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This paper presents an assessment and comparison of the core degradation progression in four Station Blackout (SBO) scenarios analysed by the ASTEC computer code. Two types of SBO scenarios namely with high- and low-pressure conditions are considered. The low-pressure conditions in the SBO scenarios were simulated by introducing a small break loss of coolant accident in the cold leg with an equivalent internal diameter of 80 mm. The selection of the break size was based on the idea to have a significantly faster primary side pressure reduction for simulation of low-pressure conditions and a significantly longer coolant injection from the hydro accumulators (passive safety system). For both types of selected scenarios, the plant behaviour was analysed without operator actions, which allows assessing the times to reach the important set points during the accident progression. The set points include; a) the prediction of the dryout of the steam generators (SG) (or loss of SG effectiveness), b) the loss of natural circulation leading to core uncovery and heat up, c) the beginning of hydrogen generation, d) different stages in core degradation, e) the actuation of the passive safety system etc. The analyses were performed until reactor vessel failure takes place in both of the investigated scenarios. The purpose of these analyses is to study the reactor core behaviour parameters and to estimate the time available to perform operator actions. In addition, this investigation is focused on the assessment of the effectiveness of operator actions as prescribed in the Severe Accident Management Guidlines (SAMGs) for the investigated reactor type. The referenced NPP is KNPP equipped with two VVER 1000 V320 reactors. The ASTECv2.1.1.0 computer code has been used for the investigation. The aim of these analyses is to assess the possibility preserving the reactor core from damage during a severe accident and to assess the hydrogen generation that occurs as a result of the overheated core reflooding at high- and low-pressure initial conditions. The injection of a coolant by an active system will start at the same core exit temperature for each scenario, but at different pressure.

Since the reactor pressure vessel (RPV) of a typical BWR features a lower head that is penetrated by a forest of control rod guide tubes (CRGTs), coolability of the debris bed formed in the lower head during a severe accident can be realized by coolant injection through the CRGTs (so-called “CRGT cooling”). This paper is concerned with performance assessment of such CRGT cooling system, whose heat removal capacity is determined by two mechanisms: (i) heat-up and boiling of coolant inside the CRGTs; and (ii) evaporation of coolant which reached the top of the debris bed from CRGTs (top flooding). For this purpose, analyses were accomplished by coupling the COCOMO and RELAP5 codes, which simulate the quenching process of the debris bed and the coolant flow inside the CRGTs, respectively. An analysis was first carried out for a unit cell with a single CRGT, whose decay heat removal was limited by heat conduction from debris to the CRGT wall. The simulation indicated that without top flooding, though the temperature of the unit cell was eventually stabilized by the cooling of the CRGT wall, remelting of metallic debris (Zr) in the peripheral region was unavoidable due to low conductivity of corium. Boiling in the CRGT was not only beneficial to heat transfer, but also contributing to a flat axial temperature profile. Given the nominal flowrate of the CRGT cooling, the coolant was not completely boiled off in the CRGT, and therefore the remaining liquid water at the outlet of the CRGT was available for top flooding of the debris bed. The subsequent simulation including the top flooding showed that the debris bed was rapidly quenched without any remelting. However, the top flooding may have a side effect which was Zr oxidation risk at high temperature, leading to production of reaction heat and H2. Finally analyses were performed for prototypical cases for a reference Nordic BWR, and the results implied that the CRGT cooling could be used as a promising strategy for severe accident mitigation. It is critical that the debris bed is sufficiently cooled down during its formation so that the oxidation risk is eliminated when the CRGT cooling is applied.

The in-vessel retention by external reactor vessel cooling (IVR-ERVC) is a severe accident management strategy to hold or mitigate the molten radioactive material in the reactor pressure vessel (RPV). Highly turbulent natural convection is one of the most important phenomenon determining the thermal behavior of the molten pool. In the safety evaluation of IVR-ERVC, thermal behavior of the corium pool has been determined using experimental correlations. Property differences between the experimental simulant and the corium have been considered negligible. The effect of property differences represented by Prandtl number (Pr) on the thermal behavior of oxide pool is numerically investigated. The algebraic heat flux model (AFM) is used to simulate the turbulent natural convection and is validated with the BALI experiment. The Pr effect on the flow characteristics and the thermal behavior is investigated in the high Rayleigh number region. The present results show that the upward heat transfer tends to increase at the same Ra ′ in the case of the molten pool which is lower Pr fluid than the simulant material.

As molten corium is ejected from a reactor pressure vessel (RPV), the melt undergoes fuel-coolant interaction (FCI). Through the FCI molten corium is fragmented into small particles, and the fragments will form a debris bed stacking on the bottom of the cavity. Heat removal performance from both fragments and the debris bed is highly related to the size of fragments. Also, as FCI is a thermal-hydraulic interaction on the interface between the corium and the coolant, the estimation of the Sauter mean diameter is required. Thus, in this study, we develop a simplified strategy to predict the Sauter mean diameter with particle size distribution. The model is proposed to calculate the particle size distribution based on upper-limit log-normal distribution and the critical Weber number criterion. The calculation results are compared with 9 cases of FARO experiments and 7 cases of TROI experiments. The particle size distribution calculated with the suggested method well follows the trend of measured distribution from experiments. The estimated Sauter mean diameter varies from 1.58 mm to 3.13 mm, while the results for the ratio of the mass mean diameter to the Sauter mean diameter is almost constant to 1.55. It turns out that the suggested model gives good predictions of the particle size distribution and the Sauter mean diameter without large computational effort.

An important topic for long-term operation of nuclear plants is aging of plant concrete structures. The containment building, biological shielding, and support concrete are examples of concrete structures that are of primary importance in the operation of a nuclear plant. These and other safety-related structures must be capable of maintaining structural capability for the operating life of the plant. Demonstration of the satisfactory condition of the concrete structures is required for safe operation, particularly when plant operation beyond 60 years is considered. The concrete biological shield (CBS) wall is particularly important because of its shielding and structural functions and that it is irradiated due to its proximity to the reactor pressure vessel. Demonstrating that the concrete remains structurally able to perform its intended function is vital in effective aging management. Neutron irradiation above certain thresholds can cause loss of tensile and compressive strength of the concrete and volumetric expansion. If the irradiation level exceeds the threshold, both loss of strength and volumetric expansion should be evaluated. Methods to reliably evaluate the effects of the loss of strength and volumetric expansion are required.

The Karlsruhe Institute of Technology (KIT) is engaged in safety-relevant investigations of a generic SMART-plant using available public data. Within this work, a boron-free core that fits into the reactor pressure vessel (RPV) of the SMART-plant has been developed and optimized at the beginning-of-life condition from the safety perspective that fulfils general regulatory requirements. In this paper, the analysis of the behavior of a boron-free core integrated into a generic SMART-plant under a steam line break (SLB) accident using the coupled code TRACE/PARCS is presented and discussed. The SLB-accident in a conventional PWR is an overcooling accident that could lead to re-criticality and return-to-power after the reactor SCRAM. It is characterized by an asymmetrical cooling behavior due to the break of one of the steam lines while others are intact leading to a strong radial non-symmetrical core power distortion. Hence, to capture this physical phenomenon, 3D neutronics and thermal-hydraulics codes are applied for the analysis. The performed investigation showed that the boron-free core does not experience any re-criticality and a return-to-power in comparisons with a conventional PWR due to the low severity impact of the SLB-accident. That is due to the following factors: (a) the unique helical steam generator (SG) design concept, in which there are eight SGs integrated within the RPV, and almost negligible coolant inventory of the secondary side; and (b) the flow mixing header assembly located around the core that is developed to reinforce the coolant mixing within the downcomer. The core decay heat has been proven to be removed passively after the reactor trip thanks to the establishment of natural circulation in both the primary and secondary-side, and the excellent performance of the passive residual heat removal system (PRHRS). Generally, it can be stated that the boron-free core integrated into the generic SMART-plant under an SLB-accident does not threaten the safety limits. Moreover, the SLB-accident is analyzed assuming the failure of all safety systems including the PRHRS. Such a hypothetical accident revealed that the grace time to reach core uncovery is about two hours without any human intervention and safety system actuation.

ALLEGRO is a demonstration reactor of the GEN IV GFR2400 gas fast reactor concept which is currently in the pre-conceptual design phase. In the present ALLEGRO design, the hot duct (leg) – which connects the reactor with the main heat exchanger – is located inside the cold duct. For this reason, the break of the hot duct does not lead to a classical loss of coolant accident (there is no depressurization of the system), but a core by-pass occurs, which may threaten the integrity of the core. The peak cladding temperature (PCT) highly depends on the pressure loss between the hot and the cold ducts at the break. In this study, by an original approach we aimed to estimate the value of the pressure loss coefficient between the hot and the cold ducts by using the ANSYS FLUENT CFD (Computational Fluid Dynamics) code as a numerical tool. The results showed that if the break is located in the hot duct halfway between the reactor vessel and the main heat exchanger, the pressure loss coefficient value falls in the range of 1.05–1.15 for break sizes up to 100%. Considering the primary importance of the hot duct break pressure loss coefficient, further studies are proposed to assess the worst case break geometry.

In-vessel retention (IVR) through passive external reactor vessel cooling (ERVC) has been adopted as one of the key severe accident mitigation strategies for Chinese-designed large-scale advanced passive pressurized water reactor. From the view of thermal load effectiveness, critical heat flux (CHF) of reactor pressure vessel (RPV) out wall has been taken as one of the key criteria for IVR-ERVC feasibility. In order to investigate the ERVC limits, as well as its performance during IVR-ERVC process, a campaign of CHF tests under real plant conditions is conducted on the full-height REPEC-II facility, which has been designed according to the prototypic ERVC configuration and verified following an integrated, hierarchical scaling approach. Based on the abundant results acquired in the tests, it is attempted, in this paper, to summarize and evaluate ERVC performances and trends under various practical engineered conditions. The main evaluation results includes: influences on CHF of RPV out wall with various non-uniform heat load distribution and ERVC channel configurations, effects of subcooling, heating surface conditions (prototypic surface material and its oxidization) and water chemistry on external cooling limits, etc. These are expected to help designers with reliable estimation on impact of some related engineered factors on real IVR-ERVC performance.

In-Vessel Retention (IVR), which arrests relocated molten core materials in the vessel during severe accident, is an appealing accident management approach to many newly-designed reactors. It is implemented in CAP1400 because it’s highly compatible with CAP1400 design philosophy. Extensive studies of relevant phenomena are carried out to investigate the possible effect on IVR strategy, which include core melting and relocation, in-vessel steam explosion, corium material interaction, heat flux to RPV (Reactor Pressure Vessel) wall under the assumption of different corium pool configuration, ex-vessel CHF (Critical Heat Flux) test, RPV structural analysis and so on. For those that may have negative impact on IVR success, the corresponding design improvements are conducted to aid and facilitate the employment of IVR, such as lowering core support plate, increasing the mass of core internals, optimizing ex-vessel insulator and so on. The benefit of taking decay heat from in to vessel cooling is also achieved by reflooding breaks or injecting RCS (Reactor Coolant System) under the instructions of SAMG (Severe Accident Mitigation Guideline). All the IVR-related phenomena and operator actions are reflected and linked together by IVR decomposition event tree (DET) to evaluate CAP1400 IVR strategy in a comprehensive manner. There is reasonable assurance that the IVR strategy implemented in CAP1400 is successful. About 93% of core damage sequences can be terminated by retaining corium in the vessel.

Realistic numerical predictions of dynamic, complex, multidimensional flow scenarios occurring in light water reactors are vital prerequisites for every reliable safety assessment study. System analysis codes employing only one-dimensional fluid-mechanics models may fail to realistically capture the complex flow conditions in large scale passive components of light water reactors. Even though today’s CFD codes are capable of accurately representing multidimensional flow behaviour, they require huge computational efforts to simulate long-term transients. The German thermal-hydraulic system code (THS) ATHLET (Analysis of Thermal-hydraulics of LEaks and Transients) part of the AC2 simulation environment, is being developed by GRS (Gesellschaft fur Anlagen- und Reaktorsicherheit gGmbH) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design-basis accidents anticipated for nuclear energy facilities. The code provides a wide range of specific models for several light water reactor designs of Generation II, III+ and IV. Currently, the code’s one-dimensional six-equation two-fluid model, is being extended towards a fully three-dimensional set of thermal-hydraulic conservation equations. Until now ATHLET’s ability to capture multidimensional flow behaviour has mainly been assessed for the flow mixing behaviour in the reactor pressure vessel. Within this work, ATHLET’s capabilities to capture the multidimensional flow behaviour will be assessed for the flow within large scale passive components of light water reactors. These analyses highlight the advantages of ATHLET’s newly developed 3D model for the simulation of large scale passive systems.

Anticipated Transients without Scram (ATWS) are evaluated to demonstrate the capability of the barriers to prevent fission product release. Swiss nuclear authority guidelines allow the use of best-estimate codes and methodologies for beyond design basis events analyses, such as ATWS, to simulate more realistic plant responses. Thus, realistic initial and boundary conditions may be applied so that all systems unaffected by the event are credited and no independent single failures need to be assumed. The RELAP5-S3K coupled thermal-hydraulics/neutronics code was used for KKM (Muhleberg BWR-4 NPP) ATWS analyses. The KKM RELAP5-S3K model was demonstrated to be valid for a wide range of transients including Loss of Coolant Accident (LOCA) and ATWS. Two ATWS cases, a main steam isolation valve (MSIV) closure and a loss of offsite power (LOOP), were simulated to demonstrate the KKM plant capability. The MSIV closure is a limiting ATWS event for reactor pressure vessel (RPV) overpressure and torus pressure and temperature. After closure of the MSIV, all RPV outlet flow is directed to the torus through the safety relief valves resulting in limiting torus pressures and temperatures comparable to other events. The LOOP event can be a limiting ATWS event for peak cladding temperature (PCT) if insufficient makeup water is available. The simulation results showed that the ATWS acceptance criteria for PCT, reactor peak pressure, and torus temperature and pressure are satisfied. The RELAP5-S3K results have been compared to the ATWS results from KKM’s fuel vendor, General Electric-Hitachi (GEH), using their TRACG code (TRACG being perhaps the most advanced system code available for BWRs). The plant response predicted by the RELAP5-S3K model is similar to that reported by GEH. Variations between the two are attributed to (1) code differences in modeling of the physical phenomena and (2) different levels of detail in modeling the KKM plant and core parameters, e.g. cross-sections. Considering that an independently developed model using the RELAP5-S3K code predicts plant behavior during an ATWS similar to the state-of-the-art code TRACG, it is concluded that the KKM ATWS analysis with RELAP5-S3K provides trustworthy information for comparison against safety criteria.

Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel to fragment and quench core melt and provide long term cooling of the debris. One of the risks associated with this strategy is early containment failure due to ex-vessel steam explosion. Assessment of the risk of steam explosion is subject to significant (i) epistemic uncertainties in modelling and (ii) aleatory uncertainties in scenarios of melt release. For quantification of the uncertainties and the risk a full model (FM) based on TEXAS-V code and a computationally efficient surrogate model (SM) have been previously developed. FM is used to provide a database of solutions that is used for development of a SM, while SM is used in extensive sensitivity and uncertainty analysis. In this work, we compare the risk of containment failure with non-reinforced and reinforced hatch door for metallic and oxidic melt release scenarios. We quantify the error of SM in the approximation of the FM and assess the effect of the approximation uncertainty on risk assessment. We analyze the results and suggest a simplified approach for decision making considering predicted failure probabilities, expected costs, and scenario frequencies.

This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.

A primary failure mechanism of reactor pressure vessel (RPV) steel is the embrittlement caused by fast-neutron irradiation. In this work, neutron irradiation tests were performed on an RPV steel at a high temperature (292 °C) using a neutron irradiation test reactor. In addition, its magnetic properties were measured before and after irradiation in a hot laboratory. These measurements show that (i) the clockwise variation of the magnetisation for the hysteresis loops of RPV steel before irradiation was lower than that after it; (ii) sample magnetisation before and after irradiation was not sensitive to changes in temperature, with only a very small hysteresis effect; and (iii) when irradiation damage did not exceed 0.154 dpa, the residual magnetisation intensity of the RPV steel was linear to the radiation fluence, and the initial magnetic susceptibility exhibited an exponential relationship with the radiation fluence. Finally, the magnetic properties of the RPV steel can be used to characterise the degree of its irradiation-induced embrittlement, and measurements of magnetic properties can be used for non-destructive evaluation in monitoring the degree of irradiation damage experienced by in-service RPV steel.

The Rossendorf Coolant Mixing Model (ROCOM) boron dilution benchmark experiment was assessed using the CUPID code in the framework of a Coordinated Research Project by the International Atomic Energy Agency, namely, “Application of Computational Fluid Dynamics Codes for Nuclear Power Plant Design.” Two-equation Reynolds-Averaged Navier–Stokes (RANS) turbulence models, such as the standard k-e model, Re-Normalization Group (RNG) k-e model, and Shear Stress Transport (SST) k-ω model, were tested, and predictions of boron concentration were compared to measurements at the downcomer and core inlet in the reactor pressure vessel. The analysis result showed that the de-borated water slug injected from one cold leg was diffused by the turbulence mixing effect as it flowed downward in the downcomer and passes through a perforated drum in the lower plenum. The applied turbulence model affected the mixing patterns of the de-borated water, which resulted in a local distribution of boron concentration at the core inlet. This was the main concern of the boron dilution phenomena from the viewpoint of safety analysis. The result showed combining the CUPID code with the RANS turbulence models could reasonably predict the boron dilution phenomena in the reactor vessel when the proper y-plus wall treatment was assured, considering the characteristics of the turbulence models.

In light water reactor (LWR) nuclear power plants, the reactor pressure vessel (RPV) plays an essential safety role, and its integrity must be ensured during a variety of transient loading conditions. These can include off-normal conditions such as a pressurized thermal shock (PTS), as well as transients encountered during normal startup, shutdown, and testing of the reactor. Exposure to irradiation and elevated temperatures embrittles the RPV’s steel over time, making it increasingly susceptible to failure due to propagation of fractures that could initiate at the locations of flaws introduced during the manufacturing process. As long-term operation scenarios are being considered for LWRs in the United States, it is important to have a flexible simulation tool that can be used to perform probabilistic evaluations of RPV integrity under a wide variety of conditions and incorporate improved predictive models of RPV steel embrittlement. The Grizzly code is being developed to meet these needs. This paper describes Grizzly’s modular architecture, provides results of benchmarking studies of various components of Grizzly, and demonstrates the application of Grizzly on a model that includes plume effects that are difficult to represent in other codes being used in current practice.

As an accident management procedure of light water (nuclear) reactors which are situated along sea shore, sea water will be injected into the reactor pressure vessel to flood the nuclear fuel which is heated by residual heat. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Precipitation behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with condensed (two and half times denser) sea water and a mixture solution of sea water and borated water. Three-dimensional salt-precipitation distributions in the rod bundles were quantified with X-ray CT system. For both solutions, salt precipitated downstream and close to the top of active fuel (TAF) height where the void fraction is the highest. The condensed sea water yields wider precipitation region in height direction than mixture solution does. Mixture solution may give localized precipitates at the same height, which is just below TAF and uniformly spread on the horizontal plane. For both solutions, dryout concentration is larger as collapsed solution level is higher. This is because that lower collapsed solution level gives longer boiling-length and higher void-fraction, which results in larger amount of salt precipitations. The proposed salt concentration is useful to evaluate dryout concentration, which is the almost constant salt concentration for heat flux levels within the experimental ranges.

The High Temperature Gas-Cooled Reactor (HTGR), as a Gen-IV reactor, adopts a Reactor Cavity Cooling System (RCCS) to remove core decay heat after a reactor trip. During a postulated accident such as a pipe break in the reactor coolant system, the core decay heat can be properly removed by a RCCS to ensure the integrity of the fuel, reactor pressure vessel (RPV), and concrete silo. In this study we are more interested in the impact of a failure beyond that of the design basis of the RCCS functions, particularly the case in which natural convection of external air is not established. Such a case could happen when a chimney is collapsed by an external event. Against a severe disaster in which air convection is completely lost, a mixed air-water RCCS concept having a water-jacket system is suggested in order to mitigate the significant overheating of the RPV. In order to quantify the effectiveness of the water-jacket RCCS functions, the Depressurized Conduction Cooldown event (DCC) of a HTGR system with a thermal power of 350 MW is analyzed using the GAMMA+ system transient analysis code. As the RCCS air-cooling function is impaired, the water-jacket system increases its role to take part in the heat removal. From the analysis results of a DCC event with complete failure of the RCCS air-cooling, it is confirmed that, compared to the case without the water-jacket system, the maximum temperature of the reactor vessel can be reduced by about 400 °C thus ensuring its integrity. Therefore, the water-jacket system can be utilized as a backup system to mitigate the consequences of a severe accident such as the collapse of the chimney of the RCCS.

The liquid sloshing behavior analysis of reactor vessel is an important topic of sodium-cooled fast reactor design. It is known that there are two pumps and four heat exchangers immersed in sodium liquid. Although lots of research work has been performed on the liquid sloshing inside simple cylindrical tank and coaxial circular cylinders, few about sloshing inside annular region separated by several cylinders (ARSSC), this paper deals with the analysis of the liquid sloshing inside the ARSSC geometry. A simplified cylindrical tank of fast reactor was designed, in which there are six internal cylinders used to model pumps and heat exchangers. By shaking table experiments, the number of internal cylinders was gradually increased, and the natural frequency and wave height of liquid sloshing were measured under each case. In numerical investigation, the VOF models were built to record wave height using the CFD simulation approach, the conditions were consistent with experiments. Much different with previous research, the characteristics of sloshing inside ARSSC geometry were strongly dependent on the number of internal cylinders. Experiment results showed that the natural frequency and wave height all decrease with the increase of internal cylinders gradually. The wave height response obtained by numerical simulation was in good agreement with the experimental results. In addition, as the increase of internal cylinders gradually, the maximum sloshing response occurred when the forcing frequency matched high order natural frequency of sloshing. Moreover, a correction factor was obtained by experiments for the natural frequency theoretical calculation formula. The conclusions can provide favorable reference for the seismic design of the sodium-cooled fast reactor.

A miniature integrated nuclear reactor design with gravity independent autonomous circulation (ACMIR) was newly proposed. The reactor core, energy transfer system of Stirling and linear electric motors are integrated in the reactor pressure vessel to achieve high power density and autonomous circulation capability. The coolant circulation is autonomously driven by gas expansion at heat end and compression at cool end thus is independent on gravity. Twelve sets of rotary drum controller and reflector are used to regulate reactivity outside the reactor pressure vessel. The physics and thermodynamic properties, as well as the safety performances are analyzed. According to these analysis, inherent safety characteristics are obtained, and the reactor is capable to shut down and remove residual heat passively without any external intervention, in accident conditions such as loss of external power supply, overtemperature/overpressure of the reactor, impact when rocket launching or landing, stagnation of the displacer or piston, et al. The integration, less pipes design, gravity independent autonomous circulation features make it a good candidate for space flight propulsion, the Moon and Mars base power supply, the deep-sea or other tilt and swing situation applications.

Advanced Power Reactor of 1400 MWe (APR1400) adopts a Direct Vessel Injection (DVI) system that injects Safety Injection (SI) water directly into the reactor vessel downcomer during accidents. As the DVI nozzles are directly attached to the reactor vessel downcomer, complex thermal-hydraulic phenomena occur in the downcomer region. Recently, in the International Standard Problem (ISP) No. 50 exercise, the phenomena of Emergency Core Cooling (ECC) water mixing in the upper downcomer during the DVI line break accident observed in Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) test were highlighted in terms of the prediction capability of the system analysis codes. Thus, to validate the results from the ATLAS study independently, an additional experimental study was performed with an integral effect test facility, Seoul National University Facility (SNUF), to observe the mixing behavior in the downcomer during a DVI line break accident. According to the SNUF test results, the ECC water mixed vigorously in the downcomer annulus. However, the temperature difference in the azimuthal direction was predicted by a system analysis code, Multi-dimensional Analysis Reactor Safety (MARS). In the MARS calculation, the momentum flux terms are set to zero for the junction between the one-dimensional volume and three-dimensional cell of the MultiD component because the axial and radial velocities are marginal in the large three-dimensional region. However, if the nozzles are attached to the downcomer with a thin gap, the axial and radial velocities are significant when the incoming orthogonal flow through the nozzles impinges against the downcomer wall. It was necessary to consider the momentum flux terms induced by the impinging flow, and to do so, an appropriate jet impingement model, for incorporation in the system analysis code, MARS, was developed in this study. To develop the jet impingement model, Computational Fluid Dynamics (CFD) calculations were carried out, and the jet impingement model was formulated based on the CFD calculations under various conditions. The momentum flux term resulting from the jet impingement phenomenon was correlated with the diameter of the nozzle, downcomer gap size, and incoming flow velocity. This model was applied to MARS by considering the momentum flux term for the junctions connected to the cell of the MultiD component. The modified MARS incorporating the jet impingement model was validated with the test results from the SNUF and ATLAS, and the analysis results exhibited reasonable agreement with the test data.

In severe accident mitigation approaches that aim to achieve In-Vessel Retention (IVR) the decay heat is removed from the corium by conduction through the Reactor Pressure Vessel (RPV) wall, and by flow boiling on the outer surface of the RPV. The boiling Critical Heat Flux (CHF) limit must not be exceeded to prevent RPV failure. Previous studies for prediction of CHF in IVR were predominantly based on data for stainless-steel heaters and de-ionized (DI) water coolant. However, the RPV is made of low-carbon steel, and its surface has an oxide layer that results from pre-service heat treatment as well as oxidation during service; this oxide layer renders the surface much more hydrophilic and rough with respect to an un-oxidized stainless-steel surface, which can have a significant influence on boiling heat transfer. In this study, test heaters were fabricated from low-carbon steel (grade 18MnD5), pre-oxidized in a controlled, high-temperature, humid-air environment, reproducing the prototypical surface oxides present on the outer surface of the RPV. The heaters were then tested in a flow boiling loop using the IVR water chemistry, i.e., DI water with addition of boric acid and sodium tetra-borate. CHF was measured in the range of pressures (100–440 kPa), mass fluxes (180–2450 kg/m2 s), inclination angles (30–90°) and equilibrium qualities (from −0.020 to +0.034) encompassing the IVR conditions. Up to 70% enhancement in CHF values was observed for pre-oxidized, low-carbon steel heaters in comparison to the stainless-steel control heaters. The effect of water chemistry on the CHF was found to be marginal. An empirical correlation fitting the CHF data for pre-oxidized, low-carbon steel surfaces with IVR water chemistry is also presented.

Residual stress in the welds that attach Control Rod Drive Mechanism nozzles into the upper head of a PWR reactor vessel can influence the vessel’s structural integrity and initiate Primary Water Stress Corrosion Cracking. PWSCC at Alloy 600 CRDM nozzles has caused primary coolant leakage in operating PWRs. We have used Deep Hole Drilling to characterise residual stresses in a PWR vessel head. Measurements of the internal cladding and nozzle attachment weld showed that although modest tensile stresses occur in the cladding, the attachment weld contains tensile residual stresses of yield magnitude. Despite the large dispersion of residual stress data for nozzle attachments of this type, all available data suggest that assuming a residual stress profile bounded by the weld material’s yield stress would be conservative for assessment purposes.

An analysis on the behavior of fission products during a station blackout (SBO) accident for the APR1400 nuclear power plant is performed using MELCOR version 2.1. The analysis is focused on investigating the characteristics of radiological releases from the damaged fuel in the reactor core and molten corium in the reactor cavity, transport of the radioactive materials in the reactor coolant system and containment, and release of the radioactive materials to the environment. It is shown that the release characteristics of radionuclides from the fuel, distributions of radionuclides in the reactor coolant system and containment strongly depend on the species of radionuclides. In cases of volatile radionuclides such as iodine and cesium, more than 95% of initial inventory is released before the reactor vessel failure. The amount of radionuclides released to the environment after a failure of the containment was significantly reduced due to the deposition of radionuclides in the form of aerosols on the surfaces of heat structures and a pool of water in the containment. A sensitivity study on the presence of water in in-containment refueling water storage tank (IRWST) shows that the water in the IRWST significantly reduced the amount of radionuclides released to the environment.

After a reactor core melts and collapses suddenly, the melting core accumulates on the base surface of the bottom head of a reactor pressure vessel (RPV), causing severe thermal ablation and thermal stress, endangering the safety of the RPV bottom head. In this study, a 1000-MW pressurized water reactor is considered an example to study the heat transfer ablation and thermal stress of an RPV lower head after a core collapses, by performing numerical simulations. A two-dimensional (2D) heat transfer model is used to analyze the coupling heat transfer between the wall surface of the RPV, two-layer melting corium pool, and outer water chamber. The transient 2D temperature and ablation of the lower head wall surface are calculated. The thermal stress of the RPV lower head and deformation are also investigated using ANSYS software. The results show that (1) The upper crust is the least thick, with a thickness of approximately 0.01 m, whereas the side crust is the thickest at approximately 0.12 m at the base of the lower head. (2) The minimum thickness of the bottom wall decreases linearly with time, starting from the collapse time of the core to 2500 s, when it becomes 0.04 m. It does not change thereafter, but the melting zone is further expanded. (3) The lower head wall starts to melt from 200 s after the core collapses. The melting mass first increases sharply, and subsequently, increases slightly with time. The total melted material is 3000 kg at 5000 s. (4) The heat flux at the inner and outer surfaces of the lower head immersed in the uranium melt layer increases with the azimuth angle, reaching a maximum value of 750 kw/m2 and 250 kw/m2, respectively, at the interface with the metal melt. The heat flux at the RPV inner wall covered by the metal melt is approximately constant at 400 kw/m2, whereas that at the outer surface decreases with the azimuth angle. (5) The stress at the lower head is concentrated at the inner surface, with a maximum value of 625.65 MPa. The radial deformation increases with time only until 2200 s, with a maximum deformation of 28.39 mm occurring in the lower part of the RPV bottom.

The development of analysis models for the Swiss reactors is a key objective of the STARS project at the Paul Scherrer Institut (PSI). Within this context there is a need for the development of computational fluid dynamics (CFD) models of the Swiss reactors in support of future high fidelity investigations of steady-state and transient scenarios. This article presents initial results for the CFD analysis of a Siemens KWU PWR with a focus on the flow behaviour and heat transfer in the gap between the core shroud and core barrel. Temperatures and densities in this region of the reactor are important, for example, for accurate estimations of fast neutron fluence and activation in the steel structures of the core shroud, core barrel and reactor pressure vessel. The flow behaviour in this region may also be relevant for better understanding of ex-core detector responses. The flow conditions in the core bypass region were found to be in the transition-to-turbulence regime, with vortex shedding taking place downstream of the core formers as a result of flow instabilities. The non-stationary nature of the flow presented a challenge in terms of obtaining a solution within a reasonable time period. Two approaches were proposed to address this challenge: time-averaging of the flow-field information before solving the conjugate heat transfer problem; time-averaging of surface heat fluxes in order to derive detailed surface heat transfer coefficients. Both approaches yielded similar results with similar computational effort. Several characteristics and features of the core bypass flow are discussed. Updated Monte Carlo simulation results show that the influence of the core bypass temperatures on the neutron fluence predictions is non-negligible. This highlights the importance of including accurate bypass temperatures in future Monte Carlo simulations focused on ex-core regions.

The main objective of this work is assessment and improvement of the COUPLE module, the lower head molten pool analysis model included in a reactor analysis code RELAP/SCDAPSIM, by using LIVE test series. This module is used to calculate the heat-up of reactor core material that slumped into the lower head of the reactor vessel. Heat transfer in the radial and axial directions to the wall structures and water surround the debris is considered in the module. In the LIVE test facility, the behavior of reactor pressure vessel in the late in-vessel phase of a postulated severe accident in a nuclear power plant is investigated. Due to the recent interest in in-vessel melt retention through external reactor vessel cooling as a key strategy of the severe accident management, the ability and applicability of the module need to be assessed and improved. The heat flux, wall temperature, crust thickness and molten pool temperature are compared with the experimental data. The code was modified to allow modelling external cooling vessel and an input deck for the LIVE facility was created. The results using other empirical correlations are also compared and, as a result, better agreement of heat flux and wall temperature is obtained.

This paper presents results from a series of integral tests performed at Framatome’s INKA test facility in Karlstein (Germany) which simulates a KERENA boiling water reactor (BWR). The scope of the test series was on the behaviour of and interaction between the different passive systems and components under the conditions of extended loss of alternating power (ELAP). These SBO-like conditions were aggravated in three out of four tests by parallel LOCA (Loss of Coolant Accident). The scenarios of all four tests fully correspond to Design Basic Conditions (DBC). They were: main steam line break, feed water line break, reactor pressure vessel (RPV) bottom leak and station blackout (SBO, non-LOCA). In the tests, the passive systems integrated in KERENA and INKA, respectively, have fulfilled their design functions fully satisfactorily and as follows: The Passive Pressure Pulse Transmitter (PPPT) triggered the RPV depressurization without delay. The Emergency Condenser (EC) system removed decay heat along with stored energy from the RPV to the containment. The Containment Cooling Condenser (CCC) system forwarded said power to a heat sink outside of the containment. The passive containment pressure suppression system kept the containment pressure within the design range, partially displacing surplus thermal energy from the drywell to the wetwell, in particular in the early phases after occurrence of LOCA. The passive core flooding system replenished the coolant inventory of the RPV thereby ensuring water levels in the RPV which are fully sufficient for core cooling. Moreover, the systems have cooperated as anticipated by the designers, quietly and without perturbing each other. Hence the test results, which are reported and discussed more in detail within this paper, soundly confirm the underlying design and its passive features. Said tests were carried out as a part of the joint research project EASY ( E vidence of Design Basis A ccidents Mitigation solely with passive safety Sy stems), the overarching objective of which was the development and validation of the code system AC2 of GRS (Gesellschaft fur Anlagen- und Reaktorsicherheit gGmbH).

The FEM simulations in the field of design and safety assessment represent very powerful tools, but they are strongly limited by available material models and material input data. If states near to ductile fracture are to be considered, more complex material description taking into account multiaxial loading conditions is necessary. As complex material models suitable to include these effects into practical simulations are still mostly phenomenological, experiments with samples of various geometries tested under various loading modes have to be used. On the basis of these tests a complex material behavior model covering elastic and plastic material behavior for various stress states can be obtained. This kind of the material behavior description allows a wide range of application from calculation of component limit loading conditions to material properties conversion for samples of different sizes e.g. This paper deals with ductile damage parameters determination for two typical Reactor Pressure Vessel (RPV) steels ferritic and austenitic. The ferritic steel is used for the RPV vessel and the austenitic one is used for internals. There are chosen appropriate samples geometries based on the preliminary FEM stress state analyses of samples at first. Subsequently, testing of proposed samples is performed and material parameters are evaluated. The obtained material plastic damage parameters are subsequently applied to FEM simulation of sharp notched samples and capabilities of applied models to describe material behavior for high stress concentrations is assessed on the basis comparison with real tests.

In the present manuscript the features of deformation of medium-power VVER-600 reactor pressure vessel (RPV) during a hypothetical severe accident (SA) in the case of worsened regimes of external vessel cooling are considered and analyzed. The analysis of thermal and deformation behavior of the RPV was performed for thermal loads on the reactor vessel that were defined before. The thermal loads acting on the reactor vessel from the corium melt were defined for two types of corium melt structures. A two-layer stratified melt, when a less dense layer of molten steel is located over the oxide heat-generating phase of the corium, was considered as the first structure of the molten pool. The inverse three-layer corium melt was considered as the second type of the melt structure. The numerical simulation of heating and creep deformation processes of the reactor vessel was performed with allowance for creep effect and failure of the RPV while varying both the value of in-vessel overpressure (from 0.2 to 0.8 MPa) and the conditions of external cooling of the RPV in the SA. The external cooling of the RPV were simulated by giving corresponding values of heat transfer coefficients (HTC) on the external surface of the vessel wall. The values of HTCs on the external surfaces of both cylindrical part of the vessel and the vessel bottom varied from 350 to 900 W/(m2K). The performed analysis resulted in the fact that under the conditions of worsened external cooling of the RPV in the SA, significant deformations of the RPV structure are observed. Particularly, during the SA the vertical displacement of the RPV lower head (LH) until the moment of its failure may attain 500 mm and more. Such considerable creep deformations of the reactor vessel structure are observed in the case of forming the inverse structure of the corium pool. Here, the LH deformations make up the bulk of total RPV deformation. Such a model of the vessel deformation may lead to partial or complete blockage of cooling gaps used for external cooling of the reactor vessel in the SA. In turn, the disturbance of the regime of external reactor vessel cooling may cause its overheating and premature RPV failure. The dependences of RPV failure time and the value of vertical displacement of the RPV structure on the value of in-vessel overpressure were obtained for the considered group of SA scenarios.

The problems touched on in this work are closely associated with the realization of in-vessel melt retention strategy through the external reactor vessel cooling and cooling of the molten corium pool inside the medium- power reactor VVER-600 (thermal power is ∼1600 MW) in the course of the SA. The general objective of the research was to determine a thermal state in two- and inverse three-layer molten corium pools, which can be formed in the reactor vessel during the SA. The second task was to estimate the efficiency of the top water flooding of corium pool for its cooling in SA by comparing the new results with those obtained in the previous investigation of the authors. Compositions and mass of the corium pools for the two-layer and inverse three-layer pool structures are analyzed and presented in the paper. Simulation of heat transfer in the molten pool was performed for time values 10, 24 and 72 h from the initiating event (IE) in the SA. To estimate the influence of decay heat generation in the bottom metallic layer of the inverse molten pool on the thermal state of molten pool, a series of model SA scenarios was considered in the work. To simplify the simulations the computation domain was bounded by only the pool with taking corresponding boundary conditions. Simulations of thermal state of the molten pool were carried out by means of the NARAL/FEM computer code in which the turbulent convection at the high-Rayleigh numbers was used through the use of the effective heat conduction properties of the corium materials. The numerical results obtained for two-layer corium pool brought out a series of features: (a) the top water flooding of the melt pool resulted in temperature decrease by ∼150 K only in the upper melt steel layer and had no effect on an essential temperature change in the oxide phase of the corium; (b) top flooding of the corium results in an essential decrease (by more than 40%) of maximal values of heat flux acting on the reactor vessel in the region of contact of the vessel wall with steel melt layer. Thus, the top water flooding of the pool surface yields an essential drop of the heat flux peak acting on the vessel wall from 1.65 to ∼1.2 MW/m2 in case at 24 h after IE; (c) the heat flux peaks acting on the vessel decrease from ∼1.65 MW/m2 (at 24 h after IE) to ∼1.15 MW/m2 (at 72 h) in case when the top flooding of the corium pool is absent, and decrease from 1.2 (24 h) to ∼0.6 MW/m2 (at 72 h) when using the top flooding. In the case of the inverse corium pool, the top water flooding essentially decreases (by more than 50%) the maximal value of heat flux in upper layer of steel melt; (d) in the case of melt inversion and redistribution of total decay heat generation in the corium pool between oxide and bottom metallic layers of the pool (parameter KOxide = QOxiide/(QOxiide + QBot_Me), the dependence of maximal values of thermal load on the lateral surface of the pool depending on KOxide value is observed. Thus, the increase of power of heat generation in the bottom metallic layer of the melt from 0.2 to 0.45 (the decrease of. KOxide from 0.8 to 0.55) causes the increase of heat flux value in the bottom layer by ∼1.5 times. Taking into account the fact that in this region of RPV lower head the CHF has low values (∼0.3…0.45 MW/m2), the probability of superheat and premature failure of the vessel bottom in this field increases. The maximal values of heat flux in the oxide phase and bottom heavy metal layer of the pool are observed near the boundary separating these layers. In this region of the VVER lower head, the heat flux attains the values that may exceed the corresponding values of CHF. Because of this, there is a high probability of superheat and the reactor vessel premature failure due to worsened heat transfer and cooling conditions on the external surface of the vessel wall. This fact should be necessarily taken into account when acting on the RPV lower head the thermal loads of moderate intensity (∼0.5…0.7 MW/m2) that may cause superheat and premature failure of the reactor vessel in the case of inverse molten pool formation during the SA in VVERs.

External Reactor Vessel Cooling (ERVC) could be accompanied by the in-vessel water injection to alleviate the thermal focusing effect of the light metallic layer and enhance the In-Vessel Retention (IVR) strategy. Through the wATer injectiOn on molten Zirconium-Stainless steel Metallic pool experiment (ATOM), the effects of different temperatures, mass flow rates of water injection, and different mass percentages of Zr on the heat transfer characteristics after water injection have been investigated. The mass flow rate of water injection ranges from 0.05 kg/s to 0.10 kg/s, and the mass percentage of Zr ranges from 13% to 60%. Eight tests were carried out, and the maximum upward heat flux removed from the melt surface ranged from around 800 to 1400 kW/m2. A higher water injection mass flow rate and a lower Zr mass percentage could lead to a higher steam generation rate. After water injection, the experimental melt temperature and concentration factor decrease. By applying the ATOM heat transfer coefficient as the input data in the nuclear reactor case, the calculated results indicate that the thermal focusing effect inducing by the decrease of the light metallic layer’s height could be highly mitigated by water injection. As a result, the integrity of the RPV could be enhanced. During the whole experiment for all the tests, no vapor explosion and melt ejection were found.

In-vessel retention (IVR) is considered a feasible technique to keep reactor pressure vessel (RPV) integrity in a severe reactor accident. For light water reactor (LWR), the effectiveness of this strategy relies soundly on the critical heat flux (CHF) distribution over the external surface of the lower plenum of RPV, whose orientation varies gradually from downward-facing horizontal to vertical. The CHF prediction capability of the liquid sublayer dryout model is efficient for high mass flux in vertical flow boiling conditions. This paper focuses on how to adapt the model to the changed orientation of the heating surface. Bubble departure diameter (dB) and net vapor generation point (NVG), the starting point for the void fraction developing in a heating channel, is one of the important key points in the CHF prediction. Therefore, to assess the predictive potential of CHF under IVR, experimental research was performed to measure bubble departure diameter and NVG for a changing heating surface orientation from downward-facing horizontal to vertical with a mechanistic model basing on the force balance. A modified liquid sublayer dryout model was then proposed where the channel orientation effect is considered to measure the bubble departure diameter (vapor blanket diameter) using the improved force balance model. The NVG is modified according to the departure diameter. The predicted departure diameter and subcooling at NVG show good consistency with the experimental data, and the modified liquid sublayer dryout model can predict the CHF data with an average relative error of 18.36% in IVR.

The effect of the Hutze (a secondary ECC injection pipe within the hot leg pipe of a Konvoi-type PWR) upon CCFL was experimentally investigated at COLLIDER test facility of the Technical University Munich (TUM). COLLIDER represents a PWR hot leg with a large dimeter of 190 mm and ~¼ scale. A Hutze pipe was built into COLLIDER’s hot leg according to a German Konvoi-type PWR reactor design data. The effect of the Hutze upon CCFL are shown through comparisons with previous investigations at COLLIDER without a Hutze. Visual observations using high-speed recordings are presented at different flow conditions: supercritical conditions at low air velocities, transition into subcritical flow, and the onset of CCFL with comparison against previous recording without a Hutze. Limits of: the onset of hydraulic jumps, onset of CCFL (flooding), and most importantly CCFL characteristics are also shown and compared. The Hutze was found to induce a large impact upon these limits: two hydraulic jumps are observed prior to the onset of CCFL instead of only one (in case without a Hutze), the limits of the onset of hydraulic jumps and onset of CCFL occur at much lower air velocities, and the hysteresis effect (difference between limits of flooding and deflooding) disappears. CCFL characteristics curve is also shifted toward lower values of gas velocities and less liquid flows into the reactor vessel at same air velocities in the case with a Hutze which means a less cooling potential during the reflux condensation mode. Ohnuki correction of the Hutze effect upon CCFL characteristics of 1988 is tested and it seems to work well at small diameter but it overestimates CCFL characteristics for COLLIDER data. The previous agreement between CCFL characteristics obtained from downscaled experiments with D > 50 mm without a Hutze (including COLLIDER data and other large scale facilities) with UPTF data (at full scale) is found to happen because of the Hutze effect which shifts the CCFL characteristics downwards in Wallis diagram and not because of the correct representation of occurring CCFL phenomena. The actual CCFL characteristics for the 1:1 scale without a Hutze are expected to be higher than those reported from data with Hutze at UPTF facility. This indicates that there is still a scale effect even at a large diameter channel, and that most previous downscaled experiments have underestimated CCFL characteristics at full scale. Based upon obtained results, a correlation of CCFL characteristics at full scale excluding the Hutze effect is given. This correlation is of a special interest for the design of new PWR reactors as most new designs does not include a Hutze.

The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. A specific novel micro-MSR, natural circulation, battery design concept involves placing all primary components within a single reactor vessel containment without the need of primary forced pumping. This small modular, integral design presents potential cost savings while producing safety, reliable, and transportable carbon-free power for decades. This investigation evaluates the neutron noise induced by density and flow fluctuations in a natural circulation MSR battery concept being developed at the University of Idaho. This study finds numerical solutions to the one-dimensional, one-group, coupled diffusion equations for a bare, homogeneous core, and evaluates neutron flux and delayed neutron precursor concentration noise due to core flow and fuel salt density perturbation sources. Noise analysis shows both point kinetic and space-dependent behavior is present despite the small size, low-flow, closely-coupled nature of the natural circulation MSR concept. Analysis shows that the low flow of the convective system results in noise behavior typically representative of a larger forced-circulation MSR system. Noise techniques may also be useful in MSR core diagnostics as a non-invasive, low-cost radiological option to traditional monitoring schemes.

The private sector’s recent interest in the active development of molten salt reactors has led to the need to develop and test advanced modeling and simulation tools to analyze various advanced reactor types under numerous conditions. This paper discusses the effort undertaken to model the Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR) design using ORNL’s Shift Monte Carlo code. The MSBR model integrates a Monte Carlo N-Particle (MCNP) MSBR core model with an MCNP model that was generated from a CAD model of the external components and the reactor building, which was subsequently run in Shift. This paper focuses on development of the fully integrated model and its use in performing neutron transport calculations in the reactor cell area. This model is intended to aid in understanding radiological dose conditions during operation, as well as the iron dpa rates in the reactor vessel. The neutron biological dose rates and flux calculated in the reactor cell are much higher in the MSBR than in typical light-water reactors. The implications of these results and future work are also discussed in this paper.

The integrity of core barrel greatly affects the secure operation of the reactor, especially under extreme condition such as earthquake. The excitation of fluid force and inertial force acting on the core barrel requires detailed analyses. Owing to the strong coupling between the inertial force and hydrodynamic pressure in the process of an earthquake, it is difficult to calculate the dynamic response performance of the core barrel in the reactor. On the basis of analyzing the dynamic vibration characteristics of the core barrel, combined with the potential flow theory of inviscid fluid, a mathematical model suitable for effectively calculating the response of the core barrel under earthquake and fluid load was established. Meanwhile, the mathematical model was validated by comparison with the finite element analyses. Using this model, the response of the core barrel under the actions of predefined sinusoidal ground motion and designed seismic spectrum were calculated. Specifically, the effects of height-width ratio and the gap between core barrel and reactor pressure vessel on the vibration behavior were observed.

In this paper, design optimization of the closure head for PWR reactor pressure vessel has been performed. The RPV closure head mainly consist of hemispherical head with openings for CRDM and QUICKLOC nozzles, and head flange with holes for closure studs. Thickness of the head and height of the flange have been optimized by employing response surface optimization methodology through design of experiments technique. The effect of thickness of head and height of flange on the stress intensity distribution of the closure head in the presence of openings were studied in order to explore the best design under the guidelines of the ASME code. The under-designing (compromise of the Tresca yield criterion) and over-designing (compromise of the fracture toughness) of the closure head are controlled in a novel way through detailed stress analysis.

Because of the accident in the Fukushima Dai-ichi nuclear power plant on March 11, 2011, the Taiwan Power Company developed the ultimate response guidelines for the owned nuclear power plants to cope with the Fukushima-type event. The guidelines had the DIVing strategy to prevent the core damage, the hydrogen generation, and the public evacuation. The strategy performed the depressurization and injection of the reactor pressure vessel and the venting of the containment simultaneously to keep the peak cladding temperature of the core fuels below 1500 °F (1088.7 K) The MAAP5 and RELAP5 models of the Kuosheng BWR/6 nuclear power plant of the Taiwan Power Company were used to assess the effectiveness of the guidelines. Based on the results of the code calculations, it was concluded that the guidelines could be used successfully to cope with the Fukushima-type event. And according to the results of the comparisons, the MAAP5 simulations were similar to the RELAP5 calculations. Based on the RELAP5 simulation results, if three safety relief valves were used to do the vessel controlled depressurization, the peak cladding temperature of the core fuels might increase and the concern of the core fuel safety might be needed. According to the sensitivity study results, if the initiation pressure of the vessel emergency depressurization was higher than 40 kgf/cm 2 (4.03 MPa), the minimum vessel alternate water injection flow would have the significant increase.

Early versions of artificial neural networks’ ability to learn from data based on multivariable statistics and optimization demanded high computational performance as multiple training iterations are necessary to find an optimal local minimum. The rapid advancements in computational performance, storage capacity, and big data management have allowed machine-learning techniques to improve in the areas of learning speed, non-linear data handling, and complex features identification. Machine-learning techniques have proven successful and been used in the areas of autonomous machines, speech recognition, and natural language processing. Though the application of artificial intelligence in the nuclear engineering domain has been limited, it has accurately predicted desired outcomes in some instances and has proven to be a worthwhile area of research. The objectives of this study are to create neural networks topologies to use Oregon State University’s Multi-Application Small Light Water Reactor integrated test facility’s data and evaluate its capability of predicting the systems behavior during various core power inputs and a loss of flow accident. This study uses data from multiple sensors, focusing primarily on the reactor pressure vessel and its internal components. As a result, the artificial neural networks are able to predict the behavior of the system with good accuracy in each scenario. Its ability to provide technical data can help decision makers to take actions more rapidly, identify safety issues, or provide an intelligent system with the potential of using pattern recognition for reactor accident identification and classification. Overall, the development and application of neural networks can be promising in the nuclear industry and any product processes that can benefit from utilizing a quick data analysis tool.

The asymmetric pressure wave caused by Loss of Coolant Accident (LOCA) transients in multi-loop type Pressurized Water Reactor (PWR) leads to great shock to the reactor core barrel. The instantaneous shock causes severe damage to the reactor vessel structure integrity. In this paper, a new method for reactor core barrel instantaneous force calculation was proposed based on the widely accepted reactor system analysis code RELAP5. The reliability of RELAP5 code for pipe blow down and pressure wave propagation simulation was assessed against Edward pipe problem firstly. Then the typical three-loop PWR system model was established utilizing Symbolic Nuclear Analysis Package (SNAP) software and the small break LOCA scenarios with different break diameters and break locations were analyzed. The maximum instantaneous asymmetric force applied on the reactor core barrel was achieved under various break conditions, providing the boundary conditions for the core barrel stress distribution and deformation calculation. Results show that the cold leg break LOCA transient leads to great instantaneous asymmetric force, while the hot leg break LOCA transient has no obvious asymmetric force generated. The core barrel instantaneous stress concentration and deformation were discovered. This research could contribute to the reliability assessment of PWR core barrel under LOCA scenarios.

Even after 6 years since the Fukushima accident, the exact accident progression for each unit and location of core debris have not been clarified, although solidified low-melting metal debris was identified at the bottom of the 1F3 reactor pressure vessel (RPV) in July 2017. Currently efforts are directed towards robotic inspection with remote cameras, as well as dose and temperature measurements of the environment inside the Primary Containment Vessels (PCV). In spite of their effort, the observed environmental temperature distribution data does not support the existence of a significant radiation heat sources attributable to the molten core at the bottom of the PCV. At this point the total decay heat of the core debris should be as large as a few hundred kilowatts after 2000 days since the initiation of the accident as summarized in Annex A. In addition, the temperature of the accumulated water inside PCVs is 10–20 °C even with a reduction of the water injection to 3.0 m 3 /h. The RPVs appear to be holding the heating core debris with water, implying that “in-vessel retention” of core debris has been achieved thanks to effective accident management. This should help greatly in the retrieval of the core debris by removing the top head of RPV. Under these circumstances the author has conducted a forensic engineering study (i.e., different fields of science working together to collect and integrate independent evidences) to clarify the most likely accident scenarios of the Fukushima Daiichi accident. Through this study the author identified that a large portion of the land contamination observed at the north-west direction is mostly the result of the accident that occurred at Unit 2. In this unit its blowout panel, provided for over-pressure protection against a main steam pipe breach in the secondary confinement building, was inadvertently activated before a leakage of radioactive effluent from the PCV. Its activation is believed due to the hydrogen explosion in Unit 1 which occurred on March 12, next day of the Fukushima accident initiation. By losing the confinement function, the radioactive effluent leaked from the 1F2’s PCV and would have been discharged without mitigation. This accident scenario explains the series of leakage events identified in two of the 24 monitoring posts which had been installed by the Fukushima Prefectural Government. A series of six large releases were repeated between March 15 and 16, behaving like a periodical actuation of the safety valves for the PCV. Such multiple release events were very likely induced by the overpressure release of the PCV due to leakage of the dry well flange. This leakage should have been induced through discharging steam and hydrogen due to the activation of Safety and Release Valves (SRV) into the suppression pool (SP) water. Unfortunately the wet well atmosphere must have been that of air, since there was no nitrogen charge line to the atmosphere of the SP surface water. The resultant air-hydrogen mixture resulted in an “internal hydrogen explosion” which should have deformed the flange. The recent robotic inspection inside PCV revealed that a gigantic water splash appears to have occurred at the bottom of the PVC dislodging the gratings installed over the platform. Next to the series of large releases from Unit 2, Unit 1 also induced two large releases on March 12. Fortunately, these releases left more than 3 orders of magnitude less soil contamination compared with the series of releases from Unit 2. Unit 3 also released a significant amount of radioactive species twice on March 13, resulting in a very small soil contamination. The main constituent of the radioactivity is likely radioactive noble gases in this unit.

In-vessel-retention (IVR) of molten core debris by external reactor vessel cooling (ERVC) technique has been proposed as the severe accident management strategy for advanced PWRs like CAP1400. CHF limits decide the maximum heat removal capacity of ERVC which have been widely tested by many investigators. Several researches have proved that the CHF limit and its mechanism are strongly related to the vapor motions and two phase flow characteristics. However, the two phase flow characteristics in the downward facing boiling process of ERVC have been rarely reported in the literature. In the present research, a three-dimensional visual experimental apparatus is developed to investigate the downward facing boiling phenomena and CHF limits for ERVC. The vapor morphology at different heat flux levels and inclination angles of the heating surface is carefully studied. The visual results show significant differences in vapor morphology between 3-D ERVC and 2-D slice ERVC experiments in the vessel bottom region. Vapor motion characteristics corresponding to the occurrence of CHF are examined by determining the boiling cycle frequency and maximum vapor height. The effect of inclination angle on the vapor morphology is identified and the local variation of CHF along the vessel outer surface is measured. Vapor motion characteristics are investigated quantitatively by advanced image processing technics. The dependence of boiling cycle on inclination angle is revealed. In addition, the effect of bottom heating on the downstream CHF behavior is investigated, and the mechanism responsible for the influence that is revealed by the vapor morphology is determined. The present study provides an in-depth physical understanding of the 3-D downward facing boiling process during ERVC that can be useful for hydrodynamic modeling of the CHF phenomenon.

The in-vessel retention by external reactor vessel cooling (IVR-ERVC) is one of the severe accident management strategies for mitigation or termination of the accident by holding the radioactive material in the reactor vessel. We made a lumped parameter based computer code to evaluate the thermal load on the reactor vessel, and applied the code for a probabilistic analysis for a high power (>1000 MWe) reactor assuming the APR1400 condition. The present analysis showed the focusing effect in most of cases. The heat transfer models, especially the critical heat flux correlation, showed a significant impact on the result. Consideration of the multidimensional heat conduction effect in the vessel wall can reduce the excessive conservatism in the vessel failure criterion by assuming the same heat fluxes at inside and outside of the vessel wall. The influence of the probability distribution profiles for the uncertainty parameters was also investigated.

We investigated the natural convection of the oxide layer in a three-layer configuration of core melts in a severe accident. In order to achieve high modified Rayleigh numbers of 10 12 –10 13 , mass transfer experiments were performed using a copper sulfate electroplating system based upon the analogy between heat and mass transfer. Four different cooling conditions of the top and the bottom plates were tested. The upward heat ratios were 14% higher for three-layer than for two-layer due to the reduced heights and the downward heat ratios were lower the same amount. The local Nusselt numbers for the top and the bottom plates were measured and compared with the two layer configuration. To explore the heat load to the reactor vessel, the angle-dependent heat fluxes at the side wall, were measured and compared with the two-layer configuration. Heat load to the side wall and peak heat at the uppermost location were intensified for the three-layer configuration.

The S calable LI quid M etal cooled small M odular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW th continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW th , the fissile production and depletion, the Beginning-Of-Life (BOL) spatial distributions of the fission power in the core at 100 MW th , and the negative temperature reactivity feedback effects. Besides decreasing the UN fuel enrichment, other parameters examined are the materials of the followers to the B 4 C rods for RSS and RC, the rods in the radial blanket assemblies, and the thickness of the scalloped BeO walls of the UN fuel and radial blanket assemblies. Despite ∼13% lower fuel enrichment (15.35%) and ∼22% lower hot-clean excess reactivity ($6.29 versus $8.06 for SLIMM-1.0 at 100 MW th ), the operation life of the SLIMM-1.2 is ∼6.8% longer (6.3 versus 5.9 FPY for SLIMM-1.0), and the total negative temperature reactivity feedback is slightly smaller. At BOL, the radial blanket assemblies in ring 4, and the six UN fuel assembles in ring 1 of the SLIMM-1.2 core generate 2% and 23% of the total reactor thermal power, respectively. The twelve and eighteen UN fuel assemblies in rings 2 and 3 of the SLIMM-1.2 core generate 38% and 37% of the reactor thermal power, respectively. At EOL, the thermal power generation by the UN assemblies in rings 1 and 2 of the SLIMM-1.2 core decreases to 21.5% and 35.9%, respectively, while that by the assemblies in rings 3 and 4 increases to 37.6% and 5%, respectively.

The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

The four different geometries of a reactor vessel auxiliary cooling system (RVACS) are designed to investigate how the heat removal capability changes by the geometries. Each geometry has different heat transfer characteristics regarding airflow and the presence of an air separator and an insulation material. The heat removal performance is evaluated with the reactor vessel (RV) temperature, the RV wall emissivity, and the airflow gap. For the same RVACS volume, 600 °C RV temperature, and 6-cm gap, the heat removal capability varies from 240.8 kW to 308.5 kW, depending on the geometry. The wall emissivity is less effective for Geometry 2, which has a large cavity volume and a small heat transfer area compared to the other geometries. The highest heat removal performance was obtained using Geometry 3 because cold air flows in from the bottom of the RVACS, improving both radiative and convective heat transfer. Reducing the gap size by 3 cm results in only 80.0 kW of heat removal capability for Geometry 1, and the heat removal dramatically decreases to 0.2 kW at an RV temperature of 800 °C. A sufficient RVACS gap size of at least 6 cm with a 0.6-m diameter intake pipe is required to provide adequate natural circulation and eventually enhance the heat removal capability.