Core Decay Heat Research Articles

Analysis of transient characteristics and design improvement of the passive residual heat removal system of NHR-200-II

NHR-200-II is a small integrated pressurized water reactor with 200 MW core thermal power. The core heat is transferred to two independent intermediate circuits via fourteen in-vessel primary heat exchangers (PHE), and the heat in the intermediate circuits is transferred to feedwater by two steam generators (SG) in the two intermediate circuits respectively. A passive residual heat removal (PRHR) branch is connected to each intermediate circuit to remove core decay heat under postulated accidents. During normal operation, PRHR branches are isolated by valves while SG branches in intermediate circuits are open. The valves in PRHR branches will be opened and the isolation valves of SG branches will be closed during decay heat removal scenarios. The decay heat removal capacity of NHR-200-II PRHRS could be seriously deteriorated once the isolation valves for SG branches fail to close, which was confirmed in a scaled integral test loop previously. Current understanding of PRHRS’s thermal-hydraulic characteristics with possible isolation failure in SG branches is limited. In this paper, the NHR-200-II PRHRS is modeled with RELAP5 considering the case of success and fail to isolate SG branches. A series of numerical simulations are carried out to study the impact of various parameters, such as the initial temperature, the size of the intermediate circuits’ header, and the initial flow direction in the intermediate circuits. Oscillatory flow is found when SG branches fail to be isolated under certain parameters combinations. An improved PRHRS design is purposed to eliminate possible flow oscillations, and the purposed improved design are tested by numerical simulations.

PWR natural circulation at a small break LOCA

In this paper, a one-dimensional quasi-steady model has been developed to analyse the potential impacts of reduced primary coolant inventory on natural circulation. This model focuses on the development of expressions for the coolant inventory, temperature distributions, and mass flow rate. The model encompasses a range of fluid flow scenarios, including single-phase, combined single-and-two-phase, and two-phase natural circulation, as well as the phenomenon of reflux condensation. The model is also utilized to examine the impact of reactor core power on natural circulation parameters. Comparisons with experimental findings and earlier research are made, highlighting a satisfactory level of agreement with the analytical findings. Based on the findings of this study, it has been observed that within the range of primary coolant inventory of 100–60% and core decay power of 1.5–5% of full power, natural circulation in a PWR type system has the potential to serve as an effective mechanism for transferring core decay heat to the secondary loop. The flow rates of natural circulation in pressurized water reactors (PWRs) are observed to have a significant dependence onthe amount ofliquid inventory, whereas their relationship with power level is comparatively less pronounced. The flow rate of a pressurized water reactor (PWR) system tends to exhibit its highest value within a specific range of inventory, typically falling between 60% and 80%. The transition between the different modes of natural circulation is observed to exhibit a smooth transition, with varying system inventories.

Unique safety features and licensing requirements of the NuScale small modular reactor

Small modular reactors (SMR) offer a novel approach to the construction and operation of nuclear power plants. The NuScale VOYGR™ plant uses a simplified SMR design that is based on proven light-water reactor technology with substantial improvements in nuclear safety. It consists of a 250 MWt reactor core housed with other primary system components in an integral reactor pressure vessel surrounded by a steel containment vessel, all of which is immersed in a large pool of water that also serves as the ultimate heat sink. At the core of the NuScale safety case are three primary safety systems: the decay heat removal system, the emergency core cooling system, and the containment. The ability of the NuScale Power Module (NPM) passive safety systems to remove core decay heat for an unlimited duration is demonstrated through analysis of a beyond-design-basis extended loss of AC power with no replenishment of water to the ultimate heat sink or operator actions. The NuScale methodology to evaluate an indefinite loss of AC power uses the proprietary NRELAP5 systems analysis computer code. Analysis results show that the reactor coolant system liquid level above the core is maintained and that containment pressure remains below the vessel design pressure. Once full passive air cooling is established, containment pressure and temperature will decrease over time with decreasing core decay heat. NuScale received standard design approval in September 2020 and design certification in January 2023 for its 50 MWe NPM configured as a 12 module plant. NuScale is currently seeking standard design approval to increase its core power to 250 MWt, nominally 77 MWe per module, in a 6-module plant (VOYGR™-6) configuration. The high-level safety of NuScale’s SMR technology is foundational to a new standard of nuclear power plant resilience.

New Approach for the Root Cause of Fukushima Meltdown Accident

The purpose of this Commentary is to clarify as much as possible the whole history of reactor and containment pressure changes in the Fukushima meltdown accident. It is based on a new approach for film boiling, which is kept after the Zr-H₂O reactions. Most important point of this approach is that the author applied film boiling based on boiling curve, which is basic theory in boiling phenomena, for the Fukushima accident phenomena. As the reaction rate is proportional to the reactor or containment pressure under film boiling, it increases rapidly and stops suddenly, keeping the film boiling. The containment pressure change consists of three phases, namely pressurizing, keeping the high pressure and de-pressurizing. The containment is pressurized by H₂ gas and steam produced by the Zr-H₂O reactions and de-pressurized by a heatsink such as the containment wall and inner shield concrete after reaction stops. The high pressure between these pressure changes is kept by balancing the H₂ gas produced by reaction with the leaked gas from the gap between the top lid and the containment. Core decay heat is large, but its change is negligibly small. So, the pressurization is calculated from H₂ gas and steam produced by the Zr- H₂O reactions. The heatsink balances with the reaction during the high pressure condition. The de-pressurization occurs after the reaction is over, so the reaction heat rate can be calculated by the heat rate of the heatsink, which is equal to the condensation rate during de-pressurization. The leak rate of the leak gas can be calculated using the reaction rate. It is very important that the rection rate is slowed by the insufficient steam supply, as the melted reactor cores in the Fukushima accident were covered with H₂ gas and steam (film boiling) at 0.8MPa or lower pressure. This is different from the rate (at approx. 7MPa) in the Three Mile Island accident, as the steam specific volume at 0.8 MPa is ten times larger than that at 7 MPa. The calculation results based on this assumption show that almost all the Zr in each core of Units 1, 2 and 3 reacted with water.

Separate-effect drainage tests in a core make-up tank

Passive safety systems are widely used in the advanced nuclear reactors, such as AP1000 and CAP1400. As an important component in the passive safety system, the core make-up tank (CMT) plays a key role for the safety injection and the core decay heat removal. Although the structure of the CMT system is simple, it involves many thermal-hydraulic phenomena, most of which significantly impact the peak temperature of the fuel cladding and the liquid level of the core. Since the CMT not only provides full-pressure injection but also triggers the auto depressurization system by its liquid level signal during small-break loss of coolant accident (SBLOCA), its thermal hydraulic behaviors significantly affect the performance of the passive core cooling system (PXS). Therefore, separate effect tests are inevitable to verify the feasibility of CMT applied to passive safety and injection systems and to study its operating characteristics. In this work, a separate-effect CMT test facility was designed and constructed. Experiments were performed to investigate the CMT's thermal hydraulic characteristics under the effects of system pressure, steam condensation, and flow friction. The study found that the delay drainage phenomenon is the result of a pressure balance between the pipe line and the gravity head. Whether delayed drainage occurs depends on the degree of subcooling of the fluid at the top of the CMT. The flowrate is not affected by the heat transfer coefficient of steam direct contact condensation. The flow friction of the drainage pipeline has no effect on the delayed drainage process, but has a great influence on the gravity drainage process.

Analysis of a Hypothetical Complete Loss-of-Coolant Accident in a Typical Research Reactor

In the present study, an analysis of a hypothetical complete loss-of-coolant accident in a typical open-pool research reactor is conducted. The reactor core is assumed to be completely uncovered and exposed to the ambient air. The possibility of passively cooling the decay heat of the exposed reactor core by natural convection to air and thermal radiation until core reflooding is investigated. A three-dimensional computational fluid dynamics analysis is conducted for the uncovered core while cooled by air natural convection and thermal radiation. The reactor core is simulated as a porous zone with decay heat generation specified as a cosine-shape distribution. The reactor core decay heat acts as a driving force for the coolant flow from the cold leg to the hot leg. The thermal equilibrium porous media model is used to represent the energy equation inside the core region. This study is conducted for core uncover time (the time interval between reactor shutdown and the moment when the reactor core is drained of water) of 10E3, 10E4, 10E5, 10E6, 10E7, and 10E8 s. Contour plots of temperature, velocity, density, and pressure at different values of core uncover time are illustrated. It’s found that for core uncover times of 10E3, and 10E4 s, the maximum core temperature exceeds the cladding melting point. The core maximum temperature is well below the melting point for uncover times of 10E5, 10E6, 10E7, and 10E8 s.

PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS

The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.

Multi-scale simulation of wall film condensation in the presence of non-condensable gases using heat structure-coupled CFD and system analysis codes

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.

Numerical study on improved design of passive residual heat removal system for China pressurizer reactor

The core residual heat removal system of the pressurized water reactor (PWR) is a vital facility to ensure the reactor safety, and its failure after shutdown will inevitably lead to the meltdown of the core. However, the existing core residual heat removal system of CPR1000 only guarantee the 72-hour grace period, after which the core decay heat can still melt the core within one day. The present study proposes an improved design of passive residual heat removal system (PRHRS) that utilizes natural forces to realize a passive self-start. Meanwhile, an innovative method is also proposed to use the atmosphere as the final heat sink in this paper. Subsequently a numerical study is conducted to verify the performance of the improved design during accident condition. Numerical results show that the improved design of PRHRS maintains a timeless safety cooling capacity during the station blackout (SBO) scenario. Finally, the parameter sensitivity study is performed to investigate a reasonable design value of chimney height. And numerical results prove that the proposed PRHRS is still capable of removing the core residual heat continuously even if the atmosphere temperature is variable. All component functions of new designed PRHRS are completely driven only by the action of natural factors without any external energy power. Therefore, the improved design of PRHRS can actually eliminate the risk of severe accident and lays the necessary technical foundation for the faster and larger-scale development of PWR.

True cause of Fukushima Meltdown Accident New Analysis about Loss of Core Cooling Based on Latest Data

The purpose of this paper is to clarify the true cause of the Fukushima meltdown accident from the viewpoint of reactor design. There are only two factors. One is that a total loss of battery power was not considered in the basic design, and the other is that the fail-close design was selected in long-term core cooling systems, as the confine function was preferred over the cool function in the three safety functions of stop （nuclear reaction）, cool （decay heat） and confine （radioactivity） in the basic criteria. After the loss of all power （battery and station power） in Unit 1, the IC system failed because of the fail-close design. After the loss of all power in Unit 2, the core-cooling function of the RCIC system failed because of the instability of two-phase flow in the turbine steam line, since the core decay heat was confined in the reactor and containment vessel owing to the fail-close design of the vent valves. As loss of station power occurred but battery power was available in Unit 3, operators were able to activate the RCIC system, which was tripped by the trip signal after one day. Then, the HPCI system was started automatically but stopped after about 10 hours because of the loss of battery power. After the loss of core cooling, core meltdown occurred in Units 1 to 3. The lessons learned from this accident are that the core-cooling system should be designed to withstand the loss of all power in its basic design and that the cool function should be preferred over the confine function.

New approach for describing reactor and containment pressure change after loss of core cooling at Fukushima meltdown accident

The purpose of this comment is to clarify the whole history of reactor and containment pressure change during the Fukushima meltdown accident. It is based on a new approach for film boiling, which is sustained after the Zr–H2O reaction. As the reaction rate is proportional to the reactor or containment vessel pressure under film boiling, it increases rapidly and stops abruptly while sustaining film boiling. The containment vessel pressure change consists of three phases, namely, pressurization, holding a high pressure and depressurization. The containment vessel is pressurized with H2 gas and steam produced by the Zr–H2O reaction and depressurized by heat removal by heatsinks such as the containment vessel wall and inner concrete after the reaction stops. The high pressure between these pressure changes is sustained by balancing the amount of H2 gas produced by the reaction and that of gas leaking from the gap of the top hat of the containment vessel. The amount of core decay heat is large, but the change of this is negligible. Thus, pressurization is calculated from the amounts of H2 gas and steam produced by the Zr–H2O reaction. The amount removed by the heatsink balances with that produced by the reaction during the high-pressure phase. Depressurization occurs after the reaction is over, so the reaction heat rate can be calculated from the heat removal rate of the heatsink, which is equal to the condensation rate during depressurization. The rate of gas leakage can be calculated from the reaction rate. It is very important that the reaction rate was slow owing to the insufficient steam supply, as the melted core in the Fukushima accident was covered with H2 gas and steam at a pressure of 0.8 MPa or lower. This is different from the rate (at approximately 7 MPa) in the Three Mile Island accident, as the specific volume of steam at 0.8 MPa is ten times larger than that at 7 MPa. The calculation results based on this assumption show that almost all the Zr in each core of Units 1, 2 and 3 reacted with water. The location of a small penetration hole produced by the contact of the high-temperature H2 gas with the suppression chamber wall, is estimated in Unit 2.

AP1000 IRWST numerical analysis with GOTHIC

The AP1000 Passive Residual Heat Removal (PRHR) system plays a significant role as it helps to remove the core decay heat, using the In-containment Refueling Water Storage Tank (IRWST) as a heat sink. The IRWST is located above the core, promoting natural circulation and allowing to be the mid-term heat-sink for the reactor core. The thermo-hydraulic pool behavior during an accident has an influence in the containment pressure and temperature, as happens in the pressure suppression pool in BWR reactors.Due to the importance of the IRWST performance on the AP1000 containment, a numerical analysis of a scaled experiment has been done using the GOTHIC code. Firstly, an IRWST scaled-down model is created, a mesh sensitivity analysis is performed, and the results obtained are compared against experimental results available in the literature. It is found the importance of the mesh discretization for the proper thermal-stratification modeling. Then, the reference model setup and mesh are applied in a full-scale prototypic model. The mesh influence on the thermal stratification is analyzed, as well as its impact on the AP1000 containment pressure. The mesh and setup for the full-scale model are selected to be implemented in a full containment 3D model in future works.

Analysis of the upflow conversion modification and influence on selected LOCA accidents in a PWR plant

The baffle jetting phenomenon, intensified by pressure difference across baffle plates, may cause excessive fuel rod vibrations and possible damage. The downward flow direction through the baffle-barrel region opposite to the upward flow through the core in pressurized water reactors affects increase of pressure gradient. Gaps between the baffle plates widen with reactor operating time and water jets become more pronounced. During the 2015 outage the NPP Krško has undergone the upflow conversion (UFC) modification in order to minimize baffle jetting and prevent fuel rod failure. The modification consisted in altering reactor vessel internals in such a way that the coolant downflow path in the baffle-barrel region was converted to an upflow path.The impact of the UFC modification on the nuclear power plant behaviour was analyzed in the paper. The double ended guillotine break of one direct vessel injection (DVI) line was selected as a representative scenario because the DVI line break is exhibiting the highest peak cladding temperature among the small break LOCA cases. The thermal-hydraulic analysis was carried out with the RELAP5/Mod3.3 code for core average channel conditions, while additional FRAPTRAN runs were performed to evaluate peak cladding temperatures for high power and high burnup fuel assemblies. Distribution of pressure and velocity fields in the reactor pressure vessel baffle-barrel region and change in bypass flow rates due to modification were determined by the ANSYS-Fluent calculation. The accident analysis has been performed for both the downflow configuration (pre-UFC) and the upflow configuration (post-UFC) to assess the influence of the modification. Additional cases with different reactor coolant pump trip times, higher core decay heat and reduced capacity of safety systems were also analyzed to evaluate the impact on reactor safety.

Integral effect test results on the effect of asymmetric supply of feedwater during a station blackout transient of pressurized water reactor

The thermal–hydraulic integral effect tests were performed to investigate the effect of asymmetric active and passive secondary cooling on the primary cool-down performance during a prolonged station blackout. The A1.1 test was performed to simulate a prolonged station blackout with asymmetric secondary cooling through the delayed supply of auxiliary feedwater only to steam generator number 2 of ATLAS. The A1.2 test was performed to simulate a prolonged station blackout with asymmetric secondary cooling through the supply of passive auxiliary feedwater only to steam generator number 2 of ATLAS. In both tests, an asymmetric cooling pattern was clearly observed during the period when the auxiliary feedwater was supplied. The primary system pressure and temperature continuously decreased during the heat removal by active and passive auxiliary feedwater supply, which indicates that the active and passive auxiliary feedwater supply can efficiently remove the core decay heat during a prolonged station blackout transient.

Numerical and experimental study on thermal-hydraulic behaviors in the secondary side of passive residual heat removal heat exchanger

Passive residual heat removal heat exchanger (PRHR HX), as a key component of the passive residual heat removal system (PRHRS) in the advanced reactors of CAP1400 and AP1000, should have sufficient heat transfer capacity to remove core decay heat during the accidents. In this paper, experiments and numerical simulations are constructed to explore the secondary side flow and heat transfer characteristics of a miniaturized PRHR HX that comprises 6 × 7 C-shaped tubes located in the center of a rectangular tank. In the three-dimensional (3D) numerical simulation, the C-shaped tubes are simplified as porous media and the drift flux model with the distributed resistance method is employed to simulate the two-phase flow in the secondary side of PRHR HX. The heat removal rate and mass transfer rate in the tank are estimated using empirical correlations. In the test, both the primary side and secondary side temperatures are measured. The bubbles generated on the tube outer walls are recorded using a high-speed camera. The local outer wall temperatures, total heat transfer rate and local water temperatures are all compared with experimental values to verify the simulation. The maximum deviation of the calculated heat transfer rate is less than 4.5%. The details about the secondary side flow conditions and bubble behaviors are analyzed.

CFD Investigation of MHTGR Natural Circulation and Decay Heat Removal in P-LOFC Accident

One of the key postulated accidents in a high-temperature gas-cooled reactor (HTGR) is the pressurized loss of forced circulation (P-LOFC) of the primary loop, which can be triggered by its primary helium circulator trip or turbine trip. If the reactor shutdown cooling system (SCS) fails during a P-LOFC accident, part of the reactor decay heat is absorbed by the reactor core materials and the rest removed by the reactor cavity cooling system (RCCS). In the extended period of P-LOFC accident compound with SCS failure, the core decay heat is supposed to be removed by conduction, natural circulation, convection, and radiation. A three-dimensional (3-D) computational fluid dynamics (CFD) simulation was performed in this research to study the long-term heat removal mechanisms in the General Atomics’ Modular High Temperature Gas-cooled Reactor (MHTGR) design in a P-LOFC accident. The reactor core temperature distribution and flow field were obtained at different decay power levels. The sensitivity of the natural circulation flow to the bypass gap width was investigated. The natural circulation flow intensity is relatively weak but very sensitive to the width of the bypass gaps.

Core makeup tank behavior investigation during ACME integral effect tests

Passive safety systems are widely used in the advance reactors, such as AP1000 and CAP1400. As an important component in the passive safety system, the core makeup tank (CMT) plays a key role for the safety injection and the core decay heat removal. Since the CMT not only provides the full-pressure injection but also triggers the auto depressurization system by its level signal during small-break loss of coolant accident (SBLOCA), the CMT thermal hydraulic behaviors affect the SBLOCA transient process. The thermal hydraulic behaviors of the CMT were investigated by conducting different SBLOCA tests on the ACME integral effect test facility, which includes the 2.5 cm, 5 cm and 20 cm breaks, and double ended direct vessel injection (DVI) line break. Based on the test observations, the CMT injection response to SBLOCA was divided into four phases: the circulation, the drain transition, the injection inhabitation and the final drain. The CMT injection characteristics and the dominate factors were scrutinized based on the phase division. It was found that the break condition has significant impacts on the CMT operating characteristics in each phase. The CMT circulation flow rate can be predicted, and the break size determines the circulation phase length. The CMT drain transition process is controlled by the PBL voiding process that has a close relation to the liquid level change in the cold leg. The ACC-CMT interaction was modeled, and the prediction result agreed with the test result. The comparison tests showed that the ACC nitrogen gas injection still affected the CMT drain behavior. A similar thermal stratification pattern formed during the different SBLCOA tests, and the thermal stratification remained not only in the CMT internal fluid but also in the CMT wall. The estimation of the flashing amount showed it did not play a significant role for the CMT liquid level decline, but it could induce the increases in the pressure difference between the CMT and the system and thus accelerate the CMT drain rate. The CMT internal fluid temperature determined the wall temperature, and a reverse heat transfer process occurred due to the flashing induced wall cooling during the ADS depressurization.

Quantification of the effect of Cr-coated-Zircaloy cladding during a short term station black out

Accident Tolerant Fuel (ATF) has the potential to improve the inherent safety of current reactors. In particular, an ATF Cr-coated Zirconium alloy (Cr-coated-Zr) cladding can be an effective short-term approach to improve the cladding oxidation resistance without changing the cladding base materials. In order to appropriately assess the effect of such an ATF concept it is important to quantify the performance of such a cladding improvement not only during normal operation but also during postulated accidents. In this work, we examine the effect of Cr-coated-Zr cladding (Cr-Zr) during a short-term station blackout (STSBO) accident as well as quantify the uncertainty in the use of ATF Cr-coated-Zr cladding. This is accomplished using a coated cladding model developed for the MELCOR severe accident systems code; i.e., MELCOR-1.8.6-UDGC model. First, we quantify the effect of the coated cladding using a set of base case simulations that show the improved oxidation resistance of the Cr-coated-Zr cladding with the UDGC model parameters. Then, we quantify the uncertainty in these simulations by a Monte-Carlo analysis that examines the effect of the ATF clad input parameters (e.g., oxidation rate and failure temperature) on key output parameters; i.e., timing of rapid hydrogen generation, clad temperatures and hot leg creep rupture. Finally, we use regression analysis methods to estimate the importance of these input parameters. These analyses show that the Cr-coated-Zr cladding can delay rapid clad oxidation and generation of hydrogen by 3600 s to 7200 s with associated delays in rapid clad temperature increases. This additional time can be of benefit for compensatory operator actions. In contrast there is little effect on the timing of hot leg creep rupture since this failure is mainly controlled by core decay heat. We also find that these results are not significantly affected by the uncertainty in input parameters, and based on the regression analysis, the high temperature Zr oxidation rate has the major influence on the selected output parameters for the Cr-coated-Zr cladding.

Decay heat analysis of system-integrated modular advanced reactor (SMART) using SCALE 6

In this study, the decay energy productions rate of the SMART reactor core and assemblies were analyzed by using SCALE/Triton depletion sequences. The fuel elements and assemblies were modeled based on technical and operational design data. Time and problem dependent cross sections were generated for the reactor and used in the fuel depletion and decay analysis. The heat generation calculations during operation showed an energy production of 21.6 MW at the beginning of the cycle, the contribution of decay heat to the total reactor power (P/Po) was 6.5% in the BOC and drops to 6.3% toward EOC. Post-shutdown calculations of the decay energy produced per assembly resulted in a decreased from 0.356, 0.369, 0.363 MW at shutdown to 0.261, 0.271 and 0.266 kW after five years, for assembly type A, B, and C respectively.Decay heat is generated mainly from short-lived fission products; the fast decay of these radionuclides results in an extreme drop in the decay heat generation hours after shutdown. In the first day after reactor shutdown, the ratio of decay heat to total reactor power (P/Po) drops from 6.3% to only 0.53%. System ability to remove this amount of heat from the reactor core during this period is crucial to the reactor safety. Loss of coolant and failure to remove decay heat is a substantial threat to fuel integrity and may lead to overheating and core meltdown.The reactor core decay heat was calculated for 30 years cooling period; results show that core energy decreases exponentially with time, from 20.8 MW to only 6.8 kW after 30 years. The decay heat as a function of time was calculated and compared to the total reactor power. The energy production of the top 30 heat generating radionuclides and their contribution to reactor decay heat were also estimated.

A comparative assessment of various single tray core catcher options for future sodium cooled fast reactors

Various single tray in-vessel core catcher options to handle whole core decay heat for a 1500 MWt capacity sodium cooled Indian fast reactor (SFR) have been analyzed towards their efficiency in rendering acceptable core catcher temperatures. For this purpose, a one dimensional transient multiphase heat transfer model that considers sodium boiling, fuel particles re-melting and slumping has been adopted. The boundary conditions for the one dimensional model are obtained from computational fluid dynamics (CFD) simulations of turbulent natural convective heat transfer within the lower sodium plenum of the SFR. It is found that the top surface of the debris bed dissipates more heat than the bottom surface of the core catcher. A single plate core catcher is seen to experience melting of the core catcher plate (CCP) for a short duration in case of CCP thickness greater than 80 mm. The heat shield plate (HSP) of the double plate core catcher is observed to undergo complete melting due to vapor locking between the HSP and CCP. The phenomena of melting and slumping of debris particles is seen in double plate core catchers with a refractory lining. However, safe temperature levels of HSP and CCP were observed in the case of double plate core catchers with refractory lining. This study suggests that a double plate core catcher with a delay bed is the potential option for future sodium cooled fast reactors, which are to be designed to handle whole core meltdown scenarios.