Plant Safety Analysis Research Articles

Simulation Analysis of Loss of Coolant Accidents Based on CNIFA Program

Abstract The loss of coolant accident in the primary pipeline of a reactor is a typical accident in the safety analysis of nuclear power plants, and it is one of the accidents with a high probability of occurrence. Due to the extremely complex structure of the lower space of the containment vessel, it increases the complexity of fluid flow during a loss of coolant accident. This article is based on the internal water flooding analysis simulation software CNIFA of a nuclear power plant to simulate the water flooding of a rupture in the main pipeline heat pipe section of a power plant, and to conduct a comprehensive water flooding scenario simulation for this working condition. The quantitative information collected during the simulation process, including dynamically displayed water volume, water level, and their changes over time, facilitate emergency response personnel to choose the correct response plan in advance, it can also provide more accurate basis for various specialties in design analysis.

APPROACH TO THE ANALYSIS OF SUCCESS CRITERIA AS A COMPONENT OF PROBABILISTIC PHYSICAL PROTECTION ANALYSIS OF NUCLEAR FACILITIES

An approach to adapting the methodology for analyzing success criteria from traditional probabilistic safety analysis (PSA) of nuclear power plants (NPPs) to probabilistic physical protection analysis (PPA) of nuclear facilities is demonstrated using the hypothetical nuclear facility SNRI as an example. Generally speaking, using the traditional PSA methodology for assessing the physical protection of nuclear facilities allows for the identification of new vulnerabilities in the physical protection systems of nuclear facilities, the development of recommendations to enhance physical protection efficiency, the performance of more substantiated cost-benefit analyses, and more.

Global Sensitivity Analysis for Segmented Inverse Uncertainty Quantification in the Safety Analysis of Nuclear Power Plants

Within the Best Estimate Plus Uncertainty framework for the safety analysis of Nuclear Power Plants, the quantification of the uncertainties affecting the Thermal-Hydraulics (T-H) codes used is crucial. For this, Inverse Uncertainty Quantification (IUQ) methodologies are being developed for determining the probability density functions of relevant T-H codes input parameters, based on experimental data from Separate Effect Tests (SETs) experimental facilities. In practice, IUQ is challenged by the large range of variability of the experimental data in terms of Initial and Boundary Conditions (ICs & BCs), because the experimental campaigns are designed to cover the widest possible domain of conditions with the smallest number of experiments, so that same or similar ICs and BCs are seldomly repeated. To address this issue, we propose to use global sensitivity analysis, to tailor the IUQ on specific sub-regions described by segmented ICs & BCs domains. The methodology proposed is exemplified on two SETs, namely Sozzi-Sutherland and Super Moby Dick, whose experimental databases have been made available in the ATRIUM (Application Tests for Realization of Inverse Uncertainty quantification and validation Methodologies in thermal hydraulics) project promoted by the OECD/NEA/CSNI. The results obtained are superior to those of traditional IUQ methodologies for models highly sensitive to ICs & BCs.

A sensitivity study on the PDFs treating uncertainties in severe accidents for pressurized heavy water reactors

This research article introduces a study regarding the uncertainties treatment during severe accidents for Pressurized Heavy Water Reactors (PHWRs). The present study is focused upon the unmitigated Station BlackOut (SBO) accident analysis for a CANada Deuterium Uranium (CANDU) type reactor emphasizing the impact of the uncertainties treatment on the relevant key timings of the SBO accident progression through different approaches for the uncertainty parameters’ Probabilistic Distribution Functions (PDFs). A comparison between the sensitivity analysis results is provided in the present research study.The uncertainty analysis is performed with the RELAP/SCDAPSIM code with the Integrated Uncertainty Analysis (IUA) package from the code.Results from the research would support the advancements on the best-practices for uncertainty analyses with respect to the parameter's uncertainties distribution functions. Data dispersion is a key element for the realistic quantification of uncertainties in nuclear power plant safety analyses, including severe accidents.

Development and application of Tractebel's multi-physics simulation capacity for advanced reactor safety evaluation

This paper provides a comprehensive overview of Tractebel’s current capabilities in transient simulation of Pressurized Water Reactors (PWRs) using multi-physics approaches. It highlights the use of advanced, industry-grade simulation tools and methodologies to tackle practical engineering and licensing studies for PWRs. The paper first introduces the dedicated coupling environment and its recent advancements, which have culminated in the CERES Multiphysics platform. It then delves into three key safety application areas where Tractebel’s multi-physics simulation capabilities are applied, with a focus on system transient analysis:•Small Modular Reactor (SMR) R&D Activities: Tractebel’s contribution to the Horizon 2020 McSAFER project is discussed.•Plant Safety Analyses: The paper discusses studies conducted within the framework of the WENRA RL DEC-A safety analyses.•Plant Operational Support: The paper details how multi-physics simulations are used to justify plant restarts after outages as a preventive measure against potential technical specification violations.The paper also outlines future developments aimed at enhancing and applying these multi-physics simulation capabilities to support new plant designs, including SMRs.

An integrated assessment method of real-time source term for high temperature gas-cooled reactor

To address high temperature gas-cooled reactor (HTR) nuclear power plant (NPP) safety analysis and emergency response, this paper proposes an integrated source term assessment (ISTA) method, guided by the principles of reality, rapidity and rationality. The fundamental concept of ISTA is to establish a connection between the theoretical model of source term (ST) calculation and actual conditions of HTR NPPs. According to the ISTA method, STs can be evaluated by mapping key parameters through monitoring variables during any abnormal operational conditions or accident conditions, resulting in a more realistic representation. Furthermore, the output of STs in various typical spaces within an HTR NPP further expands the application scope beyond the development of nuclear emergency protection strategies. In addition, a corresponding tool (ISTAT) for engineering application is developed and the case is given to verify the feasibility. This research highlights the technological advancement affecting radiological release scenarios and avoiding unnecessary conservatism, which also can give reference for other advanced reactor NPPs.

Development of the Reliability Assurance Program in a Brazilian nuclear power plant subsidized by a Reliability, Availability and Maintainability Model

The main objective of this work is to present a methodology for the development of a Reliability Assurance Program (RAP) specific to a PWR experimental nuclear installation, through the analysis of the installation and the development of a preliminary RAP subsidized by a Reliability, Availability and Maintainability (RAM) model. The study of an evaluation was carried out in the long-term decay heat removal of the studied experimental plant, whose data were used for application of the RAP. The necessary steps for applying the developed RAP are followed, using the data from the assessment of the studied plant, resulting in a list of components of significant risk for the Program, and in the following steps of sending the list to the experts panel, ranking of SSCs by the panel and development of the final list of significant risk SSCs for using the list in the optimization of the plant. The RAP subsidized by a RAM model will be able to work with the logical relationships between each component of the plant for their effects on energy generation and with the quantitative prediction of the magnitude of each contributor to the occurrence of high-level events, and the developed methodology can be applicable throughout the experimental plant. In this way, it will be possible to implement the RAP in the plant, which will provide a structured way to meet the regulatory requirements for its licensing. Also, it will be possible to complement the plant safety analysis report, which must contain the RAP.

Analysis of main steam line break accident on a BWR test facility using TRACE

PUMA-E is an integral test facility, which is used for extensive experimental investigations to study the behavior of Boiling Water Reactor (BWR) plants under accident conditions. The TRACE code is the best estimate thermal hydraulic system code developed by the U.S. Nuclear Regulatory Commission for performing nuclear power plant safety analysis. The aim of the present work is to conduct a short validation study of the TRACE code applied to the PUMA-E facility. Firstly, a TRACE model of the PUMA-E facility was developed based on the existing RELAP5 input deck. Then, an additional facility TRACE model with reactor pressure vessel modeled by three dimensional “VESSEL” component was also established. After that, a Main Steam Line Break accident test was simulated, and the calculation results were compared with experimental data and RELAP5 simulation results. The TRACE simulation capability to predict the accident phenomenon of BWR test facility was evaluated using both qualitative and quantitative methods. On the basis of the above research, the base case accompanying with one Gravity Driven Cooling System tank failure accident was carried out based on the three facility models. The validity and reliability of the simulation results of the different facility numerical models were mutually verified through a qualitative comparison of the simulation results.

A Reduced Order Model Based on ANN-POD Algorithm for Steady-State Neutronics and Thermal-Hydraulics Coupling Problem

The neutronics and thermal-hydraulics (N/TH) coupling behavior analysis is a key issue for nuclear power plant design and safety analysis. Due to the high-dimensional partial differential equations (PDEs) derived from the N/TH system, it is usually time consuming to solve such a large-scale nonlinear equation by the traditional numerical solution method of PDEs. To solve this problem, this work develops a reduced order model based on the proper orthogonal decomposition (POD) and artificial neural networks (ANNs) to simulate the N/TH coupling system. In detail, the POD method is used to extract the POD modes and corresponding coefficients from a set of full-order model results under different boundary conditions. Then, the backpropagation neural network (BPNN) is utilized to map the relationship between the boundary conditions and POD coefficients. Therefore, the physical fields under the new boundary conditions could be calculated by the predicated POD coefficients from ANN and POD modes from snapshot. In order to assess the performance of an ANN-POD-based reduced order method, a simplified pressurized water reactor model under different inlet coolant temperatures and inlet coolant velocities is utilized. The results show that the new reduced order model can accurately predict the distribution of the physical fields, as well as the effective multiplication factor in the N/TH coupling nuclear system, whose relative errors are within 1%.

Integrated Safety and Security Analysis of Nuclear Power Plants Using Dynamic Event Trees

Nuclear security relies on the method of vital area identification (VAI) to determine which locations within the nuclear power plant (NPP) need to be protected from radiological sabotage. The VAI methodology uses fault trees (FTs) and event trees (ETs) to identify locations in the NPP that contain vital equipment: structures and components that may result in reactor significant core damage if direct or indirect sabotage occurred. However, the traditional FT/ET process cannot fully capture the dynamics of NPP systems and mitigating measures at play. Existing safety systems or possible operator procedures may be able to avert or mitigate core damage despite the loss of one or more vital areas. Dynamic probabilistic risk assessment (DPRA) methodologies are those that, unlike traditional probabilistic risk assessment, explicitly consider time effects when modeling a system. One common DPRA methodology is that of the use of dynamic event trees (DETs) that drive computer models of a system with user-specified branching conditions to account for uncertainties in a scenario. The DPRA process allows analysts to explore the uncertainties and state space of a scenario in a systematic fashion. A scenario was developed that uses the novel leading simulator/trailing simulator methodology to perform a DET analysis of a combined nuclear safety and nuclear security analysis. The scenario under consideration models the successful sabotage of a vital area by adversaries and determines the effects of timing and the extent of sabotage, as well as possible recovery actions, on the state of the plant. The results of this integrated analysis include the timing and extent of core damage as well as the extent of any radiological release that may occur as a result of sabotage.

Application of the COCOSYS code in the safety evaluation of Czech nuclear power plants

Abstract Safety analyses of nuclear power plant accidents are a key tool in demonstration of nuclear safety and reliability. The chain of the safety analyses comprises several follow-up activities like neutronic core calculation, primary system simulation and containment response to a loss-of-coolant accident. The analyses can be done with numerous computational tools like the GRS codes ATHLET and COCOSYS. The COCOSYS and its predecessor RALOC have been used in a variety of safety analyses at UJV conducted in the framework of technical support of the Czech nuclear power plants. Application of the code in the field of nuclear power plant safety analyses is ruled by the Czech legislative, which defines a process called standardization. The standardization procedure equivalent to validation and verification process provides further independent validation on the level of user organization. In past decades, COCOSYS has been successfully used in numerous DBA and DEC-A analyses of VVER-1000/V-320 containment, evaluating both thermal hydraulic conditions as well as fission product source term. This paper summarizes independent COCOSYS validation at UJV, including setup of the iodine dry paint deposition model to Ameron Amerlock paint. Furthermore, a VVER-1000/V-320 containment model is presented and analysis of DBA and DEC-A LB LOCA at VVER-1000/V-320 is conducted. The evaluation of the results aims at confirmation of acceptance criteria related to containment and impact of containment spray system. An emphasis is given to evaluation of containment source term, where comparison of both events is conducted for release of iodine, cesium, and xenon.

Application of data driven modeling and sensitivity analysis of constitutive equations for improving nuclear power plant safety analysis code

Constitutive equations in a nuclear reactor safety analysis code are mostly empirical correlations developed from experiments, which always accompany uncertainties. The accuracy of the code can be improved by modifying the constitutive equations fitting wider range of data with less uncertainty. Thus, the sensitivity of the code with respect to the constitutive equations is evaluated quantitatively in the paper to understand the room for improvement of the code. A new methodology is proposed which first starts by dividing the thermal hydraulic conditions into multiple sub-regimes using self-organizing map (SOM) clustering method. The sensitivity analysis is then conducted by multiplying an arbitrary set of coefficients to the constitutive equations for each sub-divided thermal-hydraulic regime with SOM to observe how the code accuracy varies. The randomly chosen multiplier coefficient represents the uncertainty of the constitutive equations. Furthermore, the set with the smallest error with the selected experimental data can be obtained and can provide insight which direction should the constitutive equations be modified to improve the code accuracy. The newly proposed method is applied to a steady-state experiment and a transient experiment to illustrate how the method can provide insight to the code developer.

Numerical investigation of the depressurization performance in blowdown pipe using method of characteristics

The coolant leaks quickly after the break of the high-pressure pipeline system. The rapid pressure decrease and fluctuation will further affect the integrity of the pressure boundary and the effective heat removal of the reactor core in the early phase of blowdown. Generally, numerical simulation plays a vital role in the safety analysis of nuclear power plants (NPPs) under severe accidents like loss of coolant accident (LOCA) conditions. In this paper, one-dimensional two-phase hydrodynamic equations and essential closure models are established to describe the internal pipe flow firstly. Then, method of characteristics (MOC) realized by a self-programming code is introduced, which is different from the numerical scheme of system analysis code RELAP5 in the simplification and discretization. Finally, combined with the Edwards blowdown experiments, the two numerical solution methods are compared on modeling, sensitivity analysis, and calculation results under various working conditions. The influence of back pressure variation on the solution of thermal–hydraulic parameters has been considered during the blowdown process. The solution result of RELAP5 is highly related to the set discharge coefficients, which are usually given according to experience. However, the solution result of MOC can be enveloped by that of RELAP5 with different discharge coefficients. In addition, the results solved by MOC are closer and more conservative to the experimental values during the early phase of blowdown with low void fraction fluid. It lays a foundation for the further calculation and safety analysis of the complex loop system under LCOA conditions.

Development of analysis program for direct containment heating

Direct containment heating (DCH) is one of the potential factors leading to early containment failure. DCH is closely related to safety analysis and containment performance evaluation of nuclear power plants. In this study, a DCH prediction program was developed to analyze the DCH loads of containment vessel. The phenomenological model of debris dispersal, metal oxidation reaction, debris-atmospheric heat transfer and hydrogen jet burn was established. Code assessment was performed by comparing with several separate effect tests and integral effect tests. The comparison between the predicted results and experimental data shows that the program can predict the key parameters such as peak pressure, temperature, and hydrogen production in containment well, and for most comparisons the relative errors can be maintained within 20%. Among them, the prediction uncertainty of hydrogen production is slightly larger. The analysis shows that the main sources of the error are the difference of time scale and the oxidation of cavity debris.

Research on the influence of chemical reaction model over MCCI heat transfer process

In the process of a severe accident that caused loss of heat removal capacity and subsequently led to reactor vessel failure, the Molten Core corium drops into cavity to Interact with Concrete (MCCI). It is essential for the safety analysis of nuclear power plants. The MCCI mechanism process is complicated and usually simplified into chemical reactions and heat-mass transfer in simulation. This paper analyzed a chemical equilibrium model and selected suitable substances attending chemical reactions, then investigated the relationship between chemical reactions and heat transfer. This model was compared via an ACE-L6 test, and the simulation results were matched well with the experiment and CORCON/VANESA code. Afterward, it was integrated to analyze the entire MCCI sequence of a PWR. The sensitivity analysis of chemical model parameters was performed. The results showed that the chemical reaction model has important influences on the MCCI heat-mass transfer process, especially at the early stage.

Control rod shadowing effect in PWR core utilizing Urania-Gadolinia fuel

Safety analysis of nuclear power plants requires the assessment of all the systems important to safety. Control rods are one of the important safety systems for nuclear reactors. The control rod worth is affected by location of control rod banks, enrichment of uranium fuel in the assembly where the bank is inserted and the enrichment of gadolinium in the assembly. Also, the Interactions between control rods, phenomena known as shadowing effect, influence the worth of control rods. In this paper shadowing effect and the factors affecting the control rod worth are investigated for four types of control rod banks of an EPR Reactor. Parameters investigated are control rod worth, degree of interaction between rods and amplification factor. Moreover, the effect of interactions between control rod banks on the fluctuation of flux distribution in the core is studied. The calculations are evaluated at BOC and EOC using MCNP6 code. The results showed that the location of control rod bank in the core has the highest effect on the worth of a certain bank, second is the number of control rod bank groups in the core, followed by the fuel enrichment. The gadolinium burnable absorber has a slight effect on the control rod worth. The results obtained represent an important contribution that would support safety analysis of the EPR reactor design.

Jet impingement in high-energy piping systems, part II: Model improvement and guidance development

The Standard model in ANSI/ANS-58.2 (1988) has been widely employed in the design and safety analysis of nuclear power plants for jet impingement under postulated pipe rupture accidents. However, the potential non-conservatisms presented in Standard Review Plan (2015) need to be addressed when evaluating this model. The current work aims to improve the model on jet prediction including jet geometry and stagnation pressure, as well as to address the potential non-conservatisms. Based on the model evaluation results in Part I, new correlations for the jet core length and stagnation pressure in subcooled water and saturated water/two-phase jets are implemented to replace the existing correlations in the Standard model. It is shown that the revised code with the newly developed correlations predicts the stagnation pressure generally higher or conservatively for all conditions investigated.On the potential blast wave effect, the current work develops an approach to quantify the over-static and reflected pressures using the equivalent weight of TNT approach in literature. To employ this approach, the released energy in the equivalent of TNT needs to be determined from jet impingement. This energy is calculated using the vessel burst energy concept with conservative assumptions in the current work. For jet geometry, dynamic models to predict the jet spreading angle for saturated steam jets are compared with the Prandtl-Meyer angle for under-expanded supersonic jets. It is found that the jet spreading angle for saturated steam jets in the Standard model generally agrees with the dynamic model in literature for a wide range of initial vessel stagnation pressure conditions. Regarding the pressure distribution within jets, it is confirmed that edge-peaking pressure profiles can be observed for saturated steam jets at far distances from the break plane (z > 3.3 D); however, the dimensionless pressures at these locations are generally less than 0.07. Regarding the jet dynamic loading effect, the literature review suggests that dynamic amplification and resonance of impinged downstream structures and components might not occur for jet impingement in nuclear power plants.

Transformation of classical PSA and DSA into the form of conditional event tree: An approach of human action in time dependent core damage risk

Although the accident progression in NPPs follows a dynamic trend, most industrial PSA studies used in the design process include static modeling with time-independent construction and without a well-defined linkage between the DSA and PSA. This is mostly due to the complicated methods for integration and post-processing procedures. This article aims to develop a straightforward combination process between DSA and PSA applicable in nuclear power plant safety analysis for considering the dynamic effects of operator task timing on the Core Damage Frequency (CDF). The case study is a Main Steam Line Break (MSLB) accident for a PWR. The combination approach includes a four-step procedure for the consideration of operator timing on the final core damage contribution. Thus, the CDFs are calculated based on different timings of human action, and the effects of timing on the CDF are presented in the form of a Conditional Event Tree (CET). According to the results, the total CDF is highly dependent on the HEP, and it can be reduced by providing additional time to the operator. The results also showed this additional time can be provided by the successful performance of the pressurizer safety valve as a passive device.

Investigating the influence of the variation of the load characteristic on the occurrence frequency of Grid-related Loss of Offsite Power using a probabilistic-deterministic methodology

Generic frequencies are used for Grid-related Loss of Offsite Power (GR-LOOP) in the Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPPs). When historical databases are used, the influence of variation in the grid conditions over time and the impact of the NPP location are not considered in the estimation of GR-LOOP frequency. Therefore, relying purely on historical data induces uncertainty in total Core Damage Frequency (CDF).This paper aims at assessing the influence of the dynamic load behavior on the GR-LOOP frequency, for different locations. For that purpose, sensitivity cases are defined. The occurrence frequency of GR-LOOP following the occurrence of a three-phase-short-circuit fault on a transmission line is evaluated via a probabilistic-deterministic methodology for all cases.The minimum and maximum estimated values of the GR-LOOP frequency, 2.98E-04/year and 9.95E-03/year respectively, indicate the dependence on the variation in either the dynamic load behaviors or the NPP locations.The results reveal that GR-LOOP frequency varies with load model in the initial grid. In the presented probabilistic-deterministic methodology it is beneficial to consider the influence of timely changes of the grid conditions on the GR-LOOP occurrence frequency for different NPPs’ locations, especially where relevant or sufficient generic databases are not available.

Application of continuous Markov-chain Monte-Carlo method to multi-unit risk evaluations considering interdependence of accident progression among multiple units

ABSTRACT The accident in Fukushima Dai-ichi nuclear power plants reconfirms the necessity of the safety assessment considering multiple nuclear reactor units. However, consideration of the interdependency among safety systems or events in multiple units as well as the time dependency of accident progressions is difficult in the conventional event tree method, which is widely used in probabilistic risk assessments. Recently, the continuous Markov-chain Monte-Carlo (CMMC) method coupled with a plant safety analysis or a severe accident analysis code has been paid attention to address these issues. In the present study, the CMMC coupling method is applied to the risk assessment of multiple units to clarify the benefits and issues to be resolved of this method. Since the CMMC coupling method requires many executions of an accident analysis code, a meta-model that simplifies systems and physical phenomena in accidents is used in this study to reduce computational cost. Furthermore, the inverse transform sampling method is newly adapted. Numerical analyses of BWR accident under station blackout with loss of cooling capability are carried out considering the correlation among the availabilities of mitigation systems. The results suggested that the CMMC coupling method can quantitatively treat the interdependency and time dependency among events in multiple units.