System Analysis Code Research Articles

Application of risk informed safety margin characterization to the analysis of a pressurized water reactor

The sequence Core Damage Frequency of a pressurized water reactor has been quantified in three initiating events that are Medium Break Loss of Coolant Accident, Small Break Loss of Coolant Accident and Steam Generator Tube Rupture, following a realistic methodology called risk informed safety margin characteristic. The surrogate plant analyzed in the study is a typical pressurized water reactor. The plant adopted two Westinghouse Three-Loop Pressurized Water Reactors with rated thermal power of 2,830 MWt. The phenomenon identification and ranking table is applied for uncertainty analysis. The mitigation actions as described in plant specific Probabilistic Risk Assessment include cooldown and depressurization, emergency cooldown and depressurization, high head safety injection, high head safety recirculation, low head safety recirculation and Refueling Water Storage Tank replenishment. These mitigation actions are analyzed by thermal hydraulic system analysis code RELAP5-3D to determine the successfulness of the actions. The uncertainties of input parameters of the plant conditions are included, and the time of mitigation action executed is treated as one of the input uncertainties. The results of realistic methodology show a decrease in Core Damage Frequency for all three analyzed events in comparison with conventional methodology. The differences between three initiating events are also discussed.

Toward the Development of a Comprehensive Nuclear System Analysis Code Based on Two-Fluid Model: Starting with an Isentropic Approach

Toward the Development of a Comprehensive Nuclear System Analysis Code Based on Two-Fluid Model: Starting with an Isentropic Approach

Development of system analysis code and its application in transient simulation of supercritical CO2 Brayton cycle direct cooled reactors

The supercritical carbon dioxide (s-CO2) Brayton cycle direct cooled reactor system has the characteristics of high cycle efficiency, compact equipment, and simple layout, which has attracted the attention of researchers recently. The use of s-CO2 Brayton cycle and the elimination of intermediate heat exchangers make the transient characteristics of this reactor different from light water reactors which have rich operating experience. Transient simulation and safety analysis with system code is an essential stage in reactor design and safety verification. Therefore, a system code, named NUSAS (NUclear Safety Analysis and Simulation), was developed in this study for the transient simulation of s-CO2 Brayton cycle direct cooled reactor system. This code is developed based on the advanced two-fluid two-pressure seven-equation model and has the capability of modelling the turbomachinery in s-CO2 Brayton cycle direct cooled reactor system. The transient characteristics during core power step and precooler cooling water mass flow rate step were analyzed. The load tracking and operation control strategy of the reactor was designed, and transient simulations of the wide range (100% to 50%) load tracking process were conducted. According to the simulation results of NUSAS, the transient characteristics of s-CO2 Brayton cycle direct cooled reactor were demonstrated when operating conditions change.

Investigation of quenching phenomena during the relooding phase against the FLECHT-SEASET experiment by using RELAP5/MOD3.3

The reflood model of RELAP5/MOD3.3 (patch 4) was assessed by using the FLECHT-SEASET tests. The tests were conducted to have a better understanding of the postulated loss of coolant accident in a light water reactor (LWR). The best-estimate system analysis code was used to simulate these accident scenarios, especially in Design Basic Accident such as Loss-Of-Coolant-Accident (LOCA). The primary purpose of this report was to assess the accuracy of a computer system analysis code by using RELAP5/MOD3.3 in comparison to actual test results taken from the FLECHT-SEASET tests in which the reflood model was built in. In RELAP5’s simulation cases, the various boundary and initial conditions, such as power supplied and reflooding rate were selected. As a result, the RELAP5 looked to be accurate in predicting the quenching time and rod surface temperature for this particular case. However, the RELAP5 code under-estimated the rod surface temperature in comparing with the experimental data of the FLECHT-SEASET tests. Accordingly, for this high flooding rate and particular reactor power level that the reflooding model in RELAP5 could be possible used for predicting the reflooding phenomena during the LOCA accident.

Validation of In-House Imaging System via Code Verification on Petunia Images Collected at Increasing Fertilizer Rates and pHs.

In a production environment, delayed stress recognition can impact yield. Imaging can rapidly and effectively quantify stress symptoms using indexes such as normalized difference vegetation index (NDVI). Commercial systems are effective but cannot be easily customized for specific applications, particularly post-processing. We developed a low-cost customizable imaging system and validated the code to analyze images. Our objective was to verify the image analysis code and custom system could successfully quantify the changes in plant canopy reflectance. 'Supercascade Red', 'Wave© Purple', and 'Carpet Blue' Petunias (Petunia × hybridia) were transplanted individually and subjected to increasing fertilizer treatments and increasing substrate pH in a greenhouse. Treatments for the first trial were the addition of a controlled release fertilizer at six different rates (0, 0.5, 1, 2, 4, and 8 g/pot), and for the second trial, fertilizer solution with four pHs (4, 5.5, 7, and 8.5), with eight replications with one plant each. Plants were imaged twice a week using a commercial imaging system for fertilizer and thrice a week with the custom system for pH. The collected images were analyzed using an in-house program that calculated the indices for each pixel of the plant area. All cultivars showed a significant effect of fertilizer on the projected canopy size and dry weight of the above-substrate biomass and the fertilizer rate treatments (p < 0.01). Plant tissue nitrogen concentration as a function of the applied fertilizer rate showed a significant positive response for all three cultivars (p < 0.001). We verified that the image analysis code successfully quantified the changes in plant canopy reflectance as induced by increasing fertilizer application rate. There was no relationship between the pH and NDVI values for the cultivars tested (p > 0.05). Manganese and phosphorus had no significance with chlorophyll fluorescence for 'Carpet Blue' and 'Wave© Purple' (p > 0.05), though 'Supercascade Red' was found to have significance (p < 0.01). pH did not affect plant canopy size. Chlorophyll fluorescence pixel intensity against the projected canopy size had no significance except in 'Wave© Purple' (p = 0.005). NDVI as a function of the projected canopy size had no statistical significance. We verified the ability of the imaging system with integrated analysis to quantify nutrient deficiency-induced variability in plant canopies by increasing pH levels.

Verification and Demonstration of One-Dimensional Freezing Model in SAM for Salt-Cooled Reactor Analysis Applications

This work presents the development and implementation of the one-dimensional freezing model in system analysis code SAM (System Analysis Module), code verification using analytical solutions, and code demonstration of a postulated overcooling transient for a fluoride salt–cooled high-temperature reactor (FHR) system and safety analysis applications. This paper first summarizes the freezing model, finite element numerical method, and special numerical treatment for handling transitions between single- and two-phase conditions. Analytical solutions are derived for two cases, with and without solid walls, for code verification purposes. As expected, the numerical results predicted by SAM agree very well with the analytical solution. A code demonstration is then performed on a postulated protected overcooling transient event of a generic reference pebble bed FHR design. The code was found to successfully predict salt freezing during such a postulated event. However, due to the lack of salt freezing testing data, code validation is not performed in this work, but will be pursued in future studies when such data become available.

Development and application of data-driven CHF model in system analysis code

In nuclear reactor systems, when the fuel rods reach the critical heat flux (CHF), a sharp increase in fuel temperature occurs due to a drastic reduction in heat transfer capacity, thus posing a considerable risk to the reactor’s safety. Consequently, accurate CHF prediction holds paramount importance in accurately simulating accident scenarios and enhancing the overall safety of reactor systems. To tackle the challenge of limited prediction accuracy in existing CHF models, this study initially established a comprehensive database by utilizing available experimental data and lookup tables. Subsequently, various methodologies, including the Back Propagation Neural Network (BPNN), Random Forest (RF), and Physics-Informed Machine Learning (PIML), were employed to develop multiple CHF prediction models, and their performance was thoroughly evaluated. Furthermore, the optimal CHF model was integrated into the self-developed analysis code ARSAC, which was then validated using the ORNL-THTF experiment. The results indicated that the BPNN-based model not only demonstrated exceptional prediction accuracy but also exhibited rapid calculation speeds. Notably, the average relative error between the experimental data points and the calculation results of the modified code is 3.64%, while for the original code, it is 22.84%. This study effectively leverages the strengths of data-driven approaches, providing a robust technical solution for high-precision, efficient, and adaptive numerical prediction and analysis of reactor accidents.

Numerical Simulation and Experimental Comparison of System Analysis Module 1D Mixing Model for Cold Shock Transients in the Gallium Thermal-Hydraulic Mixing Facility

Liquid metals are being investigated as coolants in many advanced reactor designs because of their high thermal conductivity and effectiveness at high temperatures. However, they often pose challenges to reactor operation and safety because of the complex thermal mixing and stratification in the plenum of pool-type reactor designs. The advanced system analysis code System Analysis Module (SAM) currently under development at Argonne National Laboratory aims to develop and implement thermal mixing models to accurately capture these complex thermal fluid behaviors. In this study, the SAM thermal mixing model was compared against experimental data from the Gallium Thermal-Hydraulic Experiment facility, a scaled liquid metal test facility that uses gallium as a surrogate fluid to investigate the stratification and thermal mixing of low-Prandtl-number fluids in the upper plenum of a liquid metal–cooled reactor. Two cold shock transient cases were used: one with stable stratified flow (Ri = 32) and one with stronger thermal mixing (Ri = 0.5). The resultant temperatures were then compared with the experimental temperatures over the entire plenum to assess the ability of the mixing models to capture the thermal behavior and to better correspond mixing parameters to various flow scenarios. Generally, the zero-dimensional mixing model was more capable of capturing the bulk temperature of the component modeled assuming that an accurate mass flow rate was provided, but it was inherently unable to capture thermal gradients in space. The one-dimensional mixing model was capable of capturing that the thermal gradients provided accurate selection of the mixing coefficients. The temperature at the outlet junction was compared over time for each of the mixing models with the recorded experimental temperature. The implemented mixing models demonstrated the ability to effectively capture the overall thermal behavior for stronger mixing scenarios but struggled with more stably stratified flows. It was found that a system analysis code’s covering of the entire range of different operating conditions still remains a challenging task, and it is suggested that further model and closure improvements are necessary to accurately capture complex thermal mixing and stratification phenomena.

Development of the thermal hydraulic analysis code, TASS/SMR-S, under multi-dimensional motion for a reactor under inclination and rolling conditions

The development of a thermal hydraulic analysis code that considers ship motion is important because it would adjust for additional motion effects that are not present in a land-based system analysis code. The TASS/SMR-S code, a one-dimensional system analysis code, was modified to predict the thermal hydraulic behavior of a nuclear reactor under the conditions of an ocean environment. To validate the improved TASS/SMR-S code, the calculation results were compared with the analytical solutions for a conceptual problem involving swaying, rolling, and compounded motions; experimental data were collected under inclination and rolling conditions. The modified TASS/SMR-S code exhibited good agreement with analytical solutions for the conceptual problem and experimental results obtained under ship motion conditions. In conclusion, the modified code was able to simulate thermal hydraulic behavior under ship motion conditions.

Numerical simulation on in-vessel molten corium behavior with external vessel cooling using smoothed particle hydrodynamics

The in-vessel retention through external reactor vessel cooling (IVR-ERVC) strategy is a key management strategy for early termination of a nuclear severe accident that can threaten the integrity of the reactor vessel. To simulate the physical phenomena of the molten corium, the smoothed particle hydrodynamic (SPH) method is utilized in this study. The SPH method is a Lagrangian computational fluid dynamic (CFD) method that can simulate multi-fluid stratification, turbulence, natural circulation, radiative heat transfer, thermal ablation, and crust formation. To address the external vessel cooling, it is coupled with a conventional 1-D nuclear system analysis method. The 1-D system analysis code can calculate the two-phase natural circulation of cooling water and the convective heat transfer on the external reactor vessel wall. These two simulation codes exchange the temperature and heat flux of the reactor vessel outer wall. This study numerically simulated the IVR-ERVC strategy for a Korean high-power reactor and compared it with the traditional lumped parameter method (LPM). Unlike LPM, this study provides localized detailed data about the thermal hydraulic behavior of molten corium and visualization of phenomena in the IVR-ERVC strategy. This enhances our understanding of the phenomena in IVR-ERVC strategy and introduces new perspectives.

Assessment of MARS-KS prediction capability for natural circulation flow in passive heat removal system

Considering that system analysis codes are used for the evaluation of the performance of Passive Safety Systems (PSSs), it is important to investigate the capability of the system analysis code to reliably predict the heat transfer and natural circulation flow, which are the main phenomena governing the performance of a PSS. Since MARS-KS has been widely validated for heat transfer models, this study focuses on evaluating its capability to predict the single and two-phase pressure drops and natural circulation flow. The straight pipe simulation results indicate that the pressure drop predictions are reliable within ±5 % error margin for the single-phase flow and the errors of pressure drop up to −30 % for the two-phase flow. Through single-phase natural circulation flow analysis, it is concluded that the use of the appropriate K-factor modeling based on the flow regimes is important since the natural circulation flow rate in MARS-KS is mainly affected by the form loss factor modeling. With two-phase natural circulation flow analysis, this study emphasizes the behavior of the system could change significantly depending on the two-phase wall friction and pressure loss modeling. With the analysis results, modeling considerations for the PSS performance evaluation with the system analysis codes are proposed.

System code validation in periodic two-phase flow from low-pressure oscillations

Low-pressure two-phase flow instabilities can potentially challenge start-up transients of water-cooled nuclear reactors. Predicting oscillations caused by flow instability is therefore paramount to reactor safety. While many thermal hydraulics system codes are developed as general-purpose tools, their performance is not guaranteed outside their optimized application ranges. The current study therefore performs validations of two thermal hydraulics system analysis codes, the Adaptive SYStem Thermal-hydraulics Version 3 (ASYST VER3) and Reactor Excursion and Leak Analysis Program MOD3.3 (RELAP5/MOD3.3), in reproducing oscillatory two-phase flow induced by low-pressure flashing instability. Benchmark conditions are selected from a novel dataset resolving void fraction with radial resolution and quantifying periodic behavior with ensemble averaging. Focusing on the prediction of transient two-phase phenomena rather than the determination of stability, validations are conducted by simulating a single channel under prescribed periodic boundary conditions. ASYST exhibits a systematic underprediction of void fraction in an adiabatic chimney downstream of a heated section. The prominent causal discrepancy is identified as a lost travelling void wave due to overpredicted condensation. RELAP5 noticeably overpredicts subcooled boiling and underpredicts flashing, which is consistent with existing validations against steady-state separate-effect tests under low pressure. This degraded code performance also suggests that in low-pressure transients prone to flashing instability, the confidence in RELAP5 derived from existing validations against integral-effect tests shall be carefully limited to validated capabilities in integral systems. In general, the current study fills the previous gap of knowledge about the performance of ASYST and RELAP5 in predicting detailed transient two-phase phenomena in low-pressure flashing-induced oscillations. The identified code defects reveal future directions for code improvements eventually towards reliable application of ASYST and RELAP5 under low pressure beyond their original calibration.

Research on thermal hydraulic characteristics of reflooding process based on core multi-channel model in system analysis code

There are four main stages in PWR Loss of Coolant Accident (LOCA): blowdown, re-irrigation, re-flooding and long-term cooling. The re-flooding stage is a critical phase to avoid core meltdown and achieve long-term cooling. RBHT reflooding experiment was simulated with the RELAP5 and COSINE program respectively.Thermal-hydraulic characteristics in re-flooding phase were analyzed using multi-channel model. Quench time is mainly influenced by system pressure and coolant subcooling. Later quench will result in a higher cladding temperature. Comparatively, the peak cladding temperature is less affected by the power of the heating rod and the re-flooding rate, which fluctuates within 7 %. The higher re-flooding rate allows quenching to occur as fast as possible and improves cooling efficiency. However, higher coolant sub-cooling and re-flooding rates may cause greater thermal stresses on the core and endanger core integrity. Based on the analysis of the re-flooding rate on the peak cladding temperature and quench time, the optimal re-flooding rate interval of 0.099–0.105 m/s was obtained.

Validation of transient SYSTEM analysis code GINKGO based on a SLB TEST

Transient System analysis code is an important tool for reactor thermal hydraulic analysis. GINKGO is a transient system analysis in-house code developed by China General Nuclear Power Corporation (CGN). The code is primarily used for the analysis of Non-Loss of Coolant Accident (Non-LOCA)transients. Steam Line Break (SLB) accident is one of the representative accident condition of Non-LOCA transients. To validate the capability of GINKGO to simulate SLB accident scenario, a test of small non-isolatable main steam line break (10 %) at hot standby conditions, PKL III B5.1 test is selected. The characteristic of PKL facility, the PKL III B5.1 test sequence and the validation process of PKL III B5.1 test by GINKGO code are briefly introduced in the paper. In validation process, GINKGO simulates the steady state first and then simulates the transient process by setting reasonable initial conditions and boundary conditions. Then, the calculation results of key parameters including temperature in hot leg, temperature in cold leg, Steam Generator (SG) secondary side pressure and Pressurizer (PZR) pressure are compared with the experimental results. Through the rationality of the calculation results, it shows that GINKGO has the capability to simulate the key phenomena in a typical SLB accident scenario.

Design study of a 450MW thermal Modified CANDLE fast reactor using helium gas as a coolant

The Modified CANDLE (MCANDLE) burnup scheme divides the reactor core into several regions in an axial or radial direction with equal volume. It works like the CANDLE burnup scheme except that, in the CANDLE burnup scheme, the reactor core is divided into three main regions in the axial direction. Namely, spent fuel, burning region and fresh fuel region. In this study, the design of a modular modified CANDLE fast reactor using helium gas as a coolant has been performed. One of the important roles of a modified CANDLE fast reactor is that it can utilize natural uranium as fuel without the need for enrichment or reprocessing. This type of reactor can be used, including in developing countries, without the nuclear proliferation problem. Using helium gas as a coolant gives hope for a fast reactor due to its properties, such as its less neutron absorption, less radioactive, and as helium is an inert gas, prevents chemical reactions with other materials. The study was performed on a rector with a thermal power of 450MWth. The active core was divided into ten regions with equal volume in radial directions. The refueling scheme has been optimized every ten years of burnup to obtain a good reactor design. In the beginning, the fuel was put in the first region after ten years of burnup was moved to the second region, after another ten years of burnup was moved to the third region and so on until the tenth region where the fuel gets out. The neutronic calculation has been performed using the SRAC (Standard Reactor Analysis Code system). The collision probability method (PIJ) was employed for cell burnup calculation, and CITATION was used for the reactor core design with JENDL 4.0 as the nuclear data library.

Development of thermo-physical property database and code of fusion materials for CFETR blankets

The conceptual design of China Fusion Engineering Test Reactor (CFETR) has been completed and is currently in the engineering design phase. In this paper, focusing on three categories of candidate blanket designs and divertor designs, the selection schemes for materials of plasma-facing components, solid structure, tritium breeding materials, neutron multiplier, coolant, and divertor were extracted. Thermo-physical property data was compiled to form a material thermo-physical property database, named Thermo-physical Property Database (ThePD), which is the first advanced thermo-physical database for materials of CFETR blankets. Besides, the callable code was generated, which can be directly applied in the system code for the blanket. Based on internationally available database from National Institute of Standards and Technology (NIST) and similar software AP1700, the correctness of the data and the transitional analysis of regions designed by International Association for the Properties of Water and Steam (IAPWS) were verified. Using the thermal-hydraulic system analysis code, integrated validation was performed for the validated thermo-physical property data and code, demonstrating the applicability and accuracy of the data. ThePD is conducive to addressing the issue of incomplete thermo-physical property data faced by blanket design and safety analysis software for CFETR, and it can fill this gap.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

Design of control strategy for reactor power control system of LBE cooled Actinide Burner Reactor (ABR) using system analysis code NUSOL-LMR

The power control strategy for the Lead–Bismuth Eutectic (LBE) cooled Actinide Burner Reactor (ABR) has been designed to assess the closed-loop performance of newly developed control blocks that were not modeled in the NUSOL-LMR code before this research effort. After establishing a sustainable model of ABR in NUSOL-LMR, validation of the model in light of the already-published research is performed. The said closed-loop control strategy (consisting of the power and temperature control blocks) is evaluated under different transient scenarios (reactivity insertion, primary coolant pump fault, and change in demand power levels). The power and temperature control blocks concurrently control reactor power during any transient occurrence. The coupling effect of the primary coolant temperature deviation on the reactor power has been minimized by the proportional and integral (PI) controller in the temperature control block. The simulation results demonstrate that the closed-loop performance of the ABR under the controller action during transient events is quite satisfactory. The closed-loop power regulation performance of the conventional sliding mode controller (SMC) with a PI sliding surface (PISM) in the power control block of NUSOL-LMR is better than that of a simple proportional, integral and derivative (PID) controller. The asymptotic stability of the closed-loop system with the PISM controller has been ensured using the Lyapunov stability theory.

SBLOCA analysis of a floating nuclear reactor under ocean conditions using MARS-KS moving reactor model

With increasing demands to reduce greenhouse gas emissions from the maritime industry, nuclear power has emerged as a potential solution. Among these options, offshore floating reactors have been demonstrated and developed by various organizations and countries. The offshore floating reactors offer advantages such as a long refueling period for enhanced safety and high mobility, enabling access to remote areas. However, when operating on a moving platform, offshore floating reactors are exposed to external forces that can affect the thermal-hydraulic properties within the reactor core and various hydraulic components. To address the safety concerns associated with offshore floating reactors, evaluating their thermal-hydraulic performance under different ocean motions is necessary. Therefore, this study focused on conducting accident analyses in an offshore floating reactor called BANDI-60, specifically investigating a Small Break Loss of Coolant Accident caused by a double-ended guillotine break in the Direct Vessel Injection line using the system analysis code MARS-KS. Various ocean conditions were considered, including static inclination, combined inclination with heaving, and rolling motion. Among these conditions, static inclination, especially when the break was aligned downward with the inclination direction, was identified as the most significant factor affecting the integrity of the reactor core.

Secondary system modeling and primary system coupled simulation using MARS-KS

The flexibility of nuclear energy can be achieved by the load following operation and the thermal energy utilization for hydrogen production, desalination, district heating, and Thermal Energy Storage (TES). Thermal energy utilization is performed by transferring the remaining thermal energy over the energy required for electric power generation without changing the reactor power level significantly. For the Pressurized Water Reactor (PWR), it becomes possible by extracting steam from the secondary system and circulating it to a thermal energy application such as the TES system. However, the steam extraction operation affects the behavior of all thermal–hydraulic (TH) variables in the secondary system and further influences the primary system in terms of coolant temperature, reactivity, and reactor power. For thermal energy utilization, it is necessary to secure a technology that can examine the overall stability of the primary and secondary systems in a Nuclear Power Plant (NPP). In this study, based on the system analysis code MARS-KS, an NPP analysis model for SMART, a small-sized integral type PWR, was developed that can simulate major TH variables behavior in the primary and secondary systems at once. Its strength is that major components in the secondary system are modeled without any boundary conditions under matching the heat and mass balance. From the verification calculation of the steady state, it was confirmed that this SMART analysis model can predict the main TH variables of the primary and secondary systems similarly to the plant data. Furthermore, an example calculation was performed for the TES application. The analysis results show that the analysis model is sufficiently applicable to the transient calculation of the primary and secondary systems during the steam extraction operation. This analysis model is expected to be used for the development of technologies such as load following operation and thermal energy utilization in the future.