Interface-System Loss-of-Coolant Accident (ISLOCA) occurs by failure of isolation valves in piping systems. In this study, transient behavior of pressure propagation in piping systems after ISLOCA has been estimated. At first, capability of TRACG code to predict pressure propagation in a simple pipe line, which has expansion, contraction and bifurcation, has been investigated. Then, sensitivity analysis of valve opening period has been performed to investigate the behavior of pressure propagation in a simple pipe line. Finally, pressure propagation inside piping systems after ISLOCA has been simulated for Hitachi-GE standard Advanced boiling water reactor (ABWR) plant. Maximum pressure inside High Pressure Core Flooder (HPCF) systems has been less than 7.8MPa. TRACG code has been shown to be useful to predict transient behavior of pressure propagation in complicated network of piping systems after ISLOCA.

Pebble bed modular high temperature gas-cooled reactors (HTR), due to their characteristics of low power density, slender structure, large thermal inertia of fuel elements and reactor component materials (graphite), have good inherent safety features. However, the reflectors consisting of large piles of graphite blocks will form huge numbers of certain bypass gaps in the radial, axial and circumferential directions, thus affecting the effective cooling flow into the reactor core, which is one of the concerned issues of HTRs. According to the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the thermal-hydraulic calculation model is established in this paper. Based on this model, considering different bypass flow, that is to say, different core cooling flow, fuel element temperature, outlet helium temperature and the core pressure drop in the normal operation, as well as the maximal fuel temperature during the depressurized loss of forced cooling (DLOFC) accident are analyzed. This study on bypass effects on the steady-state and transient phases can further demonstrate the HTR safety features.

In the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs), fuel relocation out of the core region decrease the potential for severe power excursions caused by recriticality and control-rod guide tubes (CRGTs) provide an effective path for fuel-relocation. Therefore, a methodology for evaluating molten-fuel relocation through CRGTs is required in order to realistically evaluate sequences and consequences of CDAs. Since the liquid sodium exists at the coolant plenum, pressurizations and coolant void development associated with fuel-coolant interactions (FCIs) are considered to affect fuel-relocation. Therefore, the objective of the present study is to develop the methodology for evaluation of molten-fuel relocation into the coolant plenum with FCIs. In the present study, the SIMMER code which has been developed for CDA analyses was utilized as a technical basis since this code can treat multi-phase, multi-component fluid dynamics with phase-changes supposed to take place in the coolant plenum during fuel-relocation. The evaluation methodology was developed through validations of the SIMMER code using experimental data. A series of fundamental experiments were selected for model validations in which an alloy with low melting temperature and water were used as simulant materials for the fuel and the coolant respectively since the experiments were performed under a simulated CDA condition of SFRs in which a liquid-liquid direct contact was maintained between the melt and water contact surface, and the visual observation on FCI process was effective to validate models based on phenomenological considerations. The code was validated by two steps: In the first step, fundamental validations of melt-discharge into the coolant were performed, namely, momentum exchange functions of flowing-melt both to the wall of relocation-path and to the coolant were validated based on experimental data in which effects of FCIs on melt-discharge into the coolant were eliminated or negligible. In the second step, comprehensive validations of melt-ejection into the coolant were performed, namely, models which affect heat-exchange between the melt and the coolant were validated based on experimental data in which the melt was relocated into the coolant with FCIs. The second step validation required model improvements for suppression of melt-coolant interfacial area based on the results of visual observation in the experiments in order to reproduce the experimental results appropriately. Through the present model validations, the methodology to evaluate molten-fuel relocation into the coolant plenum with FCIs was successfully developed.

The problem of reactor pressure vessel (RPV) integrity during in-vessel melt retention for VVER type reactor is considered. Comprehensive numerical analysis contains the modeling of full severe accident scenario including thermal and mechanical loads on the RPV. The RPV stress calculation takes into consideration vessel thermal erosion and creep with a long-term strength, and error estimations: uncertainty analysis, sensitivity study. These problems are solved with the presented HEFEST-URAN package including: HEFEST code (2D FEM analysis of melt-structure interaction), HEFEST-M code (2D FEM stress analysis) and Varia code (multivariate calculations with uncertainty/sensitivity statistical assessment). HEFEST-URAN is a subset of the SOCRAT code developed in the Russian Federation. In the paper the HEFEST-URAN methodology and validation are described with some examples. The results of multivariate calculations of IVMR for VVER-440 reactor are presented with sensitivity study of the effects of essential factors defining the RPV thermal loading and long-term strength.

The possibility of a spent fuel severe accident has received increasing attention in the last decade, and in particular following the Fukushima accident. Several large scale experiments and also separate effect tests have been conducted to obtain a data base for model development and code validation. The outcome of the Sandia BWR Fuel Project was used to define the flow parameters adjusted for the low pressure and the increased flow resistance due to the presence of the spent fuel racks which resulted in reduced buoyancy driven natural circulation flow compared with reactor geometry. The possibility of a zirconium fire, using the flow parameters obtained from the spent fuel experiments, is investigated in the present work. The important outcome of the study is that spent fuel will degrade if temperatures above 800 K are reached. In partial loss of coolant accidents the flow through the lower bottom nozzle is blocked and because there is no cross flow possible due to the spent fuel racks the coolant flow in the upper dry part of the spent fuel is limited by the steam production in the lower still wetted part of the fuel. This accident scenario leads to the fastest heat up in a postulated spent fuel accident. The influence of different kind of spent fuel storage (hot neighbour and cold neighbour) is investigated. An important factor in these calculations is the radial heat transfer to the neighbouring fuel assemblies. In the present work limits of the spent fuel storage under accident conditions (minimum allowed water levelin the pool) and total loss of coolant (maximum coolable decay heat per fuel assembly) are shown and explained.

Emergency Core Cooling System (ECCS) analyses using Loss of Coolant Accident (LOCA) codes utilize two-phase Reactor Coolant Pump (RCP) performance models formulated on the basis of data from tests conducted on the Semi-scale pump (Reference 1) operating at 60 Hz frequency. In some PWRs, the RCPs operate at a frequency of 50 Hz. This paper presents the results of an evaluation performed to determine the applicability of RCP two-phase performance models developed on the basis of data from the Semi-scale tests for analyzing ECCS performance of new generation PWRs. The evaluation addressed two major issues: (1) the applicability of the two-phase RCP performance model developed using the data from the Semi-scale pump tests (Reference 1) for full scale Pressurized Water Reactor (PWR) LOCA simulations, and (2) the relevance of the two-phase RCP performance model developed on the basis of test data for the Semi-scale pumps running at 60 Hz frequency to PWR RCPs running at 50 Hz frequency with higher specific speeds. Reviews of pump performance test data available in the open literature identified two-phase performance data appropriate for use in substantiating the validity of current PWR pump performance models. These data supported the conclusion that the two-phase head performance degradation for the Semi-scale Mod-1 pump is conservative compared to the two-phase pump performance data generated from testing of pumps representative of full scale PWR RCPs. A review of ECCS analyses results available in the literature determined that the use of the current RCP two-phase performance model (developed using the Semi-scale Mod-1 pump test data) for a typical PWR plant resulted in about a 100 °F increase in the Peak Clad Temperature (PCT) for a Large Break LOCA (LBLOCA) in comparison to the PCTs calculated using the two-phase pump performance model developed on the basis of test data for pumps representative of full scale PWR RCPs. It was determined from the current study that the frequency (50 Hz vs. 60 Hz) of the electrical power that drives the pump motor is not of much consequence for two-phase RCP performance modeling, since (1) the RCP performance model is characterized via normalized pump performance parameters, and (2) for the LBLOCA analysis of interest, the RCPs are assumed to lose power at the start of the event.

The main steam line break (MSLB) is an overcooling accident that may lead to an over criticality, and so to a power increase, after the reactor trip. The most penalizing single failure is a RCCA bank stuck out of the core when the reactor trip occurs. This configuration leads to a strong asymmetry of the radial power shape combined with a strong asymmetry of the core inlet temperature that results in a strongly distorted 3D power distribution. In the original design, the MSLB accident was studied with a simplified and conservative 0D method. The point kinetics approach requires the use of extremely conservative assumptions in order to account for the asymmetry in the core region that takes place during the transient. The use of the coupling between a three-dimensional neutronic code (SMART), a 3D core thermal-hydraulic code (FLICA cf. ref [4]) and a reactor coolant system code (MANTA cf. ref [3]) allows representing the 3D heterogeneity of the power shape and also of the resulting cross flows. In addition, this coupling allows determining moderator and Doppler feedback effects in a much more realistic way thus limiting accident consequences estimated. A methodology, called MTC3D (for Méthode Totalement Couplée 3D in French), has been developed using the coupling between the three codes to perform the MSLB analysis. The physical dominant parameters of the transient are identified through a comprehensive sensitivity analysis. Then, a deterministic approach is used in the entire transient simulation considering dominant parameters in a penalizing way. In a first step, neutronic data are determined with SMART calculations. In a second step, MANTA/SMART/FLICA transients are performed with penalized neutronic and thermal-hydraulic data. In a third step, as the steam line break transient is a relatively slow transient, the core power distribution is evaluated with a steady state SMART/FLICA calculation without penalization. In a last step, safety criteria, such as minimum DNBR (Departure from Nucleate Boiling Ratio) are calculated with FLICA calculations based on core power distribution calculated at the third step and boundary conditions calculated at the second step. The use of 3D neutronic and detailed thermal-hydraulic codes to model the reactor core allows considering a more physical representation of the core configuration for transient analysis. The coupling between 3D neutronic and core thermal-hydraulic codes allows exhibiting intrinsic margins without over penalizations related to a simplified 0D method.

In the event of a severe accident in a pressurized water reactor, the core of a reactor melts and forms corium, a mixture that includes molten UO 2 and ZrO 2 . If the reactor pressure vessel fails, corium can be relocated in the containment cavity and interact with concrete forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question is what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using a new computer code MOCO. The code considers the heat transfer behavior in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, and incorporates phenomenology that is deemed to be important for analyzing debris cooling behavior. The interaction between thermalhydraulics and physic-chemistry is modeled in MOCO. The main purpose of this paper is to present the modeling used in MOCO and some validation calculations using the data of experiments available in the literature.

In severe accident, the assurance of long term coolability of the corium melt is one of the key issues of the severe accident management (SAM) for the mitigation and termination of the accident progression. The Korean NPPs adopts the SAM strategy that water floods into cavity levels up to the upper part of the reactor vessel to provide the external reactor vessel cooling for the in-vessel retention of corium (IVR-ERVC). However, if the IVR-ERVC strategy fails and the corium relocation onto the cavity occurs, the long term coolability of the ex-vessel corium on the cavity floor should be assured. However, this scenario takes a little account of the effect of the two-phase flow from the corium particles and pre-settled debris bed on the cavity floor. Recently there begins research concerning the effect of two-phase flow on the bed formation, like the particle relocation by self-leveling phenomenon reporting that the cavity pool convection by the two phase flow from the debris bed largely distorts the settling trajectories of the debris by their own sizes. In the perspective, the present study aims to investigate the pore clogging by the fine debris particles in two-phase conditions at the DAVINCI facility, POSTECH, Korea. In the experiments, pre-defined debris particles with various diameters and size distributions are plunged into a water pool in where two-phase flow regime is simulated by the gas injected from the bottom of the test section. The configuration of the resultant debris bed is analyzed to identify the role of two-phase flow on the settlement of fine particles and clogging of debris bed pores. In the test, it was observed that the resultant bed shape and particle distribution was significantly affected by the two-phase bubble behavior during the particle settlement. It is also showed that in the two-phase test the fine particles are more settled in the outer region compared to the pre-settled debris bed region, illustrating the potential positive effect against the coolability degradation due to the fine particle pore clogging phenomena.

The source information of the radionuclide release in nuclear accidents is a key issue of the nuclear emergency response. One way to estimate the source information is by inversing the radionuclide transportation process based on environment radiation monitoring data. The advantage of this method is that the required data are easy to obtain in accident. But it is vulnerable to large uncertainties in both data and transport model. To solve the problem, a source term estimation method based on four-dimensional variational (4DVAR) data assimilation technique was proposed for source term estimation in this study. The proposed method couples 4DVAR with the RIMPUFF air dispersion model. It formulates the inverse modelling of source term estimation as an optimization problem that is trying to find an optimal balance between real observation data and the background field. The advantage of this method is that the radionuclide transport in every time step is included in data assimilation and the result is global optimum in the whole assimilation period. The gradient for cost function is calculated by the backward integration of the adjoint model. Practical imperfectness of measurement were considered and integrated into the cost function. The proposed method was verified using numerical simulation for both homogeneous and heterologous atmospheric condition. The performance of the source term estimation method was also investigated with respect to different release profile, wind speed and atmospheric stability class. The simulation results demonstrate that the estimate matches the true release status well for both homogeneous and heterologous wind field. Also, the experimental results show that the proposed method has strong robustness to wind speed and atmospheric stability.

The phenomenon of the partial flow blockage of a fuel assembly in a reactor core is investigated with a coupled 3D neutronics/thermal-hydraulics code in order to account for the space reactivity feedback effect which is of great importance during hypothetical blockage scenarios. This paper identifies the neutronics thermal-hydraulics coupled response in the blocked assembly during the transient and analyzes the details of the phenomenon.

Critical heat flux (CHF) is one of the most crucial design criteria in other boiling systems such as evaporator, steam generators, fuel cooling system, boiler, etc. This paper presents an alternative CHF prediction method named projection support vector regression (PSVR), which is a combination of feature vector selection (FVS) method and support vector regression (SVR). In PSVR, the FVS method is first used to select a relevant subset (feature vectors, FVs) from the training data, and then both the training data and the test data are projected into the subspace constructed by FVs, and finally SVR is applied to estimate the projected data. An available CHF dataset taken from the literature is used in this paper. The CHF data are split into two subsets, the training set and the test set. The training set is used to train the PSVR model and the test set is then used to evaluate the trained model. The predicted results of PSVR are compared with those of artificial neural networks (ANNs). The parametric trends of CHF are also investigated using the PSVR model. It is found that the results of the proposed method not only fit the general understanding, but also agree well with the experimental data. Thus, PSVR can be used successfully for prediction of CHF in contrast to ANNs.

In atmospheric dispersion models of nuclear accident, the empirical dispersion coefficients were obtained under certain experiment conditions, which is different from actual conditions. This deviation brought in the great model errors. A better estimation of the radioactive nuclide’s distribution could be done by correcting coefficients with real-time observed value. This reverse problem is nonlinear and sensitive to initial value. Genetic Algorithm (GA) is an appropriate method for this correction procedure. Fitness function is a particular type of objective function to achieving the set goals. To analysis the fitness functions’ influence on the correction procedure and the dispersion model’s forecast ability, four fitness functions were designed and tested by a numerical simulation. In the numerical simulation, GA, coupled with Lagrange dispersion model, try to estimate the coefficients with model errors taken into consideration. Result shows that the fitness functions, in which station is weighted by observed value and by distance far from release point, perform better when it exists significant model error. After performing the correcting procedure on the Kincaid experiment data, a significant boost was seen in the dispersion model’s forecast ability.

Further development of nuclear engineering is inseparably linked with the requirement of vast application of the passive systems of heat removal running without human intervention. Creation of such systems is impossible, if only conventional engineering solutions are used. As known, to prevent propagation of the fission products into the environment there are three safety barriers. To provide operation of the third safety barrier (containment shell), in particular, of the reactor cavities both in operational and emergency modes a passive evaporation-and-condensation (EC) system of heat removal is proposed. The features of thermal design of the EC systems for thermal shielding of the reactor cavities are considered. They make it possible to determine the optimal main design variables of the EC systems and prove reasonability and efficiency of their application. The performed study validates engineering feasibility of an efficient EC system for thermal shielding of the reactor equipment.

When accident events are caused by a large-scale natural disaster, conditions beyond those at the plant site may affect the accident. As well, quick diagnosis and recognition of damaged equipment are necessary. We have been developing inherently safe technologies for boiling water reactor (BWR) plants in response to these. An operation support system for plant accident events is one of these technologies. Our operation support system identifies accident events and predicts the progression of plant behavior. The system consists of three main functions: sensor integrity diagnosis, accident event identification, and plant simulation functions. The sensor integrity diagnosis function diagnoses whether sensor signals have maintained their integrity by correlating redundant sensors with the plant design information. The accident event identification function extracts a few of candidate accident events using alarm and normal sensor signals received by the sensor integrity diagnosis function. The scale and position of the accident event are determined by comparing plant simulation results with normal sensor signals. The plant simulation function uses a detailed three-dimensional model of the nuclear reactor and plant. This simulation can predict future plant behavior on the basis of identified accident events. This proposed operation support system provides available results of accident event identification and plant condition prediction to plant operators. This system will reduce the occurrence of false identifications of accident events and human errors of operators.

The application of the Best Estimate Plus Uncertainty (BEPU) method is made to analysis of the “Intentional depressurization of steam generator secondary side” which is an accident management procedure in a small-break loss-of-coolant accident (SBLOCA) with high pressure injection (HPI) system failure. RELAP5/MOD3.2 is used as the analysis code. By applying the BEPU method, the uncertainties of the analysis results can be estimated quantitatively. However, the accuracy of the analysis results depends primarily on the base case result predicted by the best estimate code. In this study, in order to investigate the appropriate base case model, simulation analyses using the RELAP5/MOD3.2 were carried out for the ROSA Large Scale Test Facility (ROSA/LSTF) secondary-side depressurization tests. It was found that the code predicted well the major event progressions such as pressure responses, core liquid level responses, and rod surface temperatures, as well as important phenomena such as formation and clearing of loop seals, accumulation of water from condensation, and countercurrent flow limitation (CCFL) at the inlet of the U-tubes, which are characteristic features of this accident scenario.

Two out-of-pile bundle tests, QUENCH-L0 and QUENCH-L1, were performed recently at Karlsruhe Institute of Technology (KIT) in the framework of the QUENCH-LOCA program devoted to the investigation of the so-called secondary hydriding of the cladding. The overall objective of this bundle test series is the investigation of ballooning, burst and secondary hydrogen uptake of the cladding under representative design basis accident conditions as well as detailed post-test investigation of cladding mechanical properties to analyze the material behavior with respect to embrittlement. The program was started in 2010 with the QUENCH-L0 commissioning test using 21 electrically heated rods with as-received Zircaloy-4 claddings followed in 2012 by the QUENCH-L1 reference test using the same material. These two tests differ in 1) heat-up rate during the first transient and 2) presence of a cool-down phase before quenching. The maximum heating rate reached during QUENCH-L0 was only 2.5 K/s, whereas the transient phase of QUENCH-L1 was performed with the maximum rate of 7 K/s. The state of the QUENCH-L0 bundle was practically “frozen” immediately after the transient phase by fast injection of two-phase fluid. The reference test QUENCH-L1, was performed with a typical cooling phase after the transient phase. It provides data on Zircaloy-4 cladding embrittlement based on more prototypical temperature history. Post-test neutron radiography and tomography revealed formation of hydrogen bands around the oxidized inner cladding surface in vicinity of the burst openings for both tests. However, the concentration of hydrogen absorbed inside these bands was different for both tests: whereas the maximum hydrogen concentration for QUENCH-L0 reached 2560 wppm, the corresponding value for QUENCH-L1 was only 1690 wppm. Complementary model calculations confirm that the differences in hydrogen concentrations are mainly related to the differences in temperature sequences. Subsequent tensile tests with tube segments at room temperature revealed the dependence of the mechanical behaviour on hydrogen concentration: tubes with hydrogen contents above 1500 wppm were simultaneously double ruptured along the hydrogen bands, whereas tubes with hydrogen concentrations below 1500 wppm failed at the middle of burst openings.

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed. The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident. The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.

PSA specialists in ÚJV Řež, a. s. maintain a Living Probabilistic Safety Assessment (Living PSA) program for Dukovany Nuclear Power Plant (NPP), a four-unit VVER-440 plant type, which is operated in the Czech Republic. This project has been established as a broad framework for all plant activities related to risk assessment and as a support for risk-informed decision making carried out at this plant. In addition to recommendations for design and operation measures in order to increase the plant safety, it provides a basis and platform for all PSA applications at Dukovany NPP. The Living PSA model for Dukovany NPP is an integrated model representing the complete scope of Level 1 and Level 2 PSA for all plant operational modes. It produces the unit specific outputs for any Dukovany NPP unit. The RiskSpectrum ® PSA software has been used for development and quantification of the PSA model. It is continuously updated and extensively used for various PSA applications at Dukovany NPP (e.g. risk monitoring, evaluation of plant Technical Specification changes, support for procedure development and training process, event analysis, etc.). The paper focuses on the important features of the Living PSA project for Dukovany NPP. It also discusses the broad experience gained during model development and update as well as possible future enhancements.

With “theory of nuclear safety (TONS)”, this paper intends to explain the Core Damage (CD) scenario of a Nuclear Power Plant (NPP) with the systematic methodology, many notions introduced here can be extended to other types of nuclear installations, as well. This systematic methodology combines the Risk-Informed Safety Margin Characterization (RISMC) Metatheory of TONS, and the basic reliability theory. A “metatheory” of such theories, here, is a theory to analyze the Theory of Nuclear Safety (TONS); in its own theory system, it is designed to summarize the safety of a NPP. Meanwhile, the basic reliability theory, which is decided by the authors, is focus on the mission reliability model (a model can be established by Reliability Block Diagram (RBD)); then the related basic concepts, is simple and clear, and quite mature in NPP field. The present work outlines the traditional reliability theory and the RISMC-based Metatheory, and these two concepts here are taken as the appropriate TONS to analyze the CD Scenario, after that, a renovate or renew TONS, from these two sides, can be introduced to analyze the fundamental safety of NPP.