Safety Analysis Of Nuclear Power Plants Research Articles

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.

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.

Numerical and experimental analysis of cylindrical-type PAR catalyst behaviour

Passive autocatalytic recombiners (PARs) are essential safety systems used in nuclear power plants (NPPs) to prevent hydrogen explosions during severe accidents. This study investigates the operational behaviour of cylindrical-type catalysts used in a PAR. The study employs experimental and computational fluid dynamics (CFD) analyses to evaluate the catalyst temperature distribution and hydrogen conversion inside the PAR channel. Experimental analysis measures the hydrogen conversion of the catalyst and temperature distribution by means of hydrogen sensors and an infrared camera. The CFD analysis uses a three-dimensional (3D) model developed in STAR-CCM+ code to simulate the flow and recombination reaction in a cylindrical-type PAR catalyst section. Results indicated that the catalyst has decent conversion efficiency. Furthermore, the temperature over the catalyst section is evenly distributed and it does not exceed the lower hydrogen ignition limit. CFD analysis of the Schmidt number demonstrates that the flow inside the PAR is highly turbulent and Sct values is in the range of 0.2–0.28. It was found that the recombination reaction occurs in the diffusion regime. The recombination reaction primarily occurs in the catalyst's lower and central parts. Finally, this paper proposes the functional dependence to estimate the efficiency of the entire PAR based on the Sherwood equation and mechanistic transport approach. The results of this study provide crucial insight into the cylindrical-type catalyst operational behaviour in PARs and the CFD model design of cylindrical-type catalysts for the safety analysis of NPPs.

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%.

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.

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.

Core Modeling and Simulation of Peach Bottom 2 Turbine Trip Test 2 Using CASMO5/TRACE/PARCS

The Purdue Advanced Reactor Core Simulator (PARCS) three-dimensional neutron kinetics code and the TRACE nuclear systems analysis code were interfaced. This provides a best-estimate coupled code system for performing transient plant calculations with reactivity feedback from a detailed core model, significantly contributing to nuclear power plant safety analyses. This study performed steady-state and transient simulations of Peach Bottom 2 Turbine Trip Test 2 (PB2 TT2) using the CASMO5/TRACE/PARCS coupled code. Consequently, CASMO5/TRACE/PARCS simulates the rapid positive reactivity addition caused by the sudden closure of the turbine stop valve. Specifically, the discrepancy in the maximum total power during the transient condition was within 3% compared with the PB2 TT2 experimental data. Furthermore, the sensitivity of the thermal-hydraulic channel (CHAN) component modeling in the coupled CASMO5/TRACE/PARCS code revealed that the number of CHAN components influenced the assembly radial power peaking factor in the PB2 TT2 transient calculation.

Inclusion of Aircraft Crash into NPP Design Bases and Probabilistic Justification of Loads on Civil Structures and Equipment

The article is about nuclear power plant (NPP) safety analysis in case of aircraft crash. Specifically, the article considers the following problems: inclusion of aircraft crash into NPP design bases regarding calculation of frequency of an aircraft crash into NPP; aspects of justification of loads on NPP structures, systems and components (SSCs) caused by mechanical action of a primary missile – aircraft fuselage impact.
 Probabilistic characteristics of such random parameters as frequency of aircraft crash and direction of aircraft trajectory are determined by the results of analysis of world statistics of aviation accidents. Method of calculation of aircraft crash frequency on structures, buildings and NPP as a whole is presented. It takes into account options of accidental and intentional aircraft crashes and various aircraft approach scenarios.
 Procedure of probabilistic justification of loads on civil structures under aircraft impact is described. The loads are specified so as not to exceed allowable value of failure probability of NPP as a whole. Calculation of failure frequency of civil structures of existing NPP is given as an example to show analysis in case of a crash of an aircraft heavier than considered in NPP design.
 Procedure of probabilistic justification of dynamic loads on NPP equipment in case of aircraft impact is described. Method of floor response spectra (FRS) calculation with the required non-exceedance probability is given. Probabilistically justified loads in case of intentional aircraft impact (act of terrorism) are also considered. Additionally it is presented how internal forces calculated with the use of FRS with the required non-exceedance probability can be summed to provide analysis of subsystems.

Best estimate plus uncertainty analysis of small break loss of coolant accidents with different break sizes for CPR1000

The best estimate plus uncertainty (BEPU) is an advanced method which has been widely adopted in nuclear power plant safety analysis. In order to investigate the transient behaviors of the pressurized water reactor CPR1000 during the cold leg small break loss of coolant accident (SBLOCA) at different break sizes, BEPU analysis of the SBLOCA was carried out in this study, and the sizes of the break were 0.2%, 0.4%, 0.6%, 0.8% and 1.0% of the cross-sectional area of the cold leg, respectively. The minimum collapsed level (MCL) in the active core section was determined to be the figure of merit (FOM). The base case calculations, uncertainty propagation calculations and sensitivity analysis calculations were performed with the NUSOL-SYS code. Based on the results, the transient behaviors were analyzed, the lower 95/95 tolerance limits of the MCL at different break sizes were quantified, and the important input parameters were determined.

The role of verification & validation process in best estimate plus uncertainty methodology development

In the process of a best estimate plus uncertainty (BEPU) methodology implementation for nuclear power plant safety analyses, the verification and validation (V&V) process including the uncertainty quantification (UQ) of the used computer codes and plant models plays an essential role. From the technical point of view, a BEPU methodology must be based on fully verified and validated codes and models, with quantified key model uncertainties. From the regulatory point of view, the adequacy of the simulation codes and plant models for the intended BEPU application must be assessed through the VVUQ process. These high level requirements of VVUQ result in high-cost for BEPU methodology development, preventing wider applications. Pragmatic, graded applications and practices of VVUQ in the BEPU methodology development are needed to allow to take full benefit of BEPU applications.

Development of good practice guidance for quantification of thermal-hydraulic code model input uncertainty

Taking into account uncertainties is a key issue in nuclear power plant safety analysis using best estimate plus uncertainty methodologies. It involves two main types of treatment depending on the variables of interest: input parameters or system response quantity. The OECD/NEA PREMIUM project devoted to the first type of variables has shown that inverse methods for input uncertainty quantification can exhibit strong user-effect. One of the main reasons was the lack of a clear guidance to perform a reliable analysis. This work is precisely devoted to the development of a first good practice guidance document for quantification of thermal-hydraulic code model input uncertainty. The developments have been done in the framework of the OECD/NEA SAPIUM project (January 2017–September 2019). This paper provides a summary of the main project outcome. Recommendations and open issues for future developments are also given.

A conservative pressure based solver with collocated variables on unstructured grids for two-fluid flows with phase change

Numerical simulation of two-phase flows based on the two-fluid six-equation model is the focus of this work. This approach is widely used in thermo-hydraulics for many industrial applications, particularly nuclear power plant safety analysis. A semi-implicit numerical method that has been successfully adopted in several industrial thermo-hydraulic codes, because of its efficiency and robustness, has been extended. In this paper, the governing equations are solved on unstructured grids with collocated variables to accommodate complicated geometries, and a deferred-correction method is used to deal with the non-orthogonality of unstructured grids. In addition, the numerical method is conservative, meaning that it is capable of capturing a shock wave exactly in single-phase flows, and ensuring the conservation of physically conservative quantities of a two-phase flow mixture. Numerical benchmarks and industrial test cases are performed in order to validate the numerical method and to evaluate its behavior with respect to different physical configurations.

Code crosswalk of Fukushima-like simulations for Chinshan BWR/4 NPP using MELCOR2.1/SNAP, TRACE/SNAP, PCTRAN and MAAP5.03

After Fukushima Daiichi disaster, the safety analysis of severe accidents became one of the safety concerns in Taiwan. At the same time, Taiwan Power Company built up an event basis strategy called Ultimate Response Guideline (URG) to cope with the accident like Fukushima. The analysis of the effectiveness of URG also became the most popular issue in Taiwan. However, the beyond design basis accident was very complicate and combined with many different aspects. Although researchers in Taiwan did lots of analysis of URG using different codes with different aspects, the differences of codes in the severe accident analysis were not mention in any of those studies. Thus, a Fukushima-like accident of Chinshan Nuclear Power Plant (NPP) was done by four codes in this study: MELCOR2.1, TRACE, PCTRAN and MAAP5.03. All four codes shared the same boundary conditions in the accident simulations. The crosswalk of these codes was discussed in detail and showed the similarity and differences among these codes. Those similarity and differences can help the improvement of URG analysis and make that this kind of analysis has more reliability. Also, the application field of each code can be clearly distinguished and help the improvement of code use in future NPP safety analysis.

Preliminary study of applying adjoint-based mesh optimization method to nuclear power plant safety analysis

Although the number of meshes and their distribution are important parameters that affect both computation resources and the accuracy of the numerical results, meshing of most of all computational problems still depends highly on the users’ experiences. In this paper, a deterministic adjoint-based method of optimizing mesh distribution is proposed. The developed method is applied to a nuclear power plant safety analysis. The mesh optimization was performed with 1D steady state cylindrical nuclear fuel problem first. Radial and axial mesh distributions are optimized respectively. With no surprise, the optimized mesh system performs superior than the same number of uniformly meshed system. However, it was unexpected that the optimized mesh retains generality and therefore, the optimized mesh system can be still the best mesh system for given number of meshes under different condition or even during transient analysis. The authors applied the optimized mesh distribution to nuclear system safety analysis. A large pressurized water reactor cold leg guillotine break (LBLOCA) scenario was analyzed and the consequence of different mesh systems is investigated and discussed. From this preliminary study the usefulness and implication of the adjoint based mesh optimization method for the nuclear safety analysis is uncovered.

Evaluation of thermal hydraulic safety of a nuclear fuel assembly in a mast assembly of nuclear power plant

We investigated the thermal–hydraulic safety of spent fuel bundles installed in a mast assembly, assuming a lock-up accident of a fuel transfer system. For this, flow analysis with Computational Fluid Dynamics (CFD) for the natural convection and subsequent safety evaluation with 1-D Nuclear Power Plant (NPP) safety analysis code were carried out. Prior to the natural convection analysis for the mast assembly using CFD code, we performed benchmark calculations against two experimental data sets obtained by Betts and Bokhari (2000) and PNL (Pacific Northwest Laboratory, 1980). This was done to select the proper physical models for a natural convection flow analysis and to ensure the reliability of prediction of natural convection flow in the fuel bundle geometry. Finally, we performed a main natural convection analysis for the fuel assembly inside the mast assembly. From this calculation, we observed a stable natural circulation flow between the mast assembly and pool side, and obtained coolant velocity at the inlet of the spent fuel bundle. This flow condition is given as a boundary condition for the 1-D NPP safety analysis code, which is used to predict the Critical Heat Flux and then departure from the nucleate boiling ratio.

Content, Completeness, and Consistency of Analytical Models for Regulatory Consideration

This paper reviews the historical and contemporary precedence regarding the development of knowledge, its reformulation in computer codes, and subsequent application in decision making. It highlights the practical challenges of this process as it applies to the investigation of engineered systems to deliver on both promised benefits and protection from postulated failures. A model for demonstrating model content, completeness, and consistency is described, invoking and extending a knowledge/content model attributed to Popper. While the specific example examining the evolution of the thermal-hydraulic knowledge base applied for nuclear power plant safety analysis and its capture in the RELAP series of computer analysis codes is presented, the framework is general, true to the scientific method, and thus broadly applicable. It concludes that while content of our knowledge base is perpetually increasing, completeness and consistency are fundamentally unattainable; however, within a well-designed evaluation methodology, measurable proof, sufficient for regulatory deliberation, is possible.

VVER-440 Steam Generator’s Two-Phase Flow Analysis

The subject of this report is creation of three-dimensional thermal hydraulic model of horizontal steam generator for Dukovany nuclear power plant. A procedure is presented for simulation and analysis of secondary side of PGV-440 steam generator for nominal and increased reactor power. A two-fluid approach is applied for modeling physical processes inside the steam generator. Physical models were implemented in ANSYS Fluent CFD environment using User Defined Functions (UDFs). Results from this thermal hydraulic numerical model can be used for various other subsequent nuclear power plant operations and safety analysis.

Vulnerability models for environmental risk assessment. Application to a nuclear power plant containment building

Environmental risk management consists of making decisions on human activities or construction designs that are affected by the environment and/or have consequences or impacts on it. In these cases, decisions are made such that risk is minimized. In this regard, the forthcoming paper develops a close form that relates risk with cost, hazard, and vulnerability; and then focuses on vulnerability. The vulnerability of a system under an external action can be described by the conditional probability of the degrees of damage after an event. This vulnerability model can be obtained by a simplicial regression of those outputs, as a response variable, on explanatory variables. After a theoretical explanation, the authors present the case study of a nuclear power plant containment building. Once a given overpressure is registered inside the containment building, three possible outputs are to be considered: serviceability, breakdown, and collapse. The study consists of three steps: (a) modelling the containment building using the finite element method; (b) given an overpressure, simulating uncertain parameters related to material constitutive equations in order to obtain the corresponding proportions; (c) performing a simplicial regression to obtain a meaningful vulnerability model. The simulation provides normalized-to-unity outputs under the overpressure conditions. The obtained vulnerability model is in definite correspondence with previous results in nuclear power plant safety analysis reports.