Emergency Core Cooling Research Articles

VIKTORIA experiments in the frame of R&D project on sump filtration during a loss of coolant accident

During a Loss of Coolant Accident (LOCA), in PWR’s, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the ECCS and CSS systems. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop. The main objectives are to investigate the head loss of the filter (physical and chemical clogging) for prototypic upstream debris source term in relevant thermal hydraulic conditions (water temperature and flow velocity on the filter surface) in compliance with the temperature profile and the chemistry of the water in sumps during a LOCA transient. For that, the VIKTORIA loop was equipped successively with two types of 2 m2 filters used in 900MWe NPP’s. The tests highlight the settling of the largest particles (concrete, painted chips) and part of the fibers; the transport of debris (roughly 55 to 65 % of the injected debris source term) leads to the physical clogging of the filter. The debris carried to the filter generate at 80 °C (with chemistry) a very quick increase of the pressure drop across the filter (≈7 kPa) that could be due to rapid chemical effects further to fibers corrosion. At the end of the 80 °C plateau, a pressure drop increase was observed due to the temperature decrease in agreement with the water viscosity evolution. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type; ≈1.5 kPa. Nevertheless, downstream debris source term appears to be more significant with the planar filter (compared to the filter with pockets) during the first hour of injection before the fibrous debris bed formation. We have also observed a more stabilized “cake”, during the (7 days) medium term evolution compared to that for rectangular pockets filter due to motion of debris inside pockets. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior.

Validation and parametric study of FFRD model in DRACCAR code

The recent revision of Emergency Core Cooling System (ECCS) regulations in Korea has necessitated the consideration of Fuel Fragmentation, Relocation, and Dispersal (FFRD) phenomena in nuclear reactor safety analyses. Consequently, the Korea Atomic Energy Research Institute (KAERI) has developed and integrated the FFRD model into the domestically licensed safety analysis code, SPACE. Globally, US NRC’s FRAPTRAN and IRSN’s DRACCAR are available for evaluating FFRD phenomena. The FRAPTRAN code incorporates the FFR model, developed by Quantum Technology, while the fuel dispersal model is currently not included. In contrast, the DRACCAR code functions as an integrated analysis platform capable of modeling multi-dimensional thermal, hydraulic, mechanical, and chemical phenomena during a Loss of Coolant Accident (LOCA). This study conducts a thorough examination of the FFRD model in the DRACCAR code and validates its applicability through analysis using the Halden IFA-650 tests. The results demonstrate satisfactory predictive capabilities. Furthermore, a parametric study of key FFRD model parameters enhances the understanding of the FFRD model in the DRACCAR code. The development of detailed physical models in the FFRD model could significantly enhance the performance of the DRACCAR code, warranting the establishment of a comprehensive framework for these advancements. In the future, code-to-code comparisons between the DRACCAR code and other domestically developed integrated analysis platforms will be conducted to investigate various phenomena in depth.

Dispersal of high-burnup fuel fragment surrogate particles during and after loss-of-coolant accident tests

The issue of fuel fragmentation, relocation, and dispersal is critical in the licensing and use of high-burnup (>62GWd/MTU) nuclear fuel in light water reactors (LWRs). In this work, two test series are reported that examine the fragment dispersal during a burst event and an additional dispersal following the burst due to vibrations in the rod, such as those induced by accident recovery systems. To examine dispersal during the balloon and burst portion of a Loss-of-Coolant Accident event, HfO2 fragments and yttria-stabilized zirconia pellets were filled into an as-fabricated cladding tube, which was then pressurized and subjected to loss-of-coolant accident testing in steam. Results from this test were found to be highly non-prototypic, and dispersal was both significantly more violent and significantly greater in magnitude than identified for actual fuel tests. These findings were attributed to the conservative (more dispersive) nature of the particles chosen for dispersal and to the details of the test conditions used that led to particularly wide bursts. The second set of tests examined dispersal following the burst when vibrations were induced in the rod, primarily via recovery activities such as Emergency Core Cooling System actuation leading to rapid water addition. Post-burst dispersal testing was performed by inducing sinusoidal oscillations with 2–25 nm peak-to-peak amplitude and 2–5 Hz frequencies in pre-burst rods that had been refilled with HfO2 fragments or high-burnup fragment surrogate mixture of HfO2 fragments and yttria-stabilized zirconia sands. Testing revealed that rods with large burst openings (7 mm wide in this work) led to unmitigated dispersal from above the burst zone but that smaller bursts (5 mm wide), although still much larger than the mean fragment size of 3 mm, led to effectively no dispersal because of interparticle locking. Additionally, mixture and moisture were found to impact the amount of dispersal: mixtures increased dispersal, and moisture drastically reduced it. The implications of these findings on likely dispersal from actual fuel are discussed.

Validation of coupled ATHLET-OpenFOAM simulation on a large-scale single- and two-phase flow experiment

Abstract System thermal-hydraulics (STH) codes are successfully used in the last decades for design and analysis of the physical behavior of nuclear power plants (NPPs). These programs provide reliable results at comparatively low computational cost, but have limited capabilities to predict complex 3D flow phenomena, which is possible with Computational Fluid Dynamics (CFD) tools. In order to combine the advantages of STH and CFD codes, algorithms to couple these are developed. At GRS, a new OpenFOAM solver, based on the volume of fluid (VOF) method, was developed together with new coupling boundary conditions to couple the CFD program OpenFOAM with the system code ATHLET. In a next step, verification and validation activities were carried out. These included the analysis of a Pressurized Thermal Shock (PTS) scenario. For this purpose, data from the OECD/NEA ROSA V Test 1.1, carried out in the Japanese Large-Scale Test Facility (LSTF) was used. The experiment investigated the temperature stratification after emergency core cooling (ECC) injection under single- and two-phase natural circulation conditions in the primary cooling circuit of a pressurized water reactor (PWR). This paper discusses the results of the coupled simulations, compares the numerical results with experimental data, and identifies areas for future improvements.

Analyses of the unavailability dynamics of emergency core cooling system

Abstract The Probabilistic Safety Assessment (PSA) studies can be used to evaluate the time-dependence of the safety systems availability. The purpose of this work is to analyze the unavailability dynamics of the Emergency Core Cooling System (ECCS). Using the PSA methodology, the ECCS was modeled. In order to show how the unavailability of the system varies as function of time (mission time, test interval and component ageing), sensitivity analyses are carried out. Following these analyses, polynomial functions of ECCS unavailability as function of time have been obtained. These functions allow to estimate the unavailability without modelling other ECCS fault trees being a simplified estimation of the system unavailability. In the same time, taking into account that the ECCS unavailability must be less than 10−2, the best time variations can be established to have the unavailability complied with this criterion. Also, the best time variations could be similarly established for any safety systems having the unavailability less than 10−3.

Post-quench ductility limits of coated ATF with various zirconium-based alloys and coating designs

Post Quench Ductility (PQD) of various Cr-coated cladding designs with respect to base cladding materials, Cr coating methods, and its thickness were investigated to explore Emergency Core Cooling System (ECCS) limits. Both inner wall and outer wall of coated (Cr and Cr/CrN) claddings were steam oxidized at ∼1204 °C and then water quenched. The post-oxidized specimens were ring compressed to attain offset strains based on which ductility was assessed in compliance with the U.S. Nuclear Regulatory Commission (U.S. NRC)’s protocols. Weight gain-based Equivalent Cladding Reacted (ECR) limits of coated specimens (Cr and Cr/CrN) are consistently lower than those of the base cladding materials due to the early load drop under Ring Compression Test (RCT). Yet, time needed to reach the ECR limit is still increased for coated specimens because it effectively undergoes single side oxidation with protective coating. Cracks imitated from ZrCr2 are primarily responsible for the early major load drop observed for tested coating thicknesses (8.9 µm and 18.9 µm), and hence reduced ECR limits compared to bare Zircaloy. The cracks from ZrCr2 was promoted by an increased oxygen concentration of the interfacing Zr matrix owing to the diffusion of oxygen through the coating. The thicker coating reduces this oxygen level, thereby delaying the load drop from ZrCr2 during RCT and increases the ECR limit. The maximum attainable ECR limit of coated ECR limits is affected by the base Zircaloy as it gives as the ECR ceiling from which ECR reduction for coated cladding occurs. For the tested base cladding materials (Opt. ZIRLO™ and HANA-6), ECR 19 %, which is the limit for Cr-coated (8.9 µm) HANA-6, may serve as the lower envelope limit. Hence, ECR 19 % (single-side ECR) can conservatively serve as the conservative, yet non-design specific, Cr-coated ECR limit for the most of the modern base Zircaloy materials (Opt. ZIRLO™ and HANA-6).

Fuel integrity evaluation for the shutdown BWR-4 plant during LOCA event

Two BWR units in the Chinshan stopped operation since 2017. However, the spent fuel assemblies still reside in the reactor core, and the plant staffs defined this configuration as the Mode 5 condition. If a Loss-of-Coolant Accident (LOCA) occurs in this condition, the water level will drop and then maintain at approximately 2/3 of the height of the reactor core. The upper 1/3 of the fuel will be exposed to air, and then the fuel cladding temperature will increase significantly. This study use RELAP5 to investigate the effect of different Emergency Core Cooling System (ECCS) to mitigate the LOCA. It is found that the core spray (CS) system can sufficiently suppress the fuel temperature. However, the temperature increase of the uncovered fuel becomes larger if the low pressure coolant injection (LPCI) system which does not provide direct core injection is operable. Based on the results, the peak cladding temperature of both cases are still below the acceptance criteria of 600 °C.

Application of REPAS to analyze the sump clogging issue following a LOCA and its impact on the reliability of the ECCS long-term core cooling function

Sump clogging has been identified as a relevant issue after an accident occurred in the Barsebäck-2 nuclear power plant in Sweden (1992). Following a steam line Loss of Coolant Accident (LOCA) due to the inadvertent opening of a safety relief valve, the jet stripped some insulation material from nearby pipes. The insulation debris were transported to the inlet of the strainers for the drywell spray system, thus clogging the intake. The accident was not serious but showed that the Emergency Core Cooling System (ECCS) could have failed. As a consequence, several actions have been undertaken by international organizations, regulatory bodies and nuclear power plant owners to characterize this issue, and propose solutions and improvements. In the present paper, the Reliability Evaluation of Passive Safety Systems (REPAS) methodology is applied to analyze the sump clogging issue and its effect on the long-term core cooling function. Originally developed to evaluate the reliability of passive systems, REPAS is here applied for the first time to an active system. The application is performed using the TRAC/RELAP Advanced Computational Engine (TRACE) code, developed by the USNRC, to simulate a generic three-loop PWR, with an active decay heat removal system.

Two-phase CFD simulations of COSI tests and evaluation of condensation distribution

This study presents the two-phase CFD simulation of experiments carried out on COSI facility in the frame of Pressurized Thermal Shock (PTS). PTS may occur during a cold-water injection, by the Emergency Core Cooling System (ECCS) during a Loss-Of-Coolant Accident (LOCA). The simulations were carried out using the Neptune_CFD code. A sensitivity study on Direct Contact Condensation (DCC) models was performed with a comparison with experimental data. At the end of this study, it appears that the turbulence of the free surface is rather low. Consequently, a low-turbulence heat transfer model is more appropriate. Then, two sensitivity studies on global and local mesh refinement are carried out, with the creation of an optimized mesh. This optimized mesh is used with the low-turbulence model to simulate the 40 test cases with no water level and co-current steam flow. Deviations in temperature and total condensation rate are less than 5% and 25% respectively. Finally, the condensation rate between stratified and non-stratified regions is evaluated. Condensation tends to occur in non-stratified zones, with imbalance that can be very high, reaching almost 80% of total condensation.

Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility

The safety injection (SI) system might be put into use, and the coolant at high temperature will mix with the supcooled water under Loss Of Coolant Accident (LOCA). The thermo-hydraulic phenomena associated with the thermal mixing of coolant and supcooled water will directly affect the judgment on core reflooding. China General Nuclear Power Group (CGN) independently developed and designed a thermal hydraulic system code named LOCUST, and the verification of thermal mixing models in LOCUST is necessary. This paper will introduce the thermal mixing tests at the T-junction based on Emergency Core Cooling System in Xi'an Jiaotong University (ECCS-XJTU) experimental facility, which were mainly conducted for the mixing between subcooled water injected from the SI pipe with a range of 25–125 kg/h in mass flow and the pure steam in the primary pipe with a range of 100–500 kg/h in mass flow. Besides, the fluid temperature and the dynamic vapor quality after mixing in tests are analyzed, and the simulations of 25 thermal mixing tests using the LOCUST 1.2 code are performed. The results show that the maximum relative error of LOCUST in mass flow of liquid is within 13.8 %, and the maximum relative error of LOCUST in temperature is within 8 %, which validates the reliability and accuracy of simulations of LOCUST for two-phase thermal mixing in LOCA.

Numerical Investigation into the Temperature Profile of a PHWR Channel Containing a Disassembled Fuel Bundle During a Postulated Accident Condition

A reactor core overheats due to decay heat generated in the fuel when an effective cooling medium is unavailable, such as in a loss-of-coolant accident combined with a loss of emergency core coolant. If the heat generated is not effectively dissipated, then at extreme temperatures, the structural strength of the bundle assembly may deteriorate, leading to slumping of fuel elements onto the inner wall of the pressure tube. It is essential to examine the temperature behavior of the channel containing fuel pins in a disassembled state in order to comprehend the impact of further thermally induced deformations in the channel during postulated accident conditions. Capturing the temperature of channel components at each circumferential position from experiments is extremely difficult; thus, a modeling tool is necessary to obtain a thorough circumferential temperature profile. This paper presents a numerical study that aims to study the temperature distributions in a 1-m-long pressurized heavy water reactor (PHWR) channel containing a disassembled fuel bundle. The channel geometry and the boundary conditions implemented were obtained from the experiment. A temperature profile for each channel element at every circumferential and axial location was obtained. A thorough comparison of the predicted and the reported experimental values was performed, and it was found that the predicted temperature behavior of the channel was consistent with the experimental data. Further simulations with different fuel element configurations and decay powers may be carried out; in addition, the results obtained may be used for coupled thermal-mechanical and thermal-mechanical-chemical simulations.

Effect evaluation on performance issues of passive safety system– Part Ⅱ. Passive emergency core cooling system

In this study, the effect of performance issues on the Passive Emergency Core Cooling System (PECCS), a type of Passive Safety System (PSS), was evaluated. The PECCS relies on natural phenomena such as gravity, density difference, and pressure difference to operate. Its driving force is lower compared to that of the active safety system, making it necessary to evaluate its performance in various transient situations. Major performance issues were identified through effect evaluation using a conceptual problem for the PECCS, and their effects on the system were analyzed using the system analysis code MARS-KS. Therefore, there are concerns about whether the PSS can sufficiently perform safety functions in various situations. In Korea, performance evaluation methodology for PSSs is being developed to dispel these concerns. As part of the methodology development, this study conducted the effect evaluation on performance issues. The study identified several major performance issues for the PECCS, including those related to pressure and heat loss of piping within the system, uncertainties in tank wall condensation modeling, and checks valve modeling. The evaluation of performance issues has led to the identification of the main factors to consider in PECCS modeling, which include: 1) the geometry and pressure drop of PECCS, 2) the occurrence of wall condensation in pipes and tanks, and 3) the behavior of check valves. Proper modeling of these factors is important in achieving accurate analysis results and preventing potential issues, such as premature check valve closure or incomplete coolant injection. In conclusion, the results of this study are expected to contribute to deriving safety analysis modeling guidelines for PSSs and design considerations for NPPs that utilize PSSs by improving the completeness of the developing performance evaluation methodology.

Investigation of the source term and radiological consequences of small break LOCA with failure of the emergency core cooling system in Bushehr nuclear power plant

In the present study, the in-containment source term derivation, environmental dispersion and deposition as well as radiological consequences of Bushehr Nuclear Power Plant (BNPP) due to Small Break Loss of Coolant Accident (SB-LOCA) have been studied. First, the calculation of in-containment release and distribution of the source terms due to the severe accident of SB-LOCA with failure of the active part of the Emergency Core Cooling System (ECCS) has been carried out. In this calculation, two scenarios, including with and without operator control actions, have been considered. Then, the environmental leakage of radionuclides as well as their dispersion and deposition considering two meteorological scenarios have been investigated. Consequently, public exposure has been calculated for each meteorological scenario and the obtained results have been reported and discussed. For two accidents with the same meteorological conditions, at a distance of 400 m from the release point, the received dose in the first 24 h is 18 and 342 mSv without and with operator action, respectively. The results show that due to the intervention of the operator and the use of emergency cooling supply systems, more water enters the reactor. Therefore, if the operator's actions do not prevent the core from melting, more source terms have entered the containment after the core melts and worse conditions than the baseline scenario without operator action occurred.

Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

The QUENCH-19 experiment was a first-of-its-kind full-bundle test simulating accident conditions followed by water quench on accident-tolerant fuel (ATF) cladding. A type of FeCrAl(Y) alloy, B136Y3, was developed at Oak Ridge National Laboratory and tested at the Karlsruhe Institute of Technology using Kanthal APM corner rods, a shroud, and Kanthal AF spacer grids. Testing conditions were similar to those in QUENCH-15—which tested ZIRLO cladding behavior—so that B136Y3 and ZIRLO cladding could be compared. QUENCH-19 consisted of an initial pre-oxidation heating followed by a transient. Then, a maximum power hold, which was not present in QUENCH-15, was executed to extend the heating period for the FeCrAl(Y) rods. Finally, a rapid water quench was executed that was similar to emergency core coolant system (ECCS) actuation. Compared with the ZIRLO rods in QUENCH-15, the bundle in QUENCH-19 released significantly less H2 (9.2 g vs. 47.6 g) and achieved a much lower maximum temperature (1455°C vs. 1880°C). Furthermore, no breakaway oxidation was observed in QUENCH-19. Metallographic mounts revealed that despite the symmetry of the setup, at elevations near the maximum temperature, cladding and thermocouples were heavily damaged, substantial melting and oxidation occurred, and the cladding underwent chemical interaction with the thermocouple sheaths. Additionally, the ZrO2 spacers detrimentally interacted with the cladding, leading to mixed oxide debris and the full destruction of some rods. Additional failure was found in certain cooler rods that may have risen due to the high thermal expansion coefficient of FeCrAl alloys. This paper presents an analysis of this work, which suggests that FeCrAl cladding can chemically survive anticipated loss-of-coolant accident events followed by rapid ECCS quench if the correct geometry and core design are present.

Numerical investigation of the external emergency coolant transportation inside reactor annular down-comer

A numerical investigation of the ECC (Emergency Core Cooling) coolant transportation in the reactor annular down-comer is performed in this paper. The density difference of coolant is paid special concern and is quantificationally analyzed. If the ECC coolant with low density and high velocity is injected into the annular down-comer, the horizontal kinetic energy decays as the azimuthal angle of down-comer increases. As a result, the ECC coolant transports through the annular down-comer along the clockwise direction. If the ECC coolant with high density and low velocity is injected, the coolant falls vertically and arrives the lower plenum as it enters the annular down-comer. As the lower plenum is filled with the ECC coolant, part of the ECC coolant moves upward due to the mass force effect. With the density difference, remarkable velocity oscillation is generated in the lower space of the down-comer. The transportation of ECC coolant insider the down-comer is decided by Froud number. When the Froud number is larger than 1.2, the inertia force dominant mixing is formed and the ECC coolant transports in the clockwise direction. When the Froud number is lower than 0.8, the buoyancy driven mixing is formed and the ECC coolant transports in the anti-clockwise direction.

Assessment of spray system capabilities in reduction of containment pressure and deposition of fission products during a severe accident in VVER-1000

In this study, we have investigated the spray system's capabilities to reduce fission products in the containment atmosphere and control the containment pressure in a VVER-1000 nuclear power plant. Large Break Loss of Coolant Accident (LB-LOCA) without access to the active part of the Emergency Core Cooling System (ECCS), as a worst-case severe accident, has been investigated as a case study. Only containment is simulated via the MELCOR code, while leakage of various materials from the nuclear power plant's core and the primary circuit is considered as the boundary conditions. Following the Fukushima accident in 2011, many nuclear safety regulatory organizations focused on using portable equipment that is essential for decay heat removal in nuclear power plants during severe accidents. Based on the literature studies, the accident management scenarios have changed recently since severe accidents inside nuclear power plants have worried many scientists and the public. Evaluating the current spray system in a severe condition based on which this system was not designed is vital and novel. We considered that the spray system would be fully restored and perform its safety function using the existing portable equipment to depressurize the containment and deposition of the fission products. To calculate the maximum allowed delay time of the spray system recovery, the accident was modeled without the operation of the spray system while a safety criterion was considered to avoid reaching the ultimate pressure of containment. For a more detailed evaluation of the spray system, two different operating modes including the activation of one spray train and the activation of both spray trains have been selected. According to the obtained results, if the spray system is revived before 4,000s from the beginning of the accident, this system would be able to reasonably mitigate the accident consequences. Moreover, without the need for any external water supply source for the spray system, this system can be permanently fed from the containment sump for at least 800,000s after the initiation of the accident, requiring the relevant heat exchanger of the spray system to be recovered for this period.

Large-eddy simulation on two-liquid mixing in the horizontal leg and downcomer (the TAMU-CFD Benchmark), with respect to fluctuation behavior of liquid concentration

Pressurized Thermal Shock (PTS) is induced potentially by the rapid cooling of the cold-leg and downcomer wall in the primary system of a Pressurized Water Reactor (PWR) due to the initiation of Emergency Core Cooling System (ECCS). Thus, fluids mixing in a horizontal cold-leg and downcomer should be predicted accurately. However, turbulence production and damping often hinders this prediction due to the presence of the density gradients. Hence, the Fifth International Benchmark Exercise, “the cold-leg mixing Computational Fluid Dynamics (CFD) Benchmark,” was conducted under the support of OECD/NEA. The experiment was designed for visualization of the mixing phenomena of two liquids with different densities. The heavy liquid was a simulant of cold water from ECCS, in a horizontal leg and downcomer. We used the Large-eddy Simulation (LES) to investigate the time fluctuation behaviors of velocity and liquid concentration. The CFD simulation was performed with two turbulence models and three different numerical meshes. We investigated the characteristics of the appearance frequency of the heavy liquid concentration with the statistical method. Based on our findings, we propose further experiments and numerical investigations to understand the fluid mixing phenomena related to PTS.

Impedance investigation of the surface film formed on aluminum alloy exposed to nuclear reactor emergency core coolant

A method was proposed for in-situ evaluating the thickness and resistivity of the oxide/hydroxide film formed on the surface of aluminum alloy exposed to sump water formed in the containment after a loss-of-coolant accident. The evaluation entailed fitting a model for the film impedance, which has film thickness and other variables describing the resistivity profile of the film along its thickness direction as fitting parameters, to the practically measured electrochemical impedance data. The obtained resistivity profiles implied that the films formed at pHs25°C 7, 8, 9, 10, and 11 all had a duplex structure; compared to the outer layer in contact with the solution, the inner layer of the film had a much higher resistivity and was inferred to be denser and provide most of the protectiveness of the film. Both the thickness and the total resistance of the film decreased with the increasing solution pH25°C, suggesting that the films formed in more alkaline solutions had less protectiveness against corrosion, consistent with the increasing aluminum alloy corrosion rates previously identified.

Water Relief Through Safety or Relief Valves is not Just a Leak

Abstract Two Nuclear Safety Advisory Letters (NSALs), advise operators of PWRs that they may assume, in accident analyses, that certain safety and relief valves can open, relieve water, and then reseat properly. The NSALs claim water that exits the reactor coolant system (RCS) can be made up by water that is delivered into the RCS by the emergency core coolant system (ECCS). For example, during an inadvertent operation of the ECCS (IOECCS) event, The NSALs claim, “since the cause of the water relief is the ECCS flow, the magnitude of the leak will be less than or equivalent to that of the ECCS (i.e., operation of the ECCS maintains RCS inventory during the postulated event and establishes the magnitude of the subject leak).” A comparison of ECCS flow to water relief, at relevant RCS pressures, indicates that water relief, during an IOECCS, cannot be dismissed as a leak. Critical flow calculations indicate that ECCS flow cannot replace RCS inventory that is relieved, as water, through the pressurizer relief and safety valves when RCS pressures are near nominal operating levels. At much lower pressures, the ECCS could offset the water relief. However, by then the IOECCS will have been either resolved, or will have developed into a loss of coolant accident (LOCA). The NSALs were published in 1993 and 2007. NRC did not question the NSALs’ advice until 2015, when it was found in a licensee’s application for a power uprating. The licensee might have submitted a false statement, in support of its application.

Prediction of radioactivity releases for a Long-Term Station Blackout event in the VVER-1200 nuclear reactor of Bangladesh

Consequences of an anticipated Beyond Design Basis Accident (BDBA) Long-Term Station Blackout (LTSBO) event with complete loss of grid power in the VVER-1200 reactor of Rooppur Nuclear Power Plant (NPP) of Unit-1 are assessed using the RASCAL 4.3 code. This study estimated the released radionuclides, received public radiological dose, and ground surface concentration considering 3 accident scenarios of International Nuclear and Radiological Event Scale (INES) level 7 and two meteorological conditions. Atmospheric transport, dispersion, and deposition processes of released radionuclides are simulated using a straight-line trajectory Gaussian plume model for short distances and a Gaussian puff model for long distances. Total Effective Dose Equivalent (TEDE) to the public within 40 km and radionuclides contribution for three-dose pathways of inhalation, cloudshine, and groundshine owing to airborne releases are evaluated considering with and without passive safety Emergency Core Cooling System (ECCS) in dry (winter) and wet (monsoon) seasons. Source term and their release rates are varied with the functional duration of passive safety ECCS. In three accident scenarios, the TEDE of 10 mSv and above are confined to 8 km and 2 km for the wet and dry seasons, respectively in the downwind direction. The groundshine dose is the most dominating in the wet season while the inhalation dose is in the dry season. Total received doses and surface concentration in the wet season near the plant are higher than those in the dry season due to the deposition effect of rain on the radioactive substances.