discussed with new geometry. Due to the negative reactive feedback, it is clear that MSFR has a high safety potential.

Safety Analysis of Small Modular Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100 Under Unprotected Transient

SNCLFR-100 is a small modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China, aiming at taking full advantage of the good economics and inherent safety of lead-cooled fast reactors to develop miniaturized, lightweight and multi-purpose special nuclear reactor technology. SNCLFR-100 is still in the conceptual design stage, in order to fully evaluate the safety features of the reactor and provide reference for the optimization design of the next stage, three typical transients are selected based on the analysis of the SNCLFR-100 initiating events by using the code Analysis of Thermal-hydraulics of Leaks and Transients (ATHLET), which are unprotected transient overpower (UTOP), unprotected loss of heat sink (ULOHS) and unprotected partial blockage in the hottest fuel assembly. For UTOP, the unexpected positive reactivity insertion of 0.7$ in 15s led to two large power peaks in the core quickly, and then the core power began to decrease and gradually stabilized under the action of various of negative feedbacks of the reactor, the peak temperatures of fuel and cladding rose rapidly with the increase of core power and eventually stabilized at a higher temperature. For ULOHS, as the reactor were driven by natural circulation, the coolant mass flow rate continued to decline after the transient, both core and cladding temperatures rose quickly and the temperature rise were smaller than that of UTOP transient, the reactor shutdown by itself and the peak temperatures of fuel and cladding were smaller than the safety limit. For unprotected partial blockage in the hottest fuel assembly, with the increase of the blockage rate of the hottest fuel assembly inlet, the coolant flow rate, the peak temperatures of coolant, fuel and cladding increased significantly, when the blockage rate increased to 0.9, the coolant flow rate of the hottest fuel assembly dropped to about 12.6% of the normal value, and the cladding peak temperature would exceed the cladding melting point, measures should be taken to avoid the happening of severe accident.

FAST (floating absorber for safety at transient) for the improved safety of sodium-cooled burner fast reactors

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

Multi-physics numerical analysis of the fuel-addition transients in the liquid-fuel molten salt reactor

On-line refueling is one of the most unique merits of the liquid-fuel Molten Salt Reactor (MSR), which can dramatically improve the capacity factor of a nuclear power plant. However, the liquid-fuel addition to the fluid will change the reactivity of the core and hence the transient behavior should be carefully analyzed to guarantee the reactor safety. Therefore, a multi-physics coupling code named TANSY has been developed based on the OpenFOAM toolkit for MSR. It is capable of modeling the drift of nuclides in the online refueling process, the drift of delayed-neutron precursors, and the multi-physics coupling. For the code verifications, two transients have been simulated, including the Unprotected Loss Of Flow (ULOF) and Unprotected Loss Of Heat Sink (ULOHS) of the Molten Salt Fast Reactor (MSFR). After verifications, some special transients caused by the on-line addition of the fissile and fertile fuel for the fast-spectrum MSR were analyzed. The numerical results demonstrate that the fast-spectrum MSRs have excellent safety performance under the fuel-addition transient process due to the large negative temperature coefficient of the salt.

Potential of FAST (floating absorber for safety at transient) as a solution for positive coolant temperature coefficient in sodium-cooled FAST reactors

Floating absorber for safety at transient (FAST) is a passive safety device for sodium-cooled fast reactors. It is designed to insert negative reactivity in case of coolant temperature rise or coolant voiding in an inherently passive way. This study investigates behavior of FAST device during the anticipated transients without scram (ATWS) in a sodium-cooled breed-and-burn fast reactor (B&BR). Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS) and unprotected transient overpower (UTOP) scenarios are simulated at both MOL (middle of life) and EOL (end of life) conditions of the B&BR to investigate the impacts of FAST on the reactor safety. It is confirmed that FAST can very successfully mitigate the consequences of ULOF, ULOHS and UTOP even in the B&BR core with strongly positive coolant temperature coefficient. In addition, this paper provides key aspects of FAST design optimization for practical applications.

Safety analysis of the FASt TEst reactor (FASTER) preconceptual design

The FASt TEst Reactor (FASTER) plant is a sodium-cooled fast spectrum test reactor design that was developed as part of an advanced demonstration and test reactor study, and provides high levels of fast and thermal neutron flux for scientific research and development. The 120 MWe FASTER plant has been designed with extended testing capabilities in mind, while keeping the design as simple as possible in order to make it efficient and cost attractive. The main function of the reactor is to provide neutrons for irradiation testing utilizing existing proven technologies to ensure a high technology readiness level. A preliminary safety analysis was performed of the FASTER design using the fast reactor safety analysis code SAS4A/SASSYS-1 to assess the reactor's safety performance during beyond design basis accidents. Three types of transient scenarios were analyzed for the FASTER design: unprotected transient overpower, unprotected loss of heat sink, and unprotected station blackout. These transients were simulated at beginning of equilibrium cycle and end of equilibrium cycle conditions. For each transient, failures of both the primary and secondary reactor scram systems were assumed. Because of the favorable reactivity feedbacks, relatively low peak fuel, cladding, and coolant temperatures were obtained during the transients and the reactor vessel temperature remained well below the limit for ensuring structural stability.

Transient thermal-hydraulic evaluation of lead-bismuth fast reactor by coupling sub-channel and system analysis codes

Abstract The lead-based fast reactor (LFR) is one of the most promising choices of the Generation IV reactors as it has better transmutation and breeding capability, as well as higher safety and economical efficiency. In this paper, the new concept LFR Pb-Bi cooled direct-contact-boiling water fast reactor (PBWFR) is studied. Coupled thermal-hydraulic analysis with the in-house system analysis code SACOL and sub-channel analysis code SUBAS is conducted using the one-direction multi-step coupling method. The unprotected transient of over power (UTOP) and unprotected loss of heat sink (ULOHS) accidents are simulated, and the key parameters are compared with the results by single SACOL, preliminary demonstrating the accuracy and rationality of the coupling method. The transient analysis shows the major concern for PBWFR’s operation is the cladding temperature. In UTOP, the maximum peak cladding temperature (PCT) reaches 834.15 °C, exceeding the cladding temperature limit of 800 °C, which suggests the reactor would be at a risk during UTOP. The maximum PCT is 679.43 °C in ULOHS. Due to the immediate reactor shutdown in ULOHS, the safety margin is relatively larger but the PCT is still higher than the limit in normal condition. Therefore, it is better to further optimize the structure or the operation condition of PBWFR. In addition, more detailed information of the core thermal hydraulic performance can be obtained. In this way, the coupled code could make up the shortage of single channel models and extend the function of the sub-channel analysis code.

Thermal-hydraulic analyses of passive reactor vault cooling system (RVCS) in PGSFR using MARS-LMR

KAERI has developed a prototype Gen-IV sodium cooled fast reactor (PGSFR), which is a pool type and metallic fueled fast reactor in Korea. Recently, a reactor vault cooling system (RVCS) was designed as a mitigation feature for severe accidents in the PGSFR. The RVCS which is located between the concrete and containment vessel is a passive cooling system which utilize an air natural circulation. The major goal of the RVCS is to keep the intactness of structures such as containment vessel and reactor vessel and to confine the radioactive materials inside of the vessel. The system has an influence on the reactor vessel temperature and the reactor vessel is an important structure in a reactivity feedback during a transient, since a control rod position is changed by the reactor vessel expansion. In this study, a performance test for the designed RVCS was conducted using an individual RVCS model. In order to check the effect of the RVCS performance under accident conditions, unprotected loss of heat sink (ULOHS) and total heat removal system failure (THRSF) accidents were analyzed with and without RVCS. The RVCS has a favorable effect on the reactivity feedback during the ULOHS because the RV expansion is reduced by enhanced cooling through the RVCS and the current RVCS design has enough capacity for mitigation of a severe accident condition. In addition, in terms of internal structure integrity, the damper opening time effect is also evaluated during the THRSF.

Transient Safety Analysis of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor Using RELAP5-3D

A transportable fluoride-salt-cooled high-temperature reactor (TFHR) design with 20-MW(thermal) rated power and 18-month fuel cycle is proposed for off-grid applications. One of the design goals of the compact reactor core is potential transport by truck, rail, or air. Full-core thermal-hydraulic analyses and improvements using three-dimensional computational fluid dynamics (CFD) were performed previously to demonstrate the feasibility of a TFHR design at a nominal power of 20 MW(thermal). In this paper, the best-estimate system code Reactor Excursion Leak Analysis Program (RELAP5-3D) is adopted to study the transient behavior of this TFHR design and the safety characteristics of the primary loop system during accident conditions. The modeling results of the steady state were verified using CFD results with consideration of radial heat conduction between heat transfer unit cells. Four most challenging accidents of anticipated transient without scram were analyzed, as well as parametric studies of some key factors. These accidents include unprotected reactivity insertion accident (URIA), unprotected loss of heat sink (ULOHS), unprotected loss of flow (ULOF), and a combination accident of ULOF and ULOHS. The results indicate that transient temperature limits are not exceeded during the most severe accidents. They indicate satisfactory transient performance of the TFHR design. The transient temperature limit of structure material Hastelloy N, based on embrittlement phenomena, poses the most limiting constraint due to the small temperature margin of about 20 K in the accident combination of ULOF and ULOHS. Overall, TFHR is a sound reactor design from a thermal-hydraulic viewpoint.

Stationary liquid fuel fast reactor SLFFR — Part II: Safety analysis

Safety characteristics have been evaluated for the stationary liquid fuel fast reactor (SLFFR) proposed for effective burning of hazardous TRU elements of used nuclear fuel. In order to model the geometrical configuration and reactivity feedback mechanisms unique to SLFFR, a multi-channel safety analysis code named MUSA was developed. MUSA solves the time-dependent coupled neutronics and thermal-fluidic problems. The thermal-fluidic behavior of the core is described by representing the core with one-dimensional parallel channels. The primary heat transport system is modeled by connecting compressible volumes by liquid segments. A point kinetics model with six delayed neutron groups is used to represent the fission power transients. The reactivity feedback is estimated by combining the temperature and density variations of liquid fuel, structural material and sodium coolant with the corresponding axial distributions of reactivity worth in each individual thermal-fluidic channel. Preliminary verification tests with a conventional solid fuel reactor agreed well with the reference solutions obtained with the SAS4A/SASSYS-1 code. Transient analyses of SLFFR were performed for unprotected transient over-power (UTOP), unprotected loss of heat sink (ULOHS) and unprotected loss of flow (ULOF) accidents. The results showed that the thermal expansion of liquid fuel provides sufficiently large negative feedback reactivity for passive shutdown of UTOP and ULOHS. The ULOF transient is also terminated passively with the negative reactivity introduced by the gas expansion modules installed at the core periphery. The maximum coolant temperatures have sufficient margins to the sodium boiling point for all three unprotected transient scenarios.

Tailoring the response of Autonomous Reactivity Control (ARC) systems

The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety of Generation-IV reactors while having a minimal impact on reactor performance and economic viability. In this study we present the transient response of fast reactor cores to postulated accident scenarios with and without ARC systems installed. Using a combination of analytical methods and numerical simulation, the principles of ARC system design that assure stability and avoids oscillatory behavior have been identified. A comprehensive transient analysis study for ARC-equipped cores, including a series of Unprotected Loss of Flow (ULOF) and Unprotected Loss of Heat Sink (ULOHS) simulations, were performed for Argonne National Laboratory (ANL) Advanced Burner Reactor (ABR) designs. With carefully designed ARC-systems installed in the fuel assemblies, the cores exhibit a smooth non-oscillatory transition to stabilization at acceptable temperatures following all postulated transients. To avoid oscillations in power and temperature, the reactivity introduced per degree of temperature change in the ARC system needs to be kept below a certain threshold the value of which is system dependent, the temperature span of actuation needs to be as large as possible.

Design study of long-life small modular sodium-cooled fast reactor

Summary This paper presents a new design for a small modular sodium-cooled fast reactor core with an optimized lifetime and reactivity swing through the analysis of various breed-and-burn strategies and its neutronic analyses in terms of active core movements, isotopic mass balance, kinetic parameters, and inherent safety. The new core design aims at a power level of 260 MW with a long lifetime of 30 years without refueling and a reactivity swing smaller than 1000 pcm. Starting from five initial candidate cores with various breed-and-burn strategies, an optimum core was selected from a combination of the two candidates that shows a proper breeding behavior with the optimized uranium enrichment in the low-enriched uranium region and the optimized size of the blanket region. The depletion analysis of the new core provides various reactor design parameters such as the core multiplication factor, breeding ratio, heavy metal mass change, power distribution, and summary of neutron balance. In addition, the perturbation analysis provides the reactor kinetic parameters and reactivity feedback coefficients for the inherent safety analysis of the core. The integral reactivity parameters of the quasi-static reactivity balance analysis demonstrate that the new core is inherently safe in cases of unprotected loss of flow, unprotected loss of heat sink, and unprotected transient over power. Copyright © 2016 John Wiley & Sons, Ltd.

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

ABSTRACTThe accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10−5 and 1×10−6 per plant operating year, respectively, which were selected based on the IAEA's safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.

CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Safety analyses of 10 MWth Forced Circulation LBE-cooled Fast Reactor during unprotected transients

In order to evaluate its safety performance, a 10MWth Forced Circulation LBE-cooled Fast Reactor (FCLFR) was simulated by a Neutronics and Thermal-hydraulics Coupled simulation program (NTC) under three transient accidents which were unprotected transient overpower (UTOP), unprotected loss of flow (ULOF) and unprotected loss of heat sink (ULOHS). The simulation results showed that the reactor could survive both ULOF and UTOP transients and a new steady state was re-established without human or active device intervention. The ULOHS simulation showed that the core power could decrease to the decay heat level due to the negative reactivity feedbacks. At the later stage of this transient, the core temperatures increased very slowly, providing sufficient time (at least 4500s) for the operator to activate the decay heat removal (DHR) system.

Transient analyses for a molten salt fast reactor with optimized core geometry

Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

Comparison of transient analysis of LBE-cooled fast reactor and ADS under loss of heat sink accident

A natural circulation LBE-cooled research reactor developed by Institute of Nuclear Energy Safety Technology (INEST) is proposed and designed to be 10MWth. The reactor has two operation modes, which are LBE-cooled critical fast reactor mode and LBE-cooled accelerator-driven sub-critical system (ADS) reactor mode. In this paper, protected loss of heat sink (PLOHS) and unprotected loss of heat sink (ULOHS) transient accidents were simulated for both the critical and ADS reactors above by NTC-2D code, aiming at the investigation of the safety advantages for the two reactor modes under this typical transient condition (loss of heat sink, LOHS). The results showed that during PLOHS for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the fast reactor increased smaller than those of the sub-critical one, which means the fast reactor had a better safety advantage under LOHS transient.